OECD LOFT-T--3804
"WW
OECD LOFT Project
Quick-Look Report on OECD LOFT
Experiment LP-FP-2
September 1985
NOTICE
This report is for the benefit of OECD LOFTparticipants and their designees only
This report has been prepared pursuant to the "Agreement on anOECD Project of the LOFT Experimental Programme." It is thepolicy of the Management Board that the Information contained inthis report be used only for the benefit of the participants and theparticipants' designees. The contents of this report should not bedisclosed to others or reproduced wholly or partially unlessauthorized in accordance with the laws, regulations, policies, orwritten permission of the appropriate project participant.
Prepared by EG&G Idaho, Inc.under the direction of the U.S. Department of Energy,Idaho National Engineering Laboratory
0
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QUICK-LOOK REPORT ON OECD LOFTEXPERIMENT LP-FP-2
Authors: - .
J. P. Adams 0J. C. Birchley z
N. Newman N
E. W. Coryell ) MiM. L. Carboneau x ( -<
S. GuntayL. J. Siefken M O
Contributors: X (.
Y. Anoda C)J. Bagues
V. T. BertaE. Borioli0. Briney
J. EstebanD. L. Hagrman
R. HesbolK. J. McKennaD. C. Mecham
S. M. Modro
D. L. Batt, SupervisorFP-2 Experiment Section
D. W. Croucher, ManagerLOFT Experiment Planning and Analysis Branch
P. North, ManagerOECD LOFT Project
QUICK-LOOK REPORT ON OECD LOFTEXPERIMENT LP-FP-2
Authors:
J. P. AdamsJ. C. Birchley
N. NewmanE. W. Coryell
M. L. CarboneauS. Guntay
L. J. Siefken
Contributors:
Y. AnodaJ. Bagues
V. T. BertaE. Borioli0. Briney
J. Esteban0. L. Hagrman
R. HesbolK. J. McKennaD. C. Mecham
S. M. Modro
Published September 1985
EG&G Idaho, Inc.Idaho Falls, Idaho 83415
Prepared for theU. S. Department of Energy
Idaho Operations OfficeUnder DOE Contract No. OE-ACO7-761DO1570
ABSTRACT
Experiment LP-FP-2 was conducted on July 9, 1985, in the Loss-of-Fluid
Test (LOFT) facility at the Idaho National Engineering Laboratory under the
auspices of the Organization for Economic Cooperation and Development
(OECD). The objectives of this experiment were to obtain information on
the release of fission products from the fuel and transport of these
fission products in a vapor and aerosol environment from the primarycoolant system. The thermal/hydraulic boundary conditions during the
release and transport of fission products were based on a V-Sequence
accident. The emergency core cooling (ECC) injection was delayed until
specified temperature limits on the thermal shroud were reached, by which
time the desired time and conditions for fission product release and
transport were achieved. The plant was then brought to a safe condition
with full ECC injection. Specially designed fission product measurements
were made in the primary coolant system and blowdown suppression systemduring the transient and also up to 44 days thereafter, during which time
the plant was maintained in a quiescent state and the two systems were
individually isolated. This document provides an initial assessment of the
experiment that covers the initial conditions, sequence of events,
preliminary results of the fission product behavior within the
thermal/hydraulic boundary conditions, and comparisons of the results with
preexperiment calculations.
ii
SUMMARY
Experiment LP-FP-2, conducted on July 9, 1985, was the second fission
product release and transport experiment and the eighth (and last)
experiment conducted in the Loss-of-Fluid Test (LOFT) facility at the Idaho
National Engineering Laboratory under the auspices of the Organization for
Economic Cooperation and Development. The principal objectives for this
experiment were to determine the fission product release from the fuel and
the subsequent transport of those fission products (in a predominantly
vapor/aerosol environment) from the primary coolant system. The initial
conditions were representative of commercial pressurized water reactor
(PWR) operations. The thermal/hydraulic boundary conditions during fission
product release and transport were based on a V-Sequence accident wherein a
low-pressure injection system (LPIS) line ruptures and the emergency core
cooling (ECC) injection is delayed until fuel rod cladding and control rod
melting and material relocation occurs. The transient was initiated by
scramming the reactor, inserting the center fuel module control rods, and,
after a specific delay, opening a break in the intact loop cold leg. A
second break (simulated LPIS line) was opened 222 s after reactor scram.
The first break was closed prior to fuel rod failure to provide a well
defined path for fission product transport. The transient continued until
control rod and fuel rod cladding melting and fission product release fromthe fuel occurred. The experiment was terminated by injection from both
ECC lines into the reactor vessel downcomer and lower plenum.
The initial assessment of data from instruments monitoring the upper
plenum (Fl) and the reactor vessel outlet (F2) sampling lines indicates
that fission products were sampled and the lines operated as expected.
Since the gamma spectrometer located on the Fl line (G6) failed prior to
the experiment, a gross gamma detector (remote area monitor) was placed on
the top of the reactor vessel and detected fission products from both the
fuel/cladding gap and the fuel as they were transported through the Fl
line. The spectrometer (G2) that monitored the combined effluent from the
Fl and F2 lines during the transient measured several isotopes of xenon and
krypton.
iii
Radiation scans of the simulated LPIS line at the time of the first
postexperiment containment entry indicate that fission products were
collected by the deposition coupons in this line. Metal temperatures at
the deposition coupon locations in both the reactor vessel and simulated
LPIS line were approximately 100 K (180'F) higher than saturation, which
was sufficiently high to ensure fission product deposition in steam.
The gamma spectrometer in the simulated LPIS line (G5) sampled various
isotopes. In order of volatility, these were xenon, iodine, cesium,
tellurium, and rubidium.
The G5 gamma spectrometer, the F2 and F3 aerosol sampling systems, and
the D2 and D3 deposition spool pieces appear to have operated as designed.
High background may limit the applicability of the GI (primary system),
G2 [blowdown suppression tank (BST) vapor], and G3 (BST liquid) gamma
spectrometers during the early part of the posttransient phase, and one or
more of the Dl protected coupons may have been exposed to reflood. There
was a loss of data from the G6 gamma spectrometer, which was partially
recovered with data from a remote area monitor. Also, significant data
were obtained from grab samples and Health and Safety instrumentation.
Experiment predictions indicated that, in order to produce the desired
fission product release and transport boundary conditions, the thermal
transient should produce cladding temperatures of 2100 K (3320°F) or higher
for a minimum of three minutes. During the experiment, cladding
temperatures exceeded 2100 K (3320*F) for at least 4-1/2 min, which was 50%
longer than the minimum 3 min identified prior to the experiment. As a
result, the final fission product concentrations in the primary coolant
system and blowdown suppression tank are expected to be higher than those
which were predicted, thus enhancing the detectability of low-yield fission
products.
Comparison with the measured thermal/hydraulic response showed that
the predictions were very adequate as a planning tool for this experiment.
The timing and extent of the core thermal response was closely predicted
iv
with the exception of the lack of steam starvation in the upper parts of
the center fuel module. This discrepancy resulted from a
larger-than-predicted center fuel module steam flow which, in turn, is
judged to have been caused by greater-than-calculated depressurization rate
during the high temperature period of the transient. The resistance in the
simulated LPIS line was much greater than modeled. This led to a higher
primary system pressure at the start of the high temperature period and a
continued depressurization during the high temperature period as opposed to
the nearly flat pressure response that was predicted. Inability to
accurately predict the flow resistance in this line was recognized prior to
the experiment as an area of experimental uncertainty, and adequate
contingency measures were included in the Experiment Operating Procedure.
Based on the preliminary information presented herein, the data
obtained from this experiment are considered adequate to meet the fission
product measurement objectives and, ultimately, the overall experiment
objectives, which were to provide data to assess the fission product
release and transport during the early phases of a risk dominant accident
and the capability of computer codes to predict the same.
v
ACKNOWLEDGEMENTS
Since this is the Quick Look Report documenting the final LOFT
experiment, it is appropriate to acknowledge and thank all who have
contributed to the LOFT program, either during the original NRC test
program or during the OECD sponsored program. In a very real sense, the
success of this, the final experiment conducted in LOFT, was the direct
result of all their efforts. Without their professional contributions,
LOFT would not have been prepared for the rigors that were imposed on both
the facility and the staff in preparing for and carrying out LP-FP-2. To
all these, present and past associates, the authors express their thanks.
Those who contributed directly to the production of this report fall intotwo basic categories: contributors--those who contributed ideas, analyses,
time, and support to determine the response of the system during the
experiment--and authors--those who, in addition to the above, also
contributed to the writing found in this report. These are all
acknowledged on the title page. In addition, the authors wish to
acknowledge Darwin Grigg, who provided an excellent service as technical
editor; without his efforts, this report could not have been completed on
the schedule achieved.
vi
CONTENTS
ABSTRACT .............................................................. ii
SUMMARY ............................................................... iii
ACKNOWLEDGEMENTS ...................................................... vi
1. INTRODUCTION ..................................................... I
1.1 Objectives ................................................. 4
1.2 Experiment Description ..................................... 5
1.3 Systems Description ........................................ 8
2. EXPERIMENT CONDUCT ............................................... 12
2.1 Initial Conditions and Operational Setpoints ............... 12
2.2 Chronology of Events ....................................... 17
3. PCS THERMAL/HYDRAULIC RESULTS .................................... 23
3.1 Blowdown Hydraulics ........................................ 23
3.2 Core Thermal Response ...................................... 27
3.3 Comparison with Calculations ............................... 33
3.4 Metal/Fluid Conditions Near the FPMS ....................... 40
3.5 Summary .................................................... 40
4. FISSION PRODUCT RESULTS .......................................... 43
4.1 ORIGEN 2 Results for the LP-FP-2 Experiment ................ 43
4.2 Results of the Elemental Release Calculations .............. 46
4.3 FPMS Performance ........................................... 51
4.4 Instrument Operation ....................................... 53
4.5 Preliminary Results ........................................ 56
4.5.1 Fl and F2 Sample Lines ............................. 574.5.2 Deposition Measurements ............................ 574.5.3 G5 Gamma Spectrometer .............................. 60
vii
4.5.4 GI, G2, and G3 Gamma Spectrometers ................. 694.5.5 Grab Samples ....................................... 69
4.6 Preliminary Analysis of the BST and G5 Data ................ 69
4.7 Potential for Meeting Fission Product MeasurementObjectives ................................................ 76
4.8 Future PIE Plans ..................................... ...... 77
5. CONCLUSIONS ...................................................... 79
6. REFERENCES ...................................................... 81
APPENDIX A--REVISIONS TO THE EXPERIMENT SPECIFICATION DOCUMENT FOREXPERIMENT LP-FP-2 ......................................... A-1
APPENDIX B--FISSION PRODUCT MEASUREMENT SYSTEM FOR LP-FP-2 ............ B-1
APPENDIX C--DESCRIPTION OF THE LOFT SYSTEM AND INSTRUMENTATION ......... C-i
APPENDIX D--PCS THERMAL/HYDRAULIC RESPONSE .......................... D-l
APPENDIX E--CORE THERMAL RESPONSE ..................................... E-l
APPENDIX F--COMPARISON OF THERMAL/HYDRAULIC DATA WITH PREEXPERIMENTCALCULATIONS................................... F-i
APPENDIX G--SPECIAL INSTRUMENTATION ................................... G-l
APPENDIX H-- SCDAP/RELAP5/TRAP-MELT CODE CALCULATION AND DATACOMPARISONS ............................. H-1
APPENDIX I--QUALIFIED TRANSIENT DATA PLOTS ............................ I-1
APPENDIX J--ORIGEN2 RESULTS FOR THE LP-FP-2 EXPERIMENT ................ J-l
FIGURES
1. Preexperiment core power history .................................. 7
2. FPMS schematic ..................................................... 9
3. Axonometric representation of the LOFT primary coolant system 11
4. Primary system pressure (short term) .............................. 22
5. Primary system pressure (full term) .............................. 22
viii
6. Intact loop hot leg density ....................................... 24
7. Conductivity level probe response above Fuel Assembly 3 ........... 25
8. Comparison of cladding temperatures at the 1.14-, 0.38-, and0.28-m (45-, 15-, and 11-in.) elevations in Fuel Assembly 2with saturation temperature. (See Appendix I for thermocouplequalification limits) ............................................. 28
9. Comparison of cladding temperatures at the 1.07-, 0.69-, and0.25-m (42-, 27-, and 10-in.) elevations in Fuel Assembly 5with saturation temperature. (See Appendix I for thermocouplequalification limits) ............................................. 28
10. Comparison of three guide tube temperatures at the 0.69-m(27-in.) elevation in Fuel Assembly 5. (See Appendix I forthermocouple qualification limits) ............................... 29
11. Comparison of two cladding temperatures at the 0.69-m (27-in.)elevation in Fuel Assembly 5. (See Appendix I forthermocouple qualification limits) ............................... 31
12. Comparison of four external wall temperatures at the 1.07-,0.81-, 0.69-, and 0.25-m (42-, 32-, 27-, and 10-in.)elevations on the south side of the flow shroud. (SeeAppendix I for thermocouple qualification limits) ................ 31
13. Comparison of cladding temperatures at the 1.24-, 0.99-,0.71-, and 0.28-m (49-, 39-, 28-, and 11-in.) elevations inFuel Assembly 2. (See Appendix I for thermocouplequalification limits) ............................................ 32
14. Comparison of two cladding temperatures at the 0.69-m (27-in.)elevation in Fuel Assembly 5 with saturation temperature.(See Appendix I for thermocouple qualification limits) ........... 34
15. Comparison of primary system pressure with preexperimentcalculations made using RELAP5/MOD2 (short term) ................. 36
16. Comparison of primary system pressure with preexperimentcalculations made using RELAP5/MOD2 (full term) .................. 36
17. Comparison of the measured cladding temperature at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BDI.(Thermocouple qualified throughout) .............................. 38
18. Comparison of the measured cladding temperature at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BDI.(Thermocouple qualified to 1720 s) ............................... 38
ix
19. Comparison of the measured cladding temperature at the 1.07-m(42-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BOl.(Thermocouple qualified to 1510.3) .............................. 39
20. Comparison of metal and fluid temperatures at the lower 01deposition coupon location with saturation temperature. (SeeAppendix I for thermocouple qualification limits) ................ 41
21. Comparison of fluid temperature in the Fl aerosol sample linewith saturation temperature. (See Appendix I for thermocouplequalification limits) ............................................ 41
22. Release rate functions taken from NUREG-0772 ..................... 47
23. Comparison of measured cladding temperatures at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits) ................ 47
24. Comparison of measured cladding temperatures at the 1.07-m(42-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGER1 calculation for this elevation. (SeeAppendix I for thermocouple qualification limits) ................ 48
25. Comparison of measured cladding temperatures at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits) ................ 48
26. Input model of the center fuel assembly for the CORSOR/TIGERIcalculation ...................................................... 49
27. Elemental cumulative fractional release inventories calculatedusing TIGERI ..................................................... 49
28. Release rate functions for Cs, I, Sb, Xe, Kr, and Tecalculated using TIGERI .......................................... 50
29. Comparison of the radiation area monitor response on the Flaerosol sample line with fuel centerline temperature(TC-5108-027). (See Appendix I for thermocouple qualificationlimits) .......................................................... 58
30. Measured pressure upstream of the critical orifice in theFl aerosol sample line ........................................... 58
31. Measured pressure upstream of the critical orifice in theF2 aerosol sample line ........................................... 59
x
32. Measured metal temperatures at the Dl deposition couponlocations. (See Appendix I for thermocouple qualificationlimits) ..........................................................
33. Measured metal temperatures at the 02 and D3 deposition couponlocations. (See Appendix I for thermocouple qualificationlimits) ..........................................................
63
63
34. Measured 1311corrected for
35. Measured 1321corrected for
36. Measured 1331corrected for
37. Measured 1341corrected for
38. Measured 1351corrected for
39. Measured 8 8 Rbcorrected for
concentration in the simulated LPIS line (notplateout) ..........................................
concentration in the simulated LPIS line (notplateout) ..........................................
concentration in the simulated LPIS line (notplateout) ..........................................
concentration in the simulated LPIS line (notplateout) ..........................................
concentration in the simulated LPIS line (notplateout) ..........................................
concentration in the simulated LPIS line (notplateout) ..........................................
40. Measured 13 5 Xe concentration in the simulated LPIS line (notcorrected for plateout) ..........................................
41. Measured 13 2Te concentration in the simulated LPIS line (notcorrected for plateout) ..........................................
42. Measured 138 Cs concentration in the simulated LPIS line (notcorrected for plateout) ..........................................
43. Estimates of the elemental Cs, I, and Rb mass concentrationsin the simulated LPIS line based on isotopic activitiesmeasured by the G5 gamma spectrometer ............................
B-l. FPMS Schematic ..................................................
B-2. Schematic of Fl and F2 aerosol sample systems ...................
B-3. Sample line probe ...............................................
B-4. Cyclone separator/isolation valve ...............................
B-5. Impactor and filter train .......................................
B-6. Three stage virtual impactor ....................................
64
64
65
65
66
66
67
67
68
75
B-2
B-4
B-6
B-7
B-7
B-8
xi
B-7. Simulated LPIS line components .................................. B-10
C-I. Axonometric representation of the LOFT system for ExperimentLP-FP-2 ......................................................... C-2
C-2. Schematic of the LOFT primary and emergency core cooling
systems .......................................................... C-3
C-3. Simulated LPIS breakline instrumentation ........................ C-4
C-4. LOFT reactor vessel ............................................. C-5
C-5. LOFT cladding and guide tube thermocouple locations ............. C-8
C-6. Center fuel assembly instrumentation locations .................. C-9
C-7. LOFT steam generator instrumentation ............................ C-10
C-8. Reactor vessel upper plenum instrumentation locations ........... C-l1
D-1. Intact loop hot leg densities ................................... D-2
D-2. Intact loop cold leg densities .................................. D-2
D-3. Broken loop hot leg densities ................................... D-3
D-4. Broken loop cold leg densities ................................... D-3
0-5. Comparison of upper plenum fluid temperature with saturationtemperature. (See Appendix I for thermocouple qualificationlimits) ........................................................... D-4
0-6. Conductivity level probe response above Fuel Assembly 3 ......... D-6
D-7. Conductivity level probe response in Fuel Assembly I ............ D-7
D-8. Conductivity level probe response in Fuel Assembly 3 ............ D-8
D-9. Response of SPND in Fuel Assembly 2 to core uncovery ............ D-10
D-10. Response of SPND in Fuel Assembly 4 to core uncovery ........... D-10
V-li. Response of SPND in Fuel Assembly 6 to core uncovery ........... D-l1
D-12. Averaged BST liquid level ...................................... D-11
D-13. Primary coolant system mass inventory .......................... D-12
D-14. Comparison of fluid temperature in the Fl aerosol sampleline with saturation temperature. (See Appendix I forthermocouple qualification limits) ............................. D-16
xii
D-15. Comparison of fluid temperature in the F2 aerosol sampleline with saturation temperature. (See Appendix I forthermocouple qualification limits) ............................. D-16
D-16. Comparison of metal and fluid temperatures at theD2 deposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... D-17
0-17. Comparison of fluid temperature at the 03 deposition couponlocation with saturation temperature. (See Appendix I forthermocouple qualification limits) ............................. D-17
D-18. Comparison of metal and fluid temperatures at the DI upperdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... 0-18
D-19. Comparison of metal and fluid temperatures at the Dl middledeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... D-18
D-20. Comparison of metal and fluid temperatures at the Dl lowerdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits) ......... D-19
D-21. Comparison of primary system pressure with pressuresmeasured in the accumulators ................................... D-19
D-22. Primary system pressure during the posttransient phase ......... 0-21
D-23. Primary system fluid temperature during the posttransientphase .......................................................... D-21
E-1. Comparison of cladding temperatures at the 1.14-, 0.38- and0.28-m (45-, 15- and 11-inch) elevations in Fuel Assembly 2with saturation temperature during early stages of heatup(600 to 1000 s). (See Appendix I for thermocouplequalification limits) ........................................... E-2
E-2. Comparison of cladding temperatures at the 1.07-, 0.69- and0.25-m (42-, 27- and 10-inch) elevations in the center fuelassembly with saturation temperature during early stages ofheatup (600 to 1000 s). (See Appendix I for thermocouplequalification limits) ........................................... E-2
E-3. Effect of flow changes on rate of temperature increasemeasured at 0.69-m (27-inch) elevation on fuel rod claddingin Fuel Assembly 4 (700 to 1400 s). (See Appendix I forthermocouple qualification limits) .............................. E-3
xiii
E-4. Effect of presence of control rods on guide tube temperatureincrease at the 0.69-m (27-inch) elevation in the center fuelassembly (1100 to 1600 s). (See Appendix I for thermocouplequalification limits) ........................................... E-5
E-5. Comparison of fluid temperatures at upper tie plate aboveFuel Assembly 4 with saturation temperature (600 to 2000 s).(See Appendix I for thermocouple qualification limits) .......... E-5
E-6. Comparison of fluid temperatures at lower tie plate belowFuel Assembly 4 with saturation temperature (600 to 2000 s).(See Appendix I for thermocouple qualification limits) .......... E-6
E-7. Effect of metal-water reaction on guide tube temperatureincrease at 0.69-m (27-inch) elevation in center fuelassembly (600 to 1600 s). (See Appendix I for thermocouplequalification limits) ........................................... E-6
E-8. Measurement of gross gamma activity near reactor vessel head(600 to 2000 s) ................................................. E-9
E-9. Comparison of fluid temperatures at different horizontallocations on center fuel assembly upper tie plate(600 to 1800 s). (See Appendix I for thermocouplequalification limits) ........................................... E-9
E-l0. Momentum flux in reactor vessel downcomer (-200 to 2000 s) ..... E-10
E-11. Cladding temperatures at 0.25-m (10-inch) elevation incenter fuel assembly during high temperature stage oftransient (1200 to 2000 s). (See Appendix I forthermocouple qualification limits) ............................. E-12
E-12. Comparison of cladding temperature at 0.38-m (15-inch)elevation in Fuel Assembly 4 with saturation temperature(600 to 1900 s). (See Appendix I for thermocouplequalification limits) .......................................... E-12
E-13. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-m(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 8 with saturation temperature (600 to 1900 s).(See Appendix I for thermocouple qualification limits) ......... E-14
xiv
E-14. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-m(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 4 (600 to 1900 s). (See Appendix I forthermocouple qualification limits) ...................... k ...... E-14
E-15. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-m(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 2 (600 to 1900 s). (See Appendix I forthermocouple qualification limits) ............................ E-15
E-16. Comparison of temperatures at 0.25-, 0.69-, 0.81- and 1.07-n(10-, 27-, 32- and 42-inch) elevations on shroud wall facingFuel Assembly 6 (600 to 1900 s). (See Appendix I forthermocouple qualification limits) ............................. E-15
E-17. Comparison of cladding temperatures at 0.28-, 0.71-, 0.99-and 1.24-m (11-, 28-, 39- and 49-inch) elevations in FuelAssembly 2 (1400 to 1900 s). (See Appendix I forthermocouple qualification limits) ............................ E-16
E-18. Comparison of cladding temperatures at 0.28-m (11-inch)elevation on fuel rods in peripheral assemblies 2, 4, and6 close to shroud (600 to 2000 s). (See Appendix I forthermocouple qualification limits) ............................. E-16
E-19. Comparison of fluid temperature at upper tie plate abovecenter fuel assembly with saturation temperature duringreflood (1750 to 1850 s). (See Appendix I for thermocouplequalification limits) .......................................... E-18
E-20. Effect of material relocation on cladding temperatures at0.69-m (27-inch) elevation in center bundle during reflood(600 to 2100 s). (See Appendix I for thermocouplequalification limits) .......................................... E-18
F-1. RELAP5/MOD2 nodalization diagram ................................ F-4
F-2. TRAC-LOFT nodalization .......................................... F-1l
F-3. TRAC-LOFT center assembly rod grouping .......................... F-13
F-4. TRAC-LOFT peripheral assembly rod grouping .................... F-14
F-5. Secondary system pressure (0 to 400 s) .......................... F-16
F-6. Secondary system pressure (0 to 2000 s) ......................... F-16
F-7. Secondary system liquid level ................................... F-17
F-8. Primary system hot leg pressure (0 to 400 s) .................... F-19
xv
F-9. Primary system hot leg pressure (0 to 2000 s) ................... F-19
F-10. Pressure drop along LPIS line .................................. F-21
F-1l. LPIS line mass flow rate ....................................... F-21
F-12. Progression of core uncovery ................................... F-24
F-13. Fuel rod cladding temperature in center fuel assembly,0.25-m (10-in.) elevation. (Thermocouple qualifiedthroughout) .................................................... F-24
F-14. Fuel rod cladding temperature in center fuel assembly,0.69-m (27-in.) elevation. (Thermocouple qualified to1720 s) ........................................................ F-26
F-15. Fuel rod cladding temperature in center fuel assembly,1.07-m (42-in.) elevation. (Thermocouple qualified to1510 s) ........................................................ F-26
F-16. Fuel rod cladding temperature in peripheral fuel assembly,0.66-m (26-in.) elevation. (Thermocouple showed possibleshunting after 1700 s) ......................................... F-28
F-li. Shroud outer wall temperature at 0.69-m (27 in.) elevation.(Thermocouple qualified to 1790 s) .......................... F-28
G-l. Response of SPND at the 0.69-m (27-in.) elevation in FuelAssembly 5 ...... .......................................... G-3
G-2. Response of SPND at the 0.28-m (11-in.) elevation in Fuel
Assembly 5 ---..................................... . ....... G-5
G-3. Iodine species sampler ...................................... G-9
G-4. Cutaway of the LOFT reactor vessel illustrating the locationof PSU detectors ... o...................... ... G-12
G-5. Planar view of the LOFT reactor vessel illustrating thelocation of PSU detectors ...... ... .................... G-13
G-6. Normalized current response of PSU detectors (0 to 120 s) ...... G-14
G-7. Normalized current response of PSU detectors (0 to 1800 s) ...... G-16
G-8. Normalized pulse height response of PSU detectors(0 to 1800 s) .............................. G-17
G-9. Normalized pulse height response of PSU detectors(1800 to 3600 s) ...... .......... ..... ............... G-18
xvi
H-I. Comparison of measured hot leg pressure with pressurecalculated using the integrated code ............................ H-6
H-2. Comparison of measured cladding temperature at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits) ........................................... H-6
H-3. Calculated cladding temperature at the 1.26-m (50-in.)elevation in Fuel Assembly 5 made using the integrated code.(See Appendix I for thermocouple qualification limits) .......... H-8
H-4. Comparison of measured cladding temperature at the 0.94-m(37-in.) elevation in Fuel Assembly 4 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits) ........................................... H-8
TABLES
1. Initial conditions for Experiment LP-FP-2 ........................ 13
2. Operational setpoints for Experiment LP-FP-2 ..................... 16
3. Chronology of events for Experiment LP-FP-2 ...................... 18
4. Selected ORIGEN2 inventory results for the Experiment LP-FP-2center fuel bundle at 430 MWD/MTU burnup ......................... 45
5. Comparison of TIGERl and CORSOR cumulative release fractions forExperiment LP-FP-2 (at 1800s) .................................... 52
6. Experiment LP-FP-2 fission product measurement system sequence
of events ........................................................ 54
7. Iodine species identified by G5 spectrometer ..................... 61
8. Non-iodine species identified by G5 spectrometer ................. 62
9. BST liquid grab sample preliminary results ....................... 70
10. 8ST vapor grab sample gamma spectroscopy results ................. 71
11. BST vapor grab sample mass spectroscopy results .................. 71
12. Cumulative release fractions to the BST ........................ 73
13. Planned postirradiation examination for Experiment LP-FP-2 ....... 78
C-l. Initial conditions for Experiment LP-FP-2 ....................... C-7
xvii
E-1. Times for the center fuel module thermocouples to reach1800 K (2780°F) ................................................. E-8
F-1. Initial conditions for Experiment LP-FP-2 ....................... F-8
G-1. Measured 131I concentrations .................................... G-7
G-2. Measured 1311 species admixture ................................. G-8
H-l. Description of modeling of reactor core ......................... H-3
H-2. Conduct of Experiment LP-FP-2 assumed in preexperimentprediction ...................................................... H-4
1-1. Listing of qualified data fiche ................................. 1-2
xviii
QUICK-LOOK REPORT ON OECD LOFT
EXPERIMENT LP-FP-2
1. INTRODUCTION
This report documents the preliminary results and analyses of
Experiment LP-FP-2, which was conducted on July 9, 1985, in the
Loss-of-Fluid Test (LOFT) facility located at the Idaho National
Engineering Laboratory (INEL). This fission product release and transport
experiment was the eighth and final experiment conducted under the auspices
of the Organization for Economic Cooperation and Development (OECD).
Experiment LP-FP-2 provides information on the release and transport
of fission products and aerosols in a severe core damage scenario. The
nature of the phenomena governing fission product and aerosol release and
transport can be linked to potential pressurized water reactor (PWR) system
thermal hydraulics and core thermal response leading to fuel failure and
fission product transport behavior. The fuel rod cladding temperatures in
the center fuel module (CFM) exceeded 2100 K (3320'F) for an estimated
4-1/2 min before temperature limits for experiment termination were
reached. The 4-1/2 min fission product release and transport transient
simulates the initial part of a severe damage transient without emergency
core cooling (ECC) system operation wherein the core thermal induced damage
originated from a V-Sequence scenario.
Probabilistic Risk Assessment (PRA) studies 1 have shown that the
interfacing systems loss-of-coolant accident (LOCA), a hypothetical event
first postulated in the Reactor Safety Study2 and labeled the V sequence,
has a significant contribution to the risk associated with PWR operation.
Therefore, this risk dominant accident sequence was selected as the
thermal/hydraulic environment in which fission product release and
transport would be measured in Experiment LP-FP-2. The specific
interfacing systems LOCA associated with the significant operational risk
is a pipe break in the low pressure injection system (LPIS), also called
the residual heat removal system (RHRS). This system typically serves two
I
functions in a commercial PWR: (a) it provides emergency coolant injection
for core recovery during intermediate and large break LOCAs and (b) it
provides for decay heat removal during normal shutdown. The RHRS
represents a potential path for release of primary coolant from the
containment. If core cooling were not maintained during such an event and
failure of the fuel rods were to occur, fission product release to the
environment could occur through concomitant failure of the auxiliary
building.
Experiment LP-FP-2 simulated the system thermal/hydraulic and core
uncovery conditions during fission product release and transport that are
calculated to occur in a four-loop PWR from rupture of an RHRS pipe as a
result of a V sequence accident. The initial conditions were
representative of commercial PWR operations. The break size resulted in a
depressurization that was bounded by previously conducted LOFT
Experiments L8-2 and L5-1 3 ' 4 on the upper end and by
Experiments L3-1, 5 ' 6 L3-5/3-5A,7, 8 and L3-6/L8-1 9 ' 10 on the lower
end. The transient was initiated by a reactor scram followed by insertion
of the central fuel assembly control rods (designed to provide typical
control rod behavior during the transient). A break line in the intact
loop cold leg was then opened to start the depressurization. A second
break path, which simulated the LPIS line, was opened in the broken loop
cold leg. The intact loop cold leg break was closed in accordance with
procedure, but the subsequent depressurization was too slow and the
pressure remained too high for FPMS operation. Therefore, both this break
and the power operated relief valve were opened for a brief period to
depressurize the primary system prior to fission product release. The core
was allowed to uncover and to heat up until a high temperature trip on the
shroud outside wall was reached. By that time, the central fuel assembly
had reached an estimated maximum temperature in excess of 2400 K (3860°F)
and had been above 2100 K (3320'F) for at least 4-1/2 min. The emergency
core cooling (ECC) system was then activated to reflood the reactor vessel
and recover the plant.
2
The requirements imposed on Experiment LP-FP-2 from the standpoint of11
facility decontamination and recovery were:
1. Experiment LP-FP-2 must be conducted with peripheral assembly
fuel rod cladding temperatures limited to 1533 K (23000 F).
2. The structural integrity of the center fuel assembly must be
maintained to facilitate removal from the reactor vessel.
To meet the above facility requirements, a center fuel assembly was
designed and fabricated specifically for this experiment. This fuel
assembly contained 9.74-wt%-enriched, 2.41-MPa-(350-psia)-prepressurized
fuel rods and was separated from the peripheral fuel rods by a 2.5-cm
(l-in)-thick, canned, zirconium-oxide thermal shroud. The center fuel
assembly was designed to enable the 9.74-wt%-enriched fuel rods to heat up
above 2100 K (3320°F) while maintaining the peripheral fuel rods below
their temperature limit sufficiently long to allow fission product release
and transport.
Section 2 presents an evaluation of the plant performance during
Experiment LP-FP-2, and includes a summary of specified11 and measured
initial conditions, a list of operational setpoints, a chronological
listing of identifiable significant events, and a description of the LOFT
system geometry. Section 3 presents a summary of the PCS thermal/hydraulic
boundary conditions measured during Experiment LP-FP-2, including summariesof the core thermal response and the comparison between the measured data
and the preexperiment calculations. Section 4 presents the fission product
measurement results, including a preliminary assessment of the
achievability of the measurement objectives. Section 5 presentsconclusions based on a preliminary examination of the results discussed in
Section 3 and relates those conclusions to the experiment objectives.
Appendix A contains revisions to the Experiment Specification Document
(ESD), which were made subsequent to its issue. That appendix may be
removed from this report for insertion of the replacement pages into the
ESO. Appendix B contains a description of the FPMS, and Appendix C, a
3
description of the LOFT system and thermal/hydraulic instrumentation.
Appendix D contains the detailed primary system thermal/hydraulic
response. Appendix E contains a discussion of the detailed core thermal
response, and Appendix F contains the detailed comparison of data with
predictions. Appendix G contains the results from three special instrument
systems which were installed specifically for this experiment and which
were beyond the scope of the experiment design. Appendix H presents a
brief summary of the comparison of the data with predictions made using a
combined RELAP5/MOD2, SCDAP, and TRAP-MELT computer code. Appendix I
contains plots of all the qualified data recorded during the transient
phase, and Appendix J contains the source term calculations. These last
two appendices are contained on microfiche inside the back cover.
1.1 Objectives
The governing objective for Experiment LP-FP-2 was:
Obtain fission product release, transport, and deposition data during
the early phases of a risk dominant reactor transient to establish a
benchmark data base for:
1. Assessing the understanding of the physical phenomena controlling
reactor system fission product behavior.
2. Assessing the capability of computer models to predict reactor
system fission product release and transport.
To support that governing objective, the following two
thermal/hydraulic and four fission product objectives were defined:
Thermal/hydraulic
1. Provide LPIS interfacing system LOCA thermal/hydraulic conditions
from the initiation of the LPIS pipe break through the early
phases of severe core damage.
4
2. Provide transient fuel rod cladding temperatures in the center
fuel assembly up to the rapid metal-water reaction temperature of
2100 K (3320 0 F) and concomitant aerosol generation from the
(Ag-In-Cd) control rods.
Fission Product
1. Determine the fraction of the volatile fission products (Cs, I,
Te, Xe, Kr) and aerosols released to and from the upper plenum
region.
2. Determine the fraction of volatile fission products and aerosols
transported out of the primary coolant system.
3. Determine the retention of volatile fission products on
representative primary coolant system surfaces in the plenum and
piping.
4. Determine the general mass balance of volatile fission products
in the fuel, primary coolant system and blowdown tank.
Due to the preliminary nature of the data analysis for this
experiment, the governing objective will not be discussed. However, a
preliminary assessment of each of the specific measurement objectives is
presented.
1.2 Experiment Description
As with OECD LOFT Experiment LP-FP-1 12 (a previously conducted
fission product experiment), Experiment LP-FP-2 consisted of four phases,
designated (1) fuel preconditioning, (2) pretransient, (3) transient, and
(4) posttransient. The four phases were contiguous and had specificphenomenologically defined beginnings and endings.
5
The purpose of the fuel preconditioning phase, in conjunction with the
pretransient phase, was to subject the 9.72-wt%-enriched fuel rods to a
minimum burnup of 325 MWD/MTU. This was done by operating the core at a
thermal power of 32 MW for 80 h, shutting down for 75 h, and operating at
26.5 MW for a period of 80 h. This phase started when the plant was heated
up just prior to power operations and ended after the required burnup in
this phase (at least 252 MWD/MTU) was reached.
The pretransient phase had, as its purposes, the completion of the
burnup in the 9.72-wt%-enriched fuel and the establishment of the initial
conditions for the experiment. Figure I shows the pretransient power
history for this experiment. The initial condition requirements included
short-lived decay heat buildup (685 kW at 200 s after reactor scram),
pressure, temperature, flow, etc., which simulated typical operation of
commercial PWRs. This phase began upon termination of the preconditioning
phase and ended upon initiation of the transient phase after 30 h at 31 MW
followed by 15 h at 26.5 MW.
The transient phase started with a reactor scram and ended when the
simulated LPIS line was closed. Plant actions taken during this phase
comprised turning off the primary coolant pumps and inserting the central
fuel assembly control rods within 20 s of reactor scram, opening first the
intact loop cold leg and then the broken loop hot leg (simulated LPIS line)
breaks, closing the intact loop cold leg break, and then recycling the
intact cold leg break and cycling the PORV prior to fission product
release. This phase was terminated when the shroud external temperature
reached 1517 K (2271*F), at which time the reflood of the reactor vessel
was initiated. The maximum cladding temperature measured during this
experiment exceeded 2400 K (3860°F), and the time at temperature [time with
cladding temperatures in excess of 2100 K (3320°F)] was 4-1/2 min.
The final, or posttransient, phase consisted of a time interval of
44 d during which time the redistribution of fission products in the gas
and liquid volumes in the blowdown suppression tank and the leaching of
fission products from the damaged fuel rods in the reactor vessel were
6
40 I I
30
20
0~1010 r
1 :
-500 -400 -300 -200 -100Time (hr)
Figure 1. Preexperiment core power history.
0
7
measured. This phase initiated upon closure of the simulated LPIS line,
which terminated the blowdown and initiated the reflood of the reactor
vessel, and ended 44 d later.
1.3 Systems Description
A fission product measurement system (FPMS) was designed and
fabricated for use during Experiment LP-FP-2. This system (illustrated inFigure 2) consisted of three basic subsystems: the gamma spectrometer
system, the deposition coupons, and the filter sampling system. Each of
these is briefly described in turn in this section. Appendix B contains a
more detailed description of the FPMS.
The gamma detection sampling system included five different sample
locations: Gl (spectrometer), which sampled from the reactor vessel lower
plenum or, alternately, from the intact loop hot leg; G2 (spectrometer),
which sampled from the blowdown suppression tank vapor spaces;
G3 (spectrometer), which sampled from the blowdown suppression tank liquid
spaces; G5 (spectrometer), which sampled from the simulated LPIS line; and
G6 (gross gamma monitor), which sampled from the upper plenum. (G4 was
used during Experiment LP-FP-l and was not used in this experiment.) Each
gamma spectrometer was designed to operate remotely and could be calibrated
using a 2 28 Th source mounted on a collimator wheel. With the exception
of G5 and G6, this system operated only during the posttransient phase.
(Additionally, the G-2 spectrometer measured the activity from the combined
Fl and F2 aerosol sample line during the transient.) G5 and G6 operated
during the transient and posttransient phases.
The deposition sampling system consisted of six stainless steel
coupons and two deposition spool pieces. Two of these were located at each
of three elevations above the central fuel assembly (for a total of six
coupons, collectively designated DI). At each elevation, both coupons were
exposed to the fluid stream during the transient phase. One coupon at each
elevation was to be isolated from the fluid prior to initiation of reflood
while the other coupon remained exposed. This system design, if
successfully operated, would have allowed differentiation between the
8
F2
5 4066
Figure 2. FPMS schematic.
9
leaching and deposition during the reflood and the deposition during the
heatup phase. During the experiment, the cover did not seal around the dry
coupons and contact with the reflood water may have occurred. Two
deposition spool pieces, located at the inlet and outlet of the simulated
LPIS line header, were designated D2 and D3, respectively. These coupons
were designed to provide a measurement of the primary coolant system
surface deposition of volatile fission products during the heatup or
transient phase. Since this line was isolated prior to reflood, these
coupons did not experience any deposition or leaching subsequent to the
transient phase.
The final FPMS subsystem was the aerosol and steam sampling system.
This system was designed to provide a sample of the vapor and aerosols
generated during the heatup phase of the experiment. The Fl filter
sampling line consisted of the following major components, in order:
sample line probe, dilution gas supply, cyclone separator and isolation
valve, dilution filter, virtual impactor, collection filters, infrared
moisture detectors, and hydrogen recombiner. The F2 sampling line was
similar except there were no dilution gas supply and moisture detectors.
The F3 filter sampling line consisted of a filter, flow venturi, and Dl
and D3 deposition coupons. The three sample locations were: Fl, 180 cm
(70.75 in.) above the top of the lower tie plate and located directly above
the center fuel assembly; F2, the broken loop hot leg spool piece just
outside of the upper plenum; and F3, the exit of the simulated LPIS line
header.
Figure 3 is an axonometric representation of the LOFT primary coolant
system. The system consists of the reactor vessel, which houses the 1.68-m
(5.5-ft) nuclear core; an intact loop, which represents three of four loops
of a four-loop PWR and which contains active components (steam generator,
pumps, pressurizer, etc.); a broken loop, which represents the fourth loop;
and the blowdown suppression tank, which collected the effluent from the
primary coolant system. The LOFT PCS is volumetrically scaled to a
commercial PWR, using the ratio of core thermal powers (LOFT/PWR) as the
scaling constant. Additional details on the LOFT system and the scaling
basis used in its design are available in Appendix C and in References 13
and 14.
10
C C C
Intact loop Broken loop
Figure 3. Axonometric representation of the LOFT primary coolant system.
2. EXPERIMENT CONDUCT
The experiment conduct is described in this section. The initial
conditions and the operational setpoints for the experiment are presented
in Section 2.1. Section 2.2 briefly describes the sequence of events that
occurred during the experiment.
2.1 Initial Conditions and Operational Setpoints
A summary of the specified and measured system conditions immediately
prior to Experiment LP-FP-2 is given in Table 1. All initial conditions
were within the limits specified by Reference 11 except for the liquid
level in the blowdown suppression tank, which was <1% low. Since no
attempt had been made to simulate the containment with the blowdown
suppression tank, this single out-of-specification value is judged not to
have affected the experiment outcome.
The operational setpoints specified in the ESD for this experiment are
listed in Table 2, together with the measured values. Two of the operator
actions occurred prior to reaching the respective operational setpoint:
isolation of the gamma densitometer sources (683 s early) and initiation of
ECCS flow (1.1 s early). The latter is judged to have no effect on the
data since all FPMS lines were closed prior to injection of the
ECCS water. There are two effects caused by the early isolation of the
gamma densitometer sources: (a) loss of loop fluid density information
subsequent to isolation of the sources and (b) increase in uncertainty of
the fluid density data prior to isolation because the in situ calibration
must now be based on one (as opposed to two) known fluid density. While
the impact of the loss of density information should not be minimized, it
does not affect other data and has only a minimal impact on the
understanding of the experiment because the loops were already partially
voided prior to this time. Thus, the initiation of voiding in the loops
was measured; the major loss is knowledge of when the loops were completely
voided. The latter can be bounded using thermocouple and level information
in the upper plenum as discussed in Section 3.
TABLE I. INITIAL CONDITIONS FOR EXPERIMENT LP-FP-2
SpecifiedaValueParameter
Primary Coolant System
Core delta T (K)(OF)
Primary system pressure(hot leg) (MPa)
(psia)14.95 + 0.12168 + 15.0
MeasuredValue
11.7 + 1.421.1 T 2.5
14.98 + 0.12173 F 15
571.6 + 0.8569.2 T 1.4
559.9 + 1.1548.2 + 2
475 + 2.53.77 + 0.02
Hot leg temperature (K)(OF)
571 + 1.1569 T 2
Cold leg temperature (K)(OF)
Loop mass flow (kg/s)
(lbm/h x 106)
Boron concentration (ppm)
Primary coolant pump injection(both pumps) (L/s)
(gpm)
479 + 193.8 ; 0.15
499 + 15
0.1272.0
+ 0.0167 0.25
0.128 + 0.0031.98 T 0.02
Reactor Vessel
Power level (MW)
Decay heat (200 s)
26.5 + 0.5
685 + 10
26.8 + 1.4
(kW) 684.6
Maximum linear heat generationrate (kW/m)
(kW/ft),4AO--12
Control rod position(above full-in position)
42.6 + 3.612.97 T 1.1
1.38 + 0.0154.3 7 2.0
Wm)(in.)
1.37 + 0.0154.0 7 2.0
13
TABLE 1. (continued)
ParameterSteam Generator
Secondary system pressure (MPa)(psia)
Water levelb (m)(in.)
Pressurizer
Liquid volume (m3(ft])
Steam volume (m3(ftl)
Water temperature (K)(OF)
Pressure (MPa)(psia)
Liquid level (m)
(in.)
Suppression Tank
Liquid level (m)
(in.)
Gas volume (m3
Water temperature (K)(OF)
Pressure (gas space) kPa)t (pma)
Boron concentration (ppm)
SpecifiedaValue
1.12 + 0.144 4
1.19 + 0.051- 0.0
47.0 + 2- 0.0
<311cl100
100 + 2014.7 W 3
MeasuredValue
6.38 + 0.08925 T 12
0.17 + 0.066.7 ; 2.4
0.57 + 0.0320.13 " 1.06
0.37 + 0.0313.07 + 1.06
616.9 + 2.1650.8 T 3.8
15.1 + 0.12190 " 14.5
1.06 + 0.0644.4 + 2.4
1.18
46.5
59.112087
295.672
9513.8
3710
+ 0.06c
+ 2.4
+ 2.02+ 71
+ 0.5Ti1
+3T 0.4
+ 15
14
TABLE 1. (continued)
SpecifiedaValueParameter
MeasuredValue
Emergency Core Cooling System
Borated waterTemperature
Accumulator A
Accumulator A
storage tank(K)(OF)
liquid level (m)(in.)
pressure (MPa)(psia)
Accumulatortemperature
Accumulator
Accumulator
Accumulatortemperature
A liquid(K)(OF)
303 +85 :
<2.17<86
>4.21>611
303 +85 +
<2.16<86
>4.21>611
303 +85 T
35
35
301.3 + 382 7 5
1.81 + 0.0271.3 ; 0.8
5.1 + 0.06740 ; 9
303.1 + 0.786 ; 1.3
1.81 + 0.0271 T 0.8
4.95 + 0.06718 T 9
305.6 + 0.790.4 T 1.3
B liquid level (m)(in.)
B pressure (MPa)(psia)
6 liquid(K)(OF)
35
a. If no value is listed, none was specified
b. Steam generator liquid level referenced to 2.95 m (116 in.) above thetop of the tube sheet.
c. This value is out of specification.
15
TABLE 2. OPERATIONAL SETPOINTS FOR EXPERIMENT LP-FP-2
Event Specified Measured
Scram reactor (s) 0.0 0.0
Turn off primary pumps (s) 8 + 2 9.7 + 0.1
Insert CFM control rodsa (s) 20 22.4 + 0.1
ILCL break opened (s)b 23 32.9 + 0.1
LPIS break opened (s) 220 + 5 221.6 + 0.1
ILCL break closed (s)d 721 735.5 + 0.1
Fl and F2 opened (s)c 905 1015.7 + 0.1
Align LPIS line filter (s)c 945 950.8 + 0.1
Isolate gamma densitometer sources (s)e 945 262 + 2
Close FPMS lines (s)f 1766 1777.1 + 0.1
Close the LPIS line (s)f 1766 1777.6 + 0.1
ECCS flow initiated (s)g 1783.6 + 0.5 1782.6 + 0.1
a. Insertion of the CFM control rods was initiated when the primarycoolant flow decreased to 189 kg/s (1.5 x 106 Ibm/h), as specified.
b. The ILCL break was opened upon verification that the CFM control rodswere fully inserted.
c. The F3 filter and the FPMS line isolation valves were opened whencladding temperatures reached 840K (1052°F), as specified.
d. The ILCL break was closed when cladding temperatures reached 566 K(560°F) or PCS pressure reached 1.2 MPa (160 psig).
e. The gamma densitometer sources were to have been isolated from thedetectors when the cladding temperatures reached 840 K (1052°F).
f. The FPMS sampling line and LPIS line isolation valves were closed whenshroud temperatures reached 1517 K (2272°F).
g. ECC flow was initiated 6 s after initiation of closure of the LPIS lineisolation valves.
2.2 Chronology of Events
Identifiable significant events for Experiment LP-FP-2 are listed in
Table 3, which compares the times of occurrence with the times predicted bythe pre-experimental calculations. Annotated intact loop hot leg pressure
histories are shown in Figures 4 and 5.
The experiment was initiated by scramming the reactor using the
peripheral control rods. The primary coolant pumps were then turned off at
approximately 10 s. After the PCS flow had decreased to 190 kg/s
(1.5 x 105 lbm/h) at 22 s, the center fuel assembly control rods were
inserted. The intact loop cold leg break was opened at 33 s to initiate
the blowdown. This was followed by the opening of the simulated LPIS line
at 222 s. The core started heating up when the liquid level decreased intothe core at 662 s and 689 s in the peripheral and central fuel modules,
respectively. The intact loop cold leg break was closed at 736 s but was
reopened to increase the depressurization rate at 878 s. In addition, the
PORV was opened at 882 s, also to increase the depressurization rate;
however, no effect was measured due to opening the PORV. After a
sufficient depressurization had been achieved, the intact loop cold leg andPORV lines were closed at 1022 and 1162 s, respectively. Fission product
release was first measured in the Fl and F2 lines at approximately 1200 s.The hottest cladding temperatures reached 2100 K (3320°F) by 1504 s. The
transient continued without intervention until the outer shroud walltemperature limitation of 1517 K (2272'F) was reached at 1766 s.
Subsequently, the FPMS lines were isolated at 1777 s and ECC injection was
initiated at 1783 s. The core was quenched at 1795 s (although a few,
isolated thermocouples indicated temperatures in excess of saturation for
several minutes thereafter), and the plant was maintained in a quiescent
state for 14 days while fission product measurements were taken using the
on-line measurements systems. In addition, batch samples were taken from
the BST and PCS for several days thereafter: BST liquid samples (21 d),BST vapor samples (28 d), and PCS liquid samples (44 d). During the early
part of the cooldown, the PORV was cycled twice (see Table 3) to preventthe PCS from overpressurizing, and a feed-and-bleed operation on the steam
generator was initiated.
17
TABLE 3. CHRONOLOGY OF EVENTS FOR EXPERIMENT LP-FP-2
Time After ExperimentInitiation
(s)
Event
Scram
Control rods fully inserted
PCP coastdown initiated
CFM control rods fully inserted
ILCL break initiated
PCP coastdown completeb
End of subcooled blowdownd
Secondary relief valve cycle
Pressurizer empty
LPIS line break initiated
Secondary pressure exceeds primary systempressure
Earliest coolant thermocouple deviationfrom saturation (voidage at that location)
upper plenumhot leg pipedowncomerlower plenum
Fuel rod cladding heatup initiated in
peripheral fuel assembly
Fuel rod cladding heatup initiated in CFM
ILCL break closed
ILCL break opened
PredictedValue
0.0
0.0
25
--a
20
27c
42
51
45
220
220
52.55464
MeasuredData
0.0
2.4 + 0.1
9.7 + 0.1
23.4 + 0.5
32.9 + 0.1
25.1 + 0.1
53 + 1
56 + I
60 + 5
221.6 + 0.1
260 + 10
300390730800
662665
+
+
+
+
+
÷
+
10101020
2
2
0.1
0.1
713.5
1007
-- a
689
735.5
877.6
18
TABLE 3. (continued)
Time After ExperimentInitiation
(s)
Event
PORV opened
F3 filter on line
LPIS bypass closed
FPMS lines opened
ILCL closed
PORV closed
First indication of (gap) fissionproducts at Fl
First indication of (gap) fissionproducts at F2
First indication of (gap) fissionproducts at F3
Peripheral fuel cladding reaches1460 K (2172-F)
Maximum upper plenum coolanttemperature reachedf
First indication of (fuel) fissionfission products at Fl, F2, and F3
Cladding temperatures reach2100 K (33200)
Shroud temperature reachestrip setpoint
1st thermocouple2nd thermocouple
Maximum cladding temperature reached
PredictedValue
-- a
-- a
MeasuredData
882.0 + 0.1
950.8 + 0.1
951.9 + 0.1
1013.1 + 0.1
1021.5 + 0.1
1162.0 + 0.1
1200 + 20
1200 + 20
1249 + 60
-- e
1495 + 5
1500 + 10
1504 + 11721.6
1743 +1766 T
__g
11
19
TABLE 3. (continued)
Time After ExperimentInitiation
(s)
Predicted MeasuredEvent Value Data
LPIS line break closed __h 1777.6 + 0.1
FPMS lines closed ._h 1778.1 + 0.1
Maximum upper plenum metal temperature __h 1780 + 5reachedf
ECCS initiated ._h 1782.6 + 0.1
Accumulator flow stopped --h 1795 + 2
Maximum LPIS line coolant _.h 1800 + 5temperature reached
Core quenched __h 1795 + 5i
Cooldown initiated ._h _.i
Steam generator --h 2600 + 10feed-and-bleed started
PORV opened --h 3350 + 10
PORV closed --h 3380 + 10
PORV opened .-h 3680 + 10
PORV closed --h 3690 + 10
Experiment terminated --h __j
a. This value was not calculated.
b. The pumps were allowed to coastdown under the influence of the motorgenerator flywheel until the pump speed reached 750 rpm. At that time, theflywheel was disconnected from the motor generator and the pumps quicklystopped adding energy to the fluid. The time at which the flywheel wasdisconnected is defined as the time the PCP coastdown was complete.
c. Due to an error in the version of RELAP5/MOD2 that was used for thiscalculation, the initial prediction indicated a sharp pump coastdown.However, a later partial calculation made using a corrected version of thecode indicated completion of the pump coastdown (pump speed below 750 rpm)at 43.5 s.
20
TABLE 3. (continued)
d. End of subcooled blowdown is defined as the time when the first measuredfluid temperature outside of the pressurizer reaches saturation conditions.
e. None of the cladding thermocouples in the peripheral fuel bundlemeasured validated temperatures above the setpoint. The two which gavereadings above this setpoint were failed prior to reaching the setpoint.
f. These temperatures represent the maximum measured temperatures prior toreflood at these locations. The thermocouple output during reflood couldnot be interpreted.
g. Due to the large number of cladding thermocouples in the central fuelmodule that failed at high temperatures during the transient, it is notpossible to determine the precise maximum temperature or the time at whichit occurred. The time is estimated to be between 1782 and 1795 s. Themaximum temperature exceeded 2400 K (3860°F) based on valid temperaturereadings prior to thermocouple failure.
h. The calculations were terminated prior to this event.
i. The peripheral fuel modules were quenched by 1793 s. Most of thecentral fuel module cladding thermocouples were quenched by 1795s. Someisolated thermocouples indicated persistent high (superheated) temperaturesa few minutes longer. Interpretation of the temperature data is complicatedby the large number of thermocouples in the center fuel module that failedduring or just prior to reflood (see Appendix I).
j. Due to the high background in the area surrounding the G1, G2, and G3spectrometers, data were collected for several weeks subsequent totermination of the thermal transient.
21
Scram PCP trip
15
CL
a-
12.5
10
7.5
CL
50 50 100 150 200
Time (s)
Figure 4. Primary system pressure (short term).I
00~
0L:3InIna)La-
15
12.5
10
7.5
5
2.5
0
0
(a
0.
a)L:3(a(aa)I.-0~
0 500 1000 1500 2000Time (s)
Figure 5. Primary system pressure (full term).
22
3. PCS THERMAL/HYDRAULIC RESULTS
This section summarizes the thermal/hydraulic boundary conditions of
the PCS prior to and during fission product release and transport.
Included are the hydraulic response during the blowdown, the fluid and
metal temperatures during fission product release and transport, and the
fuel rod cladding response during the blowdown and heatup phases. Also
included is a brief comparison of the thermal/hydraulic response with
preexperiment calculations made using the RELAP5/MOD2 1 5 and TRAC-BD1 16
computer codes. An additional comparison is made in Appendix J, which
compares the experimental results with calculations made using a unified
code made up of SC)AP, RELAP5/MOD2, and TRAP-MELT. The detailed
discussions of the PCS hydraulics, core thermal response, and comparison of
data with RELAP5/MOD2 and TRAC-BDI calculations are included in
Appendices D, E, and F.
3.1 Blowdown Hydraulics
This section discusses the reactor vessel liquid level, PCS mass
inventory, center fuel module mass flow rate, and PCS reflood.
The experiment hydraulics resulted in a gradual PCS level decrease
and, ultimately, in a slow core boil-off. The loops began to void at
approximately 50 s (intact loop hot leg) as shown in Figure 6, which
compares the individual average chordal densities measured by the gamma
densitometer in this leg. The level decreased until the loops were
completely voided by 470 s (based on dryout of thermocouples in the upper
plenum). The upper plenum was voided by approximately 600 s and the level
continued to drop, entering the top of the core by 700 s. The entire core
was voided by approximately 1355 s as indicated by the level probe in the
#3 fuel module. The data from this probe is shown in Figure 7. As
discussed below, the completion of voiding as indicated by the level probe
occurred more than 300 s after all the cladding thermocouples in the core
indicated heatup.
23
I
__ JUL-r-U-UUZti-- DE-PC-002C
50E 0.75IN E
" I" 40 .,'II
" 0.50 30t30
C II
-20
0 0.25 ' ,,
010 A,-
0 50 100 150 200 250 300
T i me (s)
Figure 6. Intact loop hot leg density.
24
*** BLOBLE * VERSICN 301 * 1OD 0C2 0 2/23/31 DATE OF RUN e5/Oq/Ilo PACE 30C3
RUBBLE PLOT OF FILE 'LE3F10 .
* CHARACTER RANGE TABLE *
{ c 1.000 • N < 999.30C ** (0) .010 •N f 1.000 ** MX) -999.00C 4 1 • .C10 *
LEVEL(M ABOVE
CORE BOTTOM)
1.78 *XXXXOXCOOOOOOCCOOCOCOOCO)XXXXXXXXXXXXXOCCOOOOXCCOCGOCO OXXXYXXXYXX1.08 *XXXXOCCOOOOOOOCOCOOCCCCCCOO000000000000C0000000 CCCOCOXXXXXXXXXXXXXXXYXYXXXXXYXXX XXX XXXXxXxxXXXXX.98 *xxxcoooooOO3COOGooccccocoCooocOo320009cooc000 ccooGooxxxxxxxXoOOcOOcOOOcccccocc3c.ý4xxxxxyxxXXXxxxU, .89 *XXXXOOCOOOOOOCOCOOCoOCCoo0O3OOO0OOOOOOXX0o0000 0OOOOOCCOOOOOOCOCCCCCcOCOOCXXXXXXXx•x.71 *XXOOOCC00OOOOOOOCO3G300COOOOOOOOOOOOOOOOOGCOOOOO OOocCXXXXXXXXXXYXXX•XXX YXXXYXXX)XXXXXXXIXXXI.61 *xXXOOCGO000000CCCOOCCCOCCOOO003OO0)OO0OOOOO0 OCCc3ooGCAXXXXCCCCCCC3c CCCCCCCOCCC CXXXxx)y Xxx.51 *XXXOOOCOOOOOOOCOOCOOCOCC OOOOOOOOOOOO3O030xXXOXOCO0 CCOOOcXX XXXXXXXXxxxXXXXX xwxCCIOO0CXxxxx xXX.41 *XXXXCXCjoo0OO3OCC33OC3COCOOOOOOOOOOOCOJOOOOXXxXXXO0 COXXCOC CCCCOCOCrCCC3 Oxx)wxxxxxxxxx.20 *XXXXXXXXCOOaOO0CCO33COCCCOCO0OOOOOXOXOXXXXXXXXXXXXXXXXXXXXXXXOO OxxxXXxXXXxxx.1c *xxxxxxxxoooooocoocoocccccccoooocoo.oooocoocooccocooccxccccccoxxxxxxxxxxxxxoccýCoo OOCCCCC5c^Z03C0xXX).•XXXA
**----------- *------- -- --- ---------------- *------ *---------*--------.053 31ý8.447 S25.'.47 492.447 54,.447 826.447 993-.4'? 1160.447 1327.447 14r4.147 t'.L7I? 1e~' j~P . 47 'ýC.4
TIM~E (5)
Figure 7. Conductivity level probe response above Fuel Assembly 3.
The PCS mass inventory declined to a minimum of approximately 500 kg
(based on the blowdown suppression tank level increase) by 1300 s. At that
time, the center fuel module mass flow had decreased to approximately
0.04 kg/s (this mass flow rate was calculated from the measured cladding
temperature response. The details of the calculation are in Appendix D)
and the flow out the LPIS line, to approximately 0.3 kg/s. This mass flow,
though very small, was sufficient to sustain a rapid metal-water reaction
in much of the central fuel module as the temperatures increased above
1700 K (2600°F). The center fuel module mass flow results in an average of
0.4 gm/s/fuel rod (0.04 kg/s per 100 fuel rods). Data from the Power Burst
Facility indicate that flows as little as 0.1 g/s/fuel rod are sufficient
to sustain the metal-water reaction without steam starvation. 1 7
When the shroud temperatures reached the experiment termination
setpoint of 1517 K (2272°F), the FPMS and LPIS lines were closed and
reflood of the plant was initiated using both ECC systems. Rapid injection
of approximately 1000 kg (2200 ibm) of water from the accumulators resulted
in a PCS repressurization from 1.2 to approximately 3 MPa (174 to
435 psia). This caused the accumulator flow to momentarily cease.
Additional cycles of-accumulator flow and PCS repressurization were
required before all of the damaged core could be quenched; the ECCS was
fully capable of accomplishing this and the plant was in a safe shutdown
condition within a few hundred seconds of ECCS injection initiation. The
peripheral fuel rods quenched rapidly, in a manner similar to previous LOFT
core uncovery experiments. Most of the center fuel module also quenched
rapidly, though more slowly than in previous experiments. A small fraction
of the center fuel module, however, took much longer to quench, indicating
the disruption of the fuel rod geometry in part of this module. Additional
details on the thermal response of the core during reflood is located in
Section 3.2 and in Appendix E.
26
3.2 Core Thermal Response
This section summarizes the fuel rod cladding temperature response,
including the initiation of dryout at various core locations, the effect of
control rod melting on the thermal response, the occurrence and propagation
of a rapid metal-water reaction, and the quench of the core during reflood.
The temperature excursion began in the upper part of the peripheral
fuel modules at 662 s and moved downwards as the coolant boiled away. The
propagation of the core heatup was generally top-to-bottom in the
peripheral module, with the dryout reaching elevations of 1.14, 0.38, and
0.28 m (45, 15, and 11 in.) above the core bottom at 662, 730, and 930 s,
respectively. This is illustrated in Figure 8, which compares cladding and
saturation temperatures at these elevations in the #2 fuel module. The
quench at the 10-in, elevation associated with the opening of the PORV is
also seen. Figure 9 is a similar figure for the central fuel module, with
temperatures shown from the 1.07-, 0.69-, and 0.25-m (42-, 27-, and 10-in.)
elevations. The dryout started a little later in this module, with the
corresponding times being 689, 740, and 938 s, respectively.
At approximately 1050 K (1430*F), the guide tube temperatures
responded to a phenomena that is thought to be connected with melting of
the absorber material (Ag-In-Cd) at the 0.69-m (27-in.) elevation. The
temperatures on guide tubes 5J13 and 5K05 both show a definite decrease in
the heatup rate (from 1.2 K/s down to 0.7 K/s) which is interpreted as
resulting from the melting of the control rod material in these guide
tubes. The argument is that the latent heat of melting absorbed some of
the decay heat, causing a decrease in the heatup rate. This is consistent
with the observation that the heatup rate of guide tube 5H08, which does
not contain a control rod, was not similarly affected. Figure 10 compares
these three temperatures. The latent heat associated with the melting of
the control rods could account for a temperature shift of up to 280 K
(504 0 F). The difference between this value and the 50 K (90 0 F) measured
shift could be explained by the metal-water reaction, which was occurring
at that time.
27
600 I I I I I I II
I / ,
I I i
.4
-. 9.It •
"- 550
CL.
E 500I--
450600
I .1
* /1/
$ . *
/.7/
I ' I
* ~I,.., * /'I,,
-'-I
-600
-550
Soo
500
~450 0
TE-2E08-045- TE-2F07-015- TE-2G14-011X Saturation temrp
I 400
I.-
eraturetI I I ~
650 700 750 800 850Time (s)
900 950 1000
Figure 8. Comparison of cladding temperatures at the 1.14-, 0.38-, and0.28-m (45-, 15-, and 11-in.) elevations in Fuel Assembly 2 withsaturation temperature. (See Appendix I for thermocouplequalification limits).
600
%--1550
E o
4..L3
E 500II--
CDCL
EQ,
I--
450 L600 650 700 750 800 850 900 950 1000
Time (s)
Figure 9. Comparison of cladding temperatures at the 1.07-, 0.69-, and0.25-m (42-, 27-, and 10-in.) elevations in Fuel Assembly 5 withsaturation temperature. (See Appendix I for thermocouplequalification limits).
28
1600
'- 1300
,L9 1200
800
I100.
E 1000
I-,900
800-
700
110.
Figure 10.
J ll"
-- 00
(2-i. elvtoni ul seby .(e Apni I for
It q
- -- 1000" II
I I I I I I I '
1150 1200 1250 1300 1350 1400 1450 1500 1550 1600Time (s)
Comparison of three guide tube temperatures at the 0.69-rn(27-in.) elevation in Fuel Assembly 5. (See Appendix I forthermocouple qualification limits).
I.
29
At about 1500 s, several control rod guide tube thermocouples at the
27-inch elevation show a small discontinuity that is thought to be
associated with the failure of the rod (see, for example, Figure 10). This
occurred at approximately 1200 K (1700'F). Once again, the effect is
absent from thermocouple TE-5H08-027, which is in an empty guide tube.
The first recorded and qualified rapid temperature rise associated
with the rapid reaction between zircaloy and water occurred at about 1430 s
and 1400 K on a guide tube at the 0.69-m (27-in.) elevation. This
temperature is shown in Figure 11. A cladding thermocouple at the same
elevation (see Figure 11) reacted earlier, but was judged to have failed
after 1310 s, prior to the rapid temperature increase. Note that, due to
the limited number of measured cladding temperature locations, the precise
location of the initiation of metal water reaction on any given fuel rod or
guide tube is not likely to coincide with the location of a thermocouple.
Thus, the temperature rises are probably associated with precursory heating
as the metal-water reaction propagates away from the initiation point.
Care must be taken in determining the temperature at which the metal water
reaction initiates, since the precursory heating can occur at a much lower
temperature. It can be concluded from examination of the recorded
temperatures that the oxidation of zircaloy by steam becomes rapid at
temperatures in excess of 1400 K (2060°F).
The temperatures in the center fuel module reached the target
temperature of 2100 K (3320°F) due to the rapid reaction between the
zircaloy and the steam, and remained above this temperature for
four-and-a-half minutes. The maximum temperature reached is difficult to
determine because of the failure of the thermocouples at the high
temperatures experienced, but it was certainly in excess of 2400 K (3860°F).
During the transient, the temperatures on the outside of the shroud
increased steadily from 740 to about 1700 s. This is illustrated in
Figure 12, which compares the temperatures on the south side of the
shroud. At approximately 1700 s, the heatup rate increases. At about the
same time, the thermocouples near the outside of the shroud also start to
heat up more rapidly. Figure 13 illustrates this by comparing the
30
2500 ý 4000
.- 2000
U,)lid
4- 1500IL
E
1000
500600
Figure 11.
1700
1500
2 1300
L- 1100.
0 900C-E
I. 700
--- s -- - I I.. - -_ 1700 800 90o 1000 1100
Time
Comparison of two claddingelevation in Fuel Assemblyqualification limits).
-.3000
-2000 L-
100
I a.', E
I I 0
- - -L1000
1200 1300 1400 1500 1600(s)
temperatures at the 0.69-m (27-in.)5. (See Appendix I for thermocouple
2500~r- i -. -/1
/1// 2000/ /
- U). *" .1500
0- -°-1000
E
.50050101
300600
I I I I I I I I L I / J700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900
Time (s)
Figure 12. Comparison of four external wall temperatures at'the 1.07-,0.81-, 0.69-, and 0.25-m (42-, 32-, 27-, and 10-in.) elevationson the south side of the flow shroud. (See Appendix I forthermocouple qualification limits).
31
1500 I I I I I I T I I
-2000
1300o . ,
-- " 1100 -"1500 %.
L L.
9001) 1000
E 700 E•D •ph- I--
TE-2G14-011
500 " TE-2H14-028 500TE-2114-039TE-2H13-049
.1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 1900Time (s)
Figure 13. Comparison of cladding temperatures at the 1.24-, 0.99-, 0.71-,and 0.28-m (49-, 39-, 28-, and 11-in.) elevations in FuelAssembly 2. (See Appendix I for thermocouple qualificationlimits).
32
temperatures at various elevations in the #2 fuel module, just adjacent to
the shroud south wall. By the time the reflood turns the temperatures
around (1785 s), all of these temperatures exceed 1400 K (2060°F). In most
cases, the fuel rod cladding temperatures exceed the shroud temperatures at
the same elevation. The cause of this rapid heatup is not presently known,
but it may be an effect caused by the thermocouple leads passing through a
hot area as they exit from the top of the core (shunting) rather than a
true local effect.
The cooling of the core took much longer than any previously measured
quench in LOFT. This was in part due to the much higher temperatures that
existed prior to quench (>2400 K [38600 F] for this experiment comparedwith the previous maximum of 1261 K [181 0 F] measured during Experiment
LP-LB-l 8). More important, however, is the geometry of the core during
reflood. Relocation of the core undoubtedly resulted in masses of core
material much thicker than normal. These masses would require much more
time to cool than would the regular fuel rod geometry. This is postulated
to be the cause of the slow cooldown manifested by thermocouple TE-5JO7.-027
(failed), shown in Figure 14. (Even though this thermocouple failed, it is
believed that the failure mode is a junction relocation and that the
thermocouple is indicating a temperature at some location in the center
fuel module.) That thermocouple was slowly cooling towards saturation
until 2010 s, when the junction apparently broke. Thus, even though the
core had been essentially quenched for more than 200 s, the temperature was
only slowly decreasing, indicating the insulating effect of a large mass ofmaterial surrounding the thermocouple. Additional evidence that the center
fuel module experienced significant control rod fuel relocation is
discussed in Appendix E.
3.3 Comparison with Calculations
This section summarizes the results of the comparison of the data from
Experiment LP-FP-2 with preexperiment calculations19 made using the
RELAP5/MOD215 and TRAC-BO116 computer codes. Appendix F contains a
more detailed presentation of these comparisons, along with a description
of the input models.
33
2500 4000
2000
0L
4-
0L00~E0I-
1500
1000
3000
2000 .
0-
05001
0600 0oo 1200 1500 1800 2100
Time (s)
Figure 14. Comparison of two cladding temperatures at the 0.69-m (27-in.)elevation in Fuel Assembly 5 with saturation temperature. (SeeAppendix I for thermocouple qualification limits).
34
The predicted pressure response agreed well with the data except for
the time after initiation of the LPIS line break at approximately 220 s.
This is illustrated in Figure 15, which compares the measured and predicted
PCS pressure response for the first 400 s of the transient. The relatively
minor differences between measured and predicted pressure response can be
explained on the basis of differences between the initial conditions and
the experiment sequence that were assumed in the calculations and the
actual conditions of the experiment. However, the opening of the LPIS line
break did not have as great an impact on the depressurization rate as was
predicted. This is emphasized in Figure 16, which compares measured and
predicted pressures for the first 2000 s of the transient. The PCS
depressurized much more slowly than predicted from 220 until 735 s, when
the intact loop cold leg break was closed. This indicates that the flow
resistance in the LPIS line was much greater than was modeled in theprediction. The LPIS line was modeled as a straight line pipe segment with
a flow resistance based on the length and pipe bends. It is evident that
this modeling was not adequate. This inadequacy is again emphasized during
the time subsequent to 735 s. Closure of the intact loop cold leg break
was predicted to have little impact on the depressurization rate, whereas
the measured depressurization almost stopped (see Figure 16). In fact, the
intact loop cold leg break and plant PORV were cycled subsequent to this
time in the experiment in order to reduce the PCS pressure to the point
that the FPMS lines could be opened. While these actions were notpredicted to be necessary, the depressurization rate was foreseen as an
area of uncertainty and these actions were designated as a contingency
action should the rate of depressurization be too small. The higher than
predicted pressure may have had an indirect impact on the fluid
temperatures. Although the depressurization rate prior to the onset of
rapid core heatup (metal-water reaction discussed below) was lower than was
predicted, the reverse was true during the period of rapid heatup. This
higher depressurization rate resulted in a larger-than-predicted steam
flow, which prevented the expected steam limitation in the upper parts of
the core.
35
CL
(f)(n
EL
15 I I i
10 -
5
00 50 100 150 200 250
Time (s)
0(0
0~
0I.-
InIn0L.
a-
300 350 400
Figure 15. Comparison of primary system pressure withcalculations made using RELAP5/MOD2 (short
preexperimentterm).
15
PE-PC-002RELAP5/MOD2(25 MW, 100 pct break flow area)RELAP5/MOD2(33 MW, 70 pct break flow area)
0-
10
5
-2000
-1500 0(/}0L
-1000 L(50(0
fL.
-500
v -
0 500 1000Time (s)
1500 2000
Figure 16. Comparison of primary system pressure with preexperimentcalculations made using RELAP5/MOD2 (full term).
36
Figure 17 presents the measured cladding temperatures at the 0.25-m
(10-in.) elevation in the center fuel assembly with the prediction for the
nearest modeled location. The observed initial temperature rise rate was
1.3 K/s (2.3 0 F/s), which was predicted exactly by TRAC-BD1, whereas
RELAP5/MOD2 predicted 2.4 K/s (4.3 0 F/s). The average temperature rise rate
until 1700 s was observed to be about 0.5 K/s (0.9 0 F/s). The rise rate was
overpredicted for most of this period in both calculations, TRAC-BDl
providing the better agreement with 0.7 K/s (1.3 0 F/s) compared with
RELAPS/MOD2 with 1.0 K/s (1.8 0 F/s). The overprediction was contrary to the
fact that the modeled decay heat level in this node was lower than that
which existed at the measurement location. The underprediction of mass
flow of steam through the core is believed to have resulted in an
underprediction of the heat transfer coefficient. The observed increase in
temperature rise rate at 1700 s occurred at too low a temperature [about
900 K (1161°F)] to be the result of metal-water reaction locally and was
not predicted. The observed behavior may be the result of thermal
radiation from high temperature material at a higher elevation or to
material relocation. Neither thermal radiation in the axial direction nor
the desired effect of material location on local temperature is modeled.
The initial heat up rate at the 0.69-m (27-in.) elevation prior to the
time when the PORV and intact loop cold leg break were opened was well
calculated by both codes. This is illustrated in Figure 18, which compares
the temperatures at this elevation in the center fuel module. After that
time, however, the measured temperature rise rate decreased and the
predicted rise rate was higher. The temperature rise rate due to metal
water reaction was well predicted by TRAC-BD1, even though it initiated
from a lower-than-predicted temperature. This lower initiation temperature
may be due to precursory heating discussed above and in Appendix E.
The initial heat up rate at the 1.07-m (42-in.) elevation was
accurately predicted by both codes. After approximately 1450 s, however,
the cladding experienced the rapid heat up due to metal water reaction, in
contrast to the prediction, as illustrated in Figure 19, which compares the
temperatures at this elevation in the center fuel module. The metal water
reaction was not predicted to occur at this elevation due to steam
37
4)
4-
1400
1200
1000
800
600
40060(
Figure 17.
2500
2000
--1500
L
1000 C
E
800 1000 1200 1400 1600 1800 2000Time (s)
Comparison of the measured cladding temperature at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with preexperimentcalculations made using RELAP5/MOD2 and TRAC-BD1.(Thermocouple qualified throughout).
I I I I 1 " 4000
TE-5J07-027 -
RELAP5/MOD2. TRAC-BD1 .. . .0
- 3000
- .- - 2000 .-
- L..
1000 E
a,I..
4-0I.Sa-ES
I-
1500
1000
500
0 L600 800 1000 1200 1400 1600 1800 2000
Ti me (s)
Figure 18. Comparison of the measured cladding(27-in.) elevation in Fuel Assemblycalculations made using RELAP5/MOD2(Thermocouple qualified to 172.0 s).
temperature at the 0.69-m5 with preexperimentand TRAC-BDI.
38
3000
2500-
-S
0LJ
-4-01.~00~E0
I-
2000
1500
1000
500
1800 2000
0.
21000
-0
1800 2000
temperature at the 1.07-m5 with preexperimentand TRAC-BD1.
00600 800 1000 1200 1400 1600
Time (s)
Figure 19. Comparison of the measured cladding(42-in.) elevation in Fuel Assemblycalculations made using RELAP5/MOD2(Thermocouple qualified to 1510.3).
39
limitation. As discussed above, the estimated center fuel module steam
flow was much higher than was predicted; thus, the metal water reaction was
sustained. This had the effect of shifting the location of maximum
cladding temperatures from the predicted 0.56-to 0.84-m (22-to 23-in.)
elevation to the estimated 0.81-to 1.07-m (32-to 42-in.) elevation. Lack
of instrumentation prevents a more precise determination of the actual
level.
In general, the calculations closely predicted both the general and
specific experimental results. Despite the differences noted above and in
Appendix F, these calculations formed a very adequate basis for experiment
planning, including the need for contingencies in the experiment operating
procedure.
3.4 Metal/Fluid Conditions Near the FPMS
The fluid conditions near the FPMS sample locations were generally
superheated. This is illustrated by Figure 20, which compares the fluid
and metal temperatures at the lower Dl deposition coupon. The other
deposition coupon environments were similar. The fluid conditions near the
locations from which fission products and aerosols were drawn into the
Fl and F2 sample lines were similarly superheated except for two temporary
quenches in the Fl line. This is illustrated in Figure 21, which compares
the fluid temperature with saturation temperature in this line. There was
no measured perturbation on the pressure at this location, indicating that
very little steam generation occurred. It is possible that a small water
droplet quenched only the thermocouple.
3.5 Summary
The core boiled dry and heated up to temperatures in excess of 2400 K
(3860°F) due initially to decay heat and ultimately to a rapid metal-water
reaction. The center fuel module control rods melted, as did a substantial
fraction of the adjacent fuel rods. Much of the center fuel module between
the 0.69-m and 1.07-m (27-and 42-in.) elevations relocated to the bottom of
the fuel module. The thermal shroud was able to adequately shield the
40
1000
"- 750
L
E soo0p
I-,
--
L-
0
E
2500 500 1000 1500 2000
T i me (s)
Figure 20. Comparison of metal and fluid temperatures at the lower D1deposition coupon location with saturation temperature. (SeeAppendix I for thermocouple qualification limits).
600
550
500
a 450
E 400I.-
450
350
0
E0)
0 500 1000 1500 2000T i me (s)
Figure 21. Comparison of fluid temperature in the F1 aerosol sample linewith saturation temperature. (See Appendix I for thermocouplequalification limits).
41
peripheral fuel rods during this time, and no fuel rod failure occurred in
the peripheral modules. The prediction calculations corresponded well with
the data. The principal discrepancies between data and prediction were
dominated by inadequate modeling of the simulated LPIS line. In general,
the thermal/hydraulic boundary conditions for fission product release and
transport adequately simulated a V-sequence accident.
42
4.0 FISSION PRODUCT RESULTS
This section describes the source term calculations, the behavior of
the Fission Product Measurement System (FPMS), and the preliminary results
of the measurements concerning fission product transport.
4.1 ORIGEN2 Results for the LP-FP-2 Experiment
The pretransient isotopic mass and activity inventories in the center
bunale were calculated using the ORIGEN220 computer code and using the
measured irradiation history. The ORIGEN2 computer code is designed to
calculate the nuclide composition in a nuclear reactor as a function of
time. This code accounts for several forms of nuclide decay, neutron
activation events, or other changes induced by time-dependent fuel cycle
operations.
The center fuel bundle for Experiment LP-FP-2 consists of an
11 x 11 fuel rod geometry surrounded by a 25.4-mm (1.0-in.) thick thermal
shroud. The 11 x 11 rod geometry contains 100 fuel rods [1.67 m (5.5 ft)
in length] and 21 zircaloy guide tubes, of which 11 contain
stainless-steel-clad control rods. The fuel rods contain 1136.7 grams of
UO2 (1001.4 grams of uranium per rod or 0.10014 MTU for the center
bundle) enriched to 9.744% U-235 (97.57 grams of U-235 per rod). The
control rods contain 1270 grams of a 80% Ag-15% In-5% Cd alloy.
For calculational purposes, the preconditioning phase of the LP-FP-2
experiment was modeled as follows: (a) the reactor was assumed to be
operating for 3.5 days at a constant power of 32 MW (52.5 kW/m or
16 kW/ft), and then was shut down for 3.12 days; (b) 3.333 days of
additional operation at 26.5 MW, followed by a 4.0 day interval of
down-time; and (c) a final irradiation period consisting of running the
reactor at 32 MW for 24 hours followed by an irradiation at 26.5 MW for
16 hours immediately preceding the initiation of the experiment. This
irradiation history approximates the actual irradiation history shown in
Figure 1.
43
Since the power generated in the LP-FP-2 center bundle accounts for
approximately 17.2% of the total core power,21 it follows that a core
power of 32 MW corresponds to a bundle power of 5.5 MW (Note: the input
values listed in this section are approximate. For exact values, see
Appendix J), and that a core power of 26.5 MW corresponds to a bundle power
of 4.6 MW. Consequently, for the above irradiation history, the center
bundle burnup is calculated to be 429.4 MWD/MTU. Another calculation of
the center bundle burnup for the LP-FP-2 experiment, based on the detailed
measured irradiation history, indicates a burnup of 430 MWD/MTU, which is
in good agreement with the previous burnup result.
The input powers (e.g. 5.7 and 4.7 MW) used in the ORIGEN2 analysis
(see Appendix J) were obtained by multiplying the calculated center bundle
powers (e.g. 5.5 and 4.6 MW) by the factor 1.041. Since the ORIGEN2 code
assumes a total fission energy of 202 MeV/fission to compute the fission
rate (based on U-235), the factor 1.041 (202/194) was used to adjust the
ORIGEN2 power so that the fission rate would be based on 194 MeV/fission,
which represents the recoverable or thermal fission energy (total released
energy minus neutrino energy), instead of 202 MeV/fission.
Selected results of the ORIGEN2 analysis for several important
nuclides are shown in Table 4. Detailed results of the ORIGEN2
calculation, showing fuel bundle activities, masses, thermal power, and
estimated coolant inventories, are identified in Appendix I. The input to
the ORIGEN2 code is also presented with the ORIGEN2 output listing.
Based on the ORIGEN2 data shown in Table 4, the cesium-to-iodine mass
ratio for the center bundle, just prior to the LP-FP-2 experiment, is
calculated to be 4.00 (the atom ratio is 3.88). The ORIGEN2 code was also
used to calculate the decay heat of the center bundle at 200 s into the
transient. The result of the calculation indicates that the center bundle
decay heat is about 115.3 kW at 200 s. Based on a peripheral bundle
44
TABLE 4. SELECTED ORIGEN2 INVENTORY RESULTS FOR THE EXPERIMENT LP-FP-2
CENTER FUEL BUNDLE AT 430 MWD/MTU BURNUP
Initial Fuel Inventorya
Material
Kr-85Total Kr
Rb-88Total Rb
Xe-131MXe-133Xe-l 33MTotal Xe
1-1311-1321-1331-134Total I
Cs-136Cs-137Cs-138Total Cs
Te-129MTe-132Total Te
Ru-103Total Ru
Ba-140Total Ba
La-140Total La
(grams)
4.408 x 10-26.654 x 10-1
1.216 x 10-36.051 x 10-1
(curies)
1.730 x 1014.039 x 105
1.461 x 1054.663 x 105
SpecificActivity
(curies/gram)b
2.600 x 101
2.414 x 105
2.8305.8977.4966.493
3.8358.6571.8951.1288.588
1.3671.6616.2373.440
2.1792.9928.564
xxxx
xxxxx
xxxx
xxx
10- 3
10-110- 3
100
10-110-310-110-210-1
10- 3
10010-3100
10-210-110-1
2.3721.1043.3626.231
4.7578.9392.1473.0119.472
1.0021.4462.6405.091
6.5699.0897.363
xxxx
xxxxx
xxxx
xxx
102105103105
104104105105105
102102105105
102104105
3.6531.7005.178
5.5391.0412.5003.506
2.9134.2037.674
xxx
x
x
101104102
104105105105
101101104
xxx
5.610 x 10-12.955 x 100
1.100 x 1002.977 x 100
1.243 x 10-11.947 x 100
.811 x 1047.095 x 104
8.025 x l047.227 X 105
6.922 x 1047.355 x 105
7.670 x 1021.061 x 105
6.128 x 103
2.696 x 104
3.555 x 104
a. Center bundle inventory at 200 s into the FP-2 experiment.
b. Curies of the nuclide per gram of the element (all nuclides).
45
calculation, the total core decay heat is approximately 5.885 a times the
decay heat of the center bundle. Hence, the total core decay heat at 200 s
is estimated to be about 678.5 kW at 200 s into the transient. Detailed
reactor physics calculations based on the actual irradiation history
indicate a core decay heat of 684.8 kW at 200 s. Both results fall within
the 675 to 695 kW range set as a pretest objective for the experiment.
4.2 Results of the Elemental Release Calculations
The elemental source terms to the primary coolant system during the
LP-FP-2 transient were calculated with the CORSOR 2 2 and TIGERl 2 3
computer programs. TIGERI is an INEL-developed program that uses the
release rate data base published in NUREG-077224 (shown in Figure 22) to
predict fission product release as a function of time, similar to the
CORSOR program. Besides the NUREG-0772 data base, the CORSOR and TIGERI
programs require time-dependent axial fuel temperatures, the axial
distribution of the fission products, and the initial fission product
inventory as computed from ORIGEN2. The time-dependent axial fuel
temperatures, shown in Figures 23, 24, and 25, indicate the measured
thermocouple temperatures during the experiment and the averaged
temperatures (the "LEVEL" data) used in the CORSOR/TIGERI calculations.
The axial fission product distribution in the center bundle was
assumed to follow the axial power distribution. A schematic of the
CORSOR/TIGERI model for the LOFT center bundle, depicting the axial fission
product distribution, axial nodalization, and thermocouple locations, is
shown in Figure 26. The time-dependent output of the TIGERI program are
shown in Figure 27 and 28.
a. The factor 5.885 that was used in the decay heat calculation is notequal to the ratio of the core to center bundle powers (e.g. 5.8139 =1/0.172). This occurs because the peripheral bundles have a differentU-235 enrichment (4% versus 9.7%), a different amount of burnup (because ofprevious experiments conducted with the peripheral bundles), andconsequently a different Pu-239 concentration.
46 1,1 - ,
Temperature (*F)
3000200010
ior
4000
0
0
L.
0cc
(M10"
larg
ioroI. i.1000 1500 2000 2500
Temperature (K)3000
L87-KMI12-00
Figure 22.
3000
Release rate functions taken from NUREG-0772.
4)
:3.6-
4)aE4)
2500
2000
1500
1000
500
0
LL..
CL
ECD
0 500 1000 1500 2000Time (s)
Figure 23. Comparison of measured cladding temperatures at the 0.25-m(10-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits).
47
3000
2500
"- 2000
L.
1500
E ioooa..
E
500
00 500 1000 1500 2000
Time (s)
Figure 24.
3000
2500
a,L.
4-01...a,0~Ea,
I-
2000
1500
1000
Comparison of measured cladding temperatures at the 1.07-m(42-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGERI calculation for this elevation. (SeeAppendix I for thermocouple qualification limits).
I I I
-- TC-5108-27--- TC-SM08-27
--- TC-5M04-27 -- 4000
X LEVEL 27
,3000
-20000
E-1000 I
-0
500 1000 1500 2000Time (s)
Comparison of measured cladding temperatures at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with input data used inthe CORSOR/TIGER1 calculation for this elevation. (SeeAppendix I for thermocouple qualification limits).
500
0
Figure 25.
1, "- , e
48
Axialnode Thermocouple
Elevation FP density number location
167.0 cm
42 in. at0.305 112, C07
88.0 cm
27 in. at0.359 2 108, MOB, M04
47.0 cm
0.336 110 in. atM07, L07
0.00 cmL11O]1KM13:5-11
Figure 26. Input model of the center fuel assembly for the CORSOR/TIGERIcalculation.
Kr, Xe, I, Cs
00".• 10 "
W
,,Sb*~10"z/
-I / Ba,> Te
F Sr
E/
o Zri ii1 U 4 I
u
0 500 1000 1500 2000imne (s) Ll10 KM135-12
Figure 27. Elemental cumulative fractional release inventories calculatedusing TIGERI.
49
10
E
0
4-C.)
W0U
10"
1 0-2
10*
1•0 500 1000
Time (s)1500 2000
L110-KM135-13
I, Sb, Xe, Kr, and Te calculatedFigure 28. Release rate functions for Cs,using TIGERI.
The CORSOR calculation25 was performed with the same input used in
the TIGER] analysis. The cumulative release results (at 1800 s) of TIGERl
and CURSOR are listed in Table 5. Since the TIGERI results only include
the transient fission product release and not the initial fuel rod gap
inventory, the CORSOR results (shown in Table 5) were modified (by
subtracting the steady state WASH-1400 2 assumed gap inventories) to
produce a net transient release fraction.
Using the TIGERI release rate results, an ORIGEN2 calculation was
performed to estimate the fission product activity released from the center
bundle to the coolant. Results of this calculation are presented in
Appendix J.
4.3 FPMS Performance
The Fission Product Measurement System (FPMS) consisted of three types
of measurement devices: the steam and aerosol sampling systems, the
deposition devices, and the gamma spectrometer systems. The aerosol
sampling system consisted of the Fl sampling line, which extracted a steam
sample from the center fuel module; the F2 sampling line, which extracted
an aerosol steam sample from the broken loop hot leg; and the F3 filter,
which extracted a sample from the LPIS line. The deposition rod (Dl) and
the deposition spool pieces (02 and D3) presented a representative
stainless steel surface to coolant flow in the center fuel assembly upper
structure and the LPIS line, respectively, for fission product plateout.
The gamma spectrometer systems can be divided into two groups, those that
provided isotopic measurements during the transient and those that provided
measurements during the posttransient period. The gamma spectrometer on
the Fl sample line (G6) and on the LPIS line (G5) were intended to provide
data during the transient, while the Gl, G2, and G3 gamma spectrometers
were intended to provide isotopic concentrations in the primary coolant
system (PCS), in the BST liquid space, and in the BST vapor space,
respectively, during the posttransient period. In addition, the G2 gamma
spectrometer, which measured the posttransient isotopic concentrations in
the BST vapor space, was used to measure the concentrations from the
51
TABLE 5. COMPARISON OF TIGERI AND CORSOR CUMULATIVE RELEASEFRACTIONS FOR EXPERIMENT LP-FP-2 (at 1800 s)
Element
Xe
TIGERI Results
Net Release
0.206
0.206
0.206
0.206
Kr
Cs
CORSOR Results
Code Result
0.275
0.275
0.290
0.265
Gapa
0.030
0.030
0.050
0.017
0.0001
Net Release
0.245
0.245
0.240
0.248
0.0079
I
Te
Sb
Ba
Ru
Sr
La
0.057b
0.020
0.006
0.00014
0.0021
0.0000001
0.020
0.007
0.0002
0.002
<0.0001
a. WASH-1400 fuel rodVII-13, WASH-1400.)
b. The CORSOR resultsthat of NUREG-0772 butoxidation holdup model.tellurium-zircaloy modE
gap inventory fractions.
for Te are based on a resomewhat similar to the
The TIGERI result corr!I is 0.006.
(See Table VII 1-2 page
lease model different than)RNL tellurium-zircaloyesponding to the
52
combined Fl and F2 sample line flows during the transient. The design and
planned operation of these systems is described in more detail in
Appendix B.
The sequence of events controlling the FPMS is shown in Table 6. All
operator actions were performed as specified.
In addition to the data obtained from the FPMS, a number of
measurements were taken by the plant operators and by Health Physics
technicians to characterize plant conditions during the posttransient
period of the experiment. Of particular interest are remote area monitors
(RAMs) placed at strategic locations inside the containment building and
the stack monitors, which extract an air sample from the exhaust of the
containment building and pass that sample through filters equipped with
radiation monitors. While these measurements do not characterize the
release from the primary coolant system, they are useful in characterizing
BST and PCS leakage during the posttransient phase of the experiment.
4.4 Instrument Operation
The FPMS instruments were operated as specified and, for the most
part, collected valid data. However, a few problems were encountered.
These problems are categorized into three distinct time frames: failure of
instruments during the preconditioning phase, difficulties during the
transient and reflood phase, and difficulties during the posttransient
phase.
During the preconditioning phase of Experiment LP-FP-2, gamma
spectrometer G6, which monitored the Fl sample line, was determined to have
irreparably failed. In its place, a remote area monitor was placed on top
of the reactor vessel to view the Fl line. While this instrument cannot be
used to quantitatively measure the activity in the Fl sample line, it can
be used as a gross gamma instrument to determine the time-dependence of the
Fl activity. The Programmatic Risk Assessment Document (PRAD) 2 6
identifies the impact of the failure of the G6 spectrometer as diminishing
53
TABLE 6. EXPERIMENT LP-FP-2 FISSION PRODUCT MEASUREMENT SYSTEMSEQUENCE OF EVENTS
EVENT
Fl DILUTION GAS LINE OPENEDF1 AND F2 VENT LINE CLOSEDREACTOR SCRAMDl MOVED (TO DROP CONTROL RODS)Dl WITHDRAWN (TO CLEAR COUPONS)LPIS LINE OPENEDD1 INITIAL PURGE STARTD1 INITIAL PURGE STOPFl STEAM ANALYZER EXTERNAL PURGE STARTFl STEAM ANALYZER EXTERNAL PURGE STOPFl ANNULUS GAS LINE OPENEDF3 LINE OPENEDF3 BYPASS LINE CLOSEDFl AND F2 SAMPLE LINES OPENEDSTEAM DETECTED IN Fl LINEFl LINE FISSION PRODUCTS DETECTEDBLHL FISSION PRODUCTS DETECTEDLPIS LINE CLOSEDFl LINE CLOSEDF2 LINE CLOSEDDl "CLOSED"D0 NITROGEN BACKUP ONFl AND F2 VENT LINE OPENEDFl DILUTION GAS LINE CLOSEDD0 "OPENED"D1 NITROGEN BACKUP BYPASS OPENEDD1 NITROGEN BACKUP BYPASS CLOSEDDl NITROGEN BACKUP BYPASS OPENEDD1 NITROGEN BACKUP BYPASS CLOSEDFl ANNULUS GAS LINE CLOSED
TIME (s)
-199.4-146.8
0.020.648.9
221.6750.6763.1878.1883.0883.1950.8951.9
1013.11013.1
-,1198.0-.1201.0
1777.61778.01778.11780.61808.01823.01833.12085.62143.12148.02933.12968.23401.6
•,• ij•
54
the number and types of radionuclides measured. Noble gas data might still
be provided by G-5 and/or G2, but potential trapping of gaseous species in
the top of the vessel and in dead spots in the piping can increase the
uncertainties of the results.
Also during the preconditioning phase of the experiment, the
calibration pulsers on the G5 spectrometer failed. While this latter
failure has delayed the processing of the G5 data, it does not impact the
quality of the data.
Just prior to reflood, the deposition coupon device was to have been
closed (to protect a coupon at each elevation from reflood water) and
placed on a nitrogen purge (to maintain a positive pressure differential).
After Experiment LP-FP-2, those actions were performed, but a positive
pressure differential could not be maintained on the deposition rod. This
indicates that some of the protected coupons may have been exposed to
either reflood or PCS water, and that some of the distinction between
protected and unprotected coupons may have been lost. However, significantdeposition data has still been collected: the PRAD identifies data from D2
and D3 as an alternative means of estimating the plateout in the upper
structure.
During the transient portion of the experiment, the Fl sample line was
heated with argon gas. During the reflood phase of the experiment,
however, the sheath through which this gas was fed to the sample line was
sealed, possibly due to high temperatures causing warpage of the line.
Since the measurement of H2 in the BST was calibrated to account for the
presence of this argon, the plugging of this line has caused the BST H2
measurement to be out of calibration and may cause some additional
uncertainty in the H2 measurement.
The third area of difficulty with the FPMS involved high background
activity for the Gl, G2, and G3 gamma spectrometers during the
posttransient phase of the experiment. These instruments were intended to
measure the posttransient isotopic concentrations in the reactor vessel
lower plenum, in the BST vapor space, and in the BST liquid space,
55
respectively. Because of background problems during Experiment LP-FP-l, it
was decided that a sealed 'tent' would be placed around the Gl, G2, and G3
gamma spectrometers, with a slight purge to maintain a positive pressure
differential. Unfortunately, it appears that there were one or more leaks
either from the G2 sampling system or from the BST directly into the tent,
and that the tent was, in reality, causing a higher Gi, G2, and
G3 background activity than would have been present without it. Therefore,
the posttransient period was extended in order to allow decay and cleanup
of fission products inside the containment building and acquisition of
reliable data. The online FPMS measurements were recorded for 14 d. A
liquid sample was taken from the BST at 21 days after the experiment.
Vapor samples were taken from the BST up to 28 days after the experiment.
Liquid samples from the PCS were taken up to 44 d after the transient.
Because of this extension, the Gl and G3 systems were able to quantify
fission product concentrations in the PCS and BST liquid. There is,
however, currently some concern that posttransient data from the
G2 spectrometer for quantifying BST vapor concentrations will be highly
uncertain. This uncertainty will primarily impact noble gas release data;
I and Cs measurements will be little affected.
The final problem concerns the F2 sample line. Measurements planned
for the F2 line are intended to include both reversible and irreversible
plateout. Therefore, the line will be kept dry until postirradiation
examination in the hot cells. However, during the posttransient phase,
FPMS personnel became aware that the isolation valve on this sample line
was leaking PCS water into the sample line. To preserve the sample line
data, a slight N2 purge was applied; all expectations are that data from
this sample line will not be affected.
4.5 Preliminary Results
This section presents preliminary results (from both FPMS and non-FPMS
instrumentation) that will give insight to fission product behavior and
transport. Due to the short time available for data analysis and due to
the incompleteness of the data, no attempt is made to present a
comprehensive picture of fission product behavior during
56
Experiment LP-FP-2. Instead, the limited data available are presented both
as an indication of instrument performance and as an example of the type of
data forthcoming from the experiment.
It should be stressed that the information presented herein Is
preliminary In nature and may change as additional data are collected and
inconsistencies are resolved.
4.5.1 Fl and F2 Sample Lines
No Fl or F2 sample line components were examined in time for inclusion
in this document. However, successful operation of the sample line can be
surmised from data now available. Figure 29 shows the response of the RAM
that was placed on top of the reactor vessel to view the Fl sample line.
As seen in this figure, the RAM begins to respond at approximately
1200 seconds to what is believed to be the fission product release from the
gap. By approximately 1500 seconds, the release from the fuel can be seen.
Figures 30 and 31 show the pressure measured upstream of the critical
orifices in the Fl and F2 sample lines, respectively. As seen in these
figures, steam did flow through the Fl and F2 sample lines during the time
of fission product transport. An additional indication of successful
operation of the Fl and F2 sample lines is given by the G2 spectrometer.
During the transient phase of the experiment, this spectrometer measured
the combined effluent from both the Fl and F2 sample lines and identified
isotopes of xenon and krypton.
4.5.2 Deposition Measurements
All information available at the time of this document indicate that
significant quantities of fission products were collected on each of the
deposition pieces. However, as discussed In Section 4.4, some of the
57
id 3000
'id
L..-c id
10
10"10
- 2500
- 2000 2C)
- 1500 TCD0.
E- 10o
- 500
- 02500
L105-KM130-04
500 1000 1500 2000
Time (s)
Figure 29. Comparison of the radiation area monitor response on the F1aerosol sample line with fuel centerline temperature(TC-5108-027). (See Appendix I for thermocouple qualificationlimits).
1.50
1.25
"" 1
_ 0.75a,o
a- 0.50
0.25
0750
5-F1-8B - 200
- 150
0.
- 100($)
- 50
2000 2250
L105-KMI30-09A
1000 1250 1500 1750
Time (s)
Figure 30. Measured pressure upstream of the critical orifice in the F1aerosol sample line.
58
1.50 1 I I 1
PTP165-F2-43 - 200
1.25
1- 150
.= 0 .7 5 -6P=~- 100
CL 0.50 a.
50
0=0.75 -10
750 1000 1250 1500 1750 2000 2250
Timne (s) L106i-KM130-08A
Figure 31. Measured pre!ssure upstream of the critical orifice in the F2aerosol sample line.
59
distinction between the protected and unprotected coupons of the Dl device
may have been lost, in that all of the protected coupons may not have been
sealed from reflood water.
Significant data is expected from the unprotected Dl coupons with
respect to irreversible plateout of fission products at each of the three
elevations. Figure 32 shows the metal temperatures that were measured at
each of the three elevations of the D0 deposition coupons. The deposition
data can be correlated to the measured temperatures and interpolated over
the length of the upper structure to characterize volatile fission product
plateout in the upper plenum.
Radiation scans of the LPIS line at the time of the first containment
entry indicate that fission products were collected on deposition
devices 02 and D3. Figure 33 shows the temperatures of each of the LPIS
line deposition spool pieces. As seen in this figure, the D2 coupon was
exposed to two-phase coolant when the LPIS line was opened, but during the
time of fission product release, was exposed to dry steam. The D3 coupon
was protected from coolant flow while the LPIS line bypass was open, but
after bypass closure the coupon was exposed to steam and fission products.
4.5.3 G5 Gamma Spectrometer
The G5 gamma spectrometer operated as expected, with the exception of
the loss of the G5 calibration pulsers. Isotopes of iodine, rubidium,
xenon, tellurium, and cesium have been identified. Tables 7 and 8 list the
indicated concentrations at various times during the test. Other isotopes
may be identified as data processing continues. While the background
activity for this spectrometer is negligible, plateout has not been
accounted for, although plateout is probably a significant fraction of the
signal by the end of the transient.
Figures 34 through 42 show the concentrations measured by the
G5 spectrometer for each of the isotopes identified in Tables 7 and 8. The
60
TABLE 7. IODINE SPECIES IDENTIFIED BY G5 SPECTROMETER
All measurements are in microCurles/cm3
Time(s)
179239300361422483544605666727788849910
1024105411151189124913101371143214931554160816681729179018512372267229733274357538764478
1-131 1-132
0.60 + 7%
1-133 1-134 1-135
0.36 17%
0.25 + 16%
0.400.500.50
4.86.8
1526483924
104365
135719452540262028002900302529153100286029002650
+ 12%+ 11%+ 17%
+
+
+
+
+
+
+
+
+
+
+
+
+4.
+
+
+
+
+
~1*
+
10%10%11%11%10%10%12%5%3%6%7%8%7%9%3%3%3%4%3%4%10%
0.41)1.32.22.42.12.72.53.13
14.1522.15
21362
111313151271012513871)
29004400655066506001)664065006481)610059005701)5700
10%5%10%10%10%10%10%15%15%10%10%10%10%10%10%10%10%10%10%10%5%6%8%10%5%3%3%4%4%3%
0.601.01.6
+ 11%. 9r+ 5X
2.2 + 20%
2.2 + 10%
16 + 10%
35 + 10%66 + 10%79 + 10%
114 + 10%
88 + 13%355 + 12%
2.214222951
÷
÷
.÷
+
÷1
0.4 + 25%
1.6 + 15%
20%10%15%10%10%
45007600
113001120010700
12800128501330013500
4.
+
+
+4.
+
+
+
4.
1)13%12%5%6%
69 + 10%65 + 10%
10600 + 10%14850 + 10%
14200 + 10%
2.22.22.84.3
22
4182
106121106
50020906760
12200105001750017400
+
+
4.
+
+
+
+
+
4.
+
4.
+
+
4.
+
4.4.
18%20%16%12%12%
10%18%18%10%30%
10%106%10%6%6%8%
5%5%5%6%
16500 + 4%
16400 + 4%
61
TABLE 8. NON-IODINE SPECIES IDENTIFIED BY 65 SPECTROMETER
All Measurements are in microCuries/cm3
Time(s)
179239300361422483544605666727788849910
10241054111511891249131013711432149315b4160816681729179018512372267229733274357538764478
Rb-88
0.5 + 34%
1.6 + 21%1.6 + 12%1.6 _ 30%
Xe-135 Te-132
1 .400.860.90
0.670.300.20
++
7%6%7%
+ 7%7 19%+ 18%
Cs-138
0.18 + 15%
0.50 + 17%0.50 T 12%
0.40 + 14%0.49 + 11%
0.60 + 10%
0.60 + 12%
0.23 + 16%1.81.81.92.4
+
++
10%11%13%10%
0.150.300.59
+ 20%+ 13%+ 12%
5.4 + 21%4.2 ; 26%
2943474751
+
7+
7
10%8%8%6%9%
21 + 40% 7.4 + 23%16 + 20%81 T 15%
3700 + 8%3800 + 10%
2000 + 10%
6.1 + 34%
8 + 30%78 + 10%
473 + 5%2395 + 10%4350 + 10%5800 + 8%5760 + 5%5500 + 5%4600 + 5%4200 + 5%3700 F 3%3600 + 2%3200 + 2%2800 7 2%2370 7 5%
1900 + 2%3360 + 3%4900 7 3%2400
25903260
+ 7%T 7%: 7%
62
CL
Ea)
750
700
650
600
550
500
450
- Boo
- 700 0
W
-600 CD0.E
500
400
2000 24000 400 800 1200 1600
Time (s) LIO5-KM130-18M
Figure 32. Measured metal temperatures at. the D1 deposition couponlocations. (See Appendix I for thermocouple qualificationlimits).
4-
CLE9
800
700
600
500
400
300
200
800
L-600 2
400
E
200
0 400 800 1200 1600
Time (s)2000 2400
L105-KM130-17M
Figure 33. Measured metal temperatures at the D2 and D3 deposition couponlocations. (See Appendix I for thermocouple qualificationlimits).
63
3500
3000
2500
2000
1500
1000
0
E)
-5(
0 -I I
500 1000 1500 2000 2500 3000 3500 4000 4500
Time (s) L1O-KM135-02
34. Measured 1311 concentration in the simulated LPIS line (notcorrected for plateout).
Figure
E
8000
7000
6000
5000
4000
3000
2000
1000
0
-10000 500 1000 1500 2000 2500 3000
Time (s)3500 4000 4500
LI1O-KM135-03
Figure. 35. Measured. 132Icorrected for
concentration in the simulated LPIS line (notplateout).
64
a
14000
12000
10000
8000E
6000
4000
2000
0
-2000
Figure 36
F
Li0 500 1000 1500 2000 2500 3000 3500 4000 4500
Time (sW L11O-KM135-04
5. Measured 133I concentration in the simulated LPIS line (notcorrected for plateout).
18000
16000
14000
12000
. 10000E
8000
6000
4000
20000 e E) O00 6
-2000 I __ _
500 1000 1500 2000 2500Time (s) L110-KM135-05
Figure 37. Measured 13;4I concentration in the simulated LPIS line (notcorrected for plateout).
65
0ES
20000
18000
16000
14000
12000
10000
8000
6000
4000
2000
0
-2000500 1000 1500 2000 2500 3000 3500 4000 4500
Time (s) L110-KM135-08
8. Measured 135I concentration in the simulated LPIS line (notcorrected for plateout).
Figure 34
4500 -
4000 -
3500 -
3000 -
2500 -E
2000 -
4- 1500 -
V)
1000
500
0
-5000 500 1000 1500 2000 2500 3000 3500 4000 4500
Time (S) L110-KM135-07
Figure 39. Measured 88Rbcorrected for
concentration in the simulated LPIS line (notplateout).
66
10 T --- r - ---- -- T --.- I
E0L
8
6
4
2
0 --L
0 250 500 750 1000 1250
Time (s'1500
L11O-KM135-OSA
Figure 40. Measured 13 5 Xecorrected for p
concentration inlateout).
the simulated LPIS line (not
60
50
40E
30
201200 1250 1300 1350 1400 1450
Time (s)
1500
L110-KM135-09
Figure 41. Measured 132 Te concentration irvcorrected for plateout).
the simulated LPIS line (not
67
,." p
E0
65006000
550050004500
40003500
30002500
20001500
1000
5000
-5000 500 1000 1500 2000 2500 3000 3500 4000 4500
Time (s) L11O-KM135-iO
Figure 42. Measured 13 8 Cs concentrationcorrected for plateout).
in the simulated LPIS line (not
68
LPIS line was isolated at: 1777.6 seconds and after that time these figures
show decay processes only. The xenon data shown in Figure 40 has been
truncated because of significant background beyond 1477 seconds.
4.5.4 Gl, G2, and G3 Gamma Spectrometers
The 62 gamma spectrometer was used to m2asure isotopic concentrations
in the combined effluent from the Fl and F2 sample lines. As yet, only
isotopes of krypton and xenon have been identified. Other isotopes may be
identified as data interpretation continues. Results from the GI and
G3 spectrometers are not yet available.
4.5.5 Grab Samples
Grab samples of the primary coolant system (PCS) and of the blowdown
suppression tank (BST) liquid and vapor space were taken both preexperiment
and postexperiment. Since postexperiment saniples were not expected prior
to the test, they represent a potential for enhanced understanding of the
experiment. The postexperiment isotopic concentrations of samples taken
21 days after the experiment are shown in TaIles 9 and 10. Mass
spectroscopy results are shown in Table 11. Preliminary analyses of tne
H2 measurements, shown in Table 11, indicate that the total mass of H2
in the BST was 237 g. This compares reasona)ly well with a postexperiment
estimate of the total H2 release of 320 g.
4.6 Preliminary Analysis of the BST and G5 Data
The purpose of this section is to present a general and preliminary
analysis concerning the 8ST grab samples and 65 spectrometer data.
The first part of the analysis concerns the BST data and is dedicated
to calculating the cumulative release rates to the BST for Xe, Kr, Cs, I,
Ba, Te, and Ru. This calculation is dependelt upon the measured data
presented in Tables 9 and 10, the postexperiment liquid and gas volumes in
the BST, and the ORIGEN2.-calculated center bundle inventory shown in
Table 4.
69
TABLE 9. BST LIQUID GRAB SAMPLE PRELIMINARY RESULTS
and have
Isotope Sample 1A
1-131 9.6 x 10-1Te-132 2.8 x 10-1Cs-137 1.2 x 10-2Cs-136 9.6 x 10-3Ba-140 7.2 x 10-1Ru-103 ND
a. Vecay correction is simple e
All Measurements are in microCuries/cm3been decay corrected to the time of the experimenta
Sample lB Sample 2A Sample 26
9.4 x 10-" 9.8 x 10"I 1.022.6 x 10-1 2.7 x 101- 2.8 x 10-11.2 x 10-2 1.1 x 10-2 1.2 x 10-29.6 x 1o-3 9.1 x 10-3 9.1 x i037.1 x 101- 6.8 x 101- 7.3 x 10-18.6 x 10-4 6.0 x 10-4 6.2 x 1(04
Uncertainty
0.6 x 10-10.2 x 10-10.1 x 10-20.9 x 10-30.6 x 10-10.9 x 10-4
xponential decay with no branching effects.
( (
TABLE 10. BST VAPOR GRAB SAMPLE GAMMA SPECTROSCOPY RESULTS
All Measurements are in microCuries/cm3
and have been decay corrected to the time of the experimenta
Isotope
Xe-133Xe-133mXe-131mKr-851-131
Sample 1
3.7 x 109.1 x 10-17.3 x 10-25.9 x 10-31.2 x 10-4
Sample 2
3.58.77.15.61.2
XXXXx
1010-110-210-310-4
a. Decay correction is simple exponential decay with no branching effects.
TABLE 11. BST VAPOR GRAB SAMPLE MASS SPECTROSCOPY RESULTS
Results are percent by volume
Element Sample 1 Sample 2
H2 3.7 3.6N2 78 7802 4.0 3.8Ar 14.2 14.3CO2 0.06 0.05
71
Prior to the LP-FP-2 experiment, the BST liquid volume was 25.48 m3
and the gas volume was 59.11 m3 (total = 84.59 mi3 ). After the
experiment, the liquid volume increased by 5.25 m 3 to 30.73 mi3, and the
gas volume decreased by 5.25 m3 to 53.86 m3 .
Assuming that the grab sample data represent average conditions
existing in the BST following the experiment, the average sample
concentrations shown in either Table 9 or Table 10 are multiplied by the
respective liquid or vapor volumes (presented above) to obtain the total
amount of curies of each nuclide in the BST. This calculation assumes no
adjustment due to deposition or plateout of fission products in the BST,
which are undetected by the current measurement process. The results of
the calculation are shown in Table 12, along with the ORIGEN2 calculated
fuel inventories and the ratio of the BST activities to the respective
center bundle fuel activities.
The release fraction data presented in Table 12 should be interpreted
as only an initial estimate of the cumulative fraction of the center bundle
inventory that reaches the BST via one of four paths: (a) LPIS, (b) Fl
sample line, (c) F2 sample line, or (d) the PORV (which was opened briefly
during the reflood portion of the experiment). Although the release rate
data in Table 12 should not be interpreted as a cumulative source term for
the experiment, it does represent a minimum, or lower bound, estimate of
the source term. Finally, notice the consistency in the release rates for
the noble gases and the two independent cesium release numbers shown in
Table 12.
The second part of the analysis in this section involves the
determination of the mass concentrations in the LPIS (near the
65 spectrometer) for cesium, iodine, and rubidium. The calculation is
based on the activity data presented in Tables 7 and 8 (or Figures 34
through 42) and the specific activity information shown in Table 4. By
dividing the measured isotopic activity concentration by its specific
activity (see Table 4 for definition), an estimate can be made for the
amount of that element that should be present near the G5 detector. For
72
C (
TABLE 12. CUMULATIVE RELEASE FRACTIONS TO THE BST
BST Data (decay corrected) CumulativeRelease FractiORIGEN2 calculated
Nuclide Fuel Inventory (Ci) Liquid (Ci) Gas (Ci) lotal Ni) to tne bmI1-131 47570.0 29.96 0.0065 29.97 0.00063
Cs-lJb 100.2 0.287 0.287 0.0029Cs-137 144.6 0.361 0.361 0.0U25
Kr-85 17.3 0.310 0.310 0.0179
Xe-131M 237.2 3.88 3.88 0.0164Xe-133 1104UO.O 1939.0 1939.0 0.0176Xe-133M 3362.0 47.9 47.9 0.0142
Te-132 90890.0 8.37 8.37 0.000092
ba-140 80250.0 21.82 21.82 0.00027
Ru-103 18110.0 0.0213 0.0213 0.0000012
a. The cumulative release fraction is defined as the ratio of the total number of curies of a nuclide inthe BST to the number of curies of this nuclide in the center bundle.
rns
example, the specific activity of 1-131 is 55,370 Ci per gram of I. The
1-131 measured activity concentration at 1790 s is
2620 micro-curies/cm 3. Therefore, the mass of I that should exist in theLPIS at 1790 s is calculated to be 0.0473 micrograms/cm3
(0.0473 = 2620/55,370). The results of these calculations, for the iodine,
cesium, and rubidium isotopes measured at the G5 detector location, are
shown in Figure 43.
From Figure 43 it is clear that the concentration of iodine, as
indicated by the five separate calculations based on the 1-131, 1-132,
1-133, 1-134, and 1-135 gamma spectrometer data and the ORIGEN2 specific
activities, are in good agreement. This tends to indicate that the G5 data
are consistent and probably good. In addition to the iodine calculation,
an estimate of the cesium mass concentration, based on the Cs-138 gamma
spectrometer data (which were decay corrected), and an estimate of the
rubidium concentration are also shown in Figure 43. Since Cs-138 has a
relatively short half-life, the gamma spectrometer data for Cs-138 was
decay corrected before the cesium mass concentration was calculated.
The above analyses assume that the relative proportions of the
isotopes within one elemental group are the same for the fuel and the
LPIS. This condition should be met for most measured isotopes that have a
long enough half-life relative to the sampling time (or at least can be
decay corrected as in the case of Cs-138), and whose concentrations do not
depend greatly upon the concentrations of more populous parent nuclides.
For the purposes of generating the results shown in Figure 43,
parent-daughter relationships have been ignored.
From the data presented in Figure 43, it is estimated that the Cs
concentration in the LPIS near the G5 detector is about 0.078 x 10-6
3 -6 3g/cm and the I concentration is approximately 0.055 x 106 g/cm
If it is assumed that all of the I is transported as CsI and that the Cs is
transported as either CsI or CsOH, the mass concentrations of CsI and CsOH
necessary to produce the above calculated concentrations of Cs and I are
computed to be: 0.1127 x 10-6 g/cm3 of CsI and 0.0229 x 10-6 g/cm3
of CsOH. These mass concentrations translate into molecular concentrations
74
E
E
0.09
0.08
0.07
0.06
0.05
0.04
0.03
0.02
0.01
0.00
-0.010 500 1000 1500 2000 2500
Time (s)3000 3500 4000 4500
LI10-KM125-15A
Figure 43. Estimates of the elemental Cs, I, and Rb mass concentrations inthe simulated LPIS line based on isotopic activities measuredby the G5 gamma spectrometer.
75
of 2.613 x 1014 molecules per cm3 for CsI and 9.201 x 1013 molecules
per cm3 for CsOH. Based on this information, it appears that there may
be as much as 2.8 times as many CsI molecules as CsOH molecules in theLPIS. Because of the low wall temperature of the LPIS, it is expected that
much of the Cs and I (in the form of CsI) is probably condensed on thewalls of the LPIS (or F3 filter) or on deposited aerosols. In either case,
much of the Cs and I probably deposited early during the experiment and
only a small amount reached the BST. This conclusion appears to be
supported by the fact that only a small fraction of the center bundleinventory of Cs and I (compared to the noble gases) was detected in the
BST, as shown by the information contained in Table 12.
4.7 Potential for Meeting Fission Product Measurement Objectives
The fission product measurement objectives were derived from the
governing objectives for Experiment LP-FP-2. Achievement of the fission
product measurement objectives depends upon the thermal/hydraulic
instrumentation, as well as the FPMS and associated PIE. Since PIE is
incomplete and FPMS data have not been fully analyzed (particularly the
gamma spectrometer data), it is not possible to conclusively state that the
fission product measurement objectives have been achieved. However, on the
assumption that the data obtained, or to be obtained, are satisfactory, thepotential for achieving the objectives can be assessed.
Objective 1
Determine the fraction of volatile fission products (Cs, I, Te, Xe,
Kr) and aerosols released to and from the upper plenum region. Achievement
of this objective will require data from the Fl sample line, the F2 sample
line, and the Dl deposition coupon device. Sufficient thermal/hydraulic
data to determine flow conditions through the center fuel module and upper
structure would also support the achievement of this objective.Achievement of this objective appears likely at this time, although having
a gross gamma monitor instead of a spectrometer at the G6 location may
affect the uncertainty of the results.
76
Objective 2
Determine the fraction of volatile fission products and aerosols
transported out of the primary coolant systen. Achievement of this
objective will require an adequate fission product mass balance as well as
data on the volatile fission product inventory in the BST. Achievement of
this objective at this time appears likely.
Objective 3
Determine the retention of volatile fission products on representative
primary coolant surfaces in the plenum and piping. Resolution of this
objective will require measurements of volatile fission product retention
on the deposition coupons in the upper structure and on deposition spool
pieces in the LPIS line. Achievement of this objective appears likely,
although if all three protected coupons on the upper plenum deposition
coupon device have been exposed to reflood, there may be some additional
uncertainty.
Objective 4
Determine the general mass balance of volatile fission products in the
fuel, primary coolant system and blowdown suppression tank. This objective
requires data from all of the fission product and PIE measurements. In
addition, some supporting thermal/hydraulic data are required for
estimating the fission product plateout on PCS surfaces. Achievement of
this objective appears likely.
4.8 Future PIE Plans
After the experiment, a variety of samp'es (fluid and FPMS components)
will be examined. These samples and the planned analyses for these samples
are identified in Table 13. Results of these analyses will provide fission
product release, transport, and deposition dzlta during the early phases of
a risk dominant reactor transient. These data will aid the understanding
of fission product behavior and will be used to assess the capability ofcomputer models to predict fission product rElease and transport.
77
TABLE 13. PLANNED POSTIRRADIATION EXAMINATION FOR EXPERIMENT LP-FP-2
Analyses
Samples
AerosolBST/PCS BST Deposition Deposition LPIS Cyclone DilutionLiquids Vapor Device Spool Piece Filter Separator Filters
AerosolAerosol Collection Hydrogen Steam
Impactors Filter Recombiner Condenser
Gammaspectrometer
Sr-89, -90
Fissile
Total mass-10o Elemental
Particle size
Compounds
x x x x x x x x x x x
X x
x
xx
x x
x
x
x
x
K
K
xK
K
K
K
x
K
K
K
( (
5. CONCLUSION:S
The conduct and results of this experiment are considered sufficient
to meet the experiment objectives. Specifically, the thermal/hydraulic
conditions during the release and transport of fission products adequately
simulated those expected to occur during a V.-Sequence accident. It should
be noted that much of the information, especially fission product data,
needed to confirm this general conclusion is not yet available. Thus, the
realization of the specific experiment objec:ives depends on data that are
currently being collected. Specific conclusions and observations based on
the limited analysis completed thus far are:
1. The thermal and hydraulic boundary conditions existing during
fission product. release and transport adequately simulated the
early phases of a V-sequence accident and were as desired for
fission product transport and depo;ition.
2. The fuel rod cladding temperatures exceeded 2100 K (3320 0 F) for
at least 4-1/2 min. This exceeded by 50% the experiment goal of
at least 3 min at those temperatures.
3. The flow resistance in the simulated LPIS line was much higher
than was expected. This delayed the depressurization into the
metal-water reaction time frame and resulted in a higher than
expected steam flow rate in the center fuel module.
4. A higher-than-predicted steam flow rate in the center fuel module
resulted in adequate steam to sustain a rapid metal-water
reaction in the upper region of the core and in the elevation of
the maximum temperature being much higher than was expected.
5. The center fuel module control rods melted and a significant
fraction of the module is judged to have been relocated to the
bottom of the module.
79
6. Results of the postexperiment ORIGEN2 analysis indicate that the
Cs to I mass ratio in the center fuel module at the beginning of
the experiment is 4.00. The fission gas release calculations,
based on the measured thermocouple data and the NUREG-0072 data
base, indicate that between 20.6 and 25% of the initial
inventories of Xe, Kr, Cs, and I were released to the coolantduring the course of the experiment.
7. Based on the limited amount of fission product data available at
the time of this report, it is judged that all experiment
objectives are achievable.
80
6.. REFERENCES
1. P. R. Davis et al., "The Risk Significant of Transient Accidents fromPRA Studies," ANS Topical Meeting on Anticipated and AbnormalTransients in Light Water Reactors, Jackson, WY, September 1983.
2. Reactor Safety Study--An Assessment of Accident Risks in U.S.Commercial Nuclear Power Plants, WASH-1400, USNRC, October 1975.
3. J. P. Adams, Quick-Look Report on LOFT Nuclear Experiments L5-1 andL8-2, EGG-LOFT-5625, October 1981.
4. D. B. Jarrell and J. M. Divine, Experiment Data Report for LOFTIntermediate Break Experiment L5-1 and Severe Core TransientExperiment L8-2, NUREG/CR-2398, EGG-2136, November 1981.
5. J. P. Adams, Quick-Look Report on LOFT Nuclear Experiment L3-1,EGG-LOFT-5057, November 1979.
6. P. 0. Bayless, J. B. Marlow, R. H. Aver'11, Experiment Data Report forLOFT Nuclear Small Break Experiment L3-1, NUREG/CR-l145, EGG-2007,January 1980.
7. J. P. Adams, Quick-Look Report on LOFT Nuclear Experiment L3-5/L3-5A,EGG-LOFT-5242, October 1980.
8. L. T. L. Dao and J. M. Carpenter, Experiment Data Report for LOFTNuclear Small Break Experiment L3-5/L3-EA, NUREG/CR-1695, EGG-2060,November 1980.
9. G. E. McCreery, Quick-Look Report on LOFT Nuclear ExperimentL3-6/L8-1, EGG-LOFT-5318, December 1980.
10. P. 0. Bayless and J. M. Carpenter, Experiment Data Report for LOFTNuclear Small Break Experiment L3-6 and Severe Core TransientExperiment L8-1, NUREG/CR-1868, EGG-207E, January 1981.
11. V. T. Berta, OECD LOFT Project Experiment Specification DocumentFission Product Experiment LP-FP-2, OEC[' LOFT-T-3802, Rev. 1, May 1985.
12. J. P. Adams et al., Quick-Look Report on OECD LOFT Experiment LP-FP-l,OECD LOFT-T-3704, March 1985.
13. D. L. Reeder, LOFT System and Test Description (5.5 ft Nuclear Core ILOCES), Change 1, NUREG/CR-0247, TREE-1208, July 1978.
14. L. J. Ybarrondo et a]., "Examination of LOFT Scaling," 74-WA-HT-53,Proceedings of the Winter Meetino of the American Society ofMechanical Engineers, New York, November 17 - 22, 1974, CUN-741104.
81
15. V. H. Ransom et al., REALP5/MOD2 Code Manual, EGG-SAAM-6377, April1984.
16. J. W. Spore et al., TRAC/BOI: An Advanced Best Estimate ComputerProgram for Boiling Water Reactor Loss-of-Coolant Accident Analysis,NUREG/CR-2178, EGG-2109, uctober 1981.
17. 0. J. Usetek et al., "Fission Product Behavior during the First TwoPBF Severe Fuel Damage Tests," ANS Topical Meeting on Fission ProductBehavior and Source Term Research, Snowbird, Utah, July l5-19, 1984.
18. J. P. Adams and J. C. Birchley, Quick-Look Report on OECD LOFTExperiment LP-LB-l, OECO LOFT-T-3504, February 1984.
19. S. Guntay, M. Carboneau, Y. Anoda, Best Estimate Prediction for OECDLOFT Project Fission Product Experiment LI-l-P-Z, ULGU LUI-I-i8UJ,June 1985.
20. A. G. Croff, A Users Manual for the ORIGEN2 Computer Code,ORNL-TM-7175, July 1985.
21. B. L. Rushton and J. B. Briggs, PDQ Calculated Results for SafetyAnalyses Evaluations for the FP-2 Reload Core at Beginning of Life,OECD LOFT-I-08-5118, December 21, 1984.
22. M. R. Kuhlman et al., CORSOR User's Manual, NUREG/CR-4173, BMI-2122,March 1985.
23. M. L. Carboneau, A Report on the Transient Isotope Generation andElemental Release (TIGER) Program, to be published.
24. USNRC, Technical Bases for Estimating Fission Product Behavior DuringLWR Accidents, NUREG-0772, June 1982.
25. R. R. Sherry letter to 0. L. Batt, "Letter Report on the LOFT FP-2CORSOR Analysis," CD-CAS-85-175, August 9, 1985.
26. P. North letter to J. E. Solecki, "Potential Programmatic RiskAssessment for LP-FP-2 Experiment," PN-106-85, May 31, 1985.
82
APPENDIX AREVISIONS TO THE EXPERIMENT SPECIFICATION
DOCUMENT FOR EXPERIMENT LF'-FP-2
( OPER. NO.
PAGE I...OF -- 3-FORM EG&G-1844R na- DOCUMENT REVISION BEQUEST
( ) REQUESTER I OAR DATE , OAR NO.V.T.Berta 5-31-85 L-7423
(5.) DOCUMENT NO. (IF APPLICABLE) DOCUMENT TITLE DOCUMENT ISSUE DATE
OECD LOFT-T-3802 Rev.1 OECD LOFT Project Experiment Specification May 19851nm mpi F'nm.4 P-rnt'lt t.xeXrm TLP-FPý2 Na 18
(D) CHECK APPLICABLE BLANK (-) MAAGEFP APPR9VAL .r DATE
PERMANENT CHANGE X TEMPORARY CHANGE __ BULLETIN 7A ,/,(,/L" " . -2.- ,AJPRINT OR TYPE PROPOSED CHANGE -- NUMBER EACH CHANGE SEOUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER FORFOR EACH CHAeNGE. WATR'S USE
STEP OR INSTRUCTIONS: REWRIrEPARAGRAPH(S) OR FOR EXTENSIVE CIIAIGFS ATTACH REVISED COPY AND STATE "REVISE PERITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT ANE STATE "PREPARE NEW (SP. DOP. ETC.) PER ATTACHEDDRAFT'.
1 8 2nd Section 3.4 2nd paragraph let sentence. Delete "At 6+-is".Begin the sentence with the word "Following".
2 7 1st Section 3.3 1st paragraph. Replace the first paragraph with thefollowing paragraph.
The transient phase of the experiment will be initiated bya reactor shutdown as specified in Section 4.3. Time zero
corresponds to the action taken to drop the center fuel module(CFM) control rods. The 1.16 in. inner diameter simulated break
in the intact loop cold leg will bE opened at 20+2s. The primarycoolant pumps will be tripped at 25i±5s and will undergo a normal
coastdown. At 220*5s the simulatec. LPIS line (also 1.16 in.diameter) will be opened. (The LPIS line filter shall be bypassedto prevent plugging prior to the fission product release.) This
line is connected to the broken loop hot leg and to a blowdown
suppression tank (BST) inlet vent. The intact loop cold legsimulated break will be closed at either a CFM cladding temperature
(CONTINUED) USE CONTINUATIIN SHEET AS REQUIREDNEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATE/EVENT
®)JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. AROVE): (1V OTHER DOCUMENTATION AFFECTED:
DOC. NO. ORR NO. DATE COMPLETEDApproved experiment planning revisions
(12) ORIGINATING OAR NO:
@ REVIEW_
NAME/SIGNATURE ORG. DATE NAMEISIGNATURE OCG _ RATE NAME/SIGNATURE ORG. DATE
-- 11 _ 4 ____ f•____ __-_ _ PUACTY
cpr 24-10 s/i______I ________ SHS
7 1, o 1ii~J ______________
(9COMMENTS: * I f(P.C 4 ,A,.-i s RADDITIONALRS IN THISDOCUMENT
REVISION
DOCUMENT CONTROLLER
Q-Mýl_) DOCMEN CTR L. 6-" ... D cM.
nMPI ETED DATE:
DOCUMENT REVISION REQUEST OPER. NO.(CONTINUATION SHEET) PAGE OF 3
FORM EG&G-1544A(A" 5-"7) E .- -74;
DOCUMENT NO. (OF APPLICABLE) DOCUMENT TITLE DOCUMENT ISSUE DATEBECD LO TPro ect Ex~erimntSeii tinOECD LOFT-T-3802 Rev.1 •ocUMen7Fsson Pro uc• Epterimen• ,f -•-E May 1985
PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBERFOR EACH CHANGE. FOR
WRITER'S
STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER USEITEM PAGE PARA. ATTACHED COPr. FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP. DOP. ETC.) PER ATTACHED DRAFT".
2 7 lst of 566 + 5K (560 + 90F) or a system pressure of 1.2 + 0.03 MPa
(160 + 5 psig). The fuel rod cladding is calculated to beginheatup at approximately 700s and to reach 2100K at approximately 17(0s.Fission product and aerosol release will occur from fuel rod
failure until the center assembly shroud outer wall reaches
1473K (21920F) or the peripheral fuel rod cladding reaches1417K (20920F), at which time the fission product filter samplingsystems and the upper plenum deposition coupons will be closedin the primary coolant system coincident with the closure of the
simulated LPIS line. Within 6 + O.5s of these actions core reflood
will commence with both accumulators.
3 13 13 Step 13 is to'read as follows:13. Close the FPMS sampling system lines, the deposition coupon
device, and the simulated LPIS line in the broken loop hot
leg at a cladding temperature of 1417K (2092OF) or a
thermal shroud outer wall temperature of 1473K (21920F).
4 13 14 Step 14 is to read as follows:14. Initiate core reflood 6 + 0.5s after initiation of the
system closures in Step 13.
5 20 6 2nd paragraph of Section 4.7 is to read as follows:
Sequence Step 7 begins the transient phase of the experiment.Reactor shutdown with all control rods in as specified in Step 7
is to be completed before proceeding to Step 8. The eleven control
rods in the CFM are the aerosol source for the experiment.
If these control rods cannot be inserted the experiment sequence-mu t be
(the remaining part of this paragraph on page 21 is unchanged)
DOCUMENT REVISION FREQUEST OPER. NO.(CONTINUATION SHEET) PAGE 3 OF 3
FORM EG&G.1i44A(Rev. 5-77) DRR NO. -__74 2 3__"_
DOCUMENT NO. (IF APPLICABLE) OECD LOFT PrO TntNW riren t S peci fi cation DOCUMENT ISSUE DATE
OECD LOFT-T-3802 Rev.1 Document Fission Product ExpEriment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEQUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER
FOR EACH CHANGE.FOR
WRITER'SSTEP OR I INSTRUCTIONS: REWRITE PARAGRAPHIS) OR FOR EXTENSIVE CHAIGES ATTACH REVISED COPY AND STATE "REVISE PER USE
ITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP, DOP, ETC.) PER ATTACHED DRAFT".
6 22 table Delete the second sentence in the course of action for item 1. The
course of action is to read as follows:
If the F1 and F2 isolation valvesdo not open within 5s terminate theexperiment and commence recoveryoperations with the ECCS. Do not openthe LPIS line filter and do not store thegamma densitometer sources.
7 22 table Add the following item to the table on page 22:
All criteria met and F1 and The experiment is to proceed.
F2 open but LPIS line filter Continue attempts to completecannot be opened and/or these actions.gamma densitometer sourcescannot be stored.
8 22 Add the following sentences after tie table on page 22. Thisaddition is riot a new paragraph.
If the condition occurs where the s'stem pressure is above themaximum allowable for F1 and F2 operation when the CFM claddingtemperature is 840K (10520F), commence actions to lower thesystem pressure. These actions may be, but not limited to, openingthe PR x xr A Terminatea othese actions before the CFM cladding temperature reaches150K 4310 F). The experiment is to continue in the event thatthe system pressure cannot be lowered below the F1 and F2operati.ng pressure limit.
opening the intact loop break path.
DRR-L-7423
Rev. 1 Chg. I
of an effective 40 hours will establish the required minimum decay heat
level of 675 kW at 200 s and will complete the burnup required on the
9.72 wt% enriched fuel rods. The plant should be operated at a steady
state power level of 26.5 ± 1.0 MW for at least 2 hrs prior to experiment
initiation.
3.3 Transient Phase
The transient phase of the experiment will be initiated by a reactor
shutdown as specified in Section 4.3. Time zero corresponds to the action
taken to drop the center fuel module (CFM) controls rods. The 1.16 in.
inner diameter simulated break in the intact loop cold leg will be opened
at 20 ± 2 s. The primary coolant pumps will be tripped at 25 ± 5 s and
will undergo a normal coastdown. At 220 ± 5 s the simulated LPIS line
(alos 1.16 in. diameter) will be opened. (The LPIS line filter shall be
bypassed to prevent plugging prior to the fission product release.) This
line is connected to the broken loop hot leg and to a blowdown suppression
tank (BST) inlet vent. The intact loop cold leg simulated break will be
closed at either a CFM cladding temperature of 566 ± 5 K (560 ± 91F) or a
system pressure of 1.2 ± 0.03 MPa (160 ± 5 psig). The fuel rod cladding is
calculated to begin heatup at approximately 700 s and to reach 2100 K at
approximately 1700 s. Fission product and aerosol release will occur from
fuel rod failure until the center assembly shroud outer wall reaches 1473 K
(2192 0 F) or the peripheral fuel rod cladding reaches 1417 K (20921F), at
which time the fission product filter sampling systems and the upper plenum
deposition coupons will be closed in the primary coolant system coincident
with the closure of the simulated LPIS line. Within 6 ± 0.5 s of these
actions core reflood will commence with both accumulators.
The beginning and end of the transient phase of the experiment are
defined as follows:
7
DRR-L-7423
Rev. 1 Chg. I
Beginning initiation of the transient hy a reactor scram
End initiation of the closure of the simulated LPIS line in
the broken loop hot leg.
3.4 Posttransient Phase
The posttransient phase consists of a time interval of 12 hr for
measurement of (a) the redistribution of fission product-inventory in the
gas and liquid volumes in the blowdown suppression tank (BST), and (b) the
leaching of fission products from the damaged fuel rods in the primary
coolant system (PCS). The beginning and end of the posttransient phase of
the experiment are defined as follows:
Beginning initiation of the closure of the simulated LPIS line in
the broken loop hot leg
End completion of the time interval specified for fission
product measurements.
Following closure initiation of the simulated LPIS line in the primary
coolant system, reflood operations will commence with initiation of both
accumulators. System refill will continue as required with the high
pressure injection system (HPIS). The governing requirements for plant
operation involving the PCS in this phase are:
1. Mass transfer from the PCS is to be minimized. The purification
system can be used in the decay heat removal mode (bypassing the
ion exchanger) for temperature control along with steam generator
feed and bleed.
2. Forced coolant circulation with the primary coolant pumps is
prohibited.
3. PCS temperature is to be reduced and maintained below 449 K
(350*F) as soon as possible after simulated LPIS line closure.
8
DRR-L-7423
Rev. 1 Chg. I
b. Center fuel assembly cladding temperature increases to
566 ± 5K (5601 ± 90 F).
The cladding temperature following time zero will decrease and
correspond to saturation temperature until core uncovery occurs.
Then, cladding heatup will commence at some temperature below the
value specified in (b). Therefore, cladding temperature must be
increasing in order for criterion (b) to be valid.
12. Open the FPMS F1 and F2 sampling systems at a center fuel
assembly cladding temperature of 840 ± 5K (1052 ± 91F) and open
only if the system pressure is less than 1.43 ± 0.03 MPa
(195 ± 5 psig). )if the pressure criterion is met, for the F1 and
F2 sampling systems, also open the LPIS break line filter and
store the gamma densitometer sources. Thermocouples to be used
in determining the temperature are listed in Table 8.
13. Close the FPMS sampling system lines, the deposition coupon
device, and the simulated LPIS line in the broken loop hot leg at
a CFM cladding temperature of 1417 K (20921F) or a thermal shroud
outer wall temperature of 1473 K (2192°F).
14. Initiate core reflood 6 ± 0.5 s after initiation of the system
closures in Step 13.
15. Continue system refill with the HPIS and maintain the PCS
temperature below 449 K (350'F) as soon as possible for the
remainder of a 12 hr minimum time after initiation of core
reflood. Maintain the PCS pressure below 8.96 MPa (1300 psig).
Mass transfer to and from the PCS, excluding system leakage and
replacement, is to be minimized. PCS energy control is to be
accomplished with steam generator feed/bleed operations and/or
13
with the purification system in the decay heat removal mode (ion
exchanger bypassed). PORV venting is permitted but is to be
minimized to the extent possible.
16. Maintain isolation of the BST (no mass transfer into or out of
the BST) for the same time interval as in Item 15 subject to
plant limiting conditions which may require operator action.
Within 10 minutes of PCS isolation purge the BST downcomer with
60 SCF of N2.2
17. Maintain PLSS data acquisition continuously over the time
interval for Items 15 and 16.
4.4 System Configuration
The general system and component configuration of the LOFT PWR for
Experiment LP-FP-2 is shown in Figure 1. Specific details are given inthe
following sections on the reactor core, primary coolant system, secondary
coolant system, blowdown system, and emergency core coolant system.
4.4.1 Reactor Core
Experiment LP-FP-2 will be conducted with a specially constructed
center fuel assembly. The cross section of this fuel assembly is shown ini
Figure 2. The fuel rods are 350 psia prepressurized and 9.72 wt%
enriched. The outer two rows of fuel rods have been replaced with a
thermal shield of zircaloy with zirconium oxide ceramic internal
insulation. The purpose of the higher than normal enrichment and the
thermal shield is to provide at least 3 minutes at peak cladding
temperatures above 2100 K (3321*F) in the center fuel assembly before the
peripheral assembly fuel cladding reaches the transient termination
temperature. This time at temperature will provide a sufficiently large
fission product and aerosol release fraction.
The aerosol release will occur from (Ag-In-Cd) control rods which will
be inserted at reactor scram in the guide tubes shown in Figure 2.
14 I
4.5.3 Simulated LPIS Pipe
Thermal-hydraulic measurements in the simulated LPIS pipe will consist
of steam flow, steam temperature, and wall temperature.
4.5.4 FPMS
The identification and location of fission product measurements are
shown in Figure 4. These measurements are described in the FPMS Functional4
and Operational Requirements (F&OR) document.
4.5.5 Postirradiation Examination
An integral part of the measurements which are necessary to meet the
experiment objectives are the postirradlation examination (PIE)
measurements. A summary of' items specified for postirradiation examination
is contained in Table 5. Examination of these items will be in accordance
with the postirradiation plan for LP-FP-2. 5
4.5.6 Critical Measurements
Sets of critical measurements, required during the transient and
posttransient phases of the experiment, have been identified and are listed
in Tables 6 and 7, respectively. The transient phase of the experiment
should not be initiated without these measurements since the experiment
objectives may be jeopardized. Appendix A lists by instrument identifier
all critical measurements which are considered necessary for the successful
conduct of the experiment. The measurement uncertainties will be equal to
or less than those specified in the document "LOFT Experimental Measurement
Uncertainty Analysis," NUREG/CR-0169.
A complete list of measurements required for Experiment LP-FP-2 is
provided on the Data Acquisition Requirements list to be published prior to
the experiment.
19 I
DRR-L-7423
Rev. 1 Chg. 1
The digital data acquisition and processing system, and analog and
digital data acquisition recording is required to begin no later than 1 min
before initiation of the transient phase of the experiment. Continuous
PLSS recording is required through the posttransient phase of the
experiment.
Measurements identified on the Data Acquisition Requirements List that
fail prior to experiment initiation should be repaired or replaced if
possible. If a failed instrument(s) cannot be repaired or replaced, the
Joint Experiment Group shall determine the course of action.
Process instruments requiring calibration prior to Experiment LP-FP-2
are listed in Table 1.
4.6 Experiment Termination
Experiment LP-FP-2 will be terminated at the end of the posttransient
phase of the experiment. The posttransient phase ends with the completion
of the time interval required for monitoring the redistribution of fission
products in the vapor and liquid volumes in the blowdown suppression
system. The time interval is specified in Section 4.3 to be 12 hr minimum
after closure of the simulated LPIS line.
4.7 Abnormal Experiment Sequence
If instrumentation, hardware components, or operating systems fail
prior to or during any of the four phases of the experiment, every effort
should be made to substitute, repair, or provide alternate actions to
safely continue the experiment and to meet the programmatic objectives.
Sequence Step 7 begins the transient phase of the experiment.
Reactor shutdown with all control rods in as specified in Step 7 is to
be completed before proceeding to Step 8. The eleven control rods in
the CFM are the aerosol source for the experiment. If these control
rods cannot be inserted the experiment sequency must be
20
stopped. On completion of the corrective actions the experiment is to be
resumed as approved by the LOFT Operations Manager, DOE Site Program
Manager and LOFT Program Division Manager.
The decision matrix on center assembly control rod insertion is as
follows:
Condition Decision
Reactivity change and one rod bottom Continue experiment sequencelight indication
Reactivity change and no rod bottom Terminate experimentlight indication
No reactivity change and no rod bottom Terminate experimentlight indication
Sequence Step 8 opens the break path in the intact loop cold leg.
This flow path is the primary blowdown path and is intended to be the path
for venting the major part of the primary coolant system fluid. High
quality steam flow only is desired for venting through the simulated LPIS
pipe in the broken loop hot leg. If the intact loop cold leg break cannot
be opened, the experiment sequence must be halted. On completion of the
corrective actions the experiment is to be resumed as approved by the LOFT
Operations Manager, DOE Site Program Manager and LOFT Program Division
Manager.
The simulated LPIS line is opened at 220 ± 5 s in Sequence Step 10.
This flow path is the designed fission product and aerosol vent path to the
BST. If this line cannot be opened within 50 s of the time specified,
close the intact loop cold leg break and commence recovery operations with
the ECCS. Resume the experiment as approved by the LOFT Operations
Manager, DOE Site Program Manager and LOFT Program Division Manager.
The intact loop cold leg break is to be closed on either a cladding
temperature value or a system pressure value as specified in Sequence
Step 11. If this break is not closed then another flow path to the BST
will exist for fission product and aerosol venting. Two vent paths are not
21 I
DRR-L-7423
Rev. 1 Chg.
provided for in the experiment plan. If this break cannot be closed before
the CFM cladding reaches 840 ± 5 K (1052 ± 90 F) then terminate the
experiment and commence recovery operations with the ECCS. If this break
has not been closed after 566 K (5601F) and the system pressure decreases
below 1.2 ± 0.03 MPa (160 ± 5 psig) before the cladding reaches 840 ± 5K
(1052 ± 91F) then, the Operations Branch is to consult with the JEG to
decide if attempts to close this break should continue or if the experiment
should be terminated. If experiment termination and plant recovery
operations commence with the ECCS, return to Sequence Step I or as approved
by the LOFT Operations Manager, DOE Site Program Manager, and LOFT Program
Division Manager after repairs are made.
The FPMS F1 and F2 sampling systems are to be opened at a cladding
temperature of 840 ± 5K (1052 ± 90 F) if the system pressure has decreased
below 1.43 ± 0.03 MPa (195 ± 5 psig). The pressure criterion corresponds
to the F1 and F2 design pressure limit of 200 psig. The LPIS line filter
is to be valved in and the gamma densitometer sources are to be stored in
conjunction with the opening of the F1 and F2 sampling systems. The
following courses of action are defined in the event that abnormal
conditions occur:
1
Abnormal Condition Course of Action
System pressure below criterionbut F1 and F2 valves fail to open
Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.
All criteria met and F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.
If the F1 and F2 isolation valvesdo not open within 5s terminate theexperiment and commence recoveryoperations with the ECCS. Do notopen the LPIS line filter and do notstore the gamma densitometer sources.
Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.
The experiment is to proceed.Continue attempts to completethese actions.
N '-, I
22
DRR-L-7423Rev. I Chg. 1
If the condition occurs where the system pressure is above the maximum
allowable for F1 and F2 operation when the CFM cladding temperature is
840 K (1052°F), commence actions to lower the system pressure. These
actions may be, but not limited to, opening the PORV or opening the intact
loop break path. Terminate all of these actions before the CFM cladding
temperature reaches 1050 K (1431°F). The experiment is to continue in
the event that the system pressure cannot be lowered belwo the F1 and
F2 operating pressure limit.
The LPIS break line filter is to be valved in when the peak cladding
temperature is 840 + 90 F). As the filter loads up the differential pres-
sure may increase. At a filter differential pressure of
22A
(59 OPER. NO
FORM EG&G-1844 ,DOCUMENT REVISION iFUUEIST PAGEfle ,. Wg- 2) P. IOF
(D REQUESTER (1 ORR DATE DAII NO
V. T. Berta June 26, 1985 L - 79 37C5 DOCUMENT NO. (IF APPLICABLE) DOCUMENT In E DOCUMENT ISSUE DATE
OECD LOFT Project Experiment Specification DocumentOECD LOFT-T-3802 Rev. 1 Fission-Iroduct-Experiment LP-FP-2 .May 1985
® CHECK APPLICABLE BLANK b(,1) M-ANAR APPROVAL/, DATE
PERMANENT CHANGE __ TEMPORARY CHANGE .-- II,.,.ETIN
(1 PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE SEQUENTIALLY IN IST COLUMN AND RFCORD PAGE AND STEP OR PARAGRAPH NUMBER 1 9 FORFOR EACH CHANGE. WR TER'S USE
v 4,%
ITEM PAGESTEP OR
PARA.INSTRUCTIONS: REWRI[E PARAGRAPH(S) ORR EX TENSIVE CHANGES AT TACH REVISED COPY AND STATE 'REVISE PERATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGHOIIA.FTANID STAI E PREFARE NEW (SP. DOP. ETC.) PER ATTACHEDDRAFr.
-4-4-4
1 5 3.1s
Place an asterisk after the sentence, "In termw of reactoroperation...."
Add the following footnote at the bottom of the page:
Any combination of preconditioning and pretransient power
operation that provides the specified initial conditions(Section 4.2) is permissable. The specific combinationdescribed is for example and for ease in attaining the initialconditions in the absence of factors affecting power operation.
USE CONTINUATION SHEET AS REQUIREDNEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATEIEVENT
0
(• JUSTIFICATION: (REASON FOR CHANGE - NUMBEFi TO CORRESPOND TO ITEM NO. ABOVE) (@1 OTIIER DOCUMENTATION AFFECTED:
Provides the Operations Branch with flexibility in DDCNO. ORRNO. DATECOMPLETEDattaining the specified initial conditions and allows for OECD-LOET-I_-I_1515- ] . h!5!/..power operation variations that may be needed to account or---factors that may arise which affect power operation.
T172); 0RIGINAIING DRR NO:
@9 REVIEW
NAMEISIGNATURE ORG DATE NAMFISIGNATIUInE ORG. RATE NAME/SIGNATIURE ORG. DATE
. m-i.cy 10 &Wjp ed•", _ QUALITY
__,_,_____/,j____,o_____.-__ ASHS
Iazc Z ~ i 0~ it _ _~ Ii_ _ _ _ _ _ PRAC __
(94 COMEWS: I q 5ADDITIONALRRS IN THISDOCUMENTREVISION
Q) DOCUMENT CONTROLLER J), RELEASE DTE: \- • .'-IR" 1 I. :OMPLEIEO DATE:JT:•A--• ", we- '\ .
)RR-L-7437Rev. 1 Chg. 2
3. EXPERIMENT DESCRIPTION
Experiment LP-FP-2 consists of four distinct phases. These phases are
designated as (a) fuel preconditioning, (b) pretransient, (c) transient,
and (d) posttransient. The four phases together represent a continuous
process and have specific beginning and ending definitions. Each phase is
described in the following sections.
3.1 Preconditioning Phase
The purpose of the preconditioning phase is to subject the fuel rods
in the new center assembly [19.72 wt% enriched, 2.41 MPa (350 psia)
prepressurization] to a burnup which, in combination with the burnup
corresponding to the EFPH requirement at the experiment initial conditions,
provides the required minimum burnup for the experiment. The minimum
burnup required for these fuel rods is 325 MWD/MTU. The minimum burnup
required in the preconditioning phase is 252 MWD/MTU. In terms of reactor
operation, the preconditioning burnup is equivalent, as an example, to
power operation at a maximum linear heat generation rate (MLHGR) of
52.2 kW/m (16 kW/ft) for 111.5 hours on the 9.72 wt% enriched fuel rods* A
core power level of 32.0 MW ± 0.5 is calculated to provide a MLHGR of
-16 kW/ft on the 9.72 wt% enriched fuel rods. Power profile data using
the traversing incore probe system (TIPS) must be obtained during the
preconditioning phase. The TIPS locations in the center assembly will be
capped during the reactor shutdown time in the pretransient phase.
The beginning and end of the preconditioning phase are defined as
follows:
Beginning the start of plant heatup prior to power operation to
establish fuel burnup
End termination of power operation after the required
burnup in this phase has been achieved.
Any combination of preconditioning and pretransient power operation that provides
the specified initial conditions (Section 4.2) is permissable. The specificcombination described is for example and for ease in attaining the initialconditions in the absence of factors affecting power operation .
5
The preconditioning phase does not require special procedures relative
to experiment specification and is to be conducted with established plant
operating procedures.
3.2 Pretranslent Phase
The pretransient phase consists of a reactor shutdown interval of
2-5 days followed by a power operation interval. The final plant
preparations are to be completed during the reactor shutdown interval. The
power operation interval is to complete the required central assembly fuel
burnup, establish the required minimum decay heat level, and establish the
required initial conditions for conduct of the experiment. The beginning
and end of the pretransient phase are defined as follows:
Beginning termination of the power operation after the required
burnup is achieved in the preconditioning phase
End initiation of the transient by a reactor scram.
Specifically, the final plant preparations required by this
specification are:
1. Operational readiness of thermal hydraulic and nuclear
measurements (experimental and process) required by this
specification for this experiment and approval by the LOFT
Operations Branch to proceed into power operation andsubsequently into the transient phase of the experiment.
2. Operational readiness of the fission product measurement system
(FPMS) in accordance with FPMS operational documents.
The power operation prior to experiment initiation is to be an
effective 40 hours at a core power level of 26.5 ± 0.5 MW (21.2 EFPH*).
This core power is the required initial condition value. Power operation
*EFPH is defined as 50 MW-hr.
6
FORM EGAG.Bd44(Ile-. 09-82)
DOCUMENT REVISION REQUEST() OPER. NO. -
PAGE 1_L OF § 5
(D REQUESTER 39 OAR DATE 49 DRR NO.
V. T. Berta June 17, 1985 L-7435() DOCUMENT NO. (IF APPLICABLE) OECD LOFT Prq95ftNT P erimen t Speci fi cation DocumPfEMENT ISSUE DATE
OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 198569 CHECK APPLICABLE BLANK (0) MAGERfPPROV DATE
PERMANENT CHANGE X TEMPORARY CHANGE - BULLETIN /t"
@9 PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE SEDUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER (i FORFOR EACH CHANGE. WRTER'S USE
STEP OR INSTRUCTIONS: REWRITE' PAGRAPHIS) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE REVISE PERI TEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP. DOP. ETC I PER ATTACHED DRAFT'.
1 10 Section 4.1 item 3 Change to read:
3. Complete the pre-experiment calibration requirements for the FPM!instruments and the pre-experiment calibration requirementsspecified in DOP-87-005, "DAVDS Experimental Measurements andTest Procedure," and in DOP-87-008, "Pre LOCE Data Verification"
2 71 Tbl 2 Change the first entry to read:
Reactor shutdown CFM CR drop t = 0Peripheral CR scram 5 s
- 3W I add "or other experiment termination events." under time/setpoint
for FPMS, Broken loop and deposition closure.
Change time/se-tpoint for FPMS sampling system isolation valvea"Open" from A89i 840 + 5 K (1052 + 9 F) to 811 + 5 K (1000 + 9 F)
USE CONTINUATION SHEET AS REQUIREDNEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATE/EVENT
6•JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. ABOVEI: (© OHER DOCUMENTATION AFFECTED:
DOC. NO. DRR NO. DATE COMPLETED
Approved experiment changes. OECD LOFT-I-11-5150 6/5/85
r(l2 ORIGINATING DAR NO:
(D3 REVIEW
NAME/SIGNATURE ORG. DATE NAME/SIGNATURE ORG. RATE NAME/SIGNATURE ORG. DATE
16o j2 QUALITY
~21 _ _SHS
PRAC
1(9 COMMENTS: 0 ADDITIONALARS IN THIS
DOUMN
'i) DOCUMENT CONTROLLER
1 0 ;Z-11
EEASE .,- , ORBCOMM ETED DATE:
K I _--_ AR
DOCUMENT REVISION REQUEST OPER. NO.(CONTINUATION SHEET) PAGE. 2_ OF -, 5
rOnRM EG&O-11M4AMe,. 11.70 1 DRR NO _L-7435
DOCUMENT NO. (IF APPLICABLE) _T S DOCUMENT ISSUE DATEDOCMENNO IF PPLCABE)OECD LOFT 'roje~c txperiment Specification uocumenTu
OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1985
PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEQUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBERFOR EACH CHANGE. FOR
WRITER'S
STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER USEITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE -PREPARE NEW (SP. fOP. ETC.) PER ATTACHED DRAFT"
Add item and action to list immediately above last item listed.
Deposition coupon device Close
The time/setpoint for close actions on the FPMS sampling system isolationvalves, the simulated LPIS pipe, and the deposition coupon deviceis as follows:
1462 217ZT4 K (2M•eF) on 1517peripheral fuel assembly 0 O2272cladding or• K (ew F)on center fuel assembly'....shroud outer wall
Change time/setpoint for core reflood to 6 + 0.5 s.
22a Replace wording at top of page 22a with the following (retain thelast paragraph on page 22a):
If the condition occurs where the system pressure is above themaximum allowable for Fl gnd F2 operation when the CFM claddingtemperature is 800 K (981 F) commence actions to lower the systempressure. These actions may be, but not limited to, opening thePORV or opening the intact loop break path. Terminate all ofthese actions before the CFM cladding temperature reaches 1050 K(1431UF). The experiment is to continue in the event that thesystem pressure cannot be lowered below the Fl and F2 operating press relimit. If the actions taken to lower the pressure are successful,terminate all actions before the system pressure decreases belowan indicated value of 1.2 MPa (160 psig).
(The replacement pages will not contain a page 22a).
ORIGINAI "
DOCUMENT REVISION REQUEST OPER. NO. _
(CONTINUATION SHEET) PAGE 3 OF _ 5FORM EG&G-,•1A i -7435
IROv. 11-T9g non unL
DOCUMENT NO. (IF APPUCABLE) OECD LOFT - -- --t periment Specification Documen IOCUMENT ISSUE DATE
OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN 'ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER
FOR EACH CHANCE. FORWRITER'S
I I STEP OR I INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER USEITEM PAGE PARA. ATTACHED COPY". FOR NI-W DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP, DOP. ETC.) PER AT TACHED DRAFT".
4 7 Ist Correct typo made in first DRR on primary coolant pump uncertainty.
Change 25 + El s to 25 + 2 s
/5 13 13 Change step 13 to read as follows:
13. Close the FPMS sampling system lines, the deposition coupon
2172 device, and the simulated LPIS line in the broken loop hotleg a per~pheral fuel assembly cladding temperature of
1462- K9 M F) or a thermal shroud outer wall temperature of1517 I K (M rF)
22726 13 11 Change (b) in step 11 to read as follows:
b. Cel assembly cladding temperature increases to
566 + 5 .K (560 +9 F). This criterion effective only after 300v7 6 3.2 Replace Be first s"ntence with the following:
The pretransient phase consists of a reactor shutdown interval(minimum of 2 days) followed by a power operation interval.
8 7 Add the following paragraph before the start of section 3.3:
The length of the shutdown interval shall be dependent on plantoperations tasks to be completed prior to the final power operation,and on project decision on Cs/I ratio.
, I
ORIGINAl
DOCUMENT REVISION REQUEST(CONTINUATION SHEET)
OPER. NO .
PAGE 4 OFFORM EG&G-o844A
nn•I D' I - 7A"Vrun,, flU - ______-,-___-"
DOCUMENT NO. OF APPLICABLE) DOCUMENT ISSUE DATEOECD LOFT roject Lxperiment Specification Document
OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1945PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER
FOR EACH CHANGE. FORWRITER'S
STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE 'REVISE PER USEITEM I PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND S ATE "PREPARE NEW (SP. DOP. ETC) PER ATTACHED DRAFT".
'1 9 21 Tble Replace the 3 conditions/decisions with the following 2 conditions/decisions:
Condition
Reactivity change
No reactivity change
Decision
Continue experiment sequence
Terminate experiment
10
11
12
-'13
22
2
8
23a
Thle
3.3
23a
Revise first abnormal condition and course of action to read asfollows:
System pressure below criterion If either the Fl or F2 isolatibut either Fl or F2 isolation valves fail to open w4"4h4.valves fail to open terminate the experiment
and commence recovery operatio0 with the ECCS. Do not open
prior to 840 K (1052°F) the LPIS line filter and do(indicated) not store the gamma densitomet
sources.
in
Is
1' Change temperature in item 1 to: 1533 t (2300 0 F)
Change centeg assembly shroud ogter wall temperature from1473 K (2192 F) to 1573 K (2272 F)
Change peripheral fuel rod cladding temperature from
1417 K (2092 F) to 1462 K (2172 F).
Add last two paragraphs:
The last situatign to be considered is if the CFM cladding hasexceeded 2100 K (3321uF) and if the temperature indications on elthethe CFM shroud outer wall or the peripheral fuel assembly cladding Lto fail before the temperature limits are reached. In the case ofthe peripheral fuel assembly cladding, a minimum of two valid temperindications are required to maintain the experiment termination crilin effect. (Two valid indications of the temperature limit willactivate experiment termination.) If the minimum number oftemperature indications is not maintained before the limit is reachEa clock function defining the time remaining to experiment terminatlwill be utilized. The specifics of the operation of the clockfunction are the responsibility of the Operations Branch.
in the hot region
regin
atureerion
d,on
ORIWNAIORIGINAl
DOCUMENT REVISION REQUEST OPER. NO.(CON INUATION SHEET) PAG __ OF 5
FORM E040-1844A(Rev. I.g NRR nE L-743
DOCUMENT NO. (IF APPLICABLE) OECD LOFT p r3WJ ý11ýT .Jeriment Specification Document DOCUMENT ISSUE DATE
OECD LOFT-T-3802 Rev. 1 Fission Product Experiment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOUENTIALLY IN IST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER
FOR EACH CHANGE.FOR
S WRITER'SI STEP OR I NSTRUCTIONS: REWRITE-PARAGRAPH(S1 OR FOR EXTENSIVE CHANGES ATTACH REVISErD COPY AND STATE REVISE PER USE
ITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP, DOP. ETC.) PER A TTACHED DRAFrT.
In the case of the CFM shroud outer wall, a minimum of twovalid temperature indications (of the total of 16 indications) arerequired to maintain this experiment termination criterion in effec:.(Two valid indications of the temperature limit will activate exper menttermination.) In the event that 15 of the 16 indications fail, thi,experiment termination criterion is deleted. Experimenttermination reverts to the criterion on the peripheral fuel assemblcladding. If one valid temperature indication exceeds thetemperature limit and supporting information on adjacentperipheral temperatures indicate a shroud heatup, the experimentwill also be terminated.
14 23 last Delete the last paragraph on page 23 (reference ESD Rev. 1)
15 22 1st Delete sentence, "If this break has not been closed ..... should beterminated."
16 22 2nd Revise to read as follows:
The FPMS Fl and F2 sampling systems arS to be opened at a CFMcladding temperature of 811 + 5 K (1000 + 9 F) if the system pressuehas decreased below 1.43 + 0:03 MPa (195-+ 5 psig). The pressurecriterion corresponds to the Fl and F2 design pressure limit of200 psig. The LPIS line filter is to be valved in and the gammadensitometer sources are to be stored following the opening of theFl and F2 sampling 8ystems at a CFM cladding temperature of840 + 5 K (1052 + 9 F). The following courses of action aredefined in the event that abnormal conditions occur:
A7 13 12 Revise to read:Open the FPMS Fl and F2 sampling systems at a center fuel assemblycladding temperature of 811 + 5 K (1000 + 9°F) and open only if thesystem pressure is less than 1.43 XAE)N + 0.03 MPa (195 + 5 psig).If the Fl and F2 sampling systems are opened, open the L-EPIS breakline filter and store the gamma densitometer sources at a center fu lassembly cladding temperature of 840 + 5 K (1052 + 9 F). Thermocou lesto be used in determining the temperature are listed in Table 8.
18 7 3.3 Change censer assembly shroud outer wall temperature from1473 (2192 F) to 1573 K (2272 0 F). Cha g ge the peripheral guel rodcladding temperature from 1417 K (2092 F) to 1462 K (2172 F).
nflIGNAI
1. INTRODUCTION
OECD LOFT Experiment LF-FP-2 is intended to provide information on the
release and transport of fission products and aerosols in a severe core
damage scenario wherein the fuel rod cladding would exceed a temperature of
2100 K for at least 3 minutes, resulting in rapid metal-water reaction.
Aerosols from the Ag-In-Cd ccntrol rods would provide the principal
environmental constituent for the transport of fission products. This
severe core damage scenario criginates from a risk dominant accident
sequence postulated for comrmercial PWR plants. Within this framework, the
nature of the phenomena governing fission product and aerosol release and
transport can be linked to pctential PWR system thermal hydraulics and core
thermal response leading to fuel failure and fission product transport
behavior.
PRA studies 1 revealed trat the interfacing systems LOCA, a
hypothetical event first postulated in the Reactor Safety Study2 and
labeled the V sequence, has a significant potential contribution to risk
from the operation of nuclear power plants. This risk dominant accident
sequence was selected as the mechanism under which severe core damage
phenomena would be studied in Experiment LP-FP-2. The specific interfacing
systems LOCA associated with the significant operational risk is a pipe
break In the low pressure injection system (LPIS), also called the residual
heat removal system. This system typically serves two functions: (a) it
provides emergency coolant injection for core recovery during intermediate
and large LOCAs, and (b) it provides for decay heat removal during normal
shutdown. The LPIS represents a potential path by which a LOCA outside
containment could occur, discharging primary system coolant external to the
containment. If core cooling cannot be maintained during such an event,
fission product release to the environment could occur through failure of
the auxiliary building.
Experiment LP-FP-2 will simulate the system thermal-hydraulic and core
uncovery conditions that are calculated to occur in commercial PWRs from
rupture of an LPIS pipe as a result of a V-sequence. The facility
I
DRR-L-7435Rev. I Chg. 3
configuration will include simulation of typical commercial PWR LPIS piping
in an experimental configuration in addition to the current LOFT LPIS. The
experimental configuration will be connected to the blowdown suppression
system.
Experiment LP-FP-2, the last of a series of eight experiments in the
OECD LOFT Program, may cause significant contamination of the Primary
Coolant System (PCS) and blowdown suppression system. The requirements
imposed on experiment LP-FP-2 from the standpoint of facility
decontamination and recovery are:
1. Experiment LP-FP-2 must be conducted with peripheral assembly
fuel rod cladding temperature limited to 1533 K (2300°F).
2. The structural integrity of the center fuel assembly must be
maintained to facilitate removal from the reactor vessel.
To meet the above project requirements, a center fuel assembly
specifically for the LP-FP-2 experiment has been designed with higher fuel
enrichment than the peripheral assemblies, and with prepressurized fuel
rods. The center assembly design also includes considerations of:
(a) providing a fuel rod-to-control rod (Ag-In-Cd) mass ratio similar to
that of a commercial PWR fuel assembly with control rods; and (b) providing
a well-defined geometry for the release and transport of fission products
and aerosols to facilitate interpretation of the results and code
assessment studies; and (c) thermally insulating the peripheral fuel
assemblies from the center fuel assembly.
2
DRR-L-7437Rev. 1 Chg. 2
3. EXPERIMENT DEs:.IPTION
Experieret LP-FP-2 consists of four discinct phases. These phases are
designated as (a) fuel preconditioning, (b) Dretransient, (c) transient,
and (d) postt-ansientl The four phases tocether represent a continuous
process and rave specific beginning and enc'ng definitions. Each phase is
described in the following sections.
3.1 Preconditioninc Phase
The purpose of the preconditioning phase is to subject the fuel rods
in the new center assembly [9.72 wt*, enriched, 2.41 MPa (350 psia)
prepressurization] to a burnup which, in co-bination with the burnup
corresponding to the EFPH requirement at the experiment initial conditions,
provides the required minimum burnup for the experiment. The minimum
burnup required for these fuel rods is 325 MYD/MTU. The minimum burnup
required in the preconditioning phase is 252 MWD/MTU. In terms of reactor
operation, the preconditioning burnup is equivalent, as an example, to
power operation at a maximum linear heat generation rate (MLHGR) of
52.2 kW/m (16 kW/ft) for 111.5 hours on the 9.72 wt% enriched fuel rods* A
core power level of 32.0 MW ± 0.5 is calculated to provide a MLHGR of
-16 kW/ft on the 9.72 wt% enriched fuel rods. Power profile data using
the traversing incore probe system (TIPS) must be obtained during the
preconditioning phase. The TIPS locations in the center assembly will be
capped during the reactor shutdown time in the pretransient phase.
The beginning and end of the preconditioning phase are defined as
follows:
Beginning the start of plant heatup prior to power operation to
establish fuel burnup
End termination of power operation after the required
burnup in this phase has been achieved.* Any combination of preconditioning and pretransient power operation that provides
the specified initial conditions (Section 4.2) is permissable. The specificcombination described is for example and for ease in attaining the initialconditions in the absence of factors affecting power operation .
5
DRR-L-7435Rev. 1 Chg. 3
The preconditioning phase does not require special procedures relative
to experiment specification and is to be conducted with established plant
operating procedures.
3.2 Pretransient Phase
The pretransient phase consists of a reactor shutdown interval
(minimum of 2 days) followed by a power operation interval. The final
plant preparations are to be completed during the reactor shutdown
interval. The power operation interval is to complete the required central
assembly fuel burnup, establish the required minimum decay heat level, and
establish the required initial conditions for conduct of the experiment.
The beginning and end of the pretransient phase are defined as follows:
Beginning termination of the power operation after the required
burnup is achieved in the preconditioning phase
End initiation of the transient by a reactor scram.
Specifically, the final plant preparations required by this
specification are:
1. Operational readiness of thermal hydraulic and nuclear
measurements (experimental and process) required by this
specification for this experiment and approval by the LOFT
Operations Branch to proceed into power operation and
subsequently into the transient phase of the experiment.
2. Operational readiness of the fission product measurement system
(FPMS) in accordance with FPMS operational documents.
The power operation prior to experiment initiation is to be an
effective 40 hours at a core power level of 26.5 ± 0.5 MW (21.2 EFPH*).
This core power is the required initial condition value. Power operation
*EFPH is defined as 50 MW-hr.
6
DRR-L-7435Rev. 1 Chg. 3
of an effective 40 hours will establish the required minimum decay heat
level of 675 kW at 200 s and will complete the burnup required on the
9.72 wt% enriched fuel rods. The plant should be operated at a steady
state power level of 26.5 ± 1.0 MW for at least 2 hrs prior to experiment
initiation.
The length of the shutdown interval shall be dependent on plant
operations tasks to be completed prior to the final power operation, and on
project decision on Cs/I ratio.
3.3 Transient Phase
The transient phase of the experiment will be initiated by a reactor
shutdown as specified in Section 4.3. Time zero corresponds to the action
taken to drop the center fuel module (CFM) controls rods. The 1.16 in.
inner diameter simulated break in the intact loop cold leg will be opened
at 20 ± 2 s. The primary coolant pumps will be tripped at 25 ± 2 s and
will undergo a normal coastdown. At 220 ± 5 s the simulated LPIS line
(also 1.16 in. diameter) will be opened. (The LPIS line filter shall be
bypassed to prevent plugging prior to the fission product release.) This
line is connected to the broken loop hot leg and to a blowdown suppression
tank (BST) inlet vent. The intact loop cold leg simulated break will be
closed at either a CFM cladding temperature of 566 ± 5 K (560 ± 91F) or a
system pressure of 1.2 ± 0.03 MPa (160 ± 5 psig). The fuel rod cladding is
calculated to begin heatup at approximately 700 s and to reach 2100 K at
approximately 1700 s. Fission product and aerosol release will occur from
fuel rod failure until the center assembly shroud outer wall reaches 1573 K
(22721F) or the peripheral fuel rod cladding reaches 1462 K (21721F), at
which time the fission product filter sampling systems and the upper plenum
deposition coupons will be closed in the primary coolant system coincident
with the closure of the simulated LPIS line. Within 6 ± 0.5 s of these
actions core reflood will commence with both accumulators.
The beginning and end of the transient phase of the experiment are
defined as follows:
7
DRR-L-7423
Rev. I Chg. 1
Beginning initiation of the transiert by a reactor scram
End initiation of the closure of the simulated LPIS line in
the broken loop hot leg.
3.4 Posttransient Phase
The posttransient phase consists of a tire interval of 12 hr for
measurement of (a) the redistribution of fission product inventory in the
gas and liquid volumes in the blowdown suppression tank (BST), and (b) the
leaching of fission products from the damaged fuel rods in the primary
coolant system (PCS). The beginning and end of the posttransient phase of
the experiment are defined as follows:
Beginning initiation of the closure of the simulated LPIS line in
the broken loop hot leg
End completion of the time interval specified for fission
product measurement).
Following closure initiation of the simulated LPIS line in the primary
coolant system, reflood operations will commence with initiation of both
accumulators. System refill will continue as required with the high
pressure injection system (HPIS). The governing requirements for plant
operation involving the PCS in this phase are:
1. Mass transfer from the PCS is to be minimized. The purification
system can be used in the decay heat removal mode (bypassing the
ion exchanger) for temperature control along with steam generator
feed and bleed.
2. Forced coolant circulation with the primary coolant pumps is
prohibited.
3. PCS temperature is to be reduced and maintained below 449 K
(350*F) as soon as possible after simulated LPIS line closure.
8
4. PCS pressure is to be maintained below E.96 MPa (1300 psig).
PORV venting is permitted but must be mirmized. The preferred
operation, if the PORV is needed, is to c::n it once for a long
enough time to bring the pressure down. This should eliminate
multiple openings which could effect the £ST fission product
measurements.
The degree of system refill with HPIS is not s:ecified by this
document but is to be utilized as required by plant operations to meet the
above four governing requirements.
Following closure of the simulated LPIS line, :he spray systems in the
BST vapor volume are not to be actuated. Mass transfer to and from the BST
is prohibited in the posttransient phase subject to the occurrence of PCS
conditions requiring operator or automatic plant actions. Measurements of
the fission product concentration in the gas and li:uid volumes, together
with temperature and pressure boundary conditions, will be used to
determine the fission product gas/liquid partition coefficients and fission
product redistribution and settling phenomena.
9
DRR-L-7435Rev. 1 Chg. 3
4. EXPERIMENT REQUIREMENTS
This section specifies experiment requirements for OECD LOFT
Experiment LP-FP-2. Included in this section are requirements for
experiment prerequisites, initial conditions, operational sequence, plant
configuration, and experimental measurements for the transient and
posttransient phases of the experiment.
4.1 Experiment Prerequisites
The following prerequisites are of programmatic concern and must be
completed prior to initiating Experiment LP-FP-2.
1. Perform a one point end-to-end check of the process instruments
identified In Table 1, within 90 days of the experiment. If a
problem is indicated, recalibrate, repair or replace the
instrument.
2. Establishment of event time procedures and parameter setpoints
for the items listed in Table 2.
3. Complete the pre-experiment calibration requirements for the FPMS
instruments and the pre-experiment calibration requirements
specified in DOP-87-O05, "DAVDS Experimental Measurements and
Test Procedure," and in DOP-87-008, "Pre LOCE Data Verification."
4. Complete testing of the center fuel assembly control rod
insertion function.
4.2 Initial Conditions
Experiment LP-FP-2 initial conditions will approximate commercial PWR
pressures and temperatures, and will be consistent with the LOFT plant
safety analysis. Specifically, three initial conditions are required for
the primary coolant system. These are intact loop hot leg temperature,
571 ± 1.1 K (569 ± 20 F), decay heat level between 675 kW and 695 kW at
10
DRR-L-7435Rev. 1 Chg. 3
b. Center fuel assembly cladding temperature increases to
566 ± 5K (560 ± 90 F). This criterion effective only after
300 s.
The cladding temperature following time zero will decrease and
correspond to saturation temperature until core uncovery occurs.
Then, cladding heatup will commence at some temperature below the
value specified in (b). Therefore, cladding temperature must be
increasing in order for criterion (b) to be valid.
12. Open the FPMS Fl and F2 sampling systems at a center fuel assembly
cladding temperature of 811 + 5 K (1000 + 9°F) and open only
if the system pressure is less than 1.43 + 0.03 MPa (195 + 5 psig).
If the Fl and F2 sampling systems are opened, open the LPIS
break line filter and store the gamma densitometer sources at a center
fuel assembly cladding temperature of 840 -I- 5 K (1052 + 90 F). Thermocouples
to be used in determinig the temperature are listed in Table 8.
13. Close the FPMS sampling system lines, the deposition coupon
device, and the simulated LPIS line in the broken loop hot leg at
a peripheral fuel assembly cladding temperature of 1462 K
( 2172'F) or a thermal shroud outer wall temperature of 1517 K
(2272 0 F).
14. Initiate core reflood 6 ± 0.5 s after initiation of the system
closures in Step 13.
15. Continue system refill with the IIPIS and maintain the PCS
temperature below 449 K (350'F) as soon as possible for the
remainder of a 12 hr minimum time after initiation of core
reflood. Maintain the PCS pressure below 8.96 MPa (1300 psig).
Mass transfer to and from the PCS, excluding system leakage and
replacement, is to be minimized. PCS energy control is to be
accomplished with steam generator feed/bleed operations and/or
with the purification system in the decay heat removal mode (ion
exchanger bypassed). PORV venting is permitted but is to be
minimized to the extent possible.
13
with the purification system in the decay heat removal mode (ion
exchanger bypassed). PORV venting is permitted but is to be
minimized to the extent possible.
16. Maintain isolation of the BST (no mass trans'er into or out of
the BST) for the same time interval as in Ite- 15 subject to
plant limitin; conditions which may require cerator action.
Within 10 minutes of PCS isolation purge the SST downcomer with
60 SCF of N2.I
17. Maintain PLSS data acquisition continuously ever the time
interval for Items 15 and 16.
4.4 System Configuration
The general syster and component configuration of the LOFT PWR for
Experiment LP-FP-2 is shown in Figure 1. Specific details are given in the
following sections on the reactor core, primary coolant system, secondary
coolant system, blowdowr system, and emergency core coclant system.
4.4.1 Reactor Core
Experiment LP-FP-2 will be conducted with a specially constructed
center fuel assembly.. The cross section of this fuel assembly is shown in
Figure 2. The fuel rods are 350 psia prepressurized ard 9.72 wt%
enriched. The outer two rows of fuel rods have been replaced with a
thermal shield of zircaloy with zirconium oxide ceramic internal
insulation. The purpose of the higher than normal enrichment and the
thermal shield is to provide at least 3 minutes at peak cladding
temperatures above 2100 K (3321*F) in the center fuel assembly before the
peripheral assembly fuel cladding reaches the transient termination
temperature. This time at temperature will provide a sufficiently large
fission product and aerosol release fraction.
The aerosol release will occur from (Ag-In-Cd) control rods which will
be inserted at reactor scram in the guide tubes shown in Figure 2.
14 I
DRR-L-7435Rev. 1 Chg. 3
inserted the experiment sequence must be stopped. On completion of the
corrective actions the experiment is to be resumed as approved by the LOFT
Operations Manager, DOE Site Program Manager and LOFT Program Division
Manager.
The decision matrix on center assembly control rod insertion is as
follows:
Condition Decision
Reactivity change Continue experiment sequence
No reactivity change Terminate experiment
Sequence Step 8 opens the break path in the intact loop cold leg.
This flow path is the primary blowdown path and is intended to be the path
for venting the major part of the primary coolant system fluid. High
quality steam flow only is desired for venting through the simulated LPIS
pipe in the broken loop hot leg. If the intact loop cold leg break cannot
be opened, the experiment sequence must be halted. On completion of the
corrective actions the experiment is to be resumed as approved by the LOFT
Operations Manager, DOE Site Program Manager and LOFT Program Division
Manager.
The simulated LPIS line is opened at 220 ± 5 s in Sequence Step 10.
This flow path is the designed fission product and aerosol vent path to the
BST. If this line cannot be opened within 50 s of the time specified,
close the intact loop cold leg break and commence recovery operations with
the ECCS. Resume the experiment as approved by the LOFT Operations
Manager, DOE Site Program Manager and LOFT Program Division Manager.
21
DRR-L-7435Rev. 1 Chg. 3
The intact loop cold leg break is to bp closed on either a cladding
temperature value or a system pressure value as specified in Sequence
Step 11. If this break is not closed then another flow path to the BST
will exist for fission product and aerosol venting. Two vent paths are not
provided for in the experiment plan. If this break cannot be closed before
the CFM cladding reaches 840 + 5 K (1052 + 9°F) then terminate the
experiment and commence recovery operations with the ECCS. If experiment
termination and plant recovery operations commence with the ECCS, return to
Sequence Step 1 or as approved by the LOF1 Operations Manager, DOE Site
Program Manager, and LOFT Program Division Manager after repairs are made.
The FPMS Fl and F2 sampling systems are to be opened at a CFM cladding
temperature of 811 + 5 K (1000+ 9gF) if the system pressure has decreased
below 1.43 + 0.03 MPa (195 + 5 psig). ihe pressure criterion corresponds
to the Fl and F2 design pressure limit of 200 psig. The LPIS line filter
is to be valved in and the gamma densitometer sources are to be stored
following the opening of the Fl and F2 sampling systems at a CFM cladding
temperature of 840 + 5 K (1052 + 9OF). The following courses of action
are defined in the event that abnormal conditions occur:
22
DRR-L-7435Rev. I Chg. 3
Abnormal Condition
System pressure below criterionbut either F1 or F2 isolationvalves fall to open
Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.
All criteria met and F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.
Course of. Action
If either the Fl or F2 isolation galvesfail to open prior to 840 K (1052 F)(indicated) terminate the experimentand commence recovery operations with theECCS. Do not open the LPIS line filterand do not store the gamma densitometersources.
Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.
The experiment is to proceed.Continue attempts to completethese actions.
If the condition occurs where the system pressure is above the maximum
allowable for F1 and F2 operation when the CFM cladding temperature is
800 K (981 0 F), commence actions to lower the system pressure. These
•, actions may be, but not limited to, opening the PORV or injecting full HPIS
for 10-20 s durations. Terminate all of these actions before the CFM
cladding temperature reaches 1050 K (14311F). The experiment is to
continue in the event that the system pressure cannot be lowered below the
F1 and F2 operating pressure limit. If the actions taken to lower the
pressure are successful, terminate all actions before the system pressure
decreases below an indicated value of 1.2 MPa (160 psig).
The LPIS break line filter is to be valved in when the peak cladding
temperature is 840 ± 5 K (1052 ± 90 F). As the filter loads up the
differential pressure may increase. At a filter differential pressure of
150 psi, the filter should be bypassed to maintain the ability to measure a
sample in the aerosol collection lines. If this occurs, the time shall be
recorded.
23
DRR-L-7435Rev. 1 Chg. 3
The FPMS sampling lines are closed, the deposition coupon devices are
closed, and the simulated LPIS line is closed in Sequence Step 13. If any
of these actions cannot be completed the experiment sequence is to
continue. Core reflood must commence in time to prevent temperatures from
exceeding 1588 K (2400 0 F) on the center fuel assembly shroud outer wall
and 1533 K (2300 0 F) on the peripheral assembly fuel rod cladding.
Continue attempts to complete Sequence Step 13 but continue through
Sequence Steps 14-17.
The possibility exists that the time interval between the occurrence
of 2100 K (3321 0 F) cladding temperature in the center assembly and the
peripheral assembly cladding temperature or the shroud outer wall
temperature limit may be excessive relative to core damage and source
release limitations. Therefore, a maximum time of 4.0 minutes is
specified, after the 2100 K (3321 0 F) temperature is reached, for
continuing the experiment sequence with Steps 13 and 14.
The last situation to be considered is if the CFM cladding has
exceeded 2100 K (3321 0 F) and if the temperature indications on either
the; CFM shroud outer wall or the peripheral fuel assembly cladding begin
to fail before the temperature limits are reached. In the case of
the peripheral fuel assembly cladding, a minimum of two valid temperature
indications in the hot region are required to maintain the experiment termination
criterion in effect. (Two valid indications ot the temperature limit will
activate experiment termination.) If the minimum number of temperature
indications is not maintained before the limit is reached, a clock function
defining the time remaining to experiment termination will be utilized. The
specifics of the operation of the clock function are the responsibility of
the Operations Branch.
23A
DRR-L-7435Rev. I Chg. 3
In the case of the CFM shroud outer wall, a minimum of two valid
temperature indications (of the total of 16 indications) are required to
maintain this experiment termination criterion in effect. (Two valid indications
of the temperature limit will activate experiment termination.) In the event
that 15 of the 16 indications fail, this experiment termination criterion is
deleted. Experiment termination reverts to the criterion on the peripheral
fuel assembly cladding. If one valid temperature indication exceeds the
temperature limit and supporting information on adjacent peri-pheral temperatures
indicate a shroud heatup, the experiment will also be terminated.
23B
5. PLANNING ANALYSIS
This section discusses scaling considerations for the LOFT system for
Experiment LP-FP-2 compared with commercial PWRs, and pressents results of
computer code calculations on which this specification do:.-ent is based.
5.1 Scaling Consideration
Scaling considerations for Experiment LP-FP-2 center c the V-sequence
phenomena that are calculated to occur in commercial PWRs. The V-sequence
scenarios include little or no operation of the PWR ECCS. The LOFT
V-sequence experiment does not include ECCS operation. Therefore, the
specifications and/or recommendations for the operation of the LOFT ECCS
contained in Section 4 are based on plant recovery plannin: specific to
LOFT. The required scaling considerations for Experiment -P-FP-2 include
only the sizing of the break path flow areas (for both break paths for the
experiment) and also sizing the length of the simulated LF:S flow path.
Break area scaling provides representative thermal-hydraulics and, in
particular, similar coolant velocities that are required f:r the transport
of fission products and aerosols. Simulated LPIS pipe ler:th scaling is
necessary to provide similar residence times for transport and retention
phenomena in the LPIS piping.
PWR LPIS pipe sizes range from 6 in. Sch 160 to 10 in. Sch 160. The
inside diameter range is 0.13-0.22 m (5.19-8.5 in.). The s:aling basis
used to size the break area in LOFT is break flow area/sys:em volume.
Using the following values for system volume:
V (PWR) = 355.10 m3 (12,540 ft 3 ) (Reference 6)
V (LOFT) = 7.36 m3 (260 ft 3) (Reference 7)
gives the LPIS pipe diameter range in LOFT of 0.019 - 0.03: m
(0.747 - 1.224 in.). The pipe size selected for the LPIS ripe simulation
in LOFT is 1-1/4 in. nominal Sch 160 which has an inside diameter of
0.0295 m (1.16 in.). The scaled flow area equivalent in a commercial PWR
is 10%1 larger than an 8 in. Sch 160 pipe and 13% less than a 10 in. Sch 160
24 I
DRR-L-7435Rev. 1 Chg. 3
TABLE 2. EXPERIMENT LP-FP-2 EVENT TIMES AND PARAMETER SETPOINTS
Item Action Time/Setpoint
Reactor shutdown CFM CR t = 0dropPeripheral 5 sCR scram
Intact loop cold leg simulated break Open 20 ± 2 s
Primary coolant pumps Tripped 25 ± 2 s
Broken loop hot leg simulated LPIS Open 220 ± 5 spipe
Intact loop cold leg simulated break Close 566 ± 5 K (560 ± 90 F)1.2 ± 0.03 MPa(160 ± 5 psig)
FPMS sampling system isolation Open 811 K ± 5 Kvalves (I1000OF ± 90 F)
LPIS Break Line Filter Open 840 K ± 5 K(10520 F ± 90 F)
Gamma densitometer sources Store 840 K ± 5 K(10520 F ± 90 F)
FPMS PCS sampling system isolation Close 1462 K (2172 0 F) onvalves peripheral. fuel assembly
cladding or 1517 K(2272°F) on CFM shroud
Broken loop hot leg simulated LPIS Close outer wall or other experimentpipe termination events.
Deposition coupon device Close
Core reflood Initiation 6 ± 0.5 s from simulatedLPIS line closureinitiation
71
TABLE 3. INITIAL CONDITIONS FOR EXPERIMENT LP-FP-2
Parameter
Primary Coolant System
Core AT
Hot leg pressure (MPa). (psia)
Hot leg temperature (K)(OF)
Primary coolant flow (Ibm/hr)(kg/s)
Boron concentration (ppm)
Power level (MW)
Decay heat at 200 s minimummaximum
Maximum linear heat
generation ratec (kW/m)(kW/ft)
Control rod position(above full-in position) (m)
(in.)
Steam Generator Secondary Side
Water level
Pressurizer
Liquid level (m)(in.)
Specified Value a
As requiredb
14.95 ± 0.12168 ± 15.0
571 ± 1.1569 ± 2
3.8 ± 0.15 x 106
479 ± 19
As requiredb
26.5 _ 0.5
675 kW695 kW
-40-12.26
1.37 _ 0.0154.0 ± 2.0
As requiredb
1.12± 0.144 ± 4
I
72 I
(® OPER. NO.
rOflM EG&G-I844 DOCUMENT REVISION REOUEST
(D REQUESTSA(® OTR NO.V. T. BErta July 2, 1985 L-7448
05 DOCUMENT NO. (IF APPCABLE) OECD LOFT PrMWRIT EJerimen t Specification DocumLPIfMENT ISSUE DATE
OECD LOFT-T-3802 Rev. 1 Fisslon Product Experiment LP-FP-2 May 1985
@ CHECK APPLICABLE BLANK ______® 7 GER.PR0 I DATE
PERMANENT CHANGE X TEMPORARY CHANGE - BULLETIN 4%/~(®) PRINT OR TYPE PROPOSED CHANGE - NUMBE EACH CHANGE SEQUENTIALLY IN IST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER FOR
FOR EACH CHANGE. WRTER'S USE
STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE "REVISE PER
ITEM PAGE PAR. ATTACHED COPY. FOR NEW DOCUMENT. ATTACH ROUGHDRAFT ANDSTATE"PREPARE NEW (SP..DOP. ETC.) PERATTACHEDORAFT".
1 23A 2nd Change 4.0 minutes to 7.0 minutes in the last sentence.
2 79 Tbl E Change the wording at the bottom of the list of thermocouples to th(following:
Any valid thermocouple may be used. All of the listedthermocouples are to be monitored. If any of the listed'thermocouples fail prior to the experiment, they may be replacedwith similar thermocouples. If replacement thermocouples arenot available, the failed thermocouples may be deleted fromthe monitors.
USE CONTINUATION SHEET AS REQUIRED
NEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATEIEVENT
(.JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. ABOVE): (9 OTHER DOCUMENTATION AFFECTED:
1. Increase detectability of fission products at all IDOC.N. DARNO. DATECOMPLETED
measurement locations. OECD LOFT-1I-I1-510 675785
2. Allows for flexibility in the event of instrumentfailures.
"1__ ORIGINATING ORR NO:
@ REVIEW
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DRR-L-7435Rev. I Chg. 3
Abnormal Condition
System pressure below criterionbut either F1 or F2 isolationvalves fail to open
Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.
All criteria met and F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.
Course of Action
If either the Fl'.or F2 isolation galvesfail to open prior to 840 K (1052 F)(indicated) terminate the experimentand coninence recovery operations with theECCS. Do not open the LPIS line filterand do not store the gamma densitometersources.
Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.
The experiment is to proceed.Continue attempts to completethese actions.
If the condition occurs where the system pressure is above the maximum
allowable for F1 and F2 operation when the CFM cladding temperature is
800 K (981'F), commence actions to lower the system pressure. These
actions may be, but not limited to, opening the PORV or injecting full HPIS
for 10-20 s durations. Terminate all of these actions before the CFM
cladding temperature reaches 1050 K (1431*F). The experiment is to
continue in the event that the system pressure cannot be lowered below the
F1 and F2 operating pressure limit. If the actions taken to lower the
pressure are successful, terminate all actions before the system pressure
decreases below an indicated value of 1.2 MPa (160 psig).
The LPIS break line filter is to bevalved in when the peak cladding
temperature is 840 ± 5 K (1052 ± 90 F). As the filter loads up the
differential pressure may increase. At a filter differential pressure of
150 psi, the filter should be bypassed to maintain the ability to measure a
sample in the aerosol collection lines. If this occurs, the time shall be
recorded.
23
DRR-L-7448Rev. I Chg. 4
The FPMS sampling lines are closed, the deposition coupon devices are
closed, and the simulated LPIS line is closed in Sequence Step 13. If any
of these actions cannot be completed the experiment sequence is to
continue. Core reflood must commence in time to prevent temperatures from
exceeding 1588 K (2400 0 F) on the center fuel assembly shroud outer wall
and 1533 K (2300 0 F) on the peripheral assembly fuel roa cladding.
Continue attempts to complete SeQuence Step 13 but continue through
Sequence Steps 14-17.
The possibility exists that the time interval between the occurrence
of 2100 K (3321 0 F) cladding temperature in the center assembly and the
peripheral assembly cladding temperature or the shroud outer wall
temperature limit may be excessive relative to core damage and source
release limitations. Therefore, a maximum time of 7.0 minutes is
specified, after the 2100 K (3321 0 F) temperature is reached, for
continuing the experiment sequence with Steps 13 and 14.
The last situation to be considered is if the CFM claddinc has
exceeded 2100 K (3321OF) and if the temperature indications on either
the CFM shroud outer wall or the peripheral fuel assembly cladding begin
to fail before the temperature limits are reached. In the case of
the peripheral fuel assembly cladding, a minimum of two valid temperature
indications in the hot region are required to maintain the experiment termination
criterion in effect. (Two valid indications ot the temperature limit will
activate experiment termination.) If the minimum number of temperature
indications is not maintained before the limit is reached, a clock function
defining the time remaining to experiment termination will be utilized. The
specifics of the operation of the clock function are the responsibility of
the Operations Branch.
23A
DRR-L-7448Rev. I Chg. 4
TABLE 8. LP-FP-2 CENTER MODULE THERMOCOUPLES FOR FPMS AND LPIS LINEFILTER - OPENING
At a peak cladding temperature of 840 K (10521F) open the valves to the F-1
and F-2 sample lines and to the F-3 filter in the simulated LPIS line.
The thermocouples to be used in this decision are present below:
TE-5C7-42TE-5C12-27TE-5D9-27TE-5DI3-42TE-5F9-27TE-5G4-27TE-5G12-27TE-5G13-27TE-5H6-27TE-5H10-27TE-5103-27TE-5104-42TE-5112-42TE-5J7-27TE-5J9-42TE-5L7-10TE-5L7-27TE-5L9-42TE-5M7-27TE-5M9-42
Any valid thermocouple may be used. All of the listed thermocouples areto be monitored. If any of the listed thermocouples fail prior to theexperiment, they may be replaced with similar thermocouples. If replace-ment thermocouples are not available, the failed thermocouples may bedeleted from the monitors.
79
ropm EG&G.1-44Jltev. •0%6 DOCUMENT REVISION REQUEST
( OPER. NO. ____
PAGE L OF 3,(® REQUESTER (I ® RR DATE (•) DOR ND.V. T. Berta July 5, 1985 L-7450
®. DOCUMENT NO. (IF APPLICABLE) DOCUMENT TITLE DOCUMENT ISSUE pATE
OECD LOFT-T-3802 Rev. 1 OECD LOFT Project Experiment SpecificationDocument Fission Product.ExpjerimientjLP-FP-2 , May 1985
@ CHECK APPLICABLE BLANK . M APP A LV DATE
PERMANENT CHANGE X TEMPORARY CHANGE -___ BULLETIN____PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE SEQUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER () FOR
FOR EACH CHANGE. WRTTER'S USE
STEP OR INSTRUCTIONS: REWRITE PARAGRAPHIS) OR FOR EXTENSIVE CHANGES ATTACH REVISED COPY AND STATE REVISE PERITEM PAGE PARA. ATTACHED COPY". FOR NEW DOCUMENT, ATTACH ROUGH DRAFT AND STATE "PREPARE NEW (SP. DOP, ETC ) PER ATTACHED DRAFT".
1 7 Add the following sentence to the wording at the top of the page:
Refer to Section 4.2 for further initial conditionspecifications.
2 7 3.3 Insert the following in place of the 2nd, 3rd, and 4th sentences:
Time zero corresponds to the reactor scram action. Theprimary coolant pumps are tripped at 8 ± 2 s. The centerfuel module (CFM) control rods are dropped whehithe'loop flow hisThe intact loop cold leg simulated break is opened Idecreased toimmediately after verification that the CFM control 11.5.x 106 ibm/hr.rods are in. This action is estimated to occur atapproximately 20-25 s.
Delete the following in the next sentence:
(also 1.16 in. diameter)USE CONTINUATION SHEET AS REQUIRED
NEXT ANTICIPATED NEED FOR DOCUMENT WITH THIS REVISION INCORPORATED: DATE/EVENT 7//5/L85
(:0) JUSTIFICATION: (REASON FOR CHANGE - NUMBER TO CORRESPOND TO ITEM NO. ABOVEI: (5 OTHER DOCUMENTATION AFFECTED:
All items 'reflect sequence Step 7, 8, 9 revisions to DOC. NO. DRR NO. DATE COMPLETED
account for CFM control rod drop equipment character- 6EEDiFi:4-ULJ-51L 6/5/85istics. New sequence steps are 7, 8, 9A, and 9B.
-_ 2 ORIGINATING DRR NO:.
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(CONTINUATION SHEET) PAGE OF
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DOCUMENT NO. (IF APPLICABLE) DOCMEDOCUMENT ISSUE DATE
OECD LOFT-T-3802 Rev. I OECD LOFT Project lxperiment Specification DocumentFission Product Experiment LP-FP-2 May 1985
PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEGUENTIALLY IN 1ST COLUMN AND RECORD PAGE AND SFP OR PARAGRAPH NUMBERFOR EACH CHANGE. FOR
. . . ...... ... _WRITER'S
ITEM PAGE STEP OR INSTRUCTIONS: REWRITE PARAGRAPH(S) OR FOR EXTENSIVE CHANGfe ATTACH REVISED COPY AND STATE 'REVISE PER USEPARA. ATTACHED COPY". FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND SIATE PREPARE NEW (SP.DOP. ETC.I PER A1IACHEDDRAFT,.
3 11 1 Revise the last sentence and add one sentence to the last paragraph
of Section 4.2 as follows:
Power operation in the pretransient phase will establish theinitial conditions specified in Table 3 and the requiredminimum burnup on the center assembly fuel rods prior totransient initiation. In the event that reactor shutdownintervals occur after the initial conditions and fuelburnup are reached, subsequent power operation will bebased on establishing only the initial conditionsspecified in Table 3.
4 12 7,8,9 Replace Sequence Steps 7,8, and 9 with the following Steps 7, 8,9A and 9B:
7. Scram the reactor. Reactor scram does not includedropping the CFM control rods (refer to Step 9A).Reactor scram establishes time zero.
8. Trip the primary coolant pumps at 8 t 2 s. The pumpsare to undergo a normal coastdown.
when the loop flow has decreased lo
9A. Drop the CFM control rodsAIwI . Verify that 1.5 x 106 bm/hy
the CFM control rods are in before proceeding toStep 9B.
9B. Open the simulated break in the intact loop cold legimmediately after verification of CFM control rods in.
This action is estimated to occur at approximately20-25 s.
5 20 last Revise 2nd paragraph of Section 4.7 to read as follows:
The CFM control rods are to be inserted in Sequence Step 9A.
Verification of the insertion of these control rods is to beobtained from proximity switch indications before proceedingto Step 9B. If insertion of the CFM control rods cannot beverified the experiment may be either (1) placed on hold for
further decisions or actions to drop the CFM control rods,or (2) terminated. If a hold is placed on the experiment,the experiment sequence must be resumed before the decayheat loss jeopardizes the experiment. If too much decay heat
is lost during the hold, the experiment must be terminated.The minimum acceptable decay heat is 641 kW at time zeroplus the hold interval plus 200 s. After termination, oncompletion of the corrective actions the experiment is tobe resumed as approved by the LOFT Operations Manager, DOESite Program Manager and LOFT Program Division Manager.
ORIGINAt
DOCUMENT REVISION REQUEST(CONTINUATION SHEET)
OPER NO. -
PAGE 3 OF .
FORM EG&G-1844AfRAv. It-.7 I n• f' I -74!fl•
L-7450V- ... - -
DOCUMENT NO. (IF APPUCABLE) Do1UMETJITLE DOCUMENT ISSUE DATEOECD LOFT Project Experiment Specification Document
)ECD LOFT-T-3802 Rev. 1 Fission Product Exoeriment LP-FP-2 May 1985PRINT OR TYPE PROPOSED CHANGE - NUMBER EACH CHANGE ITEM SEOIJENTIALLY IN 1ST COLUMN AND RECORD PAGE AND STEP OR PARAGRAPH NUMBER
FOR EACH CHANGE. FOR
WRITERSSTEP OR INSTRUCTIONS: REWRITE PARAGRAPHIS) OR FOR EXTENSIVE CHANGF-, ATTACH REVISED COPY AND STATE 'REVISE PER USE
ITEM PAGE PARA. ATTACHED COPY" FOR NEW DOCUMENT. ATTACH ROUGH DRAFT AND S I A IE 'PREPARE NEW (SP, DOP, ETC.) PER A TTACHED DRAFT".
6
7
8
21
21
71
1st
2nd
Tabl e2
Delete the paragraph on the decision matrix.
Change Sequence Step 8 to Sequence Step 9B in the first sentence.
Replace the first three items with the following four items:
Item Action Time/Setpoint
Reactor scram Peripheral 0CR scram
Primary coolant pumps Tripped t 2 s
CFM control rods Dropped ' t t l oop flow1.5 x 106 lbm/hr
Intact loop cold leg Open Verification of
0 f
simulated break CFM control rods ir
UNIGINA1
DRR-L-7450Rev. 1 Chg. 5
of an effective 40 hours will establish the required minimum decay heat
level of 675 kW at 200 s and will complete the burnup required on the
9.72 wt% enriched fuel rods. The plant should be operated at a steady
state power level of 26.5 ± 1.0 MW for at least 2 hrs prior to experiment
initiation. Refer to Section 4.2 for further initial condition specifications.
The length of the shutdown interval shall be dependent on plant
operations tasks to be completed prior to the final power operation, and on
project decision on Cs/I ratio.
3.3 Transient Phase
The transient phase of the experiment will be initiated by a reactor
shutdown as specified in Section 4.3. Time zero corresponds to the reactor
scram action. The primary coolant pumps are tripped at 8 t 2s. The
center fuel module (CFM) control rods are dropped when the loop
flow has decreased to 1.5 x 106 Ibm/hr. The intact loop
cold leg timulated break is opened immediately after veri-
fication that the CFM control rods are in. This action is estimated to
occur at approximately 20-25 s. At 220 ± 5 s the simulated LPIS line
will be opened. (The LPIS line filter shall be bypassed to prevent
plugging prior to the fission product release.) This line is connected
to the broken loop hot leg and to a blowdown suppression tank (BST)
inlet vent. The intact loop cold leg simulated break will be closed
at either a CFM cladding temperature of 566 ± 5 K (560 ± 90 F) or a
system pressure of 1.2 ± 0.03 MPa (160 ± 5 psig). The fuel rod cladding is
calculated to begin heatup at approximately 700 s and to reach 2100 K at
approximately 1700 s. Fission product and aerosol release will occur from
fuel rod failure until the center assembly shroud outer wall reaches 1573 K
(2272*F) or the peripheral fuel rod cladding reaches 1462K (2172'F), at
which time the fission product filter sampling systems and the upper plenum
deposition coupons will be closed in the primary coolant system coincident
with the closure of the simulated LPIS line. Within 6 ± 0.5 s of these
actions core reflood will commence with both accumulators.
The beginning and end of the transient phase of the experiment are
defined as follows:
7
DRR-L-7423
Rev. 1 Chg.
Beginning initiation of the transient by a reactor scram
End initiation of the closure of the simulated LPIS line in
the broken loop hot leg.
3.4 Posttransient Phase
The posttransient phase consists of a time interval of 12 hr for
measurement of (a) the redistribution of fission product inventory in the
gas and liquid volumes in the blowdown suppression tank (BST), and (b) the
leaching of fission products from the damaged fuel rods in the primary
coolant system (PCS). The beginning and end of the posttransient phase of
the experiment are defined as follows:
Beginning initiation of the closure of the simulated LPIS line in
the broken loop hot leg
End completion of the time interval specified for fission
product measurements.
Following closure initiation of the simulated LPIS line in the primary
coolant system, reflood operations will commence with initiation of both
accumulators. System refill will continue as required with the high
pressure injection system (HPIS). The governing requirements for plant
operation involving the PCS in this phase are:
1. Mass transfer from the PCS is to be minimized. The purification
system can be used in the decay heat removal mode (bypassing the
ion exchanger) for temperature control along with steam generator
feed and bleed.
2. Forced coolant circulation with the primary coolant pumps is
prohibited.
3. PCS temperature is to be reduced and maintained below 449 K
(350 0 F) as soon as possible after simulated LPIS line closure.
B
DRR L-7450Rev. 1 Chg. 5
200 s, and maximum linear heat generation rate, 40.0 kW/m (12.26 kW/ft), in
the high power fuel rods. The latter initial condition (not measurable)
can be met, according to core physics calculations, by a reactor power
level of 26.5 ± 0.5 MW. Table 3 contains the initial condition
specifications for the experiment. Systems or controllable parameters not
identified in Table 3 shall be operated as specified in the Plant Operating
Manual.
Prior to the preconditioning phase of the experiment, the necessary
hardware configurations will be established and the required prerequisites
identified in Section 4.1 will be completed. The high power fuel rod
burnup will be achieved in the preconditioning phase and pretransient
phase. The fission product measurement systems operational readiness will
be completed during the reactor shutdown interval in the pretransient
phase. Also, the TIPS locations in the center assembly will be capped.
Power operation in the pretransient phase will establish the initial
conditions specified in Table 3 and the required minimum burnup on the
center assembly fuel rods prior to transient initiation. In the event
that reactor shutdown intervals occur after the initial conditions and
fuel burnup are reached, subsequent power operation will be based on
establishing only the initial conditions specified in Table 3.
4.3 Experiment Sequence
The following items will be completed prior to experiment initiation
but have no requirement to be done in the order listed:
1. Complete prerequisites established in Section 4.1.
2. Complete acquisition of pre-transient primary coolant system and
blowdown suppression tank liquid samples. These pre-transient
liquid samples will be analyzed by Chemical Sciences Branch and
LOFT Operations to establish a baseline data set.
3. Establish the initial conditions specified in Table 3.
The following actions should be performed in sequence:
11
DRR-L-7450Rev. 1 Chg. 5
NOTE: Refer to Section 4.7 for abnormal event procedures, as
required, for the following actions.
4. Start the gross gamma detector systems not later than 300 s
before experiment initiation. Start the data acquisition and
visual display system (DAVDS) and the FPMS data system not later
than 60 s before experiment initiation.
5. Secure the pressurizer cycling and backup heaters.
6. Initiate primary coolant pump injection flow. Terminate primary
coolant pump injection flow at to +60 s ± 5 s.
7. Scram the reactor. Reactor scram does not include dropping the
CFM control rods (refer to Step 9A). Reactor scram establishes
time zero.
8. Trip the primary coolant pumps at 8 t 2s. The pumps are toundergo a normal coastdown.
9A. Drop the CFM control rods when the loop flow has decreased to1.5 x 106 lbm/hr. Verify that the CFM control rods are inbefore proceeding to Step 9B.
9B. Open the simulated break in the intact loop cold leg imnediately
after verification of CFM control rods in. This action is estimated
to occur at approximately 20-25 s.
10. Open the simulated LPIS line in the broken loop hot leg at
220 ± 5 s. The filter in this line shall be bypassed to prevent
plugging until the core reaches a temperature of 840 ± 5K
(1052 ± 9*F). Thermocouples to be used in determining valve
opening are listed in Table 8.
11. Close the simulated break in the intact loop cold leg when either
of the following occurs:
a. System pressure decreases to 1.2 ± 0.03 MPa (160 ± 5 psig).
1?
4.5.3 Simulated LPIS Pipe
Thermal-hydraulic measurements in the simulated LPIS pipe will consist
of steam flow, steam temperature, and wall temperature.
4.5.4 FPMS
The identification and location of fission product measurements are
shown in Figure 4. These measurements are described in the FPMS Functional
and Operational Requirements (F&OR) document. 4
4.5.5 Postirradiation Examination
An integral part of the measurements which are necessary to meet the
experiment objectives are the postirradiation examination (PIE)
measurements. A summary of items specified for postirradiation examination
is contained in Table 5. Examination of these items will be in accordance
with the postirradiation plan for LP-FP-2. 5
4.5.6 Critical Measurements
Sets of critical measurements, required during the transient and
posttransient phases of the experiment, have been identified and are listed
in Tables 6 and 7, respectively. The transient phase of the experiment
should not be initiated without these measurements since the experiment
objectives may be Jeopardized. Appendix A lists by instrument identifier
all critical measurements which are considered necessary for the successful
conduct of the experiment. The measurement uncertainties will be equal to
or less than those specified in the document "LOFT Experimental Measurement
Uncertainty Analysis,* NUREG/CR-0169.
A complete list of measurements required for Experiment LP-FP-2 is
provided on the Data Acquisition Requirements list to be published prior to
the experiment.
.19 I
DRR-L-7450
Rev. 1 Chg. 5
The digital data acquisition and processing system, and analog and
) digital data acquisition recording is required to begin no later than i min
before initiation of the transient phase of the experiment. Continuous
PLSS recording is required through the posttransient phase of the
experiment.
Measurements identified on the Data Acquisition Requirements List that
fail prior to experiment initiation should be repaired or replaced if
possible. If a failed instrument(s) cannot be repaired or replaced, the
Joint Experiment Group shall determine the course of action.
Process instruments requiring calibration prior to Experiment LP-FP-2
are listed in Table 1.
4.6 Experiment Termination
Experiment LP-FP-2 will be terminated at the end of the posttransient
phase of the experiment. The posttransient phase ends with the completion
of the. time interval required for monitoring the redistribution of fission
products in the vapor and liquid volumes in the blowdown suppression
system. The time interval is specified in Section 4.3 to be 12 hr minimum
after closure of the simulated LPIS line.
4.7 Abnormal Experiment Sequence
If instrumentation, hardware components, or operating systems fail
prior to or during any of the four phases of the experiment, every effort
should be made to substitute, repair, or provide alternate actions to
safely continue the experiment and to meet the programmatic objectives.
The CFM control rods are to be inserted in Sequence Step 9A.
Verification of the insertion of these control rods is to be obtained
from proximity switch indications before proceeding to Step 9B. If
insertion of the CFM control rods cannot be verified the experiment may
be either (1) placed on hold for further decisions or actions to drop
20
DRR-L-7450Rev. 1 Chg. 5
the CFM control rods, or (2) terminated. If a hold is placed on the
experiment, the experiment sequence must be resume(] before the decay
heat loss jeopardizes the experiment. If too much decay heat is lost
during the hold, the experiment must be terminated. The minimum
acceptable decay heat is 641 kW at time zero plus the hold interval
plus 200 s. After termination, on completion of the corrective actions
the experiment is to be resumed as approved by the LOFT Operations
Manager, DOE Site Program Manager and LOFT Program Division Manager.
Sequence Step 9B opens the break path in the intact loop cold leg.
This flow path is the primary blowdown path and is intended to be the path
for venting the major part of the primary coolant system fluid. High
quality steam flow only is desired for venting through the simulated LPIS
pipe in the broken loop hot leg. If the intact loop cold leg break cannot
be opened, the experiment sequence must be halted. On completion of the
corrective actions the experiment is to be resumed as approved by the LOFT
Operations Manager, DOE Site Program Manager and LOFT Program Division
Manager.
The simulated LPIS line is opened at 220 ± 5 s in Sequence Step 10.
This flow path is the designed fission product and aerosol vent path to the
BST. If this line cannot be opened within 50 s of the time specified,
close the intact loop cold leg break and commence recovery operations with
the ECCS. Resume the experiment as approved by the LOFT Operations
Manager, DOE Site Program Manager and LOFT Program Division Manager.
21
DRR-L-7435Rev. 1 Chg. 3
The intact loop cold leg break is to he closed on either a cladding
temperature value or a system pressure value as specified in Sequence
Step 11. If this break is not closed then another flow path to the BSTwill exist for fission product and aerosol venting. Two vent paths are not
provided for in the experiment plan. If thi- break cannot be closed before
the CFM cladding reaches 840 + 5 K (1052 1 90F) then terminate the
experiment and commence recovery operations with the ECCS. If experiment
termination and plant recovery operations commence with the ECCS, return to
Sequence Step 1 or as approved by the LOFT Operations Manager, DOE Site
Program Manager, and LOFT Program Division Manager after repairs are made.
The FPMS Fl and F2 sampling systems are to be opened at a CFM cladding
temperature of 811 + 5 K (1000+ 90 F) if the system pressure has decreased
below 1.43 + 0.03 MPa (195 + 5 psig). The pressure criterion corresponds
to the Fl and F2 design pressure limit of 200 psio. The LPIS line filter
is to be valved in and the gamma densitometer sources are to be stored
following the opening of the Fl and F2 sampling systems at a CFM cladding
temperature of 840 + 5 K (1052 + 9OF). The following courses of action
are defined in the event that abnormal conditions occur:
22
DRR-L-7435Rev. 1 Chg. 3
Abnormal Condition
System pressure below criterionbut either F1 or F2 isolationvalves fail to open
Cladding temperature criterionreached but system pressurehigher than permitted for openingF1 and F2.
All criteria met ana F1 and F2open but LPIS line filter cannotbe opened and/or gammadensitometer sources cannot bestored.
Course of Action
If either the Fl or F2 isolation nalvesfail to open prior to 840 K (1052 F)(indicated) terminate the experimentand comTmence recovery operations with theECCS. Do not open the LPIS line filterand do not store the gamma densitometersources.
Hold opening of F1 and F2 untilpressure criterion is satisfied.Proceed with opening the LPIS linefilter and store the gammadensitometer sources.
The experiment is to proceed.Continue attempts to completethese actions.
If the condition occurs where the system pressure is above the maximum
allowable for F1 and F2 operation when the CFM cladding temperature is
800 K (981°F), commence actions to lower the system pressure. These
actions may be, but not limited to, opening the PORV or opening the
intact loop break path. Terminate all of these actions before the CFM
cladding temperature reaches 1050 K (1431'F). The experiment is to
continue in the event that the system pressure cannot be lowered below the
F1 and F2 operating pressure limit. If the actions taken to lower the
pressure are successful, terminate all actions before the system pressure
decreases below an indicated value of 1.2 MPa (160 psig).
The LPIS break line filter is to be valved in when the peak cladding
temperature is 840 ± 5 K (1052 ± 90 F). As the filter loads up the
differential pressure may increase. At a filter differential pressure of
150 psi, the filter should be bypassed to maintain the ability to measure a
sample in the aerosol collection lines. If this occurs, the time shall be
recorded.
23
DRR-L-7448Rev. 1 Chg. 4
The FPMS sampling lines are closed, the denosition counon devices are
closed, and the simulated LPIS line is Closed in Seauence SteD 13. U any
of these actions cannot be comoleted the experiment seauence is to
continue. Core refiood must commence in time to orevent tomperatures frcm.
exceeding 1583 K (2400 0 F) on tne center fuel assetflo sn, oud outer wa,.I
and 1533 K (2300 0 F) on tne peripneral assemoly fuel rca cladding.
Continue attempts to complete Sequence Step 13 but continue tnrough
Sequence Steps 14-17.
The possibility exists that the time interval between the occurrence
of 2100 K (3321 0 F) cladding temperature in the center assembly and. tne
peripheral assembly cladding temperature or the shroud outer wall
temperature limit may be excessive relative to core damage and source
release limitations. Therefore, a maximum time of 7.0 minutes is,
specified, after the 2100 K (3321 0 F) temperature is reached, for
continuing the experiment sequence with Steps 13 and 14.
The last situation to be considered is if the CFM cladding has
exceeded 2100 K (3321 0 F) and if the temperature indications on either
the CFM shroud outer wall or the peripheral fuel assembly claddina begin
to fail before the temperature limits are reached. In the case of
the peripheral fuel assembly cladding, a minimum of two valid temperature
indications in the hot region are required to maintain the experiment termination
criterion in effect. (Two valid indications ot the temperature limit will
activate exper.ime'nt termination.) If the minimum number of temperature
indications is not maintained before the limit is reached, a clock function
defining the time remaining to experiment termination will be utilized. The
specifics of the operation of the clock function are the responsibility of
the Operations Branch.
23A
DRR-L- 7450Rev. I Chg. 5
TABLE 2. EXPERIMENT LP-FP-2 EVENT TIMES AND PARAMETER SETPOINTS
Item Action Time/Setpoint
Reactor scram Peripheral t 0
CR scram
Primary coolant pumps Tripped 8 ± 2s
CFM control rods Dropped 1.5 x 106 lbm/hr
Intact loop cold leg simulated break Open Verification of CFMcontrol rods in
Broken loop hot leg simulated LPIS Open 220 ± 5 spipe
Intact loop cold leg simulated break Close 566 ± 5 K (560 ± 90 F)
FPMS sampling system isolationvalves
LPIS Break Line Filter
Gamma densitometer sources
FPMS PCS sampling system isolationvalves
Broken loop hot leg simulated LPISpipe
Deposition coupon device
Core reflood
Open
Open
Store
Close
Close
Close
Initiation
1.2 ± 0.03 MPa(160 ± 5 psig)
811 K ± 5 K(10000 F ± 90 F)
840 K ± 5 K(1052 0 F ± 90 F)
840 K ± 5 K(1052 0 F ± 90 F)
1462 K (2172'F) onperipheral fuel assemblycladding or 1517 K(2272°F) on CFM shroudouter wall or other experimenttermination events.
6 ± 0.5 s from simulatedLPIS line closureinitiation
71
TABLE 3. INITIAL CONDITIONS FOR EXPERIMENI LP-FP-2
Parameter
Primary Coolant System
Core 6T
Hot leg pressure (MPa)(psia)
Hot leg temperature (K)(OF)
Primary coolant flow (Ibm/hr)(kg/s)
Boron concentration (ppm)
Power level (IMW)
Decay heat at 200 s minimummaximum
Maximum linear heat
generation ratec (kW/m)(kW/ft)
Control rod position(above full-in position) (m)
(in.)
Steam Generator Secondary Side
Water level
Pressurizer
Liquid level (m)(in.)
Specified Value a
As requiredb
14.95 ± 0.12168 t 15.0
571 ± 1.1569 ± 2
3.8 ± 0.15 x 106
479 ± 19
As requlredb
26.5 ± 0.5
675 kW695 kW
-40-12.26
1.37 _ 0.0154.0 ± 2.0
As requiredb
1.12 1 0.144 1 4
I
I72
APPENDIX B
FISSION PRODUCT MEASUREMENT SYSTEM FOR LP-FP-2
APPENDIX B
FISSION PRODUCT MEASUREMENT SYSTEM FOR LP-FP-2
This appendix describes the design and operation of the Fission
Product Measurement System (FPMS) that was used during Experiment LP-FP-2.
The FPMS consisted of three basic systems: the aerosol sampling system,
which was operated during the transient phase only, and the gamma detection
and deposition coupon systems, which collected data during both transient
and post-transient phases. Each of these systems is described herein.Figure B-l is a schematic of the FPMS showing the location of each of the
individual measurements.
Deposition Coupon System
Stainless steel deposition coupons were positioned in the reactor
vessel upper plenum region to provide postexperiment information on fission
product plateout. These are designated D-l on Figure B-l. Two coupons
were located at each of three axial elevations, corresponding to 0.152,
0.61, and 1.65 m (6, 24, and 65 in.) above the upper tie plate. Both
coupons at each elevation were exposed to the reactor environment during
the heatup. One coupon at each elevation was isolated and sealed prior to
initiation of reflood, while the second coupon remained exposed. Thus, the
plateout during the heatup phase was to have been distinguished from the
plateout/leaching during the reflood phase.
The D-l deposition device is a hollow rod containing deposition
coupons. At experiment initiation, the 0-1 deposition device was full of
liquid water. A nitrogen purge gas system was connected to the rod to
ensure dry coupons for fission product plateout. The hollow rod was pusheddown before reflood to isolate the protected coupons. At that time, the
nitrogen gas purge was restarted to remove steam, which could condense onto
the coupons. The nitrogen gas supply to the rod was then to have been
controlled at 1.4 MPa (200 psia) above reactor pressure to ensure that anyleakage of the deposition rod seals was outward, thereby maintaining a dry
atmosphere for the protected coupons.
8-1
F2Aerosol saper..
%A I'
SuppressioveslG G2
G3
5 4066
Figure B-i. FPMS Schematic.
B-2
The D-2 and D-3 coupons were located upstream and downstream of the
simulated LPIS header, respectively. To allow only high quality steam toflow in the line, it was not opened until the primary system mass inventory
had decreased. In addition, the line was isolated prior to reflood so that
these deposition coupons were protected from water flow.
Filter Sampling System
There were three filter sampling systems installed for this
experiment. These systems provided samples of the vapor and aerosols
generated during the heatup phase of the experiment. Both of these
constituents were expected to combine to provide the medium for transportof the fission products. Figure B-2 is a schematic representation of thedesign of the Fl and F2 sample lines. The filter sample locations were:
0 Fl--in the reactor vessel upper plenum at 1.80 m (70.75 in.)above the top of the lower tie plate
o F2--in the broken loop hot leg spool piece just outside of the
upper plenum
o F3--in the exit of the broken loop hot leg
The Fl system consisted of the following major components:
o sample line probe
o cyclone separator/isolation valve
o dilution filter
o virtual impactor
o collection filters
6--3
F1 sample line F2 sample line
I f iI
I I j filter
Collectionfilters
co
Figure B-2. Schematic of F1 and F2 aerosol sample systems.
( (
0 infrared moisture detectors
o recombiner
o critical flow orifice
o gamma spectrometer.
The sample line probe, shown in Figure B-3, diluted the vapor/aerosol
sample with an inert gas to minimize sample line deposition and to inhibit
interactions within the sample.
The cyclone separator/isolation valve, shown in Figure B-4, isolated
the filter assembly before and after the heatup phase and removed particleswith an aerodynamic diameter larger than 20 to 30 micrometers.
The dilution filter reduced the mass loading of the aerosols to
prevent plugging of the virtual impactor.
The filter train, shown in Figure B-5, consisted of the dilution
filter, the three stage virtual impactor, shown in Figure B-6, and the
collection filters. The train separated the aerosols into size ranges of6 to 20 micrometers, 1.7 to 6 micrometers, and less than 1.7 micrometers
with each size range being collected on a separate filter.
The recombiner contained cupric oxide, which converted the hydrogen to
water. The infrared moisture detectors then provided quantitative data onthe amount of argon, hydrogen, and steam entering and exiting the
recombiner. These data provide the necessary input for calculating the
amount of hydrogen and the dilution ratio of argon/hydrogen sampled during
the transient.
The critical flow orifice provided a mass flow out of the line during
the transient.
B-5
V
Sheath gas in
Mixture out
iling
Heating/coolinggas out
I-
gas in
I1F,4= Il
Sample flow INEL 4 4962
Figure B-3. Samiple line probe.
9w
B-6
Pneumatic operator
rator/'
From sampleprobe
5 4060
Figure B-4. Cyclone separator/isolation valve.
Three-stagevirtual Collectionimpactor filters
Dilution -- @- 6-20 ýJm H
filter
Sample in ""• b "w 1.7-6 pm
F--O[< <1.7 j•m I
Flow-controlorifices
I , Efi
I ou
I
fluentt
5 4064
Figure B-5. Impactor and filter train.
B-7
.113 in.
0.040 in.
1.600 in.
/ 0.048 In.
0.110 in.t
id = 0.071 In. Id - 0.025 in. dia 0.026 in.
IW - --o.325 in. 0 .500 in. .I II
5 4065
Figure B-6. Three stage virtual impactor.
8-8
The F2 line was similar to the Fl line except for the deletion of the
moisture analyzers and inert dilution gas. The F3 line, also designatedthe simulated LPIS line and shown in Figure B-7, contained the following
components:
o Ueposition samples upstream and downstream of the gamma
spectrometer (Dl & D2)
0 Gamma spectrometer (G5)
o Filter (F3)
o Flow venturi.
FPMS Gamma Detection Sampling System
Four gamma spectrometers and one gross gamma monitor were used in the
FPMS to provide a real time quantitative measurement of the radioisotopes
present in the LOFT system during the 12-hour post-transient sampling
phase. Two of the five were operated during the transient phase. The
sample points are shown on Figure 8-1 and are as follows:
o Gl--spectrometer operated only during postexperiment: reactor
vessel lower plenum at 0.584 m (23 in.) below the core, or
alternately from the primary coolant hot leg in the horizontal
PC-3 flange.
o G2--spectrometer operated during transient (combined Fl and
F2 sample lines effluent) and postexperiment: vapor space of the
BST.
o G3--spectrometer operated only during postexperiment: liquid
space of the BST.
o G5--spectrometer operated during transient: upstream of the
filter in the simulated LPIS line.
B-9
Removablespool
y-spectrometer
Removablespool
Filter Venturispool spool
Broken loophot leg
Blowdownheader
5 4061
Figure B-7. Simulated LPIS line components.
B-10
o G6--gross gamma monitor operated during transient: viewed sample
being drawn by Fl sample line located in the reactor vessel upper
plenum, which is 1.80 m (70.75 in.) above the lower tie plate.
The gamma spectrometer sample systems included valves for isolation
and sample point selection, pumps to provide flow, and pressure and
temperature instruments. The samples were returned to the same source that
was being sampled. The Gl, G2, and G3 spectrometers were enclosed in a
tent, to which an inert gas purge was applied to minimize the buildup of
background contamination that occurred during Experiment LP-FP-l.Additionally, the liquid and gas sample lines were purged with clean water
and inert gas, respectively, to measure plateout.
Each gamma spectrometer was designed to operate remotely over a broad
range of sample intensities. To improve accuracy, the spectrometer was
calibrated during the experiment (also remotely) using a thorium 228 source
mounted on the collimator wheel and background radiation levels were
recorded.
B-11
APPENDIX C
DESCRIPTION OF THE LOFT SYSTEM
AND INSTRUMENTATION
APPENDIX C
DESCRIPTION OF THE LOFT SYSTEM
AN) INSTRUMENTATION
The LOFT facility was designed to simulate the major components and
system responses of a commercial pressurized water reactor (PWR) during a
LOCA. The experimental assembly includes five major subsystems that have
been instrumented such that system variables can be measured and recorded
during a LOCA simulation. The subsystems include the reactor vessel, the
intact loop, the broken loop, the blowdown suppression system (BST), and
the ECC systems. Complete information on the LOFT system is provided in
Reference C-l, and a discussion of the LOFT scaling philosophy is provided
in Reference C-2.
The arrangement of the major LOFT components is shown in Figures C-l
and C-2. The intact loop simulated three loops of a commercial four-loop
PWR and contains a steam generator, two primary coolant pumps in parallel,
a pressurizer, a Venturi flowmeter, and connecting piping. A spool piece
was connected to the intact loop cold leg downstream of the pump
discharge. This provided the initial break path during the blowdown. This
line was closed prior to fission product release so the fission product
transport would be solely in the simulated LPIS line.
The broken loop consists of a hot leg and a cold leg. For this
experiment, the broken loop cold leg was flanged off and the broken loop
hot leg pump and steam generator simulators were removed. The simulated
LPIS line was connected to the end of the broken loop hot leg and provided
the path for fission product transport from the primary system to the BST.
The simulated LPIS line and associated flow instrumentation are illustrated
in Figure C-3.
The LOFT reactor vessel, shown in Figure C-4, has an annular
downcomer, a lower plenum, lower core support plates, a nuclear core, and
an upper plenum. The downcomer is connected to the cold legs of the intact
and broken loops, and the upper plenum, to the hot legs. The core consists
C-1
Intact loop Broken loop
I .4 1• Vapor monitoro ©(gamma spectrometer)
~' Pumps
ReactorveslSuppression Downcomer
Lower plenum
Reactor vessel5 0962
Figure C-1. Axonometric representation of the LOFT system for ExperimentLP-FP-2.
( C
Morded Wket.,e
3 LPIS
HPIS 4 LpwPSe B
t1,• LJ.' -4argeo B
Aoolitpcantmpnto
.way fron TE-SG-3. 4. 5pwlhcWt n
.yeSt.. PT-P139-2, 3. 4
generat P ary ,Coolan
TE-SG-2-2A TE-SG-I-1A ventut, FT-P13-271.2, 3 T.PT139- 4
Spray I..C• 1 3 pE.PC.2 EL C-2 ME-PC-2A. 8, C - r
.old g 314 D.-PC-2. BD. DD inje.l1o 1 PE -L-IAcO-d le DTE-PC--2A. . P0,
P139 1w-6. 7.8 TE-PI139 -- -- -
Pd Cor. CV-PC38-O7A
TE-PI39-20
439- Upper
14PIT- platn..
Pre, iP13en30d1
Cydltg - - - -IB I r-L -b a c k u p2 T E -B L H . ? T E -B L ) . T E -B L H -5 5• . L -
reaterST 14 PE-11-2
no TE-BLPm2 •2 TE-B- PE-BL-I
p oPI3Pmp8 TE-Pt 39- S. C. 0, 6o 0a HL SL. .H C.D0 2 .9E.L2A B. C. B 1
BI down supeso tankCV.TI 33-57 2noze
to PCP-2
PT-Pt20-43"• -- cnetd tkprSue
.
I- CV-PE-P503-153 0 00-P 3e-1 PE.P33- F-Zt39-139EI -
Accueruia'o PuPA .e opunt~,_n,
I
!
il (epetent •r==n
L ! q I ~~~~~3~~~~---
kU ýG...ACrAC1
Croseonn fine - 8-Pipe diameten in IncthesD Rd.UC.
380.94
iLi
III'
59673
Figure C-2. Schematic of the LOFT primary and emergency core coolingsystems.
Steam temperature PE-BLH-003Pressure
Shield tank N.penetration
N.-
CV-P-138-192
Gammaspectrometerspool
K
Wall temperature TE-BLH-004
D2 (deposition sample) spool
!
out offilter
5 0961
Figure C-3. Simulated LPIS breakline instrumentation.
( ( (
ft
DowncomerLLTLE-1 ST-1LE-1ST-2
-Upper plenum LLTLE-3UP-1
-Core LLTLE-11F10LE-3F10
INEL-MCL-3000
Lower plenumLLT-
Figure C-4. LOFT reactor vessel.
C-5
of approximately 1200 enriched uranium fuel rods arranged in five square
and four triangular (corner) fuel assemblies. The fuel rods were designed
to commercial PWR specifications except that they are only 1.68 m (5.5 ft)
long and several fuel rods have special instrumentation. All 100 fuel rods
in the central fuel assembly were enriched to 9.74 wt% 235U and were
prepressurized at cold conditions to 2.41 MPa (350 psia). All fuel rods in
the peripheral fuel assemblies were unpressurized and were enriched to
4 wt% 235U. Figure C-5 shows the fuel cladding thermocouple locations
and Figure C-6 shows all the central fuel assembly instrumentation.
Figure C-6 also illustrates the insulating flow shroud which took the space
of the outer two rows of fuel rods. This shroud protected the peripheral
fuel assemblies from excessive temperatures while allowing the central fuel
assembly to reach the requisite temperatures for fuel rod failure and
fission product release.
The two LOFT ECC systems are capable of simulating the emergency
injection of a commercial PWR. They each consist of an accumulator, a
high-pressure injection system, and a low-pressure injection system. There
were no programmatic considerations inherent in ECC operation; therefore,
the ECC injection was not scaled to represent commercial PWR operations
during Experiment LP-FP-2.
The LOFT steam generator, located in the intact loop, is a vertical
U-tube design steam generator. Operation of the secondary coolant system
during Experiment LP-FP-2 approximated that of a commercial PWR.
Figure C-7 is an illustration of the steam generator and its
instrumentation.
Figure C-8 illustrates the thermocouples in the upper plenum.
Table C-1 lists the instrument nomenclature.
C-6
TABLE C-1. NOMENCLATURE FOR LOFT INSTRUMENTATION
Designations for the Different Types of Transducers a
RPE - Pump speedPE - Pressure transducerPdE - Differential pressure
transducerLE - Coolant level transducerPS - Pressure switch
FE - Coolant flow transducerDE - DensitometerME - Momentum flux transducerFT - Flow rate transducerTC - Fuel centerline
thermocoupleTE - Thermocouple
Designations for the Different Systems, Except the Nuclear Core
PCBLRVSvUP
Primary coolant intact loopBroken loopReactor vesselSuppression tankUpper plenum
LP - Lower plenumST - Downcomer stalkP120 - Emergency core coolant
systemP128 - Primary coolant
addition and control
Designations for Nuclear Core Instrumentation
Transducer location (inches from bottom of fuel rod)
Fuel assembly row
Fuel assembly column
Fuel assembly number
Transducer typeIIITE-3B 1-28
a. Includes only instruments discussed in this report.
C-7
Broken 1oo0cold leg
Intact loophot leg
1
2@@@ZI( r@ 2 ®20@@@2 &
®@®@'A
3 ri(D@@P®A
135" 225'
Q-- Identification number ?SF-45" 0 315 Height of thermocouple above
\core bottom (in.)
Thermocouple
G5 TO® © •@42
/c77©WC©®EN
\._ @1®©7
W ©2
©
@2®424(241
0
6Z
N3@ar32\r-'Typicalrod guide
tubeQ
X 9JSD79
7.l2 12 I 1 J
®©®5 0970
Figure C-5. LOFT cladding and guide tube thermocouple locations.
( ( (
LP-FP-2 Center Fuel BundleInstrumentation
A IB JC ID I E IF IG IH III J IK IL IM IN 107
TZ-58-
10 27 32 42m
1
2
3
4
5
6
7TZ-SE- 10
278 32
42
9
10
11
12
13
14
-1N15
I
- 0_C227
00 ®Q2®
27 G) 6
©QQO N ® 27
Thermal sh.edflow shroud
I
TZ-5W-11027
3242
I10 2" 32 42
TZ-SN-
Instrumented guide tube
Instrumented fuel pin
Q Neutron flux scan tube (tip)Note: Thermocouple at location F7-42 failed prior to bundle installation
5 0963
Figure C-6. Center fuel assembly instrumentation locations.
C-9
HF-SG-001
Differential pressureFeedwater transducer for feedwaterinlet• liquid level LT-P004-8A
& 8AA
Secondary sidecoolant temperaturethermocouple, TE-SG-5
pressurefor feedwaterLT-P004-BB
Secondary sidecoolant temperaturethermocouple, TE-SG-4
FRC-202le Secondary side
le 'coolant temperaturenperature thermocouple, TE-SG-3p ie s , -,IA • Primary side
coolant temperaturethermocouples
~TE-SG-2 & 2A
inlet Primary coolant outlet
INEL-MCL-1803
Figure C-7. LOFT steam generator instrumentation.
C-10
( C CFuel
Assembly1
FuelAssembly
2
Fuel FuelAssembly % Assembly
3 4
FuelAssembly
5
FuelAssembly
6Station*
290-
280 -
270 -
260 -
250 -
240 -
0- TE-1UP-4
("3I 230 -
220
210 -
200 -
190 -
180-
0- TE-1UP-6
0- TE-1UP-7DTT
DL4TE.1UP.50
TTE-iUP-i"•TE-lUP-2
--A TE-3U.1P-48 • TE-3UP-8
0- LE-3UP-1-1
I 0 LE-3UP-1-2
Io -TE-3UP-10 I
LE-3UP-I-3 I0 'xTE-3UP-11 0 TE-5L
1 - LE-3UP-1-4 TE-5LI TE-3UP-12 O TE-5L
O- LE-3UP-1-5 - TE-5LI - TE-3UP-13
0- TE-2UP-4 0 LE-3UP-1-6 -TE-4UP-4
LýE-3UP-i-7I •TE-3UP-14 K TE-5L
U XLE-3UP-1-8 7 E5
LE-3UP-1- 110 TE-51.TE-3UP-15LE-3UP-1-8 0O TE-5ULE-3UP-1-9 OTE-5L
O---TE-2UP-250 I o31 TE-4UP-005. TE'5UI DTT 1 0OTE-5U
-TE-2UP-1 1 TE-4UP-10 E50 --,TE-2UP-210 0<TE-4UP-2 1I TE-5L•TE-2UP-3 I `'TE-4UP-3 i?¶ TE-5I
TE-3UP-1 I 1t4S/rTE-sII. TE-5[
P-31 A, B"
JP-30A, B1
IP-215B1, B2IP-29A, B & 250G1, G2
JP-33A,IBI
JP-32A, B!JP-251B1, B2JP-250G1, G2
I,. r."1- n*I I~ A
Nozzlecenterline
Top of uppertie plate
IP-197B1 ,B20TE 6 UPIP-194G1,G2 T.....IP-28A, B I'l TE-6UP-
P4" 0 TE-6UP-2JP-4, 17, TE-6U"P-
19,23JP-24, 25, 26, 27JP-188A, B, C, D
T
*Station numbers are a dimensionless measure ofrelative elevation within the reactor vessel. Theyare assigned in increments of 25.4 mm withstation 300.00 defined at the core barrel supportledge inside the reactor vessel flange.
5 0971
Figure C-8. Reactor vessel upper plenum instrumentation locations.
REFERENCES
C-1 0. L. Reeder, LOFT System and Test Description (5.5-ft Nuclear Core ILOCEs), NUREG/CR-0247, TREE-1208, Change 1, Sept. 1980.
C-2 L. J. Ybarrando et al., "Examination of LOFT Scaling," 74-WA/HT-53,Proc. 95th Annual Winter Meeting of the ASME, New York,November 17-22, 1974.
C-12
APPENDIX D
PCS THERMAL/HYDRAULIC RESPONSE
APPENDIX 0
PCS THERMAL/HYDRAULIC RESPONSE
This appendix describes the thermal/hydraulic boundary conditions in
the primary coolant system (PCS) during the transient and posttransient
phases of the experiment. Particular emphasis is placed on the estimated
core hydraulics and on the fluid conditions in the vicinity of the fission
product measurement system (FPMS) measurement locations during the time of
fission product release and transport. This appendix is broken into two
parts, the first describing the PCS response during the transient and the
second, the PCS response during the posttransient phase.
1. RESPONSE DURING BLOWDOWN AND REFLOOD
Experiment LP-FP-2 was dominated early by a gradual PCS level decrease
leading to a slow core boil-off. The level in the intact and broken loops
can be inferred by examining the densitometer data. Figures D-1 to D-4
show the average chordal densities measured by the individual gamma
densitometer beams in the intact loop hot leg, intact loop cold leg, broken
loop hot leg, and broken loop cold leg, respectively. In each case, the
three chordal densities are shown except for the A-beam at the intact loop
hot leg location. The data from this detector were failed due to
inadequate background correction. As explained in Section 2, the 6 0 Co
source was prematurely isolated. Thus, density data are only available for
about the first 260 s of the transient. These data show that the loops
started voiding at approximately 50 s in the intact loop hot leg, 85 s in
the intact loop cold leg and broken loop hot leg, and 120 s in the broken
loop cold leg. The voiding continued but was not complete when the
densitometer sources were isolated at 262 s. While the level decrease in
the loops could not be directly monitored beyond this time, it is clear
from thermocouple data in the upper plenum that the loop voidage was
complete by 470 s. This is illustrated in Figure D-5, which compares the
response of an upper plenum thermocouple with saturation temperature. The
time at which the thermocouple response deviates from saturated (470 s) is
an indication of the time when the level had decreased below that
0-1
_0
'0
LA-
0.75
0.50
0.25
'4-
(nC0
~0
-I,
U-
0 50 100 150 200 250 300Time (s)
Figure D-1. Intact loop hot leg densities.
I
N"0.75
0.50
0.25
.00 50 100 150
Time (s)
if)4-.4-NE
>5
to
0-o0
U-
200 250 300
Figure D-2. Intact loop cold leg densities.
0-2
I
C,,C0,-D
-o:3
LL
0.75
0.50
0.25
-I-
E
"3CU,-o
-D
:3
LI-
00 50 100 150 200 250 300
Time (s)
Figure D-3. Broken loop hot leg densities.
0,
C0-o-o:3
0.75
0.50
0.25
I')
E.0
U,
U,
-U
:3
0 50 100 150 200 250 300Time (s)
Figure D-4. Broken loop cold leg densities.
0-3
625 I , I
FFTEi-011Oll 650I --- T-PC-02
600
.4-0
-55
CL550E
525
0 50 100 150 200 250 300 350 400 450 500 550 600Time (s)
Figure D-5. Comparison of upper plenum fluid temperature with saturationtemperature. (See Appendix I for thermocouple qualificationlimits).I
D-4
elevation. Since this thermocouple is located at an axial elevation just
below the bottom of the loops, this time represents an approximation for
the time the loops were completely voided, with the exception of the loop
seals.
Figure D-6 shows the upper plenum level and indicates that the upper
plenum was voided at least down to within 1.1 m (43 in.) of the top of the
core by approximately 600 s. (This liquid level probe has been in the
reactor vessel since the initial core fuel load was installed in 1978. The
conductivity probes at several of the elevations had already failed prior
to conduct of Experiment LP-FP-2.) The level dropped into the core region
by about 700 s. This can be inferred from Figures D-7 and D-8, which show
the level in the Fuel Modules 1 and 3, respectively. The heatup, which
initiated in Fuel Modules 4 and 5 as discussed in Appendix E, preceeded
this measured level drop by approximately 1 min. This apparent discrepancy
could have been caused by a combination of two phenomena. First, the level
probes are located in the two instrumented corner fuel modules; the radial
decay heat distribution could have resulted in a depressed level near the
center of the core with a higher level near the edge. Based on0-1
calculations using the traversing in-core probeD, the ratio of initial
linear heat generation rates between the hot fuel rod in the center fuel
module and the fuel rod adjacent to the level probe is 2.3. The ratio
between the hottest part of the peripheral fuel assembly and the fuel rod
adjacent to the level probe is 1.4. The same values should exist for the
ratios of decay heat levels at these locations. Second, a liquid level
probe operation depends on the fluid between the closely spaced probes
being at the same void fraction as that of the bulk fluid. In the
environment of lower decay heat in the corner assemblies, it could be
expected that the response of the probes lags the voiding of the bulk
fluid. The boil-off continued until the entire center fuel module was
dry. Figure D-8 indicates that Fuel Module 3 was essentially uncovered by
approximately 1000 s. However, Figure 0-7 indicates the persistence of
fluid in Fuel Module I throughout the transient. Again, interpretation of
the data must take into account the limitations discussed above.
0-5
*** BUBBLE * VERSION. 001 * MOO 002 * OZ/Z3/81 *** DATE OF RUN 85/09/10. PAGE 00C2
BUBBLE PLOT OF FILE ILE3UP 0.
* CHARACTER RANGE TABLE *
* ( ) 1.000 K5 < 999.000 ** (0) .010 5 -C 1.CCC ** (X) -999.000 5 N ..010 *
4=• LEVELI (CM ABCVESCORE TOP)
19?.00 *XXXXXXXXXXXXXXXXXXXXXXXXXXX0132aOC *XXOOOOCOOOOOO0000000COCO00000000000o112.00 *XO000CO0000000000000COOOOCO
-. 053 165.147 339*147 513.147 687.147 861.147 1C35.147
oXXXXXXXXXXXXXXxOXXXXXXXXXXXXXXXOXXXXXXXXXXXXXXXx
1209.147 1363.147 1557.147 1731.147 1905.147 2079.147
TIME (S)
Figure D-6. Conductivity level probe response above Fuel Assembly 3.
I ( (
C C
J
* BUBBLE * VERSION 3001 M OD 002 * 02/73/81 C DATE CF RUN 85109/10. PAGF COCZ
BUBBLE PLOT OF FILE 'LE1FIO '.
* CHARACTER RANGE TABLE
* C ) 1.000' • N 999.000 *• (0) .010 N H 1.000 *
( CX) -999.000 • N • .010 *
LEVEL(M ABOVE
CORE BOTTOM)
1-68 *XO0000CO000aO00300O0COCC000 DOOCCO00000 coooo000x~xxxxx~xxxxxyxxxxxxxxxxx
1.67 *XOOOOOCOOOO3OOOOOOOOCOCOCOOOOOOOOOOOOOOOO 00 000C00CCCCxOxx XXXXXXXXXXXx.71 *XXXGC0COO003030000C0CCOOOCOOO000OOOXXOOO oootcXXXXXXx 00xxoc0coCCCOC~CC~COOCcc GyxxYxxxxxxxx.51 *XXXXXXXXXXXXOCOcoca cooxoxooxxxxxxxxxxxxxxxxxxxxxxx XXXXX • x•xx•cc•ccxocc xxxyx • X.41 *XXXXOXCXoX•CXCXCXCYC0 00 X xoCXx xXXXXxxX X COx•×•x X oxוCocoxxxxYx xx xXXXK.20 *XXXXXXOOOOOOOOOOOOOOcocooooooooooooooOooooooxxxxxxOOCOooCOCCOoO GCOOCCCCCCCcCCCCOCCXXXXXXXXXX
-. 053 158.447 325.447 492.447 659.447 826.447 993.447 1160.447 1327.447 1494.447 1661.447 IF26.447 1995.447
TItE (S)
Figure D-7. Conductivity level probe response in Fuel Assembly 1.
*** BUBBLE * VERSICH 001 * MOO 002 *' 02/23/81 **E DATE OF RUN e5/09/10, PACE 0003
RUBBLE PLOT OF FILE 'LE3F1O '.
* CHARACTER RANGE TABLE *
* ( ) 1.000 1 N 4 q99.000** (0) .010 i H 4 1.000 *
M (Xl -999.000 i N ' .010 *
LEVELtM ABOVE
CORE BOTTOM)
1.78 *XXXXOX000000000O0000COOCOYXXXXXXXXXXXXX000000OXCC000OCO OxxxxxxxxxXxx0o 1.08 *XXXXOO0OOOOOOOOOCOCOCCOCOCOOOooooooo0000000 0 0 0 0 cOcO0OxxxxxxxxxXXXXXXXXXxXXXXXXXXXXXXXXxxxxXXxxXXXXxxXXX.98 *XXXXGOOoooooo0ooooooCOC00COoo000000000000000000 oCO0000XXXXXXXXOOOOOOOCOOOOOOOcOOCOOOOXXxXXXXX XXXXX.89 *XXXXOOOOOOOOOOOOOOOQOOOOOCOOOOOOOOOOOOOOCXXOOOOOO OOOOOOCCOOOO0OOCOCCCOCCOCOOCXXXxxxXXxxxx.71 *XX00000C00000000C000000C0000O000000000000C0O000 OOOCXXXXXXXXXXXXXXXXXXXXXXXXXKXXXXXXXXXXXXXXXXXXXk.61 *XXXOOCOOOOOOOOCOC0ooCCOCCCOOO000000000000000000 OCC0C0OGCcXXXXxOCCCCCO000CCCCCCOCCC Coxxxxxxxxxx.51 *XXXOOOOO0OOOOOOOOCOOCOOOOO0000000O003000XXXOXOCO0 OcooooxxXxxxxxxxxxxxxxxxxxxxcoooocxxxXxxixxxx.41 *XXXXOXCOOOOOOOOOCOOO0CCOCOOOOOOOOOOOOOOOOOOXXXXXXOO OOXXCOC ccccoC00coCCCo oxxxxxxxxxxxx.20 *XXXXXXXXOOOOOOOCCOOOCOOCCOOO00000000XXXXXXXXXXXXXXXXXXXXXXXXOo oxxxxxxxxxXXX.10 * oocc00ccoccoooc • 0xx•00xxxxcoxCxxxxxxxxx•c• OOCCCCCCCCO00Cxxxxxxxxxx
-.053 1!8.447 325.447 492.447 b5Q.447 826.447 993.447 1160.447 1327.447 14q4.447 letl.447 1e2P..147 IC,.447
TIME (S)
Figure 0-8. Conductivity level probe response in Fuel Assembly 3.
The in-core self powered neutron detectors (SPNDs) have previously
been shown to be sensitive to changes in liquid level.D-2 Figures 0-9
to 0-11 show the response of the SPNDs in the peripheral fuel modules.
These detectors are centered at an elevation of 0.66 m (26 in.) above the
bottom of the core and each indicates that the local void fraction
increased to near 1.0 at that elevation between 690 and 740 s. Cladding
temperatures in the vicinity of the SPNDs indicates heatup initiation at
approximately 730 s, or 40 s after initiation of the SPND response to the
density decrease. This time lag between SPND response to density decrease
and heatup under decay heat conditions is consistent with previous
experiments.D-2 The SPND output exhibited a sharp decrease at
approximately 735 s, or approximately when the intact loop cold leg break
was isolated. This could be indicative of a rapid collapse in liquid level
as the depressurization and steaming rates suddenly decreased. The SPNDs
which were installed in the peripheral fuel modules are all prompt
detectors utilizing a Co emitter. The SPNDs which were installed in the
center fuel module in Experiment LP-FP-2 are delayed detectors and utilize
a Rh emitter. The response of these latter detectors to the transient is
discussed in Appendix G.
The PCS mass inventory was derived from the mass increase in the
Dlowdown suppression tank. The suppression tank level is shown in
Figure D-12. This level is an average of two independently measured levels
in the suppression tank. It should be noted that due to the large
oscillations in the data and the nonphysical offset between the two level
measurements, these data are not qualified during the transient. Thus, the
derived PCS mass inventory is useful for trend information though not for
absolute magnitudes during the transient. The trends are reasonable,
however, and a single point check of the mass inventory can be made since
the levels were qualified both for initial conditions and for the time
after isolation of the PCS. The derived mass inventory is shown in
Figure D-13 and indicates that the inventory decreased from an initial
value of 4700 kg (10360 Ibm) to a minimum of just over 500 kg (1100 Ibm) by
13uO s. After this time, the mass flow was high quality steam and the
overall mass inventory remained approximately constant (at least within the
D-9
E
Y0.6
0-
~O5°•0 .
L
0.4
ED
0.7
0.3aU
-.J
0.2600
Figure 0-9.
• 0.9C
"- 0.8L0CS
S0.7
0
-- 0.6-00
-J
NE-2H08-26 700
N-
•600
C
0-500 :-a
$..
Q)
400 o
-300
0
:5 650 675 700 725 750 775 800
Time (s)
Response of SPND in Fuel Assembly 2 to core uncovery.
62
-1000 I.. B
9D
-900 cn
C
0
600
C
'700 0)•
4-
600 _
0C.,.
0.5 L600 625 650 675 700 725
Time (s)750 775 800
Figure D-1O. Response of SPND in Fuel Assembly 4 to core uncovery.
0-10
0.6
E
"0.5
0
-04c
0) 0.3
0l-0.
0
0
0.1600 6
Figure D-11.
1.35 r
25
NE-6H08-26 600I...
- C
500 M
0-400
0L.
-300 ,
-- 200
0o
5 650 675 700 725 750 775 800T i me (s)
Response of SPND in Fuel Assembly 6 to core uncovery.
4.4
4.3
-.4-
4.2 v
4.1 -0
4.0 '
-j
3.9
3.8
E
OD
-)
1.30
1.25
1.20
1.150 500 1000 1500 2000
T i me (s)
Figure D-12. Averaged BST liquid level.
D-1l
I. Mor1
5000 I
- '4500-401000000
4000
03500 -8000 0
3 3000 U-,
E -60002500 - E
4-2000
C1000_ -2000 -
500
0 500 1000 1500 2000Time (s)
Figure D-13. Primary coolant system mass inventory.
0-12
sensitivity of the suppression tank level transducers). The final mass
inventory, based on the (qualified) suppression tank level after isolation
of the PCS, was approximately 200 kg (440 Ibm), which is 300 kg (660 Ibm)
lower than the minimum value calculated using the transient calculation
method.
A key parameter affecting the core thermal response is the steam mass
flow in the center fuel, module during the period of metal-water reaction.
One of the concerns prior to the experiment was that blockage of the center
fuel module inlet might result in a steam-starved environment which would
limit the metal-water reaction. That, in turn, woula have limited the core
heatup and subsequent fission product release and transport. Although
there was no direct measure of the mass flow, there were two independent
indirect indications of the flow. The first was the thermal response
itself, which indicated that no steam starvation occurred (see
Appendix E). The second was provided by an analysis of the heat flux from
the cladding to the fluid. That heat flux is given by:
"(z) - h(Tclad - Tfluid)
Tfluid = TSAT f 2T a Q"(z) dz
where
a = fuel rod radius
cp steam thermal capacity
z axial elevation
S= mass flow
D-13
h h= heat transfer coefficient
-" surface heat flux
T clad cladding temperature
T fluid fluid temperature
T SAT saturation temperature
Data for heat generation rates and fuel rod parameters furnish the
adiabatic heatup rate and the surface heat flux under isothermal
conditions. Data for cladding temperatures, differentiated with respect to
time and compared with the adiabatic heatup rate and isothermal surface
heat flux, provide an estimate of the actual surface heat flux. Combining
the two equations to eliminate the fluid temperature results in one
equation with two unknowns, the heat transfer coefficient and the mass
flow. An assumption was made that these two parameters did not vary with
elevation at any given time, and data from the three thermocouple
elevations in the center fuel module (0.25, 0.69, and 1.07 m, 10, 27, and
42 in.) were used to solve for these two unknowns with one check for
consistency. The calculation was performed using the data at 1300 s, or
just prior to initiation of the metal-water reaction. The resulting heat
transfer coefficient did not vary more than 25% at the three elevations,
which is considered reasonable for this analysis. The resulting total mass
flow rate for the center fuel module was 0.04 kg/s (0.09 lbm/s) or 0.4 g/s
(9 x lU"4 Ibm/s) per fuel rod which is approximately 3 times the value
calculated prior to the experiment. The mass flow rate was sufficient to
allow the metal-water reaction to proceed without steam starvation. This
result is consistent with testing performed at the Power Burst Facility
wherein steam starvation was not measured for comparable steam flow
rates. D-3
The mass flow rate in the simulated LPIS line was derived from the
pressure differential across the calibrated flow orifice. During the time
D-14
of fission product release and transport, the flow rate was approximately
constant at 0.2 kg/s. This is approximately twice the flow rate predicted
prior to the experiment and, again, is consistent with the
higher-than-predicted flow rate in the center fuel module and with the lack
of steam starvation.
The fluid in the vicinity of the FPMS measurement locations generally
was superheated steam. Figures 0-14 through 0-17 compare the fluid and
saturation temperatures at the Fl, F2, D2, and D3 locations, respectively.
Figures D-18 through D-20 compare the fluid, metal, and saturation
temperatures for the upper, middle, and lower coupons in the Dl deposition
coupon rod, respectively. The fluid temperatures at all three deposition
coupon locations as well as the metal temperatures at the Dl coupon
locations were superheated throughout the period of fission product release
and transport. The fluid temperature at the F2 location is superheated
from approximately 1300 to 1800 s. The fluid temperature at theFl location indicated general superheat with two periods of quench. There
were no perturbations measured by a pressure transducer at the same
location. This implies that the amount of water present in the line was
probably very small, such that steam generation was not measurable. It is
possible that a small droplet of water (in, perhaps, a mixed flow regime)
quenched the thermocouple without affecting the steam or metal temperature.
When the shroud external temperature trip setpoint of 1517 K (2270*F)
was reached, the FPMS and LPIS lines were isolated and the ECCS injection
was initiated. Performance of the ECCS was excellent and the PCS was
rapidly brought to a stable shutdown condition. The low pressure pumps
injected approximately 25 kg of water during the initial 5 s of ECCS
operation. The accumulators injected over 1000 kg of water over the same
time interval. The resulting generation of steam caused the pressure to
increase from 1.2 to over 3 MPa (174 to over 435 psia). This
repressurization caused a temporary cessation of accumulator flow as the
PCS pressure exceeded that of the accumulator. Flow restarted at 1910 and
1940 s from Accumulators B and A, respectively, again causing a PCSrepressurization. This is illustrated in Figure D-21 which compares
D-15
600
550
v- 500
S-
-- 450La,a-
E 4001--
350
3000
Figure D-14.
500 1000 1500
Time (s)
Comparison of fluid temperature inline with saturation temperature.thermocouple qualification limits)
-400 qj
" 300p--
2O00
100
2000
the F1 aerosol sample(See Appendix I for
-400
-300
2 oE
300 a"
1--
-100
2000
the F2 aerosol sample(See Appendix I for
500
, 450
0
400La)a-
E
350
3000 500 1000 1500
Time (s)
Figure D-15. Comparison of fluid temperature inline with saturation temperature.thermocouple qualification limits).
D-16
750
v
0
EVD
I.--
625
500
375.
TE-LH00
TE-OLH-004
-. ST-OLH-003
-- -- - -- -- -
-800
-600
0-400 L
a-
EI-
-200
.02500 500 1000
Time (s)1500 2000
Figure D-16.
550
500
450
4-0L,
a)400
EI-.
Comparison of metal and fluid temperatures at the D2deposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).
500
400 ~~
300 oL
0.
E(D
200 1-
100
350
3000 500 1000 1500 2000
T i me (s)
Figure D-17. Comparison of fluid temperature at the D3 deposition couponlocation with saturation temperature. (See Appendix I forthermocouple qualification limits).
0-17
"- 600
550
L
E soo
450
4000
Figure D-18.a
Soo
.•.-700'--
E 6000
0-
E4500
4000
Figure D-19.
-600
I "I
. 500
E
-300
500 1000 1500 2000Time (s)
Comparison of metal and fluid temperatures at the D1 upperdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).
I I
- TE-5UP-21581TE-5UP-029A
X Saturation temperature
,-800
JA-
,,•0.
L --
E
-400
I !
500 1000 1500 2000Time (s)
Comparison of metal and fluid temperatures at the D1 middledeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).
0-18
1000
8000
L
5. 700L.S
E oo
0
I-
500
400-0
Figure 0-20.
5.5
57-
4.5
V.' 40~
a. S3.5
.. 3
u 2.5
D. 2
100
I • L
500 1000 1500 2000Time (s)
Comparison of metal and fluid temperatures at the D1 lowerdeposition coupon location with saturation temperature.(See Appendix I for thermocouple qualification limits).
|I I
"_- PE-PC-002--- PT-P120-029
PT-P120-043
600 ,-,0
Co
a.0.
4200
I !
1800 1900 2000 2100Time (s)
• Comparison of primary system pressure with pressuresmeasured in the accumulators.
1.5
0.51700
Figure 0-21
0-19
primary system pressure with the pressures measured in the two
accumulators. The damaged core required several hundred seconds of reflood
before all the metal had quenched, but the continued ECCS injection was
able to overcome the core thermal inertia and cool the core. The core
thermal response during reflood is discussed in more detail in Appendix E.
The PCS was recovered to cold shutdown conditions by secondary feed
and bleed operations. The times of significant events during this phase
are indicated in Section 3 of this report. The PCS was filled with
subcooled water and was solid throughout most of this phase. Figures D-22
and D-23 show the primary system pressure and upper plenum fluid
temperature during this phase. There were no unexpected thermal/hydraulic
events which were measured during the posttransient phase.
2. RESPONSE DURING THE POSTTRANSIENT PHASE
There was excessive background contamination at the G1, G2, and G3
spectrometer stations during the early part of the posttransient phase.
For that reason, the plant was maintained in a quiescent state for 44 d (to
allow for background reduction). During this time, data were collected
from the on-line gamma spectrometer systems (14 d) as well as from grab
samples taken from the PCS (44 d) and blowdown suppression tank (liquid
sample, 21 d; vapor sample, 28 d). Those data are discussed in Section 4
of this report.
A calculation of the mass flow in the PCS during the first 40 h of
this phase was made, using the decay heat, reactor vessel temperature rise,
and known fluid conditions. The resultant mass flow was in the range of
2 to 3 kg/s (4.4 to 6.6 lbm/s). This resulted in a loop transit time of
1900 to 2800 s. There were no unexpected events during this phase and thefluid boundary conditions for operation of the data collection systems were
as expected.
1ý .- ý IV
0-20
15
V)(L
12.5
10
7.5
5
2.5
0
-2000
-1500 0
(j-
a)-1000 L
(0)
a,,
CL
-500
0 500 1000 1500 2000 2500T-i me (s)
3000 3500 4000
Figure D-22. Primary system pressure during the posttransient phase.
800
750
,1-700
CD 650
-O 600L
o-550E
500
450
400
LL-60
40-I--
- 40
0 500 1000 1500 2000 2500Time (s)
3000 3500 4000
Figure D-23. Primary systemphase.
fluid temperature during the posttransient
D-21
REFERENCES
D-1. G. E. Putnam and B. L. Rushton, Three-Dimensional Pin Power from LOFTCore-I TIP Data, LTR-l1l-87, RE-Z-77-OOI, Sept. 1977.
D-2. J. P. Adams and V. T. Berta, "Monitoring Reactor Vessel Liquid Levelwith a Vertical String of Self-Powered Neutron Detectors," NuclearScience and Engineering, 88, 1984, pp. 367-375.
0-3. D. J. Osetek et al., "Fission Product Behavior During the First TwoPBF Severe Fuel Damage Tests," EGG-M-11284, ANS Topical Meeting onFission Product Behavior and Source Term Research, Snowbird, Utah,July 15-19, 1984.
D-22
APPENDIX E
CORE THERMAL RESPONSE
APPENDIX E
CORE THERMAL RESPONSE
The measured temperature excursion began in the upper part of the
peripheral fuel modules at 662 s (on TE-2E08-045) and moved downwards as
the coolant boiled away; the fifteen-inch elevation dried out at about
730 s (on TE-2F07-015); and the ten-inch elevation dried out at about 930 s
(on TE-2G14-Oll). The departure of these temperatures from saturation is
shown in Figure E-1. (The quench of the ten-inch elevation thermocouple
caused by the closure of the PORV can also be seen.) The center fuel
module remained at saturation for a longer time. This is believed to be
due to higher decay heat levels and higher steam flows in this module,
which caused the froth level inside the shroud to be higher than that
outside. The first recorded departure from saturation is at 689 s (on
TE-5M09-042); the 27-inch elevation dried out at about 740 s; and the
thermocouples at the ten-inch elevation in the center fuel module have a
minor [about 10 K (18°F)] departure from saturation starting at 780 s,
followed by a much more significant excursion at 940 s. Thesethermocouples are shown in Figure E-2. The dryout was therefore top-down,
although a few guide tube thermocouples did not follow this pattern.
In order to control the system pressure, the intact loop cold leg
break and PORV were cycled during the early part of the temperature
excursion. The effects of all of these changes (except the opening of the
intact loop cold leg break at 877.6 s, which is followed too closely by the
opening of the PORV at 882 s) can be seen on the heatup rates of some
thermocouples in the core. The greatest effect by far was caused by the
opening of the PORV at 882 s, which caused cooling throughout the core, and
quench in the lower parts of the peripheral modules as the level responded
to the changing pressure gradient in the reactor vessel. In order tobetter see the small changes in temperature gradient caused by the various
events, Figure E-3 shows TE-4G08-021 data after subtraction of the average
rate of increase of 0.538 K/s (O.9684°F/s),
600
550
450
600
Figure E-1.
600
55-
E 5oo
450 -600
Figure E-2.0L
D.-
I-
I'-,
450O600
Figure E-2.
I•TE-2EO8-045I--- TE-2F07-015 r• -400
m- TE-2G14-011-X Saturation tenperature
5/ II I I..
650 700 750 800 850 900 950 1000Time (s)
Comparison of cladding temperatures at the 1.14-, 0.38- and0.28-m (45-, 15- and 11-inch) elevations in Fuel Assembly 2with saturation temperature during early stages of heatup(600 to 1000 s). (See Appendix I for thermocouplequalification limits).
-600
-550// -500
•, .-. • •'/ 450 C#E
TE-5MO9-042
--- TE-5D09-027 -400. TE-5C12-010
X Saturation temperatureI I I I I I I
650 700 750 800 850 900 950 1000Time (s)
Comparison of cladding temperatures at the 1.07-, 0.69- and0.25-m (42-, 27- and 10-inch) elevations in the center fuelassembly with saturation temperature during early stages ofheatup (600 to 10001s). (See Appendix I for thermocouplequalification limits).
E-2
160
150
-' 140
L
1300
0.
E 12oI-
110-
100 -700
Figure E-3.
-150
-200
L.-220
0
-240 0LE
- -260
800 900 1000 1100 1200 1300 1400Time (s)
Effect of flow changes on rate of temperature increasemeasured at 0.69-m (27-inch) elevation on fuel rod claddingin Fuel Assembly 4 (700 to 1400 s). (See Appendix I forthermocouple qualification limits).
E-3
When the guide tube thermocouples at the 27-inch elevation reach about
1050 K (1430°F), which occurred at about 1300 s, the temperature gradient
decreased from about 1.2 K/s to about 0.7 K/s, as is shown in Figure E-4.
This is judged to have been caused by melting of the control rod material
and the resulting adsorption of latent heat. This hypothesis is supported
by the fact that TE-5H08-027 is the only 27-inch thermocouple which does
not exhibit this effect, and this guide tube does not contain a control
rod. An approximate calculation indicates that the latent heat could cause
a temperature shift of 280 K (504°F), which is considerably more than the
observed shift of about 50 K (90°F). That difference can be explained by
the heat input from metal-water reaction, which initiated during this time.
At 1330 s some water appears at the top of module 4, as shown by the
upper plenum thermocouples in Figure E-5. At 1360 s, as shown in
Figure E-6, the lower plenum thermocouples in this module are cooled. Close
examination of the fuel rod cladding thermocouples in module 4 show a very
small effect at about 1360 s, as can be seen on Figure E-3. The cause of
this is not known, but may be due to water, from a presently unknown
source, either dripping from the upper tie plate to the lower tie plate
without touching the fuel rods or running down guide tubes between the tie
plates. Of these two possibilities, the latter is probably more reasonable.
The first recorded (and qualified) rapid temperature rise caused by
the exothermic reaction between the steam and the zircaloy is at about
1430 s on guide tube thermocouple TE-5H08-027. (Thermocouple TE-5EIl-027
was judged to have failed at 1311 s, but the mode of failure suggests that
temperatures reached 1800 K (2780 0 F) at some location in the core by
1381 s.) The rapid temperature rise began from approximately 1400 K
(2060°F). (These figures must necessarily be approximate because there is
no abrupt change to characterize the start of the reaction). The data from
these two thermocouples are shown in Figure E-7.
Because of the relatively few locations in the center fuel module at
which the fuel rod cladding temperatures are measured, it is considered
unlikely that the first occurrence of the rapid reaction between steam and
E-4
1600
"-" 13006,L.: 1200
,4--0L 1100,6L
E 1oo04)
I--900
800'-,
7001100
Figure E-4.
g00
750
700
* 650L
4o 600L
550E
P-500
450
400 L600
Figure E-5.
/-2000
*1 Li.. 1500 L
E/ ° ._..- *° ,, f•
115 120 125 1300 13010E4010 5010
/~~ Tm (s)' ' I
,'|,I-
fuel assembly (1100 to 1600 s). (See Appendix I forthermocouple qualification limits).
I..
-600 0
CL0 )
4O00
8oo 1000 1200 1400 1600 1800 2000Time (s)
Comparison of fluid temperatures at upper tie plate aboveFuel Assembly 4 with saturation temperature(600 to 2000 s). (See Appendix I for thermocouplequalification limits).
E-5
550
"- 500
L
E 450I-.
400 -600
Figure E-6.
2500 -
- TE-4LP-003 500Saturation temperature
,450
0
400
0.
35o E
300
I I I I l. .
800 1000 1200 1400 1600 1800 2000Time (s)
Comparison of fluid temperatures at lower tie plate belowFuel Assembly 4 with saturation temperature(600 to 2000 s). (See Appendix I for thermocouplequalification limits).
CL
0
2000
1500
1000
4000
3000
2000 LID.
EQD
10-
-1000
I I I I I 1700 800 900 1000 1100 1200 1300 1400 1500 1800
Time (s)600
Figure E-7. Effect of metal-water reaction on guide tube temperatureincrease at 0.69-m (27-inch) elevation in center fuelassembly (600 to 1600 s). (See Appendix I for thermocouplequalification limits).
E-6
the zircaloy occurred at a thermocouple location. Once the reaction has
begun in some location, however, the resultant high temperatures will
influence nearby surfaces by radiation and conduction, thus causing an
increase in the rate of temperature rise. It is almost certainly this
effect that is being measured in the center fuel module.
The course of the rapid reaction between the zircaloy and the steam
can be tracked by noting the times at which indicated cladding temperatures
exceed 1800 K (2780 0 F), that being a reasonable indication that the rapid
reaction has occurred. The results, for those thermocouples which had not
failed by that temperature, are shown in Table E-l. The reaction probably
started between the 0.64-m (27-in.) and the 1.07-m (42-in.) elevations.
The reaction then spread across the entire center fuel module at the 1.07-m
(42-in.) elevation between 1480 and 1530 s, and then across the 0.69-m
(27-in.) elevation. The few thermocouples at the 0.69-m (27-in.) elevation
that reacted early (TE-5H08-027 and TE-5H06-027) seem to be exceptions to
the pattern. There is no evidence of a rapid temperature rise due to the
reaction between steam and zircaloy at the ten-inch elevation.
At about 1500 s several instruments show effects of some event that
has taken place. These instruments include the gross gamma monitor (shown
in Figure E-8), upper plenum thermocouples (shown in Figure E-9), the
momentum flux meter in the downcomer (shown in Figure E-l0), and guide tube
thermocouples. A possible initiating event for these effects is the
rupture of the control rod cladding: the sudden release of aerosols could
explain the rapid increase in gamma activity; the flow redistribution
caused by blockage would affect upper plenum temperatures; and the flashing
of water (caused by molten absorber material falling into the water below
the lower tie plate) could cause the downcomer flow.
A more direct indication of the guide tube behavior can be obtained
from an examination of the guide tube thermocouples. Several of the guide
tube thermocouples at the 0.69-m (27-in.) elevation show a discontinuity in
temperature at about 1500 s, as can be seen in Figure E-4. This may be an
indication of their rupture, perhaps caused by absorber material from a
E-7
TABLE E-1. TIMES FOR THE CENTER FUEL MODULE1800 K (2780-F)
THERMOCOUPLES TO REACH
Identifier
TE-5H08-027TE-5H06-027TE-5112-042TE-5104-042TE-5L09-042TE-5C07-042TE-5M09-042TE-5Dl3-042TE-5H12-027TE-5K05-027TE-5KlI-027TE-5103-027TE-5M06-027TE-5F03-027TE-5C09-027TE-5J07-027TE-5L07-027TE-5C12-027TE-5G12-027
Time to 1800 K Type of thermocouple
1451147514871488149114951513152915361538154515491564158016711674168616861695
S
SS
S .S
S
S
SS
S
SS
S
SS
SSS
Guide TubeInternal CladInternal CladInternal CladInternal CladInternal CladInternal CladInternal CladGuide TubeGuide TubeGuide TubeInternal CladGuide TubeGuide TubeInternal CladInternal CladInternal CladInternal CladInternal Clad
This table lists the times at which the center fuel modulethermocouples reach 1800 K (2780 0 F). Only thermocouples which reach 1800 K(2780*F) before they are judged to have failed are listed.
E-8
12
10
8
6
40
02600 800 1000 1200 t400 1600 1800 2000
Time (s)
ement of gross gamma activity near reactor vessel headFigure E-8.
1000
'-" 8006)
L
04--
0LU,r
E 600I--
4O0600
Measur(600 to 2000 s).
0U,L-
U,.0-
EI-
Figure E-9.'
800 1000 1200 1400 1600 1800Time (s)
Comparison of fluid temperatures at different horizontallocations on center fuel assembly upper tie plate(600 to 1800 s). (See Appendix I for thermocouplequalification limits).
E-9
2 E
1000X
0 E
CC
0E0
-21 -1000
-200 0 200 400 600 800 1000 1200 1400 1600 1800 2000
Time (s)
Figure E-IO. Momentum flux in reactor vessel downcomer (-200 to 2000 s).
E-10
hotter region flowing past the thermocouple. Once again, the effect is
absent from TE-5H08-027, the thermocouple on the guide tube without a
control rod. The temperature at the 0.69-m (27-in.) elevation is
approximately 1200 K (1700°F) at 1500 s, which is lower than the
temperature at which the control rod material release was expected [1400 K
(2060*F)]. However, the temperatures were probably higher at higher
elevations.
Between 1520 and 1680 s some of the thermocouples at the ten-inch
elevation in the center fuel module measured small temperature increases,
some examples of which are shown in Figure E-11. Those may be due to
molten material running down the rod or down a nearby rod.
The peripheral fuel rods exhibited some cooling around the time of
1550 s (as shown in Figure E-12); it was particularly strong further away
from the center fuel module, and also at lower elevations. This may be an
effect of the steam flow that is diverted from the center fuel module,
which is now partially blocked. Added to this steam flow may be steam
produced from hot materials from the center fuel module falling into the
water in the lower plenum.
At 1640 s there was apparently another sudden flow change in the
peripheral modules, and the cooling becomes more severe. The cooling is
strong enough to make some thermocouples exhibit a drop in temperature (see
Figure E-12).
It is possible that there is a connection between the small
temperature increases noted at the ten-inch level in the center fuel module
and the cooling of the peripheral modules, which occurred at the same time.
This would be consistent if the control rod material was running past the
ten-inch level, through the tie plate, and into the water, thus creating
steam, which cools the peripheral modules.
During most of the transient, the thermocouples on the outside of the
shroud increased steadily in temperature from saturation at about 740 s
(940 s for the ten-inch elevation) until about 1700 s. All of the shroud
E-11
1100
L
E
4..
1000
900
800
TE-5G04-010 1400
1200 L
4W
. IL
1000
300 1400 1500 1600 1700 1800 1900Time (s)
Cladding temperatures at 0.25-m (10-inch) elevation incenter fuel assembly during high temperature stage oftransient (1200 to 2000 s). (See Appendix I forthermocouple qualification limits).
7W01200
Figure E-11.
I
1000
900
I,
.4-
UL
a.E4)
I-
600
700-
600
TE 4I I I I 7 IF
I
1200
1000
0 0 "
BoL.8600 e4"
400500
V- I
600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900Time (s)
Figure E-12. Comparison of cladding temperature at 0.38-m (15-inch)elevation in Fuel Assembly 4 with saturation temperature(600 to 1900 s). (See Appendix I for thermocouplequalification limits).
E-12
thermocouples are shown in Figures E-13 to E-16. The thermocouples at the
1.07-m (42-in.) elevation deviated from this a little in that their
temperature rise rates increased at about 1540 s (TE-5E-042 and TE-5S-042)
or 1620 s (TE-5N-042 and TE-5W-042). These changes in rise rate may be
another facet of the effect that caused cooling of the peripheral rods at
about these times, or it may be due to the shroud becoming less efficient
as an insulator. The temperature of the shroud at 1700 s varied from about
8UO K (980°F) at the ten-inch elevation to about 1400 K (2060 0 F) at the
42-inch elevation. At 1700 s several of the shroud thermocouples exhibited
an increase in temperature rise rate, the effect being strongest on the
south side (next to module 2) and particularly on thermocouple TE-5S-0lO.
Also at about 1700 s, the thermocouples near the outside of the
shroud, particularly at lower elevations, began an extraordinary
temperature excursion. By 1780 s (just before reflood) all of the
thermocouples in module 2 near the south wall of the shroud, with
elevations ranging from 0.28 to 1.24 m (11 to 49 in.) reached approximately
1400 K (2060'F). A selection of these thermocouples are shown in
Figure E-17. The shroud wall thermocouples at or below 0.81 m (32-in.)
were cooler than 1400 K (2060°F). The thermocouples near the shroud in
modules 4 and 6 behaved similarly, but not in such an extreme manner. A
comparison of modules 2, 4 and 6 are shown in Figure E-18. (The
thermocouples in the peripheral modules away from the shroud see a little
of a similar effect just before reflood, as can be seen in Figure E-12.
However the time of the temperature rise in Figure E-12 corresponds closely
to the closure of the LPIS line break and the isolation of the FPMS system,
so the effect may be unrelated.)
The cause of the rapid temperature rise is somewhat uncertain. The
exothermic reaction between zircaloy and water is not a possibility because
the initiation temperatures for the rapid rise are too low; nor is
radiation from the shroud wall a possible cause because the wall
temperature is less than that reached by the fuel rod thermocouples at this
elevation. It is judged that the rapid temperature rise is caused by
E-13
1700
1500
, 13000
L 11004-I
0~-900
E,. 700
-2500TE-5N-010TE-5N-027 ,1TE-5N-032 /TE-5N-042 •20Saturation temperature
-1500 '
-1000 E0
50
800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900T ime (s)
Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-rn (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 8 with saturation temperature(600 to 1900 s). (See Appendix I for thermocouple
qualification limits).
500
300 -I-
600 700
Figure E-13.
1700
1500
IiL.
4-a
U)a.EU)
I-
1300
tIo 40
I-TE-5E--0104--TE-5E-027A
X Sauraiontemperature
4-
X XKX
-2500
-2000
0
- 1500
L
-1000 CLE
900
700
500 -500
300
600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900Time (s)
Figure E-14. Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-m (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 4 (600 to 1900 s). (SeeAppendix I for thermocouple qualification limits).
E-14
1700
1500
1300
CL
E4)
I-
11004
I I I I I I I I ! I !
TE-5S-010TE-5S-027
-- TE-5S-032- TE-5S-042
X Saturation temperature /
/ X XI
-2500
-2000
-1500 L
-1000 0-
I--
- 500
900
700
500,
.0U,600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900
Time (s)
Figure E-15. Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-m (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 2 (600 to 1900 s). (SeeAppendix I for thermocouple qualification limits).
1700
1500
2 1300
1 100
900-.
,~700
-2500
-2000
-1500 .
-1000 .
E
-500500
300 1600 700
Figure E-16.
800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900Time (s)
Comparison of temperatures at 0.25-, 0.69-, 0.81- and1.07-n (10-, 27-, 32- and 42-inch) elevations on shroudwall facing Fuel Assembly 6 (600 to 1900 s). (SeeAppendix I for thermocouple qualification limits).
E-15
1500
1300
-" 1100
L
E 900L.4)a-E 700
I,-
500
3001400
Figure E-17.
2000
f - ).... ...- "
-1500
0CD
L
1000 L0.-EI.-
- TE-2G4-011Soo--- TE-2H14-028 500
TE-2114-039-TE-2H13-049
I I I I I I I I J
1450 1500 1550 1600 1650 1700 1750 1800 1850 1900Time (s)
Comparison of cladding temperatures at 0.28-, 0.71-, 0.99-and 1.24-m (11-, 28-, 39- and 49-inch) elevations in FuelAssembly 2 (1400 to 1900 s). (See Appendix I forthermocouple qualification limits).
I I I I I I
TE-2G14-011- TE-4G14-011 -2000
-- TE-6G14-011
-0
"".1500 0 L
S-I,=
1000 L
E0I--
500
, I I I I II ,
800 1000 1200 1400 1600 1800 2000T I me (s)
Comparison of cladding temperatures at 0.28-m (11-Inch)elevation on fuel rods in peripheral assemblies 2, 4, and6 close to shroud (600 to 2000 s). (See Appendix I forthermocouple qualification limits).
1500
I300
v 1100
900L
E 700F--
500
300 i600
Figure E-18.
E-16
shunting of the thermocouple leads where they pass through an area of high
temperature (near the top of the core). This is reflected in the
qualification statements for those transducers in Appendix J.
Reflood was initiated at 1782.6 s, and the peripheral modules quenched
very rapidly between 1789 s (on TE-4F07-015) and 1793 s (on TE-2HI5-032 andTE-4114-021). The shroud outer wall also quenched during this period; all
thermocouples on the south side of the shroud failed simultaneously at
1790 s, presumably due to some mechanical strain associated with the
reflood process.
In the center fuel module, only the ten-inch elevation thermocouples
were qualified as accurate through the reflood, and the last of these to
quench was TE-5J07-0lO at 1795 s. The other center fuel module
thermocouples, including those on the upper tie plate, failed at or before
reflood. The nearest thermocouple above the center fuel module that
survived the reflood was TE-5UP-028B, which is shown in Figure E-19. This
thermocouple quenched at 1793 s, then heated rapidly to 1900 K (2960*F)
before quenching again at 1801 s. This gives some evidence that the center
fuel module remained hot for at least a short period after the rest of the
core quenched.
More information on the cooling of the center fuel module can be
inferred from the thermocouples which failed during the transient. One
mode of failure that is expected is for the thermocouples to melt and
rejunction at some hot location that the thermocouple lead passes through.
If this happens, then the thermocouple may still give qualitative
temperature information about the location where the new junction is
formed. Two examples of this appear to be TE-5K1l-027 and TE-5J07-027,
both of which are shown in Figure E-20. Thermocouple TE-5K11-027 was
evidently reading a true temperature because it-quenched at 1880 s and then
remained subcooled, as might be expected after reflood. This is therefore
fairly reliable evidence that some location in the center fuel module was
hot until at least 1880 s. Thermocouple TE-5J07-027 was cooling towardssaturation until 2010 s, at which time the junction apparently broke
E-17
2000
1500
vL
CD
0
EI,-
!i
I I
TE-5UP-028BSaturationtemperature
1000
3000
-2000
1000 0
E
'D0uu [-- - -
0L1750
I I I I I I I I I
1760 1770 1780 1790 1800 1810 1820Time (s)
1830 1840 1850
Figure E-19. Comparison of fluid temperature at upper tie plate abovecenter fuel assembly with saturation temperature duringreflood (1750 to 1850 s). (See Appendix I for thermocouplequalification limits).
0L.idCD
E.
0
I.-
2500
2000
1500
1000
500
- 4000
-3000
CD-2000
-0a.-
-1000 E"
I--
-0
0 LI600
Figure E-20.
900 1200 1500 1800 2100Time (s)
Effect of material relocation on cladding temperatures at0.69-m (27-inch) elevation in center bundle during reflood(600 to 2100 s). (See Appendix I for thermocouplequalification limits).
E-18
again. In order to cool so slowly, the thermocouple junction must be well
insulated from the reflood water. It is therefore reasonable to assume
that there is a mass of material in the center of the center fuel module
which is in a compact, and thus difficult to cool, geometry. This stayed
above the saturation temperature for several hundred seconds.
Additional information concerning the configuration of the center fuel
module after the transient is provided by experiments conducted at
Kernforshungs Karlsruhe (KfK) in the Federal Republic of Germany. The rate
of temperature rise in Experiment LP-FP-2 is bounded by the rates measured
in KfK Experiments ESBU-l [2K/s (3.6 0F/s)] and ESBU-2A [0.5/K/s
u.9 0F/s)]. E- The maximum temperatures reached in these experiments were
approximately 2525 K (4086°F) and 2450 K (3951 0 F), respectively. These
maximum temperatures approximate those that are believed to have occurred
in Experiment LP-FP-2, so it is reasonable to believe that some aspects of
the final configuration may be similar. In the KfK experiments, the center
part of the core melted and flowed downwards, liquidifying some of the fuel
pellets; the test bundle was almost completely blocked by the resolidifiedmelt. A similar configuration is expected to have resulted in this LOFT
experiment.
E-19
REFERENCES
E-1 S. Hagen and B. J. Buescher, "Out-of-Pile Experiments on PWR FuelBehavior Under Severe Accident Conditions," BNES Conference on NuclearFuel Performance, Stratford-on Avon, England, March 25-29, 1985.
E-20
APPENDIX F
COMPARISON OF THERMAL/HYDRAULIC DATA
WITH PREEXPERIMENT CALCULATIONS
APPENDIX F
COMPARISON OF THERMAL/HYDRAULIC DATA
WITH PREEXPERIMENT CALCULATIONS
This appendix presents comparisons of the data with preexperiment
thermal/hydraulic calculations for the entire system using the computer
code RELAP5/MOD2,F-I'a and for the core thermal response using a special
version of TRAC-BDi. F 2 ,b The major features of the codes and the
respective input model descriptions are summarized in Section 1. Several
sets of calculations, employing a range of values for initial and boundary
conditions, were performed as part of the planning analyses and are
described in some detail in the Experiment Prediction Document (EPD).F- 3
None of the calculations employ initial and boundary conditions identical
to those that applied in the experiment itself, since the experiment
operation was finally specified after consideration of the results of the
planning analyses. The comparisons described in this section relate to the
most recent calculations, on which the experiment operation was based and
which provide the closest agreement (among the calculations) with the
initial and boundary conditions. The effect of variations in initial and
boundary conditions is discussed in the EPD in relation to sensitivity
studies based on a range of operating conditions. The comparisons between
predictions and measured data are described in Section 2.
a. This analysis was performed using RELAP5/MOD2 Cycle 36, a productionversion of the RELAP5/MOD2 code which is filed under INEL Computer CodeConfiguration Management (CCCM) Archival Number A05844.
b. TRAC/BDI modified Version 8, INEL CCCM Archival Number F01498.
F-l
1. DESCRIPTION OF CALCULATIONAL METHODS
This section presents comparisons of the experiment data with
preexperiment thermal/hydraulic calculations using RELAP5/MOD2 and TRAC-BDI.
1.1 RELAP5/MOD2 Computer Code
RELAP5/MOD2 is an advanced, best-estimate computer program developed
at the Idaho National Engineering Laboratory (INEL) for the analysis of
Loss-of-Coolant Accident (LOCA) and other PWR transients. The specific
application of the code to the Experiment LP-FP-2 prediction is discussed
in this section.
1.1.1 RELAP5/MOD2 Description
RELAP5/MOD2 employs a finite-difference fluid cell representation of
the primary and secondary coolant systems. The six-equation hydrodynamic
formulation employs separate equations to describe the conservation of
mass, momentum, and energy for liquid and steam within each fluid cell.
The description of the hydrodynamics is essentially one dimensional within
each fluid cell. The inclusion of a simplified treatment of the
conservation of momentum in the direction perpendicular to the main stream
flow, where cross flow occurs between parallel volumes and in branches,
brings a special treatment of two-dimensional effects.
Descriptions of the hydrodynamics of choked flow, stratified flow, and
abrupt area changes are carried out with special process models. Special
models are included for simulation of particular components, such as pumps
and accumulators. Flow-regime-dependent constitutive equation and heat
transfer packages are incorporated to complement the hydrodynamic
description. Conduction of heat within metalwork and fuel rods is
calculated with a one-dimensional (two-dimensional in fuel cladding for a
reflooding simulation) finite difference formulation. An extensive control
and trip logic capability is built into the code.
F-2
1.1.2 RELAP5/MOD2 Input model for Experiment LP-FP-2
The nodalization used in RELAP5/MOD2 for this calculation is based ona
a standard LOFT nodalization, with changes that were necessary to
represent the particular system configuration for Experiment LP-FP-2.
Several updates were made to overcome code difficulties encountered or to
better represent phenomena expected to occur during Experiment LP-FP-2.
The final version of the nodalization model used for the calculations
presented in this report is shown in Figure F-l.
The nodalization differs from the standard RELAP5 LOFT model in the
following aspects:
1. The broken loop hot leg pump and steam generator simulators and
the quick-opening blowdown valve were replaced by a pipe
simulating the LPIS break line with two valves attached at both
ends.
2. The quick-opening valve on the broken loop cold leg and its
connection piping to the cold leg were deleted. The broken loop
cold leg is a dead end volume.
3. The reactor vessel was extensively remodeled to represent the
special core configuration and to better simulate the flow
splitting and mixing. Special emphases were given to peak
cladding temperature behavior in the center and peripheral fuel
assemblies and also to the thermal responses of the guide tubes,
control rods and thermal shroud surrounding the center fuel
assembly.
a. The standard LOFT input model Version 131 was used as the basis for theinput deck for Experiment LP-FP-2. The model is continually being updatedand improved. However, complete traceability of each version is maintainedin the model and by the LOFT Program Division.
F-3
Steam valve leak--,I A
r I - 116 114 11--2 Intact loop hot leg 720 261.€= Broken
1 Pum Intact loop cold log 03 22Boe
- 13S•T 14i4 7 32 118 P16 p ,10 1 1 2180 185 2 230 72236
Fir 160. 182 d"[-1120 • .. BSB-3 break -7
Accumulator 222 23
q-1- 600 214 215
1 1zI zII I
Figure F-1. RELAPS/MOD2 nodalization diagram.
( ( (
4. The cross flow model was applied to the junctions connecting the
cold legs to the vessel and to the junction connecting the
pressurizer to the intact loop hot leg.
5. Although the reflood phase was not simulated, the emergency core
coolant (ECC) system and its injection locations (one into the
lower plenum and the other into the downcomer) were shown for a
complete system nodalization.
6. The blowdown piping was attached to the ILCL leg with a
nodalization similar to that used in the Experiment LP-SB-3
prediction calculation.F4
The input model contains a total of 43 fluid cells for the vessel and
95 cells for the remainder of the primary and secondary systems. Detailed
models were developedF-5 to better simulate the flow splitting (from the
lower plenum into the channels representing the peripheral and center
assemblies) and flow mixing (in the upper plenum). Seven fluid cells are
used to represent the lower plenum and lower core support structure. The
flow splitting is represented by cross flow junctions between the cells at
three elevations below the core. The diffuser plate is specifically
modeled by the junctions connecting Volumes 222 and 224 and Volumes 223
and 225. The detailed upper plenum model specifically considers the mixing
in the upper end box represented by Volumes 240 and 241 with a cross flow
junction between these volumes. The mixing between the flows from the
center assembly and the peripheral assemblies below the 5.69-m (224-in.)
elevation, as referenced to the bottom of the reactor vessel, is also
modeled by the cross flow junction between Volumes 245 and 246. No mixing
is allowed between Volumes 252 and 253 due to the geometry of the upper
plenum between the 5.69-m (224-in.) elevation and the nozzle level.
The nozzle area is modeled by four fluid cells. The hot legs are
connected to Volume 250. The split downcomer approach was chosen
especially to simulate the effect of liquid level in the downcomer on the
void distribution in the core at the time of ILCL break valve closure. The
two downcomer channels are horizontally connected with cross flow junctions
F-5
at five different axial elevations. The core is divided into two channels,
each containing six axial fluid cells of equal length. The channels are
hydraulically isolated. The thermal shroud, which is represented by a heat
structure, is the thermal link between the two core channels. The leak
path between the upper downcomer annulus and the upper plenum is modeled by
a cross flow junction connecting Volumes 700 (upper downcomer annulus)
and 256 (upper portion of the nozzle area above the peripheral bundles).
The eight hot rods in the center assembly and the remaining 9.74%
enriched fuel rods are represented by two heat structures. The 10 guide
tubes and 11 control rods are separately represented by two heat
structures. The fuel rods in the peripheral assemblies are represented by
two heat structures. One structure represents the four rows of rod groups
surrounding the thermal shroud outer surface. The remaining fuel rods are
represented by the second heat structure. The control rods in the
peripheral fuel assemblies are not simulated.
Other input features include the containment modeled as a
time-dependent volume. The decay heat power was based on ORIGEN2
calculations specifically performed for the peripheral and center fuel
bundles. A burnup of 500 MWD/MTU was used for the calculations. The
RELAP5/MOD2 code does not include a metal-water reaction model. However,
as the fuel rod cladding temperature rises, metal-water reaction becomes an
increasingly important, and eventually the dominant heat source.
Therefore, a metal-water reaction model was included using the RELAP5
control system.
Heat generation was calculated using the Cathcart-PawelF-5 model for
cladding temperature in the range 1273 to 1853 K (1832 to 2876°F) and the
UrbanicF-5 model for cladding temperatures above 1853 K (2876°F). A
steam limitation model was included to account for the steam availability
for the reaction. Weaknesses as of the model are (a) the energy generated
by the metal-water reaction is deposited in the body of the fuel rods
rather than in the cladding surface, (b) no hydrogen is generated, and
(c) the center assembly flow is required to be positive. The metal-water
reaction was also calculated on the cladding of the guide tubes and the
F-6
inner surface of the thermal shroud. These models were included in the
input deck and can be seen in Appendix B of the LP-FP-2 Experiment
Prediction Document. F-3
The calculation was initiated from a power level of 25 MW, which
corresponds to a calculated peak power density in the 9.74 wt% enriched
fuel rods of approximately 38.8 kW/m (11.8 kW/ft). The important initial
conditions are presented in Table F-1. All of the initial conditions used
in the calculation are close to the experimental values, with the exception
of core bypass. The modeled bypass results in a calculated initial bypass
flow of about half the estimated value for the experiment; however, this
exception is judged to have only a minor effect on the LP-FP-2 transient.
1.2 TRAC-BDI Computer Code
The TRAC (Transient Reactor Analysis Code) is an advanced best
estimate system analysis computer program designed for the analysis of
postulated accidents in light water reactors. TRAC-BDl is designed
primarily for the simulation of design basis LOCAs and transients in
boiling water reactors (BWR). This version of the code is suitable for
analysis of the thermal behavior of the LOFT core with two separated flow
channels during Experiment LP-FP-2.
1.2.1 TRAC-BOl Description
Unique features of the code include (a) a full nonhomogeneous,
nonequilibrium two-fluid thermal-hydraulic model of the two-phase flow in
all portions of the BWR system and (b) a detailed model of BWR fuel
assemblies, which includes a radiation heat transfer model for thermal
radiation between multiple fuel rod groups, inner surface of the fuel
channel wall, and liquid and steam phases within the bundle.
Extensive development work was carried out at the INEL to improve the
TRAC-BDI computational capabilities to more closely predict the LOFT core
thermal response during Experiment LP-FP-2. The detailed description of
F-7
TABLE F-1. INITIAL CONDITIONS FOR EXPERIMENT LP-FP-2
Parameter
Core power
Maximum linear heatgeneration rate
Core AT
Primary systempressure - hot leg
Intact looptemperature
hot leg
cold leg
Primary coolant massflow rate
Units
MW
kW/mkW/ft
KOF
MPapsia
KOF
KOF
kg/salbm/h
Measured Calculated
26.8
42.612.97
11.721.1
14.982173
571.6569.2
559.9548.2
25.0
38.811.8
9.917.8
14.902161
567.8562.4
557.9544.5
475.03.77x106
476.83.78xi06
Total bypass flow kg/s a 28.0Ibm/h 2.2xi05
Pressurizer liquid m 1.12 1.12level in 44.1 44.1
Steam generator MPa 6.38 6.06secondary pressure psia 925 879
a. Not measured but estimated to be approximately 63 kg/s (5 x 1O5 ibm/h).
F-8
the models (the first five items indicated below) developed has already
been published in Reference F-6. The models which are important for this
calculation are:
o LOFT full core radiation model
o Thermal shroud model, which thermally links the two separated
flow channels
o Extended metal-water reaction model
o Steam limitation model
o Conductance across the gap between the guide tube and control rod
0 Capability of starting post dryout heat transfer at a given time
and axial elevation, as defined in input.
The sixth model indicated above is built to start the postdryout heat
transfer calculation at a given axial elevation and a time. This is a
necessity due to (a) deletion of several code models (as indicated below)
which must be used for a boil-off simulation and (b) the simple input model
developed for the simulation which allows only a heatup calculation. The
model employs a 20,000 W/m2K liquid heat transfer coefficient for the
nucleate boiling heat transfer regime so that the initial temperature at
that elevation can be maintained steady until the transient time passes the
given time. The code is then allowed to pass into the postdryout heat
transfer calculation.
The metal-water reaction model developed considers the reaction
occurring on the rod groups simulating fuel or control rods and on the
inner surface of the thermal shroud. The lack of heat source due to
metal-water reaction on the outer surface of the shroud will yield a
slightly lower temperature excursion on this surface and also on the
peripheral bundle fuel rods surrounding the thermal shroud because of the
radiation heat loss to the shroud.
F-9
In addition to the models above, numerical stability of the
calculation was further enhanced by improving the calculational strategy of
the code.
All of the code models that were not of use for this analysis, such as
jet pump model, valve model, etc., were removed to offset the extra
dimensions required for the radiation model. New material properties were
added to analyze the control rods, guide tubes, and the thermal shroud
responses. This special version of the code, called TRAC-LOFT, was used to
calculate the thermal behavior of the LOFT core.
1.2.2 TRAC-LOFT Nodalization for Experiment LP-FP-2
The TRAC-LOFT nodalization used for this calculation is presented in
Figure F-2. The LOFT core is represented by two CHAN components (a special
TRAC model for the simulation of a BWR assembly), one representing the
center assembly, and the other representing the peripheral assembly. The
thermal shroud is modeled by the wall of the CHAN component representing
the center assembly. Boundary conditions at the outer surface of the
thermal shroud wall are taken from the CHAN component representing the
peripheral assembly. Both CHAN components are divided into six equally
spaced axial volumes for consistency with the RELAP5/MOD2 nodalization.
Two flow boundary conditions (FILL components) simulating the steam flow
entering the center and peripheral assemblies are separately modeled. The
outlet flows from the CHAN components are mixed in a TEE component. A
pressure boundary condition at the core exit is provided with a BREAK
component attached to the TEE component. The wall of the peripheral
assembly CHAN component represents the core filler plates and the flow
shroud.
The fuel rods, control rods, and guide tubes are represented by rod
groups arranged by rod powers and geometry. The individual members of a
rod group are assumed to be equivalent to all other members of the rod
group. The model used for this analysis employs three groups for the fuel
rods in the center assembly. These additional groups are used for the
center assembly guide tubes and control rods as follows: the center guide
F-1O
LOI-KMI116-03
Figure F-2. TRAC-LOFT nodalization.
F-li
tube, by one group;the other 10 guide tubes, by one group; and the guide
tubes with the control rods, by one group. The peripheral assembly rods
are grouped into nine rod groups according to their radial peaking
factors. Due to excessive computer memory requirements, the guide tubes
with and without the control rods in the peripheral assemblies are not
modeled. The rod grouping established for the center and the peripheral
assemblies are presented in Figures F-3 and F-4, respectively. In addition
to the thermal/hydraulic boundary conditions required from the RELAP5
calculations, an adiabatic boundary condition is assumed on the outside
surface of the wall belonging to the peripheral assembly CHAN component.
This assumption added some conservatism to the predicted temperature
transient in the peripheral assemblies. A listing of the TRAC-LOFT input
model is provided in Appendix C of the LP-FP-2 Experiment Prediction
Document. F-3
F-12
1ý
C (
S66 6666X66633333_ 3333663 3 3333 3 6
63311111 336331 1T1 13366 _311 :iT3 6
66331 1 11 13363311111 1 33
63 33333 3663333 33336
66 666 M 66
l W E-6] Fuel rods(24) (48) (28)
MM Guide tubes(1) (9)
• Control rods(l1)
"11
(A
LOI-KM116-04
Figure F-3. TRAC-LOFT center assembly rod grouping.
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Figure F-4. TRAC-LOFT peripheral assembly rod grouping.
F-14
2. COMPARISON BETWEEN PREDICTION AND EXPERIMENT DATA
This section presents a comparison between pre-experiment predictions
and experiment data. The RELAP5/MOD2 and TRAC-BOI calculations, described
in some detail in the Experiment Prediction Document (EPD), were terminated
shortly prior to reaching the criterion for initiation of reflood.
The initial power of 25 MW used in the calculations is sufficiently
close to the experimental value of 26.8 MW to provide meaningful comparison
with the data. Other differences between the calculated and actual
operating conditions include a slightly lower primary system initial
temperature and secondary side pressure, a different experiment initiation
sequence and trip times for the calculation , and reopening of the cold leg
break and operation of the PORV to assist the depressurization. The
probable effect of these differences on the calculations are considered in
the discussion that follow. Comparison between the predicted and measured
data for system hydraulic response is presented in Section 2.1. A
discussion of the comparison for core thermal response is presented in
Section 2.2.
2.1 Comparison of System Hydraulics
The measured and predicted secondary system pressure during the first
400 s are shown in Figure F-5. The measured steam generator secondary
pressure, after termination of feedwater and steam flows, increased to the
main steam valve cycling setpoint of 7.11 MPa (1031 psia) at 56 s compared
with 50 s predicted. The pressure increase and time of steam valve cycling
were correctly predicted, with some minor differences due, possibly, to the
slightly different initial conditions and experiment initiation sequence.
The secondary system continued to act as a heat sink until the primary
pressure had dropped below the secondary pressure. This was predicted at
224 s compared with the observed time of 260 s. Figures F-6 and F-7 show
the measured and predicted secondary side pressure for the entire
transient, and the collapsed level in the steam generator secondary. The
rates of depressurization and liquid depletion were overpredicted due to an
F-15
8 I I I I I I
7
00.v
L
C-5
-- PT-PO04-O1OA--- RELAP5/MOD2
# I I I I I
-1000 -"0Ut)-.7
-900:3'I,
-800 6
a)
-700
-600
-1100
40 50 100 150 200 250
Time (s)300 350 400
Figure F-5. Secondary system pressure (0 to 400 s).
8
7a
C-,
6
:3_
5
4
PTPO4-10
RELAP5/MODLJ 1100
-1000
-900
-800
-700
-600
aU,
0.
4)L.:30U,U)1~
0~
0 400 800 1200Time (s)
1600 2000
Figure F-6. Secondary system pressure (0 to 2000 s).
F-16
C" 2.5 o
.7
3 .5I I I I I I I I I
-200 0 200 400 600 800 1000 1200 1400 160o 1800Time (S)
Figure F-7. Secondary system liquid level.
F-17
overestimate, based on data from Experiment LP-SB-3, of the leak rate
through the main steam valve. The leak rate tends to differ each time the
the valve is cycled and can vary by a factor of two or more.
Figure F-8 shows the measured and predicted primary system pressures
for the first 400 s of the transient. Both curves show a slight drop in
pressure following scram and a subsequent rapid decrease down to saturation
pressure following break initiation. A minor interruption in the early
depressurization observed in the data at 37s does not correspond to any
operational or measured event. Examination of the predicted cold leg breakfluid conditions indicates the possibility that liquid that was subcooled
in the cold leg flashed to steam as it depressurized while accelerating
through the break line. The end of subcooled blowdown was predicted at
42 s, compared with 53 s indicated from measurements. A slightly lower
pressure was reached in the calculation due to the slightly lower initial
fluid temperature. The predicted pressure response agreed well with data
for the period until initiation of the LPIS line break at 220 s, except for
minor differences that were a direct result of different initial conditions
and initiation sequence. Figure F-9 compares the measured and predicted
pressure for the transient to 2000 s. In contrast with the good agreement
for the period prior to LPIS line break initiation, the subsequent
depressurization rate was considerably overpredicted. The LPIS line and
break characteristics had previously been considered to be a major source
of uncertainty. An attempt was made to estimate the effect of the
uncertainty by performing a sensitivity calculation with the break flow
areas reduced by 30%. This provided slightly better agreement, but still
overpredicted the depressurization rate. Because the cold leg break area
was also reduced, and the decay heat levels were considerably higher (due
to the initial power of 33 MW used), the depressurization rate should have
been underpredicted given reasonably representative modeling of the LPISline and break. From this it is concluded that the straight pipe volume
used to represent the LPIS line, which in reality contains numerous bends,
does not provide a realistic representation of the line. The closure of
the cold leg break resulted in almost no change in the predicted pressure
response, yet the data show that this action terminated the
F-18
15
a-
a.
10
5
0 I I
0 50 100 150 200 250Time (s)
0
0.
'I,1...:3
In0La-
300 350 400
Figure F-8.
15
a-
CL
Primary system hot leg pressure (0 to 400 s).
II I
PE-PC-002 2000--- RELAP5/MOD2
(25 MW, 100 pct break flow area)-- RELAP5/MOD2
(33 MW, 70 pct break flow area)1500 0
-1000
-50L
• \ -500
500 1000 1500 2000Time (s)
Primary system hot leg pressure (0 to 2000 s).
10
5
00
Figure F-9.
F-19
depressurization until the cold leg break was reopened and the PORV was
operated. It is postulated that the complicated network of bends in the
LPIS line resulted in a higher flow resistance under single phase
conditions and also inhibited the draining of liquid from the line under
two phase conditions. There is indication from measurements of fluid
temperature that the LPIS line was not completely drained until after about
1200 s. The latter effect differs from the prediction in which the LPIS
line was completely void after about 350 s. The venting of steam,
calculated by the code, would not readily take place with liquid remaining
in the line. The higher system pressure observed and the use of the PORV
and cold leg break to effect depressurization affects all the comparisons
of system hydraulics and core thermal response beyond 300 s.
The fact that the primary system pressure was much higher during the
heat up and core damage phase means that there was a much greater driving
head to sustain the break flow. The measured and predicted differential
pressures between the PCS and BST are compared in Figure F-l0 for the
period from 1200 to 1700 s. The measured values for the pressure drop were
in the range 1.0 to 1.3 MPa (145 to 189 psia), compared with predicted
values of 0.15 to 0.25 MPa (22 to 36 psia). The LPIS line flow calculated
from measured variables and the predicted flow are compared in Figure F-ll
for the same time period, for which the flow of single phase vapor was both
predicted and indicated by measurement. The experimental values for mass
flow were approximately twice the predicted values. This result is
consistent, qualitatively at any rate, with the fact that the measured PCS
depressurization rate was about 2.5 times the predicted rate during this
time frame. The rate of vapor generation is approximately proportional to
the depressurization rate for a given liquid mass, provided the heat input
from the fuel and metalwork is small, as was the case with all or most of
the core uncovered. Also, the estimated steam flow rate in the center fuel
assembly of 0.04 kg/s, obtained from analysis of the core thermal response
presented in Section 3 of this report, exceeds the predicted center
assembly flow by about a factor of 2.5. Despite the mass flow in the LPIS
line being higher than predicted at this time, the momentum flux, as
F-20
Ký1 A 2
1.75
1.50
T
0
CL
1.25
1
0.75
0.50
0.25
III I
RELAP5/MOD2 -2:
21
-5-
i I i 1 10
5o0
00
0.
0~
01200 1300 1400 1500
Time (s)1600 1700
Figure F-1O. Pressure drop along LPIS line.
0.5
0.4 1-
(#7
0
M
0.3
- -- RELAP5/MOD2)-1.0
L0.8
E0.6
0
0.4
0.2
0.0
0.2
0.1 t
0.01000 1200 1400 1600
Time (S)1800 2000
Figure F-11. LPIS line mass flow rate.
F-21
calculated from measured variables was similar. Therefore, the effective
flow resistance between the PCS and BST was greater than the predicted
value by a factor of about six.
The areas of agreement and disagreement between the predicted and
observed system hydraulic response indicate that most of the important
features of the primary and secondary systems were correctly modeled, the
major exception to this being the LPIS line flow characteristics. Some
differences between predicted and measured data are a result of differences
in operational initial and boundary conditions.
2.2 Comparison of Core Thermal Response
The major objectives in performing the pre-experiment analyses were
(a) the assessment of the achievability of the required hydraulics and core
thermal response and (b) of maximizing the likelihood of achieving them.
In order to make as good an assessment as possible of the core thermal
response, system calculations were carried out using RELAP5/MOD2, in
conjunction with estimates for blockage in the center fuel assembly. Known
limitations in the capability of RELAP5/MOD2 to model the core thermal
response with the necessary accuracy and detail meant that these objectives
could not be met by using RELAP5/MOD2 alone. Therefore the core hydraulic
conditions calculated by RELAP5/MOD2 were input into a special version of
TRAC-BDI. Certain aspects of the RELAP5/MOD2 calculated hydraulics
(negative core flow at certain times) were judged to be unphysical and were
suppressed in the TRAC-BDI input in order to seek a more realistic
representation. The comparisons presented in this section include both the
RELAP5/MOD2 and TRAC-BDI calculations. However, the comparisons with
RELAP5/MOD2 are considered to be less meaningful and of limited
usefulness. Differences between experiment and calculation in respect of
operation, and in the primary system hydraulics have a bearing on any
conclusions that may be drawn from the comparisons and should be borne in
mind by the reader. The reader should also recall that many of the
cladding thermocouple readings are not qualified during the portion of the
transient at which high temperatures occurred. The time after which the
data is no longer qualified is indicated on the figures.
F-22
Figure F-12 shows the progression of core uncovery in the center and
peripheral fuel assemblies, as indicated by the observed initiation of heat
up and as predicted by RELAP5/MOD2. The times of initiation of uncovery
and for uncovery of the whole core were quite well predicted, remarkably
so in view of the degree of disagreement in the system hydraulics. The
data did not, however, exhibit the significantly later uncovery of the
center fuel assembly (compared with the peripheral assemblies) that was
predicted. The observed progression of uncovery also differed from
prediction in the following respect. The cold led break was closed in the
experiment at 735 s, compared with 1007 s in the calculation. Although the
the predicted and observed rates of uncovery are similar up to 735 s, the
data show an increase in the core uncovery rate after 735 s down to the
0.38 m (15 inch) elevation, followed by a very slow uncovery shortly
thereafter. It is hypothesized that closure of the cold leg break, in
terminating the system depressurization, caused a sharp reduction in the
rate of vapor generation and thereby brought about a total or partial
collapse of the froth level in the vessel. The system pressure then
remained almost constant until the cold leg break was reopened at 878 s, sothat the continuation of core uncovery until then was solely dependent on
heat input from the fuel and metalwork. As a result, the remainder of the
core uncovered extremely slowly, 180 s elapsing between the initiation of
heat up at the 0.38- and 0.28-m (15- and 11-inch elevations).
Figure F-13 presents the measured cladding temperatures at the 0.25 m
(10 inch) elevation in the center fuel assembly with the prediction at the
nearest modeled location. The average temperature rise rate until 1700 s
was observed to be about 0.5 K/s (0.9°F/s). The rise rate was
overpredicted for most of this period in both calculations, TRAC-BDI
providing the better agreement with 0.7 K/s (l.3 0 F/s) compared with the
RELAP5/MOD2 prediction of 1.0 K/s (l.8 0 F/s). The better agreement using
the TRAC-BDI was due in part to the effect of finer (13-node) axial
nodalization on fluid conditions. The local power factors were similar in
the two calculations. The overprediction was contrary to the fact that the
modeled decay heat level in this node was lower than that which existed at
the measurement location because of the lower elevation of the node used in
the model. The underprediction of mass flow of steam through the core is
F-23
ILCL closed (736 a)
0W
w
160
140
120
100
80
60
40
20
0
N( 4"
>0(
.
I II IILCL open (878 $) PORV open (882 s)
&I
x
x
v Center measured* Center calculatedx Peripheral measured* Peripheral calculated
60
50
40
30
20
10
0)00
0
S,W,
10600 700 800Time (s)
900
LIIO-KM135-01
Figure F-12. Progression of core uncovery.
.- I-
E E'0
400 ' I -600 800 1000 1200 1400 1600 1800 2000
T i me (s)
Figure F-13. Fuel rod cladding temperature in center fuel assembly,0.25-m (10-in.) elevation. (Thermocouple qualifiedthroughout). ,
F-24
believed to have resulted in an underprediction of heat transfer
coefficient. The observed increase in temperature rise rate at 1700 s
occurred at too low a temperature, about 900 K (1161°F), to be the result
of rapid metal-water reaction at this location and was not predicted. The
observed behavior may be the result of thermal radiation from high
temperature material at higher elevation or to material relocation.
Neither thermal radiation in the axial direction nor the direct effect of
material relocation on local temperature is modeled.
Figure F-14 compares the fuel rod cladding temperature measured at the
0.69 m (27 inch) elevation in the center assembly with the corresponding
predicted temperatures. Good agreement with the initial heat up rate of
2.2 K/s (4.0 F/s) was obtained in both calculations, during the period
prior to PORV operation and reopening of the cold leg break. The observed
temperature rise rate then decreased due, apparently, to flashing of liquid
in the lower plenum induced by the depressurization. The rise rate was
then overpredicted for the remaining period until about 1630 s when the
metal-water reaction initiated at this elevation. The TRAC-BDI calculation
achieved good agreement with the rate of temperature rise during the time
of metal water reaction, despite the fact that the heat up rate was
observed to increase from a lower temperature than predicted. This, again,
may be due to processes not calculated.
Figure F-15 shows the measured fuel rod cladding temperature at the
1.07 m (42 inch) in the center assembly with the nearest corresponding
calculated temperatures (0.84- to 1.12-m (33- to 44-inches) elevation).
The average rate of temperature rise was observed to be about 1.3 K/s
(2.3 F/s) until 1450 s ( after which the temperature increased very rapidly
due to the metal-water reaction). To this'point in time the temperature
increase was fairly accurately predicted in both calculations despite the
difference in system hydraulic conditions between calculation and
experiment. The observed temperature rise rate increased rapidly after
1450 s, at which time the cladding temperature was about 1500 K (2240°F),
to about 22 K/s (40°F/s). Only a small increase in temperature rise rate
was predicted because the comparatively low mass flow rate in the center
assembly, about 0.015 kg/s (0.033 Ibm/s), resulted in steam limitation at
F-25
2500
1500
.4,-
L
CD 10000.
EIV
500
0600
Figure F-14.
3000
2500 L- 2000CD
i 1500
E 1000I--
CD
-2000 =
-1000 E
0.0
II I ,I I I
800 1000 1200 1400 1600 1800 2000Time (s)
Fuel rod cladding temperature in center fuel assembly,0.69-m (27-in.) elevation. (Thermocouple qualified to1720 s).
w
a,
-4-
0L.a,a.2a,I-
500
0Boo
Figure F-15.
800 1000 1200 1400 1600 1800 2000Time (s)
Fuel rod cladding temperature in center fuel assembly,1.07-m (42-in.) elevation. (Thermocouple qualified to1510 s).,
, F-26
this elevation. The maximum cladding temperatures that were measured
occurred at this elevation, whereas the maximum predicted temperatures
occurred at the 0.56- to 0.84-m (22- to 33-inch) elevation.
The maximum cladding temperatures measured in the peripheral fuel
assemblies occurred at the 0.66 m (26 inch) elevation on fuel rods adjacent
to the insulating shroud. Figure F-16 shows the temperature history
recorded by thermocouple TE-2H15-026 together with the corresponding
temperatures predicted by RELAP5/MOD2 and TRAC-PDI. The predictions were
in good agreement with the data until about 1600 s, after which the
cladding temperature increased more rapidly. A similar temperature rise
was observed on many of the peripheral fuel rods and was particularly"
noticeable at the lower elevations. This phenomenon was not calculated,
and a completely satisfactory explanation has not, as yet, been found. It
is possible that the observed behavior is the result of the thermocouple
forming a new junction at a hotter location. The temperature measured on
the outer wall of the shroud at the location close to TE-2H15-026 and the
temperature calculated by TRAC-BDI are shown in Figure F-17. The
calculation using RELAP5/MOD2 underpredicted the shroud temperature
measured at this location and also those measured at the 0.81 m (32 inch)
1.07 m (42 inch) elevations due to the lack of a model for thermal
radiation, an important mechanism controlling the temperature rise of
unheated structures. TRAC-BDl gave much better agreement, with a slight
overprediction in temperature rise rate for most of the transient. The
increase in measured temperature rise rate may have been due to
deterioration in shroud insulation and was not calculated. Both data and
calculation with TRAC-BOl show that the peripheral fuel rod temperatures
tended to be somewhat higher than the shroud temperatures, except at the
elevations at which the highest temperatures occurred in the center fuel
assembly. Heat conduction through the shroud raised the outer wall
temperatures at these elevations to above those of the nearby peripheral
assembly fuel rods. The predicted relationship between the center and
peripheral fuel rod temperatures and the shroud temperatures was in good
agreement with the data. As a result, the time above 2100 K (3321'F) in
the center bundle (about 270 s) was very close to expected time of
F-27
2000
-S
5-
V
L4-
L.4~a-EV
I-
1600
1200
800
400 1--600
Figure F-16.
-. 2000
-1500 LW0.
E1000
"500
800 1000 1200 1400 1600 1800 2000Time (s)
Fuel rod cladding temperature in peripheral fuel assembly,0.66-m (26-in.) elevation. (Thermocouple showed possibleshunting after 1700 s).
/,/•2000- TE-5S-027 /- RELAP5/MOD2 /
TRAC-BD1 -
1500
1000
CL•- -1000 0, oa.
-. E
500
I I I I I _ _ _ _
800 1000 1200 1400 1600 1800 2000T i me (s)
17. Shroud outer wall temperature at 0.69-m (27 in.)elevation. (Thermocouple qualified to 1790 s).
-S
VI.-
.6-a1~00~E0
I-
1400
1200
1000
80
I.
600
400600
Figure F-
F -28
280-300 s based on the predictions. The programmatic experiment
termination criterion was reached on the shroud outer wall temperature, as
had been predicted.
The core thermal response during Experiment LP-FP-2 was, in general,
fairly accurately predicted by TRAC-BOI for the period prior to metal-water
reaction. Fairly good agreement was also obtained for the period during
the metal-water reaction, within the limited capability of the codes to
model the processes that take place at the high temperatures. However, it
should be remembered that many of the cladding thermocouple readings are
not qualified for this latter portion of the transient. In view of
discrepancies between calculated and measured system hydraulics, the extent
of the agreement is somewhat surprising and may be to some extent
fortuitous. Several aspects of the predictions that differ from the data
can be traced to known limitations in the modeling, for example thermal
radiation in the axial direction and effects of material relocation. Other
areas of disagreement result from differences between the predicted and
measured system hydraulics. Of these, the most significant effect was the
prediction of steam limitation at the higher elevation, in contrast to the
observed behavior. Also, the temperature rise rate prior to metal water
reaction tended to be slightly overpredicted at many locations due,
apparently, to the predicted steam flows being lower than those obtained in
the experiment.
F-29
REFERENCES
F-1. V.H. Ransom and R.J. Wagner, RELAP5/MOD2 Code Manual Volume 1:Code Structure, System Models, and Solution Methods, EGG-SAAM-6377,April 1984.
F-2. J. W. Spore et al., TRAC-BDl: An Advanced Best.Estimate Computer Codefor Boiling Water Reactor Loss-ot-Goolant Accident Analysis,NUREK/CR-2I/8, EGG-2109, October 1981.
F-3. S. K. Guntay et al., Best Estimate Prediction for OECD LOFT ProjectFission Product Experiment LP-tP-Z, UtLU LUFI-I-JdU3, June I98b.
F-4. M. Tanaka et al., Quick Look Report on OECD LOFT Experiment LP-SB-3,OECD LOFT-T-36U4, March 1984.
F-5. D. L. Hagrman et al., MATPRO-Version 11 (Revision 2), A Handbook ofMaterials Properties for use in the Analysis of Light Water ReactorFuel Rod Behavior, NUREG/CR-0497, TREE-1280, Rev. 2, August 1981.
F-6. G. A. Dineen et al., LP-FP-2 Supplement to the LOFT Integral TestSystem Final Safety Analysis Report, OECD LOFT-I-ll-5113,December 1984.
F-30
APPENDIX G
SPECIAL INSTRUMENTATION
APPENDIX G
SPECIAL INSTRUMENTATION
Experiment LP-FP-2 represented a unique opportunity to study
PWR response to a very severe nuclear accident. Because of this, specially
designed instrumentation was installed in the reactor vessel and in the
containment to take advantage of this opportunity. This instrumentation
was in addition to and independent of that installed by the project for
measuring the thermal, hydraulic, and fission product transport response of
the system. The instruments were: Three Mile Island (TMI)-design
rhodium-emitter self powered neutron detectors (SPNDs) installed in the
central fuel module; a multi-stage iodine species sampler; and nuclear
detectors installed in the reactor vessel shield tank. The information
contained in this appendix came from References G-l, G-2, and G-3 for the
SPN~s, iodine species sampler, and nuclear detectors, respectively. As is
the case with all data from this experiment, these results are preliminary;
in-depth analyses of these data will be reported in the future.
TMI Type SPNDs
The TMI core contained 364 SPNDs which were installed in 52 guide
tubes to measure the local power densities. All of these transducers were
monitored by the plant computer with data being recorded only when they
changed states. In addition to the plant computer, signals from 36 of
these detectors were recorded on two multipoint back-up recorders. The
computer data were lost during the time interval corresponding to the
initial core heatup of the November 1979 accident. Thus, the only SPND
data for this time interval came from the two back-up recorders.
The SPNDs responded normally to the reactor scram early in the
transient. At 2 h 15 min into the accident, some of the SPNDs produced a
negative signal (of unmeasured magnitude), indicating that they were being
heated. Fifteen minutes later, the polarity of several of these detectors
switched to positive. Eventually the output current exceeded the upper
G-1
recording range of 1500 nanoamps. Since the nuclear reaction had been
terminated at the beginning of the transient (more than 2 h prior to
generation of the large positive signals), the residual neutron flux was
much too low to account for this large positive signal. Thus, the signals
are generally judged to have been caused by a combination of high
temperatures and the physical and chemical environment surrounding the
SPNDs. In an effort to determine after the fact what the core temperatures
were during the accident, attempts were made in the laboratory to reproduce
these large positive signals. Those attempts, reported in References G-4
through G-7, failed to reproduce the phenomena even though temperatures up
to 1700 K (2600*F) and gamma radiation fields up to 2 x l05 R/h were
produced in laboratory testing.
A vertical string of 4 TMI-design SPNDs were installed in the LOFT
center fuel module for Experiment LP-FP-2. These detectors contained a
rhodium emitter and were encapsulated in a zirconium alloy sheath. They
were located in the center guide tube (location H-8) and their sensitive
lengths were centered at 0.28-, 0.69-, 1.11-, and 1.55-m (11-, 27-, 44-,
and 61-in.) above the bottom of the core. The detectors at the 1.11 and
1.55-m (44- and 61-in.) elevations were damaged during installation of the
fuel module into the core and did not operate during the experiment. The
two lower SPNDs remained intact and functional during the transient.
The response of the SPND at the 0.69-m (27-in.) elevation to
Experiment LP-FP-2 is shown in Figure G-l. The detector output was
initially at 340 nanoamps and decayed to near zero within 500 s of scram.
Starting at 1125 s, the output became negative, increasing in magnitude
until 1325 s, when it was -10,O00 nanoamps. The detector output remained
negative at this value until 1425 s, when it suddenly increased to more
than 10,000 nanoamps with positive polarity, remaining at this value for
approximately 20 s. A second large positive pulse occurred at 1450 s,
though for a shorter time. The fuel rod cladding temperatures at the
0.69-m (27-in.) elevation during this time interval were approximately
2200 K (3500°F). This temperature is close to that at which the zirconium
alloy sheath material melts. The output from the SPND at the 0.28-m
G-2
15000 iL. NE-5H8-7 -
o 12000 o
o 10000 0
9000
6000-0.. 6000 " ,----
r < 5000 c<L 3000 L
0-- 0 -
CD -3000 - D-5000
-6000 -0 0
- -900010000
ch -12000 f w
0 500 1000 1500 2000Time (s)
Figure G-1. Response of SPND at the 0.69-m (27-in.) elevation in FuelAssembly 5.
G-3
(11-in.) elevation, shown in Figure G-2, remained negative throughout the
transient (except for a brief positive "spike" during reflood). The
difference in responses between these two detectors is believed to be due
to the much lower cladding temperatures [approximately 1000 K (18000 F)
lower than at the 0.69-m (27-in.) elevation], which persisted at this
elevation.
The following preliminary conclusions can be drawn from these
observations. First, the conditions which existed in the LOFT core during
Experiment LP-FP-2 approximated those of the TMI core during the November
1979 accident. (Previous attempts to duplicate these conditions in the
laboratory had failed.) Second, the temperature at which the SPND output
became positive corresponds approximately to the melting point of the
zirconium sheath material.
Iodine Species Sampler
Samples of reactor containment air were taken and analyzed for
radioactive iodine species using an iodine species sampler. The samples
were taken from the Heating and Ventilating Systems (H&V) 8 and 9. It
should be noted that the repository for most of the fission products
released during Experiment LP-FP-2 was the blowdown suppression tank. The
fission products measured by this iodine species sampler were those which
were leaked from the blowdown suppression tank and primary coolant system
into the LOFT containment. Thus, the fission product concentrations
measured by this sampler were not intended to be representative of those
that would be expected in the containment of a commercial PWR during a
V-Sequence accident. A two-in, diameter line is used to vent the
containment into H&V 8, which is a 10-in, line. The exhaust from H&V 8 runs
into a silver zeolite cleanup filter before it is vented to the
atmosphere. H&V 9 consists of a 24-in. diameter line that exhausts into a
silver zeolite cleanup filter and then is either vented back into thecontainment (recirculation mode) or to the atmosphere (vent mode). The
venting rate was not constant with time. In each case, the samples
described in this appendix were taken from locations upstream of the
cleanup filters.
G-4
3000
L
0-I-()
a,1,
L
L-D ja, 0L0,
0Q.
a,C,,
0
-3000
-6000
-g000
-120000
L.
0"4--
0 C.2D.,-o
z<•
0)
-10000
o )'CA
500 1000 1500 2000Time (s)
Figure G-2. Response ofAssembly 5.
SPND at the 0.28-m (11-in.) elevation in Fuel
G-5
The iodine species samplers used in this experiment were those
described in Reference G-8 and shown in Figure G-3. These consist of five
media: a particulate filter to retain particulate material; cadmium iodide -
on Chromosorb to adsorb elemental iodine; 4-iodophenol on alumina to adsorb
hypoiodous acid; silver zeolite to adsorb organic iodine species; and a
backup filter to adsorb any species possibly escaping the other media.
After each sample was drawn through the iodine species sampler, the sampler
was removed and taken to the laboratory for analysis. A new sampler was
installed prior to taking the next sample. Two samples were taken from
H&V 9 during the power operations prior to initiation of the experiment.
Six samples were taken from H&V 8 and four, from H&V 9 after experiment
initiation. Table G-1 lists the samples and indicates the location and
time of sampling. Table G-l also lists the total radioactive iodine
concentrations for each sample.
Preliminary results for 1311 are shown in Tables G-l and G-2.
Cursory examination of the results indicates the following. Just prior to
initiation of the experiment, when H&V 9 was rapidly purging the131
containment, more than 50% of I was in the form of 12 and 18% was
organic. The sample taken at 3 h 39 min was 70% 129 12% HOI, 12%
organic, and 7% particulate. The particulate and 12 fractions then
decreased with time; the organic fraction increased with time; and the HOI
fraction first increased, peaked, and then decreased with time. All this
occurred as the H&V 8 system was slowly venting the containment.
Ventilation of the containment was changed from H&V 8 to H&V 9 between
samples 6 and 7. The iodine species admixture in these samples are very
similar, indicating that the smaller diameter H&V 8 line was taking a
representative sample and that there was no excessive plate-out. Results
from samples 7 through 10 (taken with H&V 9) indicate that the relative
fractions tended toward the preexperiment sample fractions as the rapid
purge of the containment continued.
Other radioactive iodine species were also measured. The iodine
species admixtures for 1331 and 1351 were similar to that measured
G-6
TABLE G-1. MEASURED 1311 CONCENTRATIONS
Sample
Pre-1
Pre-2
1
2
3
4
5
6
7
8
9
10
H&VSystem
9
9
8
8
8
8
8
8
9
9
9
9
Time After Initiation(h, min)
3
5
7
31
42
68
75
77
138
211
39
50
33
17
55
19
49
40
45
11
Concentration a
(microCi/cc)
1.38 + 0.08 x 10 9
9.7 + 0.8 x 10-10
4.18 + 0.03 x 10-7
1.37 + 0.02 x 1O"5
1.55 + 0.02 x 1O-5
2.84 + 0.02 x 1O-5
7.85 + 0.04 x 10-6
6.62 + 0.03 x 10-6
6.33 + 0.03 x 10-5
1.43 + 0.008 x 10-5
9.85 + 0.05 x 10-7
5.79 + 0.04 x 1O- 7
a. Concentration measured in the H&V system. A dilution factor must beapplied to correct these values to the concentration in the containment air.
G-7
TABLE G-2. MEASURED 1311 SPECIES ADMIXTURE
Species Percent "" V
Sample Particulate12 HOI Organic
Pre-1 6 + 2 55 + 7Pre-2 4 T 2 58 121 6.6 + 0.2 70 +12 2.9T0.1 64 23 2.2 + 0.] 57 + 14 1.04 + 0.01 45.2 + 0.75a 8.0 + 0.1 22.1 + 0.26 0.26 + 0.04 24.1 + 0.27 0.64 + 0.01 23.5 + 0.28 2.13 T 0.04 30.2 + 0.49 3.62 + 0.09 57.0 + 0.7
10 9.2 • 0.2 60.0 + 1.0
a. Results from this sample are suspect.
520
11.816
21.016.55.7
13.79.89.1
21.015.2
+
T++
T++
240.210.40.20.10.20.10.30.30.3
3418
12.017.020.037.364.261.966.158.618.415.1
+5+4
+0.3+ 0.4+0.3* 0.5+ 0.8¥0.6
70.8+0.8+ 0.3¥0.3
G-81, IV
.0 - Particle filter
Elemental Iodine (12)adsorber
..- Hypoiodous acid (HOIW)edsorber
Organic iodineadeorber
Backup charcoaladsorber
P-ST-OO6O-18
Figure G-3. Iodine species sampler.
G-9
for 131I. Results for 1321, however, differed. The predominate
fraction for this isotope was elemental 12 (56 to 71% for all samplestaken after experiment initiation) and the organic fraction remained
relatively low (10 to 28%). Thus, this isotope did not exhibit a reduction
in the elemental fraction with time with a corresponding increase in the
organic fraction, as was the case with the other three isotopes.
Non-Intrusive Level Transducers
The final part of this appendix deals with a string of nuclear
detectors that were installed in the reactor vessel shield tank. These
detectors, which are very sensitive to neutron flux, were used to determine
the reactor vessel liquid level in a nonintrusive manner. They have been
used in several previous LOFT experimentsG-9 and have been shown to be
sensitive to changes in liquid level, correlating very well with level
determinations made using in-core instruments such as in-core SPNDs and
cladding temperatures as well as in-core conductivity probe level
transducers.
A detailed description of the measurement system is found inReference G-1O and is briefly summarized here. The detectors are Reuter
Stokes Model R/SP6-0805-135 fission chambers with a 235U core. Each
fission chamber has a 13-cm (5-in.) sensitive length and is surrounded by a
15-mm (0.59-in.) thick blanket of polyethylene and encapsulated in a 0.5-mm(0.02-in.) thick can of cadmium. The cadmium allows only epithermal
neutrons to be transmitted to the polyethylene, which then thermalizes
them, thus enhancing their detectability by the 2 3 5U. Each detector was
recorded using three different discriminator modes. These are:
1. Current mode--this is the straight current out of the amplifier
and has no discrimination between neutron and gamma events;
G-l0
2. Pulse mode--in this mode, a pulse height discriminator is used to
discriminate between neutrons which result in fission products
with 160 - 179 MeV energy and gamma rays which have only a few
MeV energy; and
3. Rise time mode--this mode takes advantage of the observation that
gamma ray events have a shorter rise time (approximately 1/3)
than neutron events.
The current mode is typically used for the first 400 to 500 s after
scram, when the radiation fields are high and the events are too closely
spaced to allow individual counting. The pulse and rise time modes are
then used to individually count the events and to discriminate between
neutrons and gamma rays.
Figure G-4 is a schematic of the LOFT reactor vessel and shows the
relative location of the Pennsylvania State University (PSU) detectors to
the core and the power range detectors. Figure G-5 is a plane view of the
reactor vessel, illustrating the azimuthal and radial location of the PSU
detectors relative to other reactor vessel instrumentation. The
instrumentation tube used for these detectors lies between the intact loop
hot and cold legs, or closest to Fuel Module 7. The closest instrumented
fuel module is #4. The approximate axial locations are: Detector A--just
above the top of the core; Detector B--just below the top of the core;
Detector C--just above the bottom of the core; Detector D--just below the
bottom of the core; and Detector E--l.O m (40 in.) above the top of the
core. Thus, Detectors B and C respond to the top and bottom approximately
20 cm (8 in.) of the core, respectively and Detectors A and D, to the
approximately 20 cm (8 in.) above and below the core, respectively.
Detector E views a section in the middle of the upper plenum.
Figure G-6 shows the normalized current output from these five
detectors for the first 120 s of Experiment LP-FP-2. The initial drop
corresponds to the reactor scram, after which the detectors followed a
normal scram curve for approximately 20 s. The four detectors near the
core responded to the insertion of the center fuel module control rods at
G-11
r
57 7 inreactor vessel inside diameter
57 2 Incore fillet outside diameter
PSU
DETECTORS -I Top Of fuel 4XusemblitsCore barrel and
•,%• ~flow skirll•J
LOFT 2 In.anularSOURCE & downcomerIR RANGE •.•Center fuel
DETECTORS module-- -- Corner luel modules•._ •\Lows,..-,.e
.. , , support s|rucluffPSU C _
DETECTORS
INSTRUMENTTUBE
Figure G-4. Cutaway of the LOFT reactor vessel illustrating the locationof PSU detectors.
G-12
PSU LEVEL GAUGE DETECTORS
RE-T-77-1A2AE-T-7-IA2 Sied lank
hot leg ECC inlet
I C) RE-TRM-86-5
R~E-RC-66-5
Drag disk E"6
turbine Ilowmeter 10
ECC inlet-L__ ' nloRE-TRM-86-6 t oken loop
RE-T-86-6HE-RC-86-6• A• E-T-87-4AI
Reacloto vesse, RE-T-87-4A2Reacor vessel V-2166support bracket Reactor head
Thermocoupt31RE-T-77-2A1 4 ,._" RE'-7-8-2
•E77-A RE-T-86-3ECC inlet •
locatiooken loop) " hOt leg
Inla *c, loop- /•RE-T-77-3AI
cold leg j(. A E-T-77.3A2
RE-T- B5-1 M~anway
RE-T-86-4E CC inletO
Figure G-5. Planar view of the LOFT reactor vessel illustrating thelocation of PSU detectors.
PSU LEVEL GAUGES
1 .0000.
0.1000
0.0100
9 JULY 85
I
0.0010
0.0001
0 20 40 60 80
TIME (SEC.)
LEGEND : CHANNEL 1 2_4. ------- 4 5
1-DETECTOR E (ABOVE ACTIVE CORE)2-DET. A 3-DET. B (UPPER CORE DETS.) 4-DET. C 5-DET. D (LOWER CORE DETS.)
Figure G-6. Normalized current response of PSU detectors (0 to 120 s).
100
------------ 3
120
( (
21 s, and all five detectors continued to follow a normal shutdown curve,
indicating the absence of any voiding during this time. Figure G-7 shows
the normalized current output for the first 1800 s. Detectors A, B, D,
and E were all relatively insensitive to the density fluctuations after
approximately 400 s, as was expected using this current mode. However,
Detector C continued to be sensitive to density fluctuations in this
recording mode, responding to decreasing density and boiling near the
bottom of the core during this time frame.
Figure G-8 shows the output from these detectors recorded in the
Pulse-mode for the first 1800 s of the transient. The pulse mode recording
for Detector C failed, so only Detectors A, B, D, and E are shown.
Detector E indicated dryout at the 1-m (40-in.) elevation in the upper
plenum at approximately 500 s into the transient. This detector also
apparently responded to the release of fission products from the gap and
fuel at approximately 1150 s and 1550 s, respectively. Detector Aresponded to the boiling in the upper plenum, also. Detector B clearly
showed the onset of dryout in the top of the core at approximately 600 s,
after which this detector continued to follow a shutdown curve, offset to
indicate continued dryout. Detector 0, however, did not indicate a similar
dryout below the bottom of the core. While boiling occurred at this
elevation, there was a continued indication of a recognizable level near
the bottom of the core throughout the transient.
Figure G-9 shows the pulse mode data for the four detectors for the
period from 1800 to 3600 s. The reflood was detected by the upper three
detectors, but was not seen as distinctly by Detector B. This gives
further credence to the conclusion that the level remained near the core
bottom throughout the transient. There was a large offset between the
post-reflood detector response and that measured in previous LOFT LOCA
experiments. This offset is perhaps indicative of fuel relocation, though
the magnitude of the relocation cannot be determined on the basis of these
detectors alone.
G- 15
;SU LEVEL CJUGES
1.0000
0.1000
0.0100
9 JULY 85
!
0.0010
0.0001
0 600 1200 1800
L
TIME (SEC.)
EGEND CHANNEL 1 245
1-DETECTOR E (ABOVE ACTIVE CORE)2-DET. A 3-DET. B (UPPER CORE DETS.) 4-DET. C 5-DET. 0 (LOWER CORE DETS.)
Figure G-7. Normalized current response of PSU detectors (0 to 1800 s).
------------ 3
( ( (
a
(PSU LEVEL GAUGE
S1 00000.
0 100000
0 01000
0 00100
C9 JULY 85
0 0 000 1 0-
0 00001
LEGEND:
0 600 1200 1800
TIME (SEC.)
DETECTO R 1.2 ------ 3 .. 4
I-DETECTOR E (ABOVE ACTIVE CORE) 2-DEFECTOR A (AT ACTIVE CORE TOP)3-DETECTOR B (4 INCHES BELOW DEL. A) 4-DETECTOR D (4 INCHES BELOW ACTIVE CORE)
Figure G-8. Normalized pulse height response of PSU detectors (0 to 1800 s).
9 JULY 85PSU LEVEL GAUGES
1 .00000
0. 10000
0.01000
0.00100
I•o
0.00010--
"-
0 .00001- 0
18002400 3000 3600
TIME (SEC.)
LEGEND: DETECTOR 12 ---------2- 3
1-DETECTOR E (ABOVE ACTIVE CORE) 2-DETECTOR A (AT ACTIVE CORE TOP)3-DETECTOR B (4 INCHES BELOW DET. A) 4-DETECTOR D (4 INCHES BELOW ACTIVE CORE)
Figure G-9. Normalized pulse height response of PSU detectors(1800 to 3600 s).
4
( ( (
To summarize, the PSU fission chambers responded to the Experiment
LP-FP-2 transient as was expected based on previous data. Deviations from
the behavior during this and previous transients are consistent with the
postulations that the level remained near the bottom of the core throughout
the transient and that fuel relocation took place during the transient.
G-19
REFERENCES
G-1. D. J. N. Taylor, "Testing of TMI-2 Type SPNDs in LOFT," privatecommunication, received August 1, 1985.
G-2. J. W. Mandler, J. T. Case, J..W. Tkachyk, "Airborne RadioiodineMeasurements - LOFT FP-2 Test Preliminary Results," privatecommunication, received August 1, 1985.
U-3. A. Baratta, private communication August 20, 1985.
G-4. H. 0. Warren, "SPND Thermal Currents in a Furnace," Appendix B,Interpretation of TMI-2 Instrument Data, NSAC 28, May 1982.
G-5. M. N. Baldwin and H. D. Warren, "SPND Thermal Currents in aFurnace and Gamma Ray Field," Appendix C., Interpretation ofTMI-2 Instrument Data, NSAC 28, May 1982.
G-6. 0. J. N. Taylor, "The Results of Separate Effects Testing onTMI-2 Type SPNDs," to be published.
G-7. C. P. Cannon, D. P. Brown, S. C. Wilkins, R. D. Meininger,Mechanisms for Anamolous Signal Outputs for Self Powered NeutronDetectors During the TMI-2 Accident, October 1984.
G-8. N. C. Dyer et al., Procedures: Source Term Management Program,NUREG-0384, October 1977.
G-9. W. A. Jester et al., Final Report, 1983 LOFT Reactor Testing ofthe Penn State Non-Invasive Liquid Level/Density Gauge,non-published report, June 1984.
G-10. A. J. Baratta et al., Feasibility Study on the Development of aNon-Invasive Liquid Level Gauge for Nuclear Power Reactors,NUREG/CR-3290, May 1983.
G-20
APPENDIX H
SCDAP/RELAP5/TRAP-MELT CODE CALCULATION AND DATA COMPARISONS
APPENDIX H
SCDAP/RELAP5/TRAP-MELT.CODE CALCULATION AND.DATA COMPARISONS
The SCDAP/RELAP5/TRAP-MELT code is an integrated code designed to
predict the damage to a reactor coolant system and the transport of fission
products and hydrogen in the event of a severe accident. The code
integrates models for calculating thermal/hydraulic response, reactor core
oxidation and meltdown, and transport of fission products released from
fuel rods and does this in a fully coupled manner. The oxidation and
meltdown models were obtained from the SCDAP codeHl, the
thermal/hydraulic models were obtained from the RELAP5 code H-2, and the
fission product transport models were obtained from the TRAP-MELTH- 3 code.
The integrated code has the capability of calculating the strong
interaction between core damage progression, reactor coolant system
thermal-hydraulics and radionuclide behavior. For example, fission product
release and the transport of iodine and cesium are strongly influenced by
the presence of steam and hydrogen. For high steam flow rates, iodine and
cesium can be transported as free iodine and as CsOH, while for low flow
rate conditions with high hydrogen concentrations, the predominant forms
are CsI, CsOH, and HI. During a severe accident, steam and hydrogen
flowrates can change from a steam-rich environment during initial heatup to
a hydrogen-rich environment during the period of maximum heatup. The
integrated code takes into account the effect of these changing mixtures of
steam and hydrogen on fission product transport.
The entire LOFT primary coolant system and a portion of the secondary
coolant system were modeled with the integrated code. The primary and
secondary systems were divided into 136 fluid cells. The nodalization is
similar to that for the RELAP5 calculations shown in Figure F-1. The heat
transfer into or from the structural components contacted by the fluid was
also modeled. A total of 167 heat structures were used to model this heat
transfer.
H-1
The reactor core was modeled by a combination of SCDAP and RELAP5 heat
structures. The rods in the center assembly were modeled by SCDAP heat
structures and the rods in the peripheral assemblies were modeled by RELAP5
heat structures. The modeling is described in Table H-l. All of the SCDAP
and RELAP5 heat structures were divided into six evenly spaced axial
nodes. The SCDAP heat structures are capable of modeling severe damage and
of calculating ballooning, oxidation, meltdown and fission product
release. Rod-to-rod and rod-to-steam radiation heat transfer is also
calculated. A total of five SCDAP heat structures were used to model the
center assembly. Two groups of RELAP5 heat structures were used to model
the rods in the peripheral assemblies: the first group modeled the 220 hot
rods and the second group modeled the 876 average rods.
Operator control of the reactor system was assumed to proceed as shown
in Table H-2. The transient was initiated by scramming the reactor. The
intact loop cold leg break valve was opened 20 s after the scram and was
closed 620 s after the scram. The LPIS line valves were opened 220 s after
scram. The assumed operator control differed somewhat from the actual
conduct of the experiment. The cold leg valve was open 115 s longer in the
actual experiment than in the preexperiment calculation. The cold leg
valve was reopened from 878 to 1022 s during the actual experiment, but was
not reopened in the preexperiment calculation. The PORV at the top of the
pressurizer was open from 882 to 1162 s during the actual experiment, but
was never open in the preexperiment calculation.
Results of the preexperiment calculation using the integrated code are
next presented. The calculations are divided into three areas: system
pressure response, fuel rod temperature response, and fuel rod ballooning.
The calculation was performed through the beginning of the high temperature
period of the transient. The end time of the calculation was 1200 s. The
preexperiment calculation was not extended beyond 1200 s because, beyond
that point, deviations from planned test conditions were expected to render
calculation/data comparisons of little value.
H-2
TABLE H-I. DESCRIPTION OF MODELING OF REACTOR CORE
Relative Power(Fraction of
Average Power inPeripheral AssembliesReactor Core Component
Heat StructureType
24 center fuel rods incenter assembly
76 outer fuel rods incenter assembly
Control rods in centerassembly (total of 11)
Hollow guide tubes incenter assembly(total of 2)
Insulated flow shroudfor center assembly
220 hot fuel rods inperipheral assemblies
876 average fuel rodsin peripheral assemblies
2.04
2.38
0
0
0
SCDAP
SCDAP
SCOAP
SCDAP
SCDAP
RELAP5
RELAP5
1.28
0.93
H-3
TABLE H-2. CONDUCT OF EXPERIMENT LP-FP-2 ASSUMED IN PREEXPERIMENTPREDICTION
Time FromReactor Scram
0.0
20.0
25.0
220
620
Reactor Operator Action
Scram reactor by inserting fuel rods incenter and peripheral assemblies
Open intact loop cold leg break valve
Turn off power to pumps in primary coolantsystem
Open valves to allow flow throughLPIS line
Close intact cold loop leg break valve inresponse to system pressure dropping to lessthan 1.2 MPa.
H-4
System Pressure Response
The calculated and measured system pressure response are compared in
Figure H-i. The calculated and measured end of subcooled blowdown occurred
at 60 and 53 s, respectively. The depressurization rates continued to be
in excellent agreement until flow through the LPIS line was actuated at
220 s. After 220 s, the calculation overpredicts the rate of flow through
the LPIS line and as a result overpredicts the rate of system
depressurization.
The difficulty in accurately modeling the LPIS line is due to its
complex configuration, with many bends and changes in flow area. No
experiments were performed prior to the LP-FP-2 Experiment to provide data
for determining loss coefficients. The preexperiment calculation used loss
coefficients ranging from 0.82 to 0.97. The results of the FP-2 experiment
indicate that, to accurately model flow through the LPIS line, the loss
factors need to be increased by a factor of six.
Reactor Core Temperature Response
Measurements show that heatup of the center fuel assembly began at
662 s and that the heatup accelerated when the intact loop cold leg break
valve was closed at 735 s. The closure of this valve reduced the rate of
depressurization and precipitated a collapse of the two-phase mixture level
in the core due to reduced flashing. The calculation also showed that
heatup of the center fuel assembly began with closure of the intact loop
cold leg break valve, but at 620 s. Thus, heatup began 42 s earlier in the
calculation than in the experiment.
The calculated and measured temperature responses of rod J07 in the
center assembly at an elevation 0.69 m (27 in.) above the bottom of the rod
are compared in Figure H-2. The measured temperature response flattened
out from 880 to 1162 s due to the opening of the PORV and the reopening of
the break valve. Oxidation began when the cladding temperature exceeded
1000 K (1340-F). The calculated and measured rates of temperature increase
prior to oxidation were both about I K/s (1.8°F/s).
H-5
17.5
--- PE-BL-002ACalculated
15
0~
S.-
VL:3C,,
0~
12.5
10
7.5
5
2.5
0
2500
-2000
-1500 O.
L
-1000
L
-500
LA
0 250 500 750 1000 1250Time (s)
1500
Figure H-i. Comparison of measured hot leg pressurecalculated using the integrated code.
with pressure
w2500- 4000
5-
VL:3
0LV0~EV
I-
2000
1500
1000
TE-5J07-027Calculated
-3000
-2000
L
lOOO0 E
I-
500
00 250 500 750 1000
Time (s)1250 1500
Figure H-2. Comparison of measured cladding temperature at the 0.69-m(27-in.) elevation in Fuel Assembly 5 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits).
yý
H-6
The calculated response of rod J07 in the center assembly at an
elevation of 1.26 m (50 in.) is shown in Figure H-3. A measured
temperature response is not available at this elevation. Rapid oxidation
is calculated to begin at 1180 s and cause a cladding temperature increase
of 35 K/s (63°F/s).
The calculated and measured response of rod H13 in a peripheral
assembly at an elevation of 0.94 m (37 in.) are compared in Figure H-4 and
are seen to be in good agreement. Both also show dryout and fuel rod
heatup beginning when the intact loop cold leg break valve is closed. Both
calculation and measurement show a heatup rate of 0.8 K/s (l.4*F/s). The
measurement shows a significant decrease in the rate of heatup beginning at
1400 s. This decrease in heatup may be due to an increase in steam flow
caused by liquefied material in the center bundle falling into the pool of
water in the lower plenum of the vessel.
Fuel Rod Ballooning
The preexperiment calculation predicted that the fuel rods in the
center bundle would balloon and rupture. The ballooning was predicted to
occur in the elevation span of 0.84 to 1.12 m (33 to 44. in.) and to cause a
78% reduction in flow area. The rods were predicted to rupture in the
period of 1000 to 1100 s at cladding temperatures of 1140 to 1160 K
(1593 to 1629 0 F). A radiation measurement at the top of the vessel
indicates that fuel rod rupture began at 1200 s.
Conclusions
The SCDAP/RELAP5/TRAP-MELT code was successfully used to perform a
preexperiment calculation of the blowdown and heatup phases of
Experiment LP-FP-2. The code accurately predicted that the time of
initiation of core heatup would begin when the intact loop cold leg break
valve was closed. The code also predicted that the initial rate of heatup
of the center bundle would be I K/s (l.8 0 F/s) and that the rate of heatup
during rapid oridation would exceed 35 K/s (63 0 F/s). The code appears to
have correctly predicted the time at which fuel rods in the center bundle
would rupture and release fission products. Experiment results clearly
indicate that modeling of the LPIS line needs improvement.
H-7
I-N
V,
E
2500
2000
1500
1000
500
0
LL
E
0 250 500 750 1000 1250 1500Time (s)
Figure H-3.
1500
1250
I 1000
-4--a
• 750
E
Calculated cladding temperature at the 1.26-m (50-in.)"elevation in Fuel Assembly 5 made using the integratedcode. (See Appendix I for thermocouple qualification limits).
U)L:3
.9-
aLU)0.EU)I-
500
2500 250 500 750 1000 1250 1500
Time (s)
Figure H-4. Comparison of measured cladding temperature at the 0.94-m(37-in.) elevation in Fuel Assembly 4 with calculations madeusing the integrated code. (See Appendix I for thermocouplequalification limits).
H-8
REFERENCES
H-1. C. M. Allison, E. R. Carlson, R. H. Smith, Proceedings of anInternational Meeting on Light Water Severe Accident Evaluation,August 1983, 5.1.1 to 5.1.5.
H-2. V. H. Ransom and R. J. Wagner, RELAP5/MOD2 Code Manual, EG&GReport EGG-SAAM-6377, April 1984.
H-3. H. Jordan, J. A. Gieseke, P. Baybutt, TRAP-MELT User's Manual,NUREG/CR-0632, February 1979.
H-9
APPENDIX I
QUALIFIED TRANSIENT DATA PLOTS
APPENDIX I
QUALIFIED TRANSIENT DATA PLOTS
This section contains qualified transient plots of all data. The
plots are on fields on the inside of the back cover of this report.
Additional files are included which contain the first report from the Data
Integrity Review Committee. A list of all the parameters on this file is
included in Table I-1.
I-1
TABLE I-1. LISTING OF QUALIFIED DATA FISCHE
Column
AAAAAAAAAAAAAAA
BB
BBBB
8
fi
BaC
BBBB
CCCCCCCCCCCCCCC
PlotNumber
001OJ2003004005036007008009010Oil012013014015
016017018019020021C2202302402502602?028029030
0310320J33(34035036037038C3 9C40C41042043
C45
MeasurementIdentification
AH2E-T75-O01AH2E-Tt 5-U02AH2E-T 5.-003CR-5UP-ACR-5UP-BCV-PC04-008CV-P004-01 0CV-PO04-09CCV-P 004-091CV-P13e-070AC V-P 13u-07 1ADE-KL-OIADE-t-1 L-OC IBOE-dL-O01CDE-8 L-C02A
DE-IlL-002BDE-BL-002CDE-B L-105DE-flt-205JE-PC-CO1ADE-PC- 016OE-PC-0G1COE-PC-002AOE-PC-0028DE-PC-GOZC)E -PC-i 05DE-PC-2C5F EP165-F1-22FE-PC-CC2OFE-IST-GO1
FE-I ST-CC2FR-PC-201FR-PC-205FR-PC-20bFT-PO04-012F T-PC04-72-2FT-PIZE-065FT-P 139-27-1FT-P 139-27-2FT-P 139-27-3L EPO T-P 139-007LE-FCC-01ALI T-Pi20-013LIT-P120-C14LIT-PlýC-08(i
Column
DDDDD0DDDDDD
D
E
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FFFFFFFFFFFFF
FF
PlotNumber
C46047048049050051052053G54055056057058059Cb0
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MeasurementIdentification
LT-PO04-OObALT-PO04-008BLT-PO04-042LT-PO04-OBAALT-P 138-033LT-Pl3E-058ME-PC-002AME-PC-0028ME-PC-G02ME-IST-001N E-2HO-26NE-4HO-26NE-6HO0-26POE-BLH-001POE-0Li-002
POE-BLIt-003POE-OLH-004PU)E-BLH-005POT-PI39-OC6POT-P 139-0C7PDT-P139-030PDT-P13 9-30AP0T-P139-3C8PE-BLH-OC1PE-BiLH-002PE-BLh-003PE-BL-OOZAPE-PC-002PE-PC-GOSPE-PC-CC6
PTPI6S-F--bBPTPl5-F2-43PT-POC4-O1CAPT-P 004-022PT-P 004-034PT-PuO4-05PT-P12C-029PT-P 120-04 3PT-P138-05OPT-P 13b-057PT-P13S-004PT-P139-042PT-P 139-05-1RE-T-77-lA1RE-T-77-1A2
Column
GGGGGGGGGGG6GGG
HHHHHHHHHH
HHH
IH
IIIIIIIIIIIII
PlotNumber
091C92093094095096097098099100101102103104135
106107108109110ill112113114115116117118119120
121122123124125126127128129130131132133134135
MeasurementIdentification
RE-T-77-?A 1RE-T-77-3A1RE-T-77-3A2RE-T-85-1RE-T-85-2RE-T-86-3RE-T-86-4RE-T-87-4A1RE-r-87-4A2RE-T4-096RP E-PC-001RPE-PC-002RP-CROM2-PTRP-CRDh2-TCRP-CPDP4-PT
RP-CkDfl4-TCRP-C RDM6-P TRP-C RDMb-TCRP-CRDMB-PTRP-C PDM8-1CSP-8LH-CO1SP-BLH-002SP-BLH-CC4SP-LLH-305SP-3LH-006SP-OLH-007ASP-BLH-008SP-PC-002BS P-P139-019SP-P13q-02C
SP-SG-C£3SO-SG-CO4SP-lST-005ST-BLH-001ST-BLH-002ST-BLH-003ST-BL-002AST-PC-002ST-PC-005ST-P139-05-1TC-5108-27TC-5M04-27TC-5M08-27TE-BLH-001TE-1LH-002
( ,( (
t tj I A
C ( CTABLE 1-1. (continued)
PlotColumn Number
KKKKKKKK
KK
LLLLL
JJJJJJ
L
K-o K
KKKKKKKKKK
LL
LtLLLLLLLLL
13713813914U14114214314414514b14714814915C
151152153154155156157158159160161
1631641b5
166167168169170171172173174175176177178179180
MeasurementIdentification
TE-BLIIt-C04TE-BLh-605TE-8LIi-CG7AT E-i3Li- 008TE-PC-OC2ATE-PC-CC28TE-PC-OC2C-TE-P CC,4-0,4FE-Pi2c.-Jo11E-t0120-027TE-PI26-ý41TE-tI 2C-132
"E-P 134-01 'TE-P139-'2 CYE-PYo2
TE-11 3q--2-2TE-P134-32-1TE-P 141-C94TE-P 141-095tE-SG-OCIATE-SG-t.O2ATE-SC-GO3TE-SG-O6CTE-SG-6C5
TE-SV-0O01TE-SV-002TE-SV-003TE-SV-OC5TE-SV-OC6TE-SV-007
TE-SV-GCETE-SV-00qTE-SV-GIG
TE-SV-0.11TE-SV-C12TE-TOS5-002TE-LA11-030TE-IB1C-037T1-111-02eTE-1B11-O3ZTE-1C1l-021TE-LC11-034TE-IF07-015TE-IF 07-026TE-IST-001
PlotColumn Number
18118218318416518618718818919019 1192193194195
1961971981992002012022032042052J 6207208209210
211212213214215216217218219220221222223224225
MeasurementIdentification
YE-1SI-C02TE-IST-003TE-1 ST-004TE-LST-O05TE-i ST-OObTE-1SI-O0eTE-iST-009TE-17ST-CITE-iST-O1ITE-1 ST-012TE-IST-1I3TE-1ST-15TE-IUP-001TE-1UP-002TE-1UP-O05
TE-I UP-006TE-IUP-OG7TE-2E08-045TE-ZFOT-Ol0T E-2 FOO-03 2TE-ZFOr--026TE-2G14-011TE-2G14-030TE-ZG14-04TE-2H14-026TE-2HG2-028TE-ZH13-021TE-2H13-049TE-2H14-032r E-2H1 -02 6
TE-2HI5-04 1TE-2114-021TE-2114-039TE-20P-001TE-ZLP-002TE-2LP-003TE-ZUP-C01TE-2UP-O02TE-2UP-C03TE-2UP-004TE-2UP-005TE-3A1-030TE-3B11-028TE-32!1-032TE-3CII-02 1
PlotColumn Number
pPPPPPPPPPPPPPP
226227228229230231232233234235236237238239240
241242243244245246247248249250251252253
2s0
255
2562372582592b0261262263264265266267268269273
MeasurementIdentification
TE-3C 11-039TE-3F07-02CTE-3UP-C01TE-3UP-CO6TE-3UP-CO0TE-3UP-010TE -3 UP-O11TE-3UP-012TE-3UP-013TE-3bP-014TE-3UP-015TE-3UP-C16TE-4ECB-04TE-4FO7-015TE-4FOb-032
TE-IGO-02 1TE-4GI4-0O 1TE-4GI4-030TE-4G14-045TE-4H13-01 5TE-4H13-037TE-41114-02 8TE-4 K15-02 tTE-4H115-04 1TE-4 114-021TE-4 114-039TE-41LP-001TE-40LP-003TE-4UP-CC1TE-4UP-C02
TE-4UP-003TE-4UP-004TE-,UP-O05TE-5C06-027TE-5CO6-066TE-5CG7-04 2TE-5COY-010TE-5CO-02 7TE-5C10-G27TE-5C12-010TE-5CI 2-027TE-5D09-027TE-5D13-042TE-5EO5-027TE-5E11-027
TABLE 1-1. (continued)
Column
AAAAAAAAAAAAAAA
BBBBBBBBBBBB
8
BB
CCCC
CCCCCCCCCC
PlotNumber
27127227327427527627727b279280281282283284285
286287Z8828 929029129229329421)29629729829q300
3u3323J33043033u6337308309310311312313314315
MeasurementIdentification
TE-5f-010TE-5E-027TE-5E-C42TE-5FG3-027TE-5F13-066TE-3GO4-O10TE-5GC4-027TE-5L12-0101 E-5(,12-027TE-5tiCt-C271iE-Ai~b-027TE-,H12-027TE-ý163-C27TE-,I'4-042TE-5• l2-L42
TE-,JC3-GCLTE-,J07-OI uTE-5J07-027TE-,.j13-027TE-5KO5-02 7TE-iK11-027TE-5L C7-CI 0TE-5L07-027TE-5 L09-04 2TE-,MG6-02 7TE-5MC7-010TE-5hO",-04 2TE-5 PIO-G-bTE-54-0ICTE-N-Cj27
SE -5U- 032T E -51-042TE-5S-010TE-5S-027FE-5S-032TE-iS-G42TE-SUP-004TE-5UP-017TE-5UP-019TE-5UP-023TE-5UP-024TE-5UP-025TE-SUP-0261E-SUP-627r E - U P--028 A
Column
D
DD0
0DDDD0DD
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EEEEEEEEEEEE
F
FFFFFFFFFFFFFFF
F
PlotNumber
316317318319320321322323324325326327328329330
331332333334335336337338339340341342343344345
3463473483493503513453523533,4355356357358
359360
MeasurementIdentification
rE-sUP-028BTE-5UP-029AT E-5IP-a 296TE-SUP-030ATE-5IP-G30BTE-5 LP-031ATE-5 UP-031BTE-SUP-C32ATE-SUP-32 BTE-5 LP-033 BrE-5uP-188ATE-SUF-1 888TE-SUP-IhBCTE-,bP-16860TE-5UP-194C1
TE-3LP-1l4G2IF-5UP-i c#7FlTE-5UP-197B2TE-SUP-21262TE-5UP-2 1G2TF-SbP-2152B1IE-5UP-21500
TE-jaUP-2 50G2TE-SUP-2 5181TE-5UP-251B2TE -W-CIOFE-S W-0 27TE-SW-L32TE-5W-C42
TE-6EG8-04,T E-6F7-037TE-b6FOI-041TE-bGGS-03csTE-bG14-011TE-6G14-030TE-5W-042TE-6G14-04 5TE-6H13-01 STE-6H13-037TE-6H14-028TE-6H14-032TE-b5,15-02bTE-SI14-GZ1TE-6114-039TE-6LP-001
Column
GGGG6GGGGGGGG
GG
HHHHHH
HHhH
HH
I-H
IIIIIIIIIIIII
PlotNumber
36136Z363364365366367368
370370371372373374375
37 b377378379380
301382383384385386387388389390
391392393394395396397398399400
41140243340440 i
MeasurementIdentification
TE-bLP-002TE-6LP-003TE-bUP-001TE-6UP-002TE-61P-003TE-6UP-004TE-6UP-005TT-PCC4-004FT-P13%-03 2TT-P139-03 3Tt-P139-G34XEP16 -Fl-42XE P165-F1-44CVP165-F 1-348CVP165-F 1-12
CVPlt5-F2-36C VP 165-Fl- 48ECVP1f5-F1-13CVP165-FI-34 AC VPI65-F 1-20CVP165-F1-14IL P165-D1-5ACVPI65-DI-4ACVP165-D1-3ACVPI .5-F 2-34ACVP1(5-FI-2PCVP165-F2-48CVP165-DI-15CVP165-F2-34BCVPI65-F1-316.
PFP165-FI-40FTP165-F 1-22PTP165-FI-5PTP165-U1-20PrP165-Fl-BAPTP165-Dl-2PT P165-D1-19TEP165-DI-21BTE P165-F I-30ATEP165-F2-45TEP165-F1-tCTEPI65-F2-38TEP165-F1-8AFEP1b5-F1-BBTEPlbS-F1-38
I, / ((
mmmm-mý
APPENDIX J
ORIGEN2 RESULTS FOR THE LP-FP-2 EXPERIMENT
APPENDIX J
ORIGEN2 RESULTS FOR THE LP-FP-2 EXPERIMENT
A microfiche copy of the ORIGEN2 calculations reported in Section 4 of
this report are attached at the end of this report. The two microfiche
cards that contain the ORIGEN2 results are titled: "FP-2 ORIGEN2
CALCULATION FOR QLR." This calculation lists the center bundle fuel and
coolant inventories in terms of grams, gram-atoms, and curies. The decayheat of the center bundle is also listed on these cards. Notice that the
listed fuel inventory results do not assume any fission product loss during
the experiment. That is, to obtain the end of experiment fuel inventory,
subtract the coolant inventory from the listed fuel inventory. Also, for
this calculation, it is assumed that the entire fission product loss to the
coolant occurs between 1740 and 1800 s.
The input deck that was used to create the ORIGEN2 calculation is
shown near the end of the microfiche listing.
J-1