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| LOCA ECCS REANALYSIS FOR D.C. COOK.

UNIT 1 USING THE ENC WREM.11Ai PWR ECCS EVALUATION MODEL:

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FEBRUARY 1981

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RICHLAND, WA 99352-

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.

XN-NF-81-07

ISSUE DATE: 02/12/81

,

LOCA ECCS REANALYSIS FOR D.C. COOK

UNIT 1 USING THE ENC WREM IIA

-

PWR ECCS EVALUATION MODEL

Prepared by:

S.E. Jensen

J.C. Cherng

W.V. Kayser

D.J. Braun .

Approved : /t / WA < # ~//' & /J{p.' Morgan,Manhser

| Licensing & Safety Engineering

Approved [6 "'NcpNbi el eering

/mb

ERON NUCLEAR COMPANY,Inc.

I._ , . . . . - . , , . - . - - .-- - - - . . - - -

. _ . . _ . . _ _ ~ _ _ _ _ - - - _ - _ - . . __ -- _ . . -_. _ . - _. _. _ _ _ _ . .- ._. _ -_ _ . _ _ _ _ . . - _ . . . ._ _ _ .

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i XN-NF-81-07

TABLE OF CONTENTS

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; SECTION PAGE

i

1.0 INTRODUCTION AND SUPMARY . . . . . . . . . . . . . . . . . 1 >

2.0 ANALYTICAL AND SYSTEM MODELS . . . . . . . . . . . . . . . 5

i

3.0 SYSTEM ANALYSIS RESULTS 7' . . . . . . . . . . . . . . . ..

4.0 FUEL EXPOSURE ANALYSIS RESULTS . . . . . . . . . . . . . . 28

29- 5.0 CONCLUSIONS .................... . . .

! 6.0 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . 30

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11

XN-NF-81-07

LIST OF TABLES

TABLE PAGE

!.1 0.C. COOK UNIT 1 EXPOSURE SENSITIVITY RESULTS . . . . . 33.1 D.C. COOK UNIT 1 REANALYSIS LIMITING BREAK

EVENT TIMES (1.0 DECLS) . . . . . . . . . . . . . . . . . 8

LIST OF FIGURES_

FICURE PAGE

1.1 0.C. COOK UNIT 1, ALLOWABLE TOTAL PEAKING FACTORAS A FUNCTION OF PEAK PELLET EXPOSURE . . . . . . . . . . 4

93.1 BLOWDOWN SYSTEM PRESSURE ................

3.2 BLOWDOWN BREAK FLOW RATE 10................:

3.3 PRESSURIZER SURGE LINE FLOW RATE 11............

3.4 ACCUMULATOR FLOW RATE TO IRTACT LOOPS . . . . . . . . . . 12

3.S BLOWDOWN CORE INLET FLOW RATE . . . . . . . . . . . . . . 13

3.6 BLOWOOWN CORE OUTLET FLOW RATE 14.............

3.7 BLOWDOWN HOT ASSEMBLY INLET FLOW RATE . . . . . . . . . . 15

3.8 BLOWDOWN HOT ASSEMBLY OUTLET FLOW RATE 16.........

3.9 BLOWDOWN PCT NODE CLADDING TEMPERATURE 17.........

3.10 BLOWDOWN PCT N0DE VOLUME AVERAGE- FUEL TEMPERATURE . . . . 18

3.11 PCT N0DE BLOWDOWN HEAT TRANSFER COEFFICIENT . . . . . . . 19,

N.

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iii XN-NF-81-07

LIST OF FIGURES (contd.)i

FIGURE PAGE

3.12 PCT NODE BLOWDOWN DEPTH OF ZIRCONIUM-WATERREACTION . . . . . . . . . . . . . . . . . . . . . . . . 20

4

3.13 NORMALIZED CORE POWER 21.................

3.14 ICECON COATAINMENT BACKPRESSURE 22............

i 3.15 REFLOOD UPPER PLENUM PRESSURE 23.............

3.16 CORE REFLOODING RATE , . . . . . . . . . . . . . . . . . 244

3.17 REFLOOD DOWNCOMER MIXTURE LEVEL 25............

3.18 REFLOOD CORE MIXTURE LEVEL . . . . . . . . . . . . . . . 26

3.19 CLADDING SURFACE TEMPERATURE DURING HEATUPENC FUEL BOL . . . . . ... . . . . . . . . . . . . . . . 27

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1 XN-NF-81-07

1.0 INTRODUCTION AND SUMMARY

In 1976, Exxon Nuclear Company (ENC) performed a LOCA ECCS analysis for ENC

' fabricated fuel in the Donald C. Cook Unit 1 reactor and established limits (I)assuring conformance to NRC 10 CFR 50.46 criteria (2) The ENC WREM II PWR ECCS.

evaluation model was used for the 1976 0.C. Cook analysis. The ENC PWR evaluation

.model has since been updated to the ENC WREM IIA model, and the ENC ice condenser

containment code, ICECON, has been reviewed and approved by the NRC. Several

features of the latest ENC PWR ECCS Evaluation Model and removal of known over-

conservatisms in input data have been shown to produce benefits with respect to

the pravious D.C. Cook LOCA ECCS results. This report details a reanalysis for

ENC fuel in the D.C. Cook Unit I reactor using the latest NRC approved ENC WREM

IIA PWR ECCS evaluation model. The analysis results show compliance with NRC 10

CFR 50.46 criteria, and establish increased allowable LOCA ECCS operating limits

for the ENC fuel in the D.C. Cook Unit I reactor.

The analysis consists of a recalculation of the previously established

limiting break LOCA, the equivalent double-ended split break of the cold leg or

reactor vessel inlet line (1.0 DECL9). A complete system calculation was

performed including the'RELAP4-EM system blowdown, accumulator discharge, ICECON

containment pressure, ae.d REFLEX reflood calculations. Multiple fuel heatup'

calculations were performed fer the ENC fuel at various linear heat rates and

exposure conditions to establish allowable ECCS limits. These analyses

consist of GAPEX, RELAP4-EM hot channel, and T000EE2 heatup calculations.

The effects of the NRC model for enhanced fission gas release and fuel rod

pressure uncertainties are also considered in the analysis.

. , .. - . . -. --

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2 XN-NF-81-07

The results of the calculations are shown in Figure 1.1 which provides

the revised maximum LOCA ECCS allowed peaking with exposure for ENC fuel in'the'

D.C. Cook Unit I reactor. Corresponding linear heat generation rates and

ECCS results are given in Table 1.1.

Details of the analytical models used and revised input are described in

Section 2.0. Section 3.0 shows the complete calculated results for the system

analysis and the beginning-of-life ENC fuel heatup analysis. Final fuel

analysis results at exposed conditions are given in Section 4.0.

The conclusion of the reanalysis is that, based on the LOCA ECCS analysis

results shown, the D.C. Cook Unit 1 reactor can be operated with ENC fuel at

or below the limits defined by Figure 1.1 and Table 1.1 which will assure

conformance with the NRC 10 CFR 50.46 criteria and 10 CFR 50 Appendix K

requi rements.

- . . . - - _ _

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Table 1.1 D.C. Cook Unit 1 Exposure Sensitivity Results

Peak Pellet Burnup (GWD/MTM) BOL 12.0 23.5 34.5 42.2*

T 2.07 2.10 2.04 1.98 1.89Total Peaking, Fg

Peak Linear Heat Generation Rate 14.24 14.45 14.03 13.6213.00

(Kw/ft)Peak Clad Temperature (PCT, *F) 2199 2177 2195 2185 2186

Max. Local Zr/H 0-Reaction, % 6.42 6.09 6.25 5.95 .5.622

Core Wide Zr/H 0-Reaction, %

- _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ ._ _______ _ ____ __ __

.

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1.80 e s i i i 1 s0 10 20 30 40 50 60 &

TPEAK PELLET EXPOSURE, GWD/MTPS E

6"

Figure 1.1 D.C. Cook Unit 1, Allowable Total Peaking Factoras a Function of Peak Pellet Exposure

.

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5 XN-NF-81-07

2.0 ANALYTICAL AND SYSTEM MODELS

The D.C. Cook Unit i reanalysis used the ENC WREM IIA PWR ECCS Evaluation

Model(3,4,5,6) as approved by the NRC and applied in several of the latest ENC

PWR analyses. Only the CONTEMPT code which computes containment backpressure was

replaced by the ICECON computer code as appropriate for the D.C. Cook Unit 1

ice condenser containment design. ICECON(7) has been reviewed and approved by

the NRC(0) for LOCA ECCS application with ice condenser containments.

The 1976 LOCA ECCS analyses for D.C. Cook Unit 1 was performed with the ENC

WREM II PWR ECCS evaluation model. During the development of the ENC WREM IIA

model, several of the changes made were shown to give small improvements in ECCS

margins for the D.C. Cook Unit I reactor. Model changes from the ENC WREM II

model include: (1) RELAP4-EM updates from RELAP4-EM/ ENC 25 to RELAP4-EM/ ENC 28;

(2) Replacing RELAP4-EM FLOOD with REFLEX; (3) Replacing the Westinghouse

calculated reflood backpressure with the ENC ICECON calculated results; and

(4) T00DEE2 updates from T00DEE2/76 to T00DEE2/DMAY79(9).

In addition to the ENC WREM IIA updates, revised system input was obtained

through ENC neutronics calculations for moderator density reactivity data.

A correction was also made to accumulator line dimensions input. The effect of

this correction with regard to Peak Cladding Temperature (PCT) results is

insignificant. The initial steady-state conditions were reestablished based on

nominal conditions for core inlet temperature and secondary pressure as per a

previous NRC request.

-.

t

6 XN-NF-81-07

The system and fuel nodalization for the O.C. Cook Unit 1 reanalysis

remains as in the previous analysis or as documented for REFLEX (5) and

ICECON(7) Only a minor change in the T00DEE2 nodalization was made ,.,

to center the maximum power node at the peak axial power location.,,

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7 XN-NF-81-071

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3.0 SYSTEM ANALYSIS RESULTS

The D.C. Cook Unit 1 ECCS reanalysis was performed for the previously

identified limiting large break, the large split break of the reactor vessel

recirculation inlet line or cold leg with the break area equal to twice the pipe&

cross sectional flow area. This break is referred to as the equivalent double-

ended cold leg split break (1.0 DECLS). System behavior is essentially un-

affected by exposure and the system calculation is performed for the highest

stored energy beginning-of-life (80L) case.

Calculated event times for the ECCS reanalysis are given in Table 3.1..,

RELAP4-EM system blowdown results are given in Figures 3.1 through 3.6. Figures

3.7 through 3.12 present results of the RELAP4-EM hot channel calculation.

Extended decay power is shown in Figure 3.13, and the ICECON computed containment

pressure is given in Figure 3.14. REFLEX reflood results are shown in FiguresT of 2.07 are shown3.15 through 3.18. T00DEE2 results for the BOL case with Fq

in Figure 3.19.

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8 XN-NF-81-07

Table 3.1 D.C. Cook Unit I Reanalysis Limiting Break EventTimes (1.0 DECLS)

Event Calculated Event Time (sec)

Start 0.0

Initiation of Break 0.05

Safety Injection Signal 0.65

Accumulator Injection, Intact Loop 15.3

Accumulator Injection, Broken Loop 2.1

End-of-Bypass 24.25

Bottom of Core Recovery 41.53

52.2Accumulator Empty, Intact Loop -

25.65Pumped Safety Infection

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Figure 3.18 Reflood Core Mixture Level

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28 XN-NF-81-07

[

4.0 FUEL EXPOSURE ANALYSIS RESULTS

Effects of exposed fuel conditions on ECCS analysis results were computed

using RELAP4-EM for blowdown hot channel calculations and T000EE2 for fuel

heatup and reflood analyses. BOL system boundary conditions were assumed.

The fuel analyses used the approved ENC Cladding swelling ind flow blockage

model. Fission gas release was computed using GAPEX, and the NRC fission gas

enhancement was employed at appropriate fuel exposures. The effects of pressure

uncertainties were included in these calculations as in previous ENC analyses.

TTotal peaking, Fq , and the corresponding peak linear heat generation rate(LHGR) were adjusted until 10 CFR 50.46 criteria were achieved. The final

heatup analyses results for the exposures calculated are presented in Table 1.1.

The ECCS allowed peaking is limited to 2.07 at BOL due to the high initial

stored energy. The allowed peaking increases with exposure as fuel swelling and

cracking reduce stored energy. At high burnup, increased fuel rod swelling

and flow blockage are calculated to occur due to increased fission gas release.

This results in reduced heat transfer during reflood and a corresponding

reduction of allowed peaking limit with exposure accumulation.

i

. .-

,

'

s' ||i

29 XN-NF-81-07 l

5.0 CONCLUSIONS

The reanalysis of the limiting break (1.0 DECLS) for the O.C. Cook Unit i

reactor with the ENC WREM IIA PWR ECCS evaluation model shows that the reactor

can operate at increased allowed peaking and continue to meet the NRC 10 CFR

50.46 criteria with analyses performed in conformance to 10 CFR 50 Appendix K

requi rements. Operation within the ECCS allowed limits as defined in Figure 1.1

and Table 1.1 assures that the NRC acceptance criteria are met. That is:,

(1) The calculated peak fuel cladding temperature does not

exceed 2200'F.

(2) The calculated local cladding oxidation does not exceed

17% of the cladding thickness during or after quenching, and

the temperature transient is terminated while the core geometry is

amenable to cooling.

(3) The calculated core-wide reaction of cladding with water or steam

does not exceed 1% of the total mass of zircaloy in the reactor.

(4) System long term cooling capabilities provided for previous

cores will also cool ENC fueled cores.

,

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, -, --

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.

.

30 XN-NF-81-07

6.0 REFERENCES

(1) Exxon Nuclear Company, Donald C. Cook Unit 1 LOCA Analyses Usingthe ENC WREM-Based PWR ECCS Evaluation Model (ENC WREM-II) , XN-76-51,October 1976 and Supplements; Flow Blockage and Exposure SensitivityStudy for D.C. Cook Unit 1 Reload Fuel Using ENC WREM-II Model ,XN-76-51 Supplement 1, January 1977; XN-NF-76-51(P) Supplement 2January 1978; XN-NF-76-51(P) Supplement 3, March 1978.

(2) U.S.N.R.C. Acceptance Criteria for Emergency Core Cooling Systemsfor Light Water Cooled Nuclear Power Reactors ,10 CFR 50.46 andAppendix K to 10 CFR 50. Federal Register, Volume 39, Number 3,January 4,1974.

(3) Letter, G.F. Owsley (ENC) to D.F. Ross (NRC), Description of RELAP4-EMENC 288, dated October 30, 1978.

(4) Letter, Thomas A. Ippolito (NRC) to Warren S. Nechodom (ENC),ENC-EM Update Evaluation, March 1979.

(5) Exxon Nuclear Company, Exxon Nuclear Company WREM-Based GenericPWR ECCS Evaluation ?odel Update ENC WREM-IIA , XN-NF-78-30, August19'8, and XN-NF-78-30 Amendment 1 February 1979.

(6) Letter, Thomas A. Ippolito (NRC) to Warren S. Nechodom (ENC,Topical Report Evaluation, dated March 30, 1979.

(7) Exxon Nuclear Company, ICECON: A Computer Program Used to CalculateContainment Break Pressure for LOCA Analysis (Including Ice CondenserPlants) , XN-CC-39 Rev. 1, November 1977.

(8) Letter,ThomasA.Ippolito(NRC)toWarrenS.Nechodom(ENC),Topical Report Evaluation, dated June 30, 1978.

(9) Letter, G.F. Owsley (ENC) to D.F. Ross (NRC), Updates of T00DEE2Program, dated April 1, 1980.

.

,- .~. e- s, - , , -

. . _ __

.'.

|

XN-NF-81-07

ISSUE DATE: 02/12/81

LOCA ECCS REANALYSIS FOR D, C. COOK UNIT 1 USING THE

ENC WREM 11A PWR ECCS EVALUATION MODEL

4

Distribution

D.J. Braun

J.C. Cherng

R.E. Collingham

G.C. Cooke

S.L. Garrett

K.P. Galbraith

S.E. Jensen

J.D. Kahn

W.V. Kayser

J.E. Krajicek

J.N. Morgan

G.F. Owsley

G.A. Sofer

H.E. Williamson

American Electric Power (5)/H.G. Shaw

Document Control (10)

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