COMPUTER CALCULATION CONTROL SHEET Page 1 of 78
plus Attachments
REV. 1 IP3 [ JAF LCALC. NO. JAF-CALC-RAD-00048
mOD/TASK NO.
QA CATEGORY OF CALCULATION: II/III
CALCULATIONAL TYPE: PRELIMINARY:PROJECT/TASK:SYSTEM NO./NAME:TITLE: Power Uprate Project -
Offsite Outdoor Recept
FINAL: x
- Radiological Impact at Onsite and
:ors Following Design-Basis Accidents
PREPARER:CHECKER:VERIFIED: N/A a
APPROVED:
NAMEA. RamachandranM. Golshani
G.C. Re'
he- ^ Q )1 NZs]w
L;& - I-4^S /I / OXe
- . LW O .. L A/I// 7_ _ _ _ _ _ _ J
PROBLEM/OBJECTIVE/METHOD
See pages 2, 11-16, and 36-37
DESIGN BASIS/ASSUMPTIONS /ANALYSIS
See pages 38-78
SUMMARY/CONCLWSIONS
See pages 17-35
REFERENCES
See pages 4-7
AFFECTED SYSTEMS/COMPONENTS/DOCUMENTS
O VOIDEDo SUPERSEDED BY: /S upersedes Revision 0 of this
(CALC NO.) calculation.
eSupers.edes Those Portions of JAF-CALC-RAD-00008 Which Deals Withthe Onsite and Offsite Outdoor Receptors.
NYPA FORM DCM-14, ATTACHMENT 4.1 (REVISION 1) Page 1 of 1
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FINAL [XI CHECKED BY 6- DATE i1/13/191TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Statement Of Problem
This calculation [prepared by Corporate Radiological Engineering
(CRE)] is in support of the power uprate program at JAF, and re-
evaluates the analyses documented in Ref. 1 (JAF-CALC-RAD-00008)
for outdoor receptors at onsite locations, and receptors at the
Site Boundary (SB) and the Low Population Zone (LPZ). The
reasons for the re-analysis are as follows:
(a) Incorporation of the recently revised atmospheric
dispersion factors (Ref. 2),
(b) Revisions to the scenarios for a Main Steam Line Break
Accident (Ref. 3), and a Control Rod Drop Accident
(Ref. 4), and
(c) Revisions to certain accident assumptions, for
consistency with those employed in the revised Control
Room Habitability analysis (Ref. 5).
The analyses documented in this calculation fall under ACTS Item
18820 (Ref. 39).
Revision 1 - Remarks
Revision 1 of this calculation was undertaken to address the
concern identified under ACTS Item 23847 (Ref. 43):
(a) evaluation of the loss of coolant accident (LOCA) and
refueling accident (RA) assuming a lowered stand-by gas
treatment system (SGTS) charcoal filter efficiency (assumed
efficiency of 90% for halogens).
In addition, the present calculation incorporates the following
changes to the assumptions and methods employed in the previous
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revision of this calculation:
(b) revision of the atmospheric dispersion factors for elevated
releases as documented in revision 2 of Reference 2, and
(c) use of the ICRP 30 (Ref. 45) dose conversion factors for the
determination of thyroid doses.
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Offsite Outdoor Receptors Following Design-Basis Accidents
References
1. CRE Calculation JAF-CALC-RAD-00008, "RadiologicalConsequences of Design Basis Accidents at James A.Fitzpatrick" (11/27/91)
2. CRE Calculation JAF-CALC-RAD-00007, Revs. 1 & 2, "PowerUprate Program - Onsite and Offsite Post-AccidentAtmospheric Dispersion Factors"
3.- CRE Calculation JAF-CALC-RAD-00039, "Revised OffsiteRadiation Exposures Followiing a Design-Basis Main SteamLine Break Accident" (12/15/94)
4. CRE Calculation JAF-CALC-RAD-00041, Rev. 0, "RadiologicalAssessment of a Control Rod Drop Accident Without MSIVClosure at pre-Uprate Conditions" (2/9/95)
5. CRE Calculation JAF-CALC-RAD-00042, Revs. 0, 1 & 2,"Control Room Radiological Habitability Under Power UprateConditions and CREVASS Reconfiguration"
6. CRE Computer Code DORITA-2, "A Computer Code for theDetermination of Radioactivity and Radiation Levels inVarious Areas of a Nuclear Power Station and OffsiteFollowing Accidental Releases of Gaseous FissionProducts," RAD-001, Release 1.5.1.2 (1/22/97)
7. CRE Computer Code QAD-CGGP, "A Combinatorial GeometryVersion of QAD-P5A, A Point Kernel Code System for Neutronand Gamma-Ray Shielding Calculations Using the GP BuildupFactor," RAD-006, Release 1.3.1.1 (3/26/92)
8. CRE Calculation-Specific Computer Code MATILDA, documentedin:
a) CRE Calculation JAF-CALC-RAD-00003, "Power UprateProgram - Reactor Building Post-LOCA EQ RadiationLevels Due to Buildup of Halogen Activity on AirFiltration Systems" (November 1991)
b) CRE Calculation JAF-CALC-RAD-00015, "EquipmentQualification Radiation Exposures Following A ControlRod Drop Accident" (7/28/92)
9. NYPA Memorandum *7AG-93-245 addressed to J. Lazarus, -_omJ. Gray, titled "'Control Rod Drop Accident (CRDA)Assumption" (9/24/93) [See JAF-CALC-RAD-00026 fo-. a copyof this ref.]
10. US NRC NUREG-0800, "Standard Review Plan for the Review ofSafety Analysis Reports for Nuclear Power Plants"
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Offsite Outdoor Receptors Following Design-Basis Accidents
11. GE letter addressed to Richard Chau, NYPA, from C. H.Stoll, GE Plant Performance Engineering, titled "J. A.FITZPATRICK (JAFNPP) Power Uprate Program - Transmittal ofNuclear Boiler Parameters and Final Reactor Heat Balance"f(2/11/91) [See JAF-CALC-RAD-00004 for a copy of thisreference.]
12. GE letter addressed to Richard Chau, NYPA, from C. H.Stoll, GE Plant Performance Engineering, titled "J. A.FITZPATRICK (JAFNPP) Power Uprate Program - FormalTransmittal of Final Source Term Analysis Results"(5/2/91) [See JAF-CALC-RAD-00008 (Ref. 1) for a copy ofthis reference.]
13. J. DiNunno, F. Anderson, R. Baker and R. Waterfield,"Calculation of Distance Factors for Power and TestReactor Sites," AEC, Division of Licensing and Regulation,TID-14844 (March 1962)
14. US NRC Regulatory Guide 1.3, "Assumptions Used forEvaluating the Potential Radiological Consequences of aLoss of Coolant Accident for Boiling Water Reactors" (Rev.2, June 1974)
15. US NRC Regulatory Guide 1.5, "Assumptions Used forEvaluating the Potential Radiological Consequences of aSteam Line Break Accident for Boiling Water Reactors"(March 1971)
16. US NRC Regulatory Guide 1.25, "Assumptions Used forEvaluating the Potential Radiological Consequences of aFuel Handling Accident in the Fuel Handling and StorageFacility for Boiling and Pressurized Water Reactors"(3/23/72)
17. US NRC Regulatory Guide 1.49, "Power Levels for NuclearPower Plants" (Rev. 1, December 1973)
18. US NRC Regulatory Guide 1.77, "Assumptions Used forEvaluating a Control Rod Ejection Accident for PressurizedWater Reactors" (May 1974)
19. US NRC Regulatory Guide 1.52, "Design, Testing andMaintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration andAdsorption Jnits of Light-Water-Cooled Nuclear PowerPlants" (Rev. 2, March 1978)
20. General Electric Report NEDO-31400, "Safety Evaluation forEliminating the BWR Main Steam Isolation Valve ClosureFunction and Scram Function of the Main Steam LineRadiation Monitor" (May 1987) (See JAF-CALC-RAD-00013 forcopy)
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FINAL [xi CHECKED BY 2 DATE j/qj/97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
21. Proposed Technical Specification Changes - Power Uprate(JPTS-91-025), and NYPA Letter to NRC JPN-92-028 (6/5/92)
22. US NRC NUREG-0123, Rev. 3, "Standard TechnicalSpecifications for General Electric Boiling Water Reactors(BWR/5)" (Fall 1980)
23. NYPA Memorandum No. MHM-91-6, addressed to J. Lafferty,from M. Mozzor, titled "Charcoal Filter Efficiencies forUse in Accident Analyses Associated with JAF Power UprateProgram" (10/2/91) [See JAF-CALC-RAD-00008 (Ref. 1) for a
- copy of this reference.]
24. Stone & Webster Engineering Calculation No. 12966-PE(N)-019-0,"High Energy Line Break Analysis in the TurbineBuilding for Class IE Electrical Equipment Qualificationin Response to IE Bulletin 79-01B" (6/9/81) [See JAF-CALC-RAD-00042 (Ref. 5) for a copy of this reference.]
25. GPU Nuclear Corporation letter 5450-95-0006, addressed toM. Karasulu, from N. G. Trikouros, titled "FitzPatrickNuclear Plant Turbine Building HELB Analysis Results"(2/17/95) [See JAF-CALC-RAD-00042 (Ref. 5), for a copy ofthis reference.]
26. JAF Emergency Plan Implementing Procedure EAP-44, "CoreDamage Estimation" (July 1991)
27. JAF Original FSAR, Supplement 25, "Effects of High EnergyPiping System Breaks Outside of Primary Containment,"(7/22/74)
28. GE Technical Report NEDE-31152P, "GE Fuel Bundle Designs,"Rev. 3 (February 1993)
29. R. G. Jaeger, Ed., "Engineering Compendium on RadiationShielding," Springer-Verlag, NY (1975)
30. SWEC Engineering Calculation #12966-RP-76-004, "LOCA Six-Month Gamma Doses for IE 79-O1B Equipment Qualifications"(9/29/80)
31. JAF Procedure AP-08.02, "Failed Fuel Action Plan" (Rev. 0,1/29/94)
32. Johnson Service Company, Test Report TLP-774-448 (02-4925-72), "FitzPatrick NPP Damper Leakage (D-1300 Series)(11/29/72) (NYPA Microfiche No. 60067, frames 025-030)[See JAF-CALC-RAD-00028 for a copy of this reference.]
33. Stone & Webster Engineering, Calculation CC-70-04,"Calculation for Air Conditioning System Cooling Load"(9/2/70) (DSR #249252)
34. JAF Process Surveillance Procedure PSP-1, "Reactor Water
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Offsite Outdoor Receptors Following Design-Basis Accidents
Sampling and Analysis" (12/4/91)
35. Niagara Mohawk Power Corporati:n letter # NMP86969addressed to J. Hamawi, from Tom Galletta, titled "TheValidity of the Nine Mile Point (NMP) Meteorological Data(1985-1990) Sent for Use in Updating the Offsite DoseCalculation Manual (ODCM)" (2/23/93) [See JAF-CALC-RAD-00007, Rev. 2, (Ref. 2) for a copy of this reference.]
36. CRE Calculation JAF-CALC-RAD-00025, Rev. 1, "AtmosphericDispersion and Deposition Parameters for Routine Releases"(5/4/95)
37. Empire State Electric Energy Research Corporation(ESEERCO) Technical Report No. EP 91-28, "Eastern LakeOntario - On-Shore Flow Field Study," prepared by GalsonCorp. (4/94) [See JAF-CALC-RAD-00007, Rev. 2, (Ref. 2) fora copy of this reference.]
38. CRE Calculation JAF-CALC-RAD-00005, "Drywell PersonnelAccess Lock Removable Wall Shielding Analysis" (12/23/91)
39. E-Mail Message addressed to D. Burch, from G. Lozier,titled "Power Uprate NRC Submittals" (12/20/95)
40. JAF Emergency Plan Implementing Procedure EAP-10,"Protected Area Evacuation" (Rev. 12, 2/6/95)
41. JAF Emergency Plan Implementing Procedure EAP-11, "SiteEvacuation" (Rev. 13, 2/6/95)
42. JAF DVP-01.02, "Radiological Effluent Controls and OffsiteDose Calculation Manual" (Rev. 0, 12/28/93)
43. * ACTS Item #23847, "Revise Calculations which use SBGTEfficiency from 99% to 95%".
44. US NRC Nuclear Safety Evaluation related to JAF AmendmentModification #239 (Power Uprate) (12/96)
45. International Commission on Radiological Protection (ICRP)Publication 30, "Limits for Intake by Workers" (VariousParts and Supplements, 1979-1982)
46. NYPA Letter JPN-96-055 addressed to the NRC, titled"JAFNPP - Additional Information Regarding Analyses atPower Uprate Conditions" (12/23/96)
* See Attachment A for a copy this reference
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Offsite Outdoor Receptors Following Design-Basis Accidents
List of Computer Programs EmDloyed
The following CRE computer programs and data libraries were used
in the analyses documented in this calculation:
Program
Name
Reference
Number
Release Date of Computer
SystemNumber Release
DORITA-2
QAD-CGGP
RAD-001
RAD-006
1.5.1.2 01/22/97 RS/6000
1.3.1.1 03/26/92 DG AViiON
----- 07/28/92 DG AViiONMATILDA(a)
(a) MATILDA was developed for use in conjunction with QAD-CGGP(Ref. 7) and the gamma spectra produced by DORITA-2 (Ref. 6)and two other CRE codes (namely, ELISA and ALLEGRA), for thecomputation of radiation exposures. It is documented inRefs. 8(a) and 8(b).
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Offsite Outdoor Receptors Following Design-Basis Accidents
Table of Contents
Page
CALCULATION CONTROL SHEET ................................. 1
STATEMENT OF PROBLEM ...................................... 2
REFERENCES ....................................... 4
LIST OF COMPUTER PROGRAMS EMPLOYED ........................ 8
TABLE OF CONTENTS ...... .................. 9
1. INTRODUCTION ......................................... 11
2. SUMMARY OF RESULTS .................................... 17
2.1 Offsite Receptors ............. .. ................ 172.2 Onsite Outdoor Receptors - Immersion Dose Rates
and Cumulative Doses ........................ 172.3 Onsite Outdoor Receptors - Reactor Building
Shine . . 18
3. METHODS OF ANALYSIS .. 36
4. RADIATION EXPOSURES FROM A LOSS OF COOLANT ACCIDENT .. 38
4.1 Drywell Leakage ................................. 39
4.1.1 Basic Data and Assumptions .... ........... 394.1.2 Results .................................. 42
4.2 ESF Component Leakage ........... .. .............. 44
4.2.1 Basic Data and Assumptions ... ............ 444.2.2 Results .................................. 45
4.3 Total LOCA Dose ................................. 48
5. RADIATION EXPOSURES FROM A MAIN STEAM LINE BREAK ..... 50
5.1 Basic Data and Assumptions ...................... 505.2 Results ......................................... 54
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TABLE OF CONTENTS (Cont.)
Page
6. RADIATION EXPOSURES FROM A CONTROL ROD DROP ACCIDENT . 57
6.1 Basic Data and Assumptions ...................... 576.2 Results ......................................... 61
7. RADIATION EXPOSURES FROM A REFUELING ACCIDENT ... ..... 63
7.1 Basic Data and Assumptions ...................... 637.2 Results ......................................... 66
8. RADIATION EXPOSURES FROM OTHER POST-LOCA SOURCES ..... 68
8.1 Direct Shine from Post-LOCA AirborneRadioactivity in the RB Refueling Level ... ...... 68
8.1.1 Basic Data and Assumptions .... ........... 688.1.2 Results .................................. 74
8.2 Direct Shine from Post-LOCA AirborneRadioactivity at El. 272' of the RB .... ......... 75
8.2.1 Basic Data and Assumptions .... ........... 758.2.2 Results .................................. 78
ATTACHMENTS
A. Excerpts from References 'ertinent to this Calculation
B. Copies of Computer Outputs
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Offsite Outdoor Receptors Following Design-Basis Accidents
1.0 Introduction
The radiological consequences of design-basis accidents at onsite
and offsite outdoor receptors under power-uprate conditions atJAF were originally assessed in CRE calculation JAF-CALC-RAD-
00008 (Ref. 1). The present calculation re-evaluates thepotential accident consequences for a number of reasons, asfollows:
(a) Incorporation of the recently revised atmospheric
dispersion factors (Ref. 2),
(b) Revisions to the scenarios for a Main Steam Line Break
Accident (MSLB, Ref. 3), and a Control Rod Drop
Accident (CRDA, Ref. 4), and
(c) Minor revisions to some of the accident assumptions,
for consistency with those employed in the revised
Control Room Habitability analysis (Ref. 5).
The calculation addressing the post-accident atmospheric
dispersion factors (Ref. 2) was recently updated to accommodate
the following:
(lj The meteorological data base for calendar years 1985-
1990 (which was made use of in JAF-CALC-RAD-00007, Rev.
0) was updated by Niagara Mohawk Power Corporation
(Ref. 35) to adjust a minor (few-degree) miscalibration
in the wind direction sensors.
(2) For consistency with the dispersion data in Ref. 42
[the JAF Offsite Dose Calculation Manual (ODCM)], the
meteorological data base was extended to include 8
years' worth of hourly values (1985 through 1992),
(3) New information on onshore flows at the site (Ref. 37)
was used to determine that only receptors at the SB and
the LPZ can be affected by the prescribed assumption of
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fumigation conditions at the time of a design-basis
LC . The location 3f the CR with respect to Lake
Ontario and the main stack is such that the potential
for having a fumigation condition affecting the CR (as
used in JAF-CALC-RAD-00008) is non-existent.
The revisions to the accident scenarios for an MSLB and a CRDA
were as follows:
MSLB:
In contrast to Regulatory Guide 1.5 (Ref. 15), which is
applicable under pre-uprate conditions at JAF, Sec. 15.6.4
of the Standard Review Plan (SRP, Ref. 10, the guiding
document in this calculation) extends the possible MSLB
locations to encompass not only the steam tunnel but all
locations "outside containment." In the original MSLB
analysis under power-uprate conditions (Ref. 1), the rupture
location was assumed to be in the steam tunnel, which is not
limiting. Recently identified information (Refs. 24 and 25)
shows that a break in the 16" bypass line leading to the
turbine bypass steam chest would release more reactor
coolant than a break in one of the 24" main steam lines.
CRDA:
Implementation of Modification F1-93-086 during the December
1994 refueling outage, which eliminated the reactor-scram
and MSIV-closure functions of the main-steam line radiation
monitors, changed the release pathway of a design-basis
CRDA. Under the old CRDA scenario, the Main Steam Isolation
Valves (MSIVs) close and release of radioactivity to tar
atmosphere was due to turbine/condenser leakage into the
turbine building. In the new scenario (without MSIV closure)
the release could also be via the offgas system. The old
scenario was used in the original power-uprate analysis
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(Ref. 1); the present calculation uses the new scenario.
Note that the latest CRDA calculation at pre-uprate
conditions (Ref. 4) was also based on the new scenario.
With respect to item (c) above, the list of o-ther minor changes
in the accident assumptions were as follows:
(1) The breathing rate at the LPZ was allowed to vary with
time after the accident, in line with the model in
Regulatory Guide 1.3 (Ref. 14); in JAF-CALC-RAD-00008,
the high breathing rat- of 3.47E-04 m3/sec was assumed
for the duration of the accident.
(2) The atmospheric release rate of airborne radioactivity
resulting from ESF component leakage into the reactor
building was reduced from a conservatively selected
value (7.2 air changes per hour) to one which reflects
the actual flow through the Standby Gas Treatment
System (SGTS) (3.3 air changes per hour).
(3) The release rate to the atmosphere associated with an
MSLB was reduced from instantaneous to 3 air changes
per hour, for consistency with the CR habitability
model (Ref. 5). (Thyroid doses differ by about 3%
between these two cases.)
(4) The post-CRDA iodine plateout fraction within the steam
lines and condenser was increased from 50% to 90% [Ref.
10 (SRP Sec. 15.4.9), Ref. 20 (Sec. 6.3.1.1), and Ref.
9].
In addition to the above, JAF-CALC-RAD-00008 considered the dose
rates to onsite outdoor receptors due to post-LOCA gamma
radiation emanating from airborne radioactivity within the RB
Refueling Level, and from gamma radiation streaming from the
drywell through the Personnel Access Lock (PAL). In the current
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calculation, the source accumulating in the refueling level was
redefined to reflect the actual RE air exchange rate though the
SGTS, in lieu of the conservatively selected rate of 1 air change
per day in JAF-CALC-RAD-00008. Also, the analysis for the
radiation streaming through the PAL was replaced with a source of
higher impact, namely airborne radioactivity within El. 272' of
the reactor building.
Results of the present study are summarized in Sec. 2, and the
methods of solution are presented in Sec. 3. Sections 4 through
7 present the data, assumptions and analytical details associated
with each of the four design-basis accidents. Section 8 looks at
the radiation fields to onsite outdoor receptors due to other
post-LOCA sources. Excerpts from references pertinent to this
calculation, and copies of the computer outputs appear in
Attachments A and B.
Revision 1 - Remarks
This calculation is being re-issued as Rev. 1 to address the
following:
(a) evaluation of the loss of coolant accident (LOCA) and
refueling accident (RA) assuming a lowered stand-by gas
treatment system (SGTS) charcoal filter efficiency (assumed
efficiency of 90% for halogens).
In addition, the present calculation incorporates the following
changes to the assumptions and methods employed in the previous
revision of this calculation:
(b) revision of the atmospheric dispersion factors for elevated
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releases as documented in Reference 2, and(c) use of the ICRP 30 (Ref. 45) dose conversion factors for the
determination of thyroid doses.
Details are presented below.
JAF ACTS Item #23847 requested the revision of the power uprateradiological analyses using an SGTS filter efficiency of 95%(instead of 99t). The intended purpose of this change is toreduce the filter efficiency test acceptance criteria from apenetration of 0.175% (as committed to the NRC in Ref. 46) to theless restrictive value of 1t.
In the present revision, an SGTS filter efficiency for theremoval of halogens is conservatively assumed as 90% for allhalogen species [Ref. 23 documents the case for use of a 99%filter efficiency for a 2" charcoal with humidity control, andtest acceptance criteria as specified for 4" beds in Ref. 33.Although the higher efficiency has been accepted by the NRCduring discussions on the power uprate analyses (Ref. 46) thepresent analysis employs a lower filter efficiency to providesome relief for testing acceptance criteria.]
Dispersion factors for elevated releases were reanalyzed toaccommodate possible short-term meandering and looping effects byplacing the control room intake at that distance from the stackwhere concentrations would peak. See revision 2 of Reference 2for details.
Thyroid dose analyses based on ICRP-2 (or, equivalently, TID-14844) yield results which are higher than those based on themore up-to-date factors in ICRP-30. Indeed, use of the latter
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can reduce the dose estimates by about 40%O. It is CRE's
understanding that radiological analy3es basec n ICRP--30, and
submitted to the NRC by various vendors and utilities, are not
uncommon. In fact, the NRC confirmatory analyses for the JAF
power uprate (Ref. 44) were based on ICRP-30. Following
discussions with Licensing, it was decided to employ the ICRP-30
dose conversion factors for the present revision.
Reference 5, Revs. 1 and 2, document the analysis for an MSLB
with an assumed pre-accident iodine spike corresponding to the
maximum iodine concentration stated in the technical
specifications as a Limited Condition of Operation (namely, 2
gCi/gm I-131 DE for JAF), in addition to the equilibrium value
for continued full power operation (namely, 0.2 gCi/gm I-131 DE).
For offsite receptors, there is no difference in the relative
exposure with respect to the guidelines, since the 10-fold
increase in the RCS concentration (from equilibrium to spiked
conditions) is offset by the 10-fold increase in the acceptable
dose. As a result, only the equilibrium value is analyzed in the
present calculation.
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Offsite Outdoor Receptors Following Design-Basis Accidents
2.0 Summarv of Results
2.1 Offsite Receptors
The following design-basis accidents were considered in the re-
assessment of the radiological consequences at JAF under power
uprate conditions:
(a) Loss of coolant accident (LOCA) (drywell leakage and
ESF component leakage pathways),
(b) Main Steam Line Break outside containment (MSLB),
(c) Control Rod Drop Accident (CRDA), and
(d) Refueling accident (RA).
The basic data and assumptions in each of the four accident
scenarios are consistent with the current licensing basis and the
models in the regulatory guides (Refs. 14 - 19) and the Standard
Review Plan (SRP, Ref. 10). Complete details for each accident
are presented in Secs. 4 through 7. A summary of the principal
assumptions associated with each DBA and the ensuing immersion
doses at the site boundary appear in Table 2.1. The doses at the
Low Population Zone (LPZ) are presented in Table 2.2.
From Tables 2.1 and 2.2, it is seen that the highest immersion
dose is 68.7 rem (to the thyroid at the LPZ following a design-
basis LOCA) which is about 23!k of the regulatory limit.
2.2 Onsite Outdoor Receptors - Immersion Dose Rates and
Cumulative Doses
Post-accident immersion dose rates an4 cumulative doses were also
calculated for onsite outdoor receptors at grade elevation in the
general vicinity of the old administration building. The results
are shown in Tables 2.3 and 2.4. The worst-case dose rates are
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE IS -OF 'ePROJECT: JAY PRELM [ ] PREPARED BY /bt_ DATE
FINAL EX] CHECKED BY oZ6- DATE _fi//y7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
31.2 rem/hr to thep thyroid following an MSLB, and 0.22 rem/hr tothe who _ body a-- the start of a LOCA. The worst- Lse cumulativedoses are 32.4 rem to the thyroid for a CRDA and 0.85 rem to thewhole body for a LOCA.
2.3 Onsite Outdoor Receptors - Reactor Building Shine
The results presented below are unchanged from revision 0 of thiscalculation.
Direct shine radiation fields at various onsite outdoor locationsdue to post-LOCA airborne radioactivity accumulating on the RBrefueling level are presented in Table 2.5. The worst-case doserate amongst the analyzed receptors is at Loc. #1 (due West fromthe RB) and amounts to 3.4 rad/hr at 4 hours after the postulatedaccident. At the farthest receptor (Loc. #10), the dose rate is0.24 rad/hr. These dose rates are sufficiently high to requireswift access to and from the plant to minimize personnelexposures. The radiation fields remain high for several days,dropping below 10% of the maxima at about 1 week after theaccident.'
For informational purposes, the refueling-level direct shine doserates at 4 hours after the postulated LOCA (from Table 2.5) arealso presented in Fig. 2.2 as a function of distance from the RBcenterline. It is seen that the results follow a smooth curve,
l The peak dose rates in Table 2.5 are lower than those inJAF-CALC-RAD-00008 (ReE. 1) by at least a factor of 3. Ini t-ecurrent calculation, t.ae source accumulating in the refuelinglevel was based on the actual RB air exchange rate of 3.3 airchanges per day (via the SGTS, at 6000 scfm), in lieu of theconservatively selected rate of 1 air change per day in JAF-CALC-RAD-00008.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 1i OF ;.
PROJECT: JAF PRELM [ I PREPARED BY 2_ DATE I110[3/
FINAL EX] CHLECKED BY -26 DATE /1/3,/g7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
with the exception of shielding edge effects which become
apparent at close-in receptors. Thus, the dose rates in Fig. 2.2
can be applied along any direction from the reactor building.
Direct shine dose rates were also calculated at outdoor receptors
adjacent to the east side of the reactor building, at distances
ranging from contact with the 21"1 wall, to 21 ft. The source
term in this case was post-LOCA airborne radioactivity within El.
2721 of the reactor building. The results are summarized in
Table 2.6. The worst-case dose rate (in contact with the wall)
peaks at about 150 mrad/hr at 4 hours after the accident. Due to
the large size of the source and the closeness of the receptor
locations analyzed to the RB, the dose rate drops relatively
slowly with distance. At the farthest location analyzed (21 ft
from the RB wall), the peak dose rate is 106 mrad/hr. However,
these dose rates are low in comparison to those due to radiation
emanating from the RB refueling level.
In summary, the post-LOCA radiation fields at onsite outdoor
receptors are expected to be high. Indeed, plant procedures
(Refs. 40 and 41) suggest that the protected area and site may
have to be evacuated during the initial period following a LOCA.
The information presented in this calculation may be used to
define the most appropriate emergency evacuation route to
minimize exposures.
For the corresponding cumulative doses at the onsite receptors
analyzed, refer to the computer outputs in Attachment B (MATILDA
Run Cases #1 and #2). The worst-case post-LOCA gamma air dose
due to shine from the refueling level of the reactor building is
237 rads, at receptor location #1.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 2.0 OF IlkPROJECT: JAP PRELM [ I PREPARED BY j3,L DATE
FINAL [X] CHECKED BY o44, DATE 11/3/9 7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.1
JA. Power Uprate Project - Site Boundary DosesFollowing Postulated Design-Basis Accidents
Design-BasisAccident
Thyroid(rem)
Wh. Body(rem)
Skin(rem)
LOCA
Regulatory Limit 300.0 25.0
Drywell LeakESF LeakageTotal 2-hr Dose
5.82E+013. 99E+006.22E+01
2.32E+002.20E-022.34E+00
4. 06E+003.40E-024. 09E+00
t of Reg. Limit
MSLB
20.7k 9.36%
Regulatory Limit (a)Total 2-hr Dose
30.05.573E-01
2.57.056E-03 1.113E-02
% of Reg. Limit
CRDA
1.86% 0.28%
Regulatory Limit (b)
Total 2-hr Dose75. 01. 855E-01
6.01.312E-02 2.514E-02
% of Reg. Limit 0.25% 0.22%
Regulatory Limit(c)Total 2-hr Dose
75. 07.376E-01
6.09.131E-02 2.108E-01
% of Reg. Limit 0.98% 1.52%
(a) Ref. 10, SRP,(b) Ref. 10, SRP,(c) Ref. 10, SRP,
Sec. 15.6.4 (10% of 10 CFR 100)Sec. 15.4.9 (25% of 10 CFR 100)Sec. 15.7.4 (25% of 10 CFR 100)
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 2.1 OF 5t8PROJECT: JAF PRELM L ] PREPARED BY JW DATE JiliM
FINAL EX] CHECKED BY pz6- DATE N//n/ 7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.1 (Continued)
BASES (Refer to the pertinent sections for references)
LOCA (Drywell Leakage)
(a) A LOCA takes place at full power (2535.8 MWt + 2%
uncertainty).
(b) All core-inventory noble gases and 25t of the halogens
become airborne within the drywell at the time of the
accident and are available for release.
(c) Leakage from the drywell is at the rate of 1.5t per day,
consisting of 1.27% per day due to containment leakage, and
0.23% per day due to MSIV leakage.
(d) Noble gases and halogens leaking from the drywell are
exhausted to the atmosphere via the Standby Gas Treatment
System (SGTS) and the main stack without holdup or mixing in
the reactor building.
(e) The SGTS filter efficiency is 90% for the removal of all
halogen species.
(f) 4-hour fumigation conditions prevail at the time of the
accident.
LOCA (ESF Leakage)
(a) 50% of the core-inventory halogens mix uniformly with the
coolant in the RHR system (113,400 ft3 ).
(b) The ESF leakage rate is 5 gpm, and is constant from the
start of the LOCA through the duration of the accident.
(c) An additional 30-minute leakage of 50 gpm (due to gross
failure of a passive component) is conservatively assumed to
begin at the time of the accident.
(d) 10k of the halogens in the leaking fluids become airborne
and mix uniformly with the reactor building atmosphere
(2.6E+06 ft3).
(e) Release from the reactor building is through the SGTS and
the main stack at the rate of 6000 scfm.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 22. -OF 3jPROJECT: JAP PRELM [ I PREPARED BY d DATE I///3/q;
FINAL [XI CHECKED BY wf& DATE Iii/3/gTITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.1 (Continued)
(f) 4-hour fumigation conditions prevail at the time of theaccident.
Main Steam Line Break
(a) A line break occurs in the 16" bypass line leading to theturbine steam chest outside containment during full poweroperation. (Note: A break in one of the 24" main steamlines is less restrictive.)
(b) The MSIVs close in 10.5 seconds after the break.
(c) The total discharge through the break prior to isolationamounts to 18,179 lb of steam and 87,118 lb of liquid.
(d) The ensuing high fuel temperatures do not lead to any fueldamage.
(e) The noble gas fission product concentrations in the steamcorrespond to the design values which would yield thestandard release rate to the atmosphere during normaloperation (i.e., 100,000 pCi/sec following a 30-minutedecay). Fifty percent of all noble gases leaving thereactor vessel during the 10.5-sec MSIV closure time (viaall four steam lines) are released through the break. Thehalogen source term in the discharged liquid was selected torepresent the limit for the maximum permissible reactorcoolant system (RCS) activity under power uprate conditions,namely 0.2 gCi/gm I-131 Dose Equivalent.
(f) 100 W of the radioactivity discharged into the turbinebuilding becomes airborne and is released to the atmosphereat ground level over a period of 2 hours. The release ratewas selected to be equivalent to 3 air changes per hour.
Control Rod Drop Accident
(a) The reactor has been operating at full power for an extendedperiod of time. It is shut down, taken critical, andbrought back to the initial temperature and pressureconditions within 30 minutes of the departure from designpower.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 23 OF 9.32PROJECT: JAF PRELM [ ] PREPARED BY < DATE J11/iLL
FINAL [XI CHECKED BY f DATE 11/13/97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.1 (Continued)
(b) A CRDA takes place leading to the failure of 850 fuel rodsat a core location with a radial power peaking factor of1.5.
(c) All activity within the gaps of the failed fuel rods isreleased to the reactor coolant and is instantaneously anduniformly mixed with the coolant in the pressure vessel atthe time of the accident. The released activity correspondsto 10% of all halogens and 10 of all noble gases (except Kr85) in each failed rod, and to 30% of the Kr 85 inventory.
(d) 10% of the iodines and 100l of the noble gases released inthe pressure vessel reach the turbine and condensers.
(e) As a result of elimination of the MSIV-closure and reactor-shutdown functions of the main steam line radiationmonitors, the pathway of post-CRDA atmospheric releases atJAF has changed. Under the new CRDA scenario, the MSIVsstay open and the release is to the offgas system.
(f) As a result of plant shutdown following a CRDA, or as aresult of offgas system automatic isolation due to highradiation fields at the offgas monitors (following a 15-minute delay, which is not considered in the analysis), thereleased radioactivity is retained within the turbine,condensers and the offgas system. Release to the environsis due to leakage from the various contaminated systems intothe turbine building. [Note: Without offgas systemisolation, releases would be via the charcoal holdup systemand the stack and would be significantly less restrictivethan the scenario analyzed.]
(g) 90% of the iodines plate out on system internal surfaces.
(h) The leakage rate from contaminated systems into the turbinebuilding amounts to it per day and lasts for 24 hours. Therelease to the atmosphere is at ground level and there is noholdup within the turbine building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 24 OF
PROJECT: JAP PRELM [ I PREPARED BY 1 DATE Jl/[f9FINAL [XI CHECKED BY /2&; DATE
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
Table 2.1 (Continued)
Refueling Accident
(a) The reactor has been operating at full power for an extendedperiod of time.
(b) The reactor is shutdown, refueling operations are initiatedand an RA takes place at 24 hours after shutdown.
(c) The accident involves the dropping of a fuel assembly andthe ensuing rupture of 125 fuel rods (a conservativeestimate).
(d) The failed fuel rods were at a core location with a radialpower peaking factor of 1.5.
(e) All activity within the gaps of the failed fuel rods isreleased to the fuel pool water. The released activity isconservatively assumed to correspond to 10 of all halogens(except I 129) and 10k of all noble gases (except Kr 85) ineach failed rod, and to 30k of the I 129 and Kr 85inventories.
(f) The halogen composition (inorganic, organic and particulatespecies) and the pool halogen retention factors are suchthat 99k of all released halogens are assumed to be retainedby the water in the fuel pool. The retention of noble gasesby the pool water is negligible.
(g) Radioactive gases which escape the pool are released to theatmosphere via the SGTS and the main stack over a 2 hourperiod. The release rate was selected to be equivalent to 3air changes per hour.
(h) 4-hour fumigation conditions prevail at the time of theaccident.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 2-5 OF ets
PROJECT: JAP PRELM [ I PREPARED BY /M_ DATE
FINAL [X] CHECKED BY <a9- DATETITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.2
JAF Power Uprate Project - Doses &t the Low Population ZoneFollowing Postulated Design-Basis Accidents
Design-BasisAccident
Thyroid(rem)
Wh. Body(rem)
Skin(rem)
LOCA
Regulatory LimitDrywell LeakESF LeakageTotal 30-day dose
300.06.317E+015.496E+006. 87E+01
25. 01.856E+003.648E-021.89E+00
75.03.124E+005.505E-023.18E+00
% of Reg. Limit
MSLB
22 . 9% 7.56%
Regulatory LimitTotal 24-hr Dose
30.06.241E-02
2.58.616E-04 1.330E-03
* of Reg. Limit
CRDA
0.21% 0.03%
Regulatory LimitTotal 24-hr Dose
75. 01.258E-01
6.04.524E-03 8.417E-03
% of Reg. Limit
RA
0.17% 0.08%
Regulatory LimitTotal 24-hr Dose
75. 02.879E-01
6.04.277E-02 9.290E-02
% of Reg. Limit
Refer to Table 2.1and other details
0.38% 0.71%
for a listing of the basic assumptions
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE _ _6 OF
PROJECT: JAF PRELM [ ] PREPARED BY DATEFINAL [X] CHECKED BY -2 Af- DATE ______
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
Table 2.3
JAF Power Uprate Project - Dose Rates atOutdoor Receptors in the General Vicinity
of the Old Administration Building
Time(hours)
ThyroidDose Rate(rem/hr)
Whole BodyDose Rate(rem/hr)
SkinDose Rate(rem/hr)
LOCADrywell Leak
0. OOOE+005.000E-011. OOE+002. OOE+00
4. OOOE+008. OOOE+001.200E+011. 800E+01
2.400E+013 . 600E+014.800E+017.200E+01
9.600E+011.680E+023.360E+027.44OE+02
5. 238E-015. 190E-015.145E-015.060E-01
4. 907E-014.648E-013.228E-013.028E-01
2. 860E-011.301E-011.196E-011. 038E-01
9.192E-022.478E-021.217E-022.179E-03
2.242E-018.675E-026.727E-025.327E-02
4. 064E-022. 929E-021.775E-021.326E-02
9. 901E-033.057E-031. 906E-031. 117E-03
8.805E-042.330E-048.667E-058.554E-06
3 .204E-011.130E-018.748E-026.927E-02
5.323E-023. 928E-022.415E-021.838E-02
1.394E-024.404E-032.849E-031.749E-03
1.396E-033.631E-041.353E-041.466E-05
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 2_- OF 5)5
PROJECT: JAF PRELM [ ] PREPARED BY /2 DATEFINAL [X] CHECKED BY *ZC- DATE 111/3j/7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.3 (Continued)
JAF Power Uprate Project - Dose Rates atOutdoor Receptors in the General Vicinity
of the Old Administration Building
Time(hours)
ThyroidDose Rate(rem/hr)
Whole BodyDose Rate(rem/hr)
SkinDose Rate(rem/hr)
LOCAESF Leak
O.OOOE+OO5.OOOE-O11.OOOE+OO2.OOOE+OO
4.OOOE+OO8.OOOE+OO1.200E+O11.800E+O1
2.400E+O13.600E+O14.800E+O17.200E+O1
9.600E+O11.680E+023.300E+027.44OE+02
MSLB
O.OOOE+OO5.OOOE-O11.OOOE+OO2.OOOE+OO
4.OOOE+OO8.OOOE+OO1.200E+O11.800E+O1
O.OOOE+OO3.901E-023.998E-024.164E-02
4.407E-024.642E-023.415E-023.342E-02
3.218E-021.485E-021.373E-021.199E-02
1.069E-022. 939E-031.511E-033.021E-04
3.118E+O16. 829E+OO1.497E+OO7.216E-02
1. 687E-04a .373E-104.507E-156.123E-23
O. OOOE+OO8.288E-048.447E-048. 998E-04
1. 051E-031.233E-039.205E-047. 638E-04
5.707E-041.536E-047.462E-052.607E-05
1.644E-054.171E-061. 962E-063.552E-07
7.587E-021.324E-022 .387E-038.367E-05
1.289E-074. 606E-131.683E-181. 666E-26
0. OOOE+001. 052E-031. 077E-031.155E-03
1.360E-031.604E-031. 197E-039.972E-04
7.482E-042.018E-049. 957E-053 .604E-05
2.307E-055.752E-062.678E-064.823E-07
2.786E-014.692E-028.542E-033. 091E-04
5.142E-072.090E-128.391E-189.234E-26
2.400E+Ol 8.447E-31 1.772E-34 1. 057E-33
NYPA - CALC.# JA1-CALC-RAD-00048 REV 1 PAGE 2A -OF
PROJECT: JAF PRELM [ ] PREPARED BY 7tt DATE 1/ 13/19
FINAL [X] CHECKED BY /2G; DATE /1/13/9TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.3 (Continued)
JAF Power Uprate Project - Dose Rates atOutdoor Receptors in the General Vicinity
of the Old Administration Building
Time(hours)
ThyroidDose Rate(rem/hr)
Whole BodyDose Rate(rem/hr)
SkinDose Rate(rem/hr)
CRDA0. OOOE+005. 0OOE-01l. OOOE+002. OOOE+00
1. 735E+001. 719E+001. 704E+001. 677E+00
6.314E-022.252E-021. 661E-021.232E-02
5. 627E-011.182E-018.534E-026. 089E-02
4 .OOOE+008. OOOE+001.200E+011. 800E+01
2 .400E+01
1. 627E+001. 542E+001. 256E+001.179E+00
1. 116E+00
8.295E-034.730E-032.803E-031.935E-03
1.465E-03
4.005E-022. 534E-021. 702E-021.339E-02
1. l1OE-02
RA0 . OOOE+005.000E-011. OOOE+002. OOOE+00
3. 931E-028.734E-031. 941E-039.585E-05
1.707E-023.995E-031.201E-034.279E-04
2.623E-026.084E-031.756E-035.650E-04
4. OOOE+008. OOOE+001.200E+011.800E+01
2.400E+01
2.338E-071.394E-126. 067E-188. 880E-26
1.303E-33
3.182E-042.103E-041.052E-045. 693E-05
3.102E-05
4.135E-042.736E-041.368E-047.432E-05
4. 075E-05
Note: The results in this table were conservatively basedon the atmospheric' dispersion factors applicable to tl.control-room outside air intake located on the roof of theold administration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 219 -OF _ _
PROJECT: JAY PRELM E I PREPARED BY fr( DATE l1l3LSFINAL EX] CHECKED BY -26- DATE (//3/97
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
Table 2.4
JAF Power Uprate Project - Integrated Doses (ContinuousOccupancy) at Outdoor Receptors in the General Vicinity
of the Old Administration Building
Design-BasisAccident
Thyroid(rem)
Wh. Body(rem)
Skin(rem)
LOCA
Drywell LeakESF Leakage
Total 30-day dose
2.45E+012.72E+00
2.72E+01
8.23E-012.77E-02
8.51E-01
1. 13E+003.63E-02
1. 17E+00
MSLB
Total 24-hr dose 1.03E+01 2.17E-02 7.76E-02
CRDA
Total 24-hr dose 3.24E+01 1.23E-01 6. 91E-01
RA
Total 24-hr dose 1.31E-02 9.24E-03 1.34E-02
Note: The results in this table were conservatively basedon the atmospheric dispersion factors applicable to thecontrol-room outside air intake located on the roof of theold administration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 3.O OF ____l
PROJECT: JAF PRELM [ I PREPARED BY X DATE LI.,FINAL [XI CHECKED BY f 6- DATE ,//hM97
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.5
JAF Power Uprate project - Onsite Outdoor Dose Rates (r.d/hr) dueto Post-LOCA Shine from Airborne Activity in the Refueling Level
Time (hr) Loc. #1 Loc. #2 Loc. #3 Loc. #4
0 .00.51. 02.0
4.08.0
12.018. 0
24. 036.048.072. 0
96. 0168 .0336.0744. 0
Time (hr)
0 .00.51.02.0
4.08.0
12 . 018. 0
24. 036.048.072. 0
96. 0168 .0336 .0744. 0
0. 000E+001. 342E+002. 059E+002. 857E+00
3 .375E+003.264E+002. 899E+002. 387E+00
1.970E+001. 387E+001.044E+007.157E-01
5.666E-013 .611E-011.559E-012 .297E-02
Loc. #5
0.0002E+007.971E-011.219E+001.683E+00
1. 964E+001. 853E+001.610E+001.290E+00
1.042E+007.064E-015.146E-013.358E-01
2.592E-011.637E-017.374E-021.183E-02
0.0 00E+001.200E+001.840E+002.550E+00
3. 002E+002. 886E+002. 549E+002.084E+00
1. 709E+001. 188E+008.854E-015.974E-01
4.694E-012.986E-011.306E-011. 977E-02
Loc. #6
0.0002E+003. 881E-015. 925E-018. 147E-01
9.423E-018.710E-017.426E-015.828E-01
4.638E-013.081E-012.210E-011.408E-01
1.071E-016.708E-023.082E-025.138E-03
0. OOOE+001.211E+001.858E+002.579E+00
3. 045E+002.942E+002.611E+002.148E+00
1.771E+001.244E+009.353E-016.395E-01
5.057E-013.222E-011.394E-012.062E-02
Loc. #7
0.000E+003.134E-014.782E-016.567E-01
7.575E-016.956E-015.894E-014.594E-01
3.639E-012.403E-011.717E-011.087E-01
8.234E-025.142E-022.374E-023.995E-03
0.0 00E+001.230E+001.888E+002 .620E+00
3.095E+002.993E+002.658E+002.188E+00
1.806E+001.272E+009.582E-016.570E-01
5.202E-013.316E-011.431E-012.106E-02
Loc. #8
0.000E+001.375E-012.093E-012.863E-01
3.266E-012.916E-012.405E-011.817E-01
1.410E-019.097E-026.408E-023.968E-02
2.957E-021.822E-028.532E-031.477E-03
Refer to Fig. 2.1 for the receptor locations.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 3f -OF 8i
PROJECT: JAF PRELM [ ] PREPARED BY & DATE ilhA/9/ IFINAL EX] CHECKED BY A96- DATE _/13//97
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
Table 2.5 (Continued)
JAF Power Uprate project - Onsite Oui:door Dose Rates (rad/hr) dueto Post-LOCA Shine from Airborne Activity in the Refueling Level
Time (hr) Loc. #9 Loc. #10 Loc. #11 Loc. #12
0.00.51.02.0
4.08.0
12. 018. 0
24. 036.048.072 . 0
96.0168 .0336.0744. 0
Time (hr)
0.00.51.02.0
4.08.0
12. 018. 0
24. 036.048.072 . 096.0
168.0336.0744. 0
0. OOOE+003.410E-015.206E-017. 157E-01
8.273E-017.637E-016.503E-015.097E-01
4.053E-012.689E-011.928E-011.227E-01
9.323E-025. 837E-022.684E-024.482E-03
Loc. #13
0. OOOE+005.314E-018.121E-011. 118E+00
1.299E+001. 212E+001.043E+008. 267E-01
6.624E-014.440E-013 .206E-012.064E-011.5-3..-019. 937E-024.528E-027.427E-03
0.OOOE+001. 019E-011.550E-012.117E-01
2.405E-012.123E-011.731E-011.291E-01
9.934E-026.346E-024.446E-022.732E-02
2.023E-021.239E-025.823E-031.015E-03
Loc. #14
0. 000E+002.861E-014.364E-015.991E-01
6.902E-016.319E-015.338E-014.147E-01
3 .278E-012.160E-011.541E-019.729E-027. 353E-024. 587E-022.122E-023. 584E-03
0. OOOE+002.763E-014.215E-015. 787E-01
6. 668E-016.108E-015. 163E-014. 013E-01
3.174E-012. 092E-011.493E-019.430E-02
7. 131E-024.449E-022. 057E-023.472E-03
Loc. #15
0. OOOE+001.038E+001.589E+002. 197E+00
2. 575E+002.450E+002.144E+001.734E+00
1.410E+009.655E-017. 096E-014.693E-013 .650E-012.313E-011.030E-011.614E-02
0. OOOE+002.023E-013.083E-014.225E-01
4. 847E-014.391E-013.673E-012.821E-01
2.213E-011.445E-011.025E-016.419E-02
4. 822E-022.993E-021.393E-022.380E-03
Loc. #16
0. OOOE+005. 638E-018. 617E-011. 187E+00
1.380E+001.290E+001. 111E+008.821E-01
7. 077E-014.750E-013.435E-012.215E-011.697E-011.068E-014.860E-027.949E-03
Refer to Fig. 2.1 for the receptor locations.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 32_ OF t
PROJECT: JAF PREL4M C ] PREPARED BY itZ. DATEFINAL [XI CHECKED BY 6y DATE L_/i3/9
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
Table 2.5 (Continued)
BASIS
LOCA (Drywell Leakage)
(a) A LOCA takes place at full power (2535.8 MWt + 2%uncertainty).
(b) All core-inventory noble gases and 25% of the halogensbecome airborne within the drywell at the time of theaccident and are available for release.
(c) Leakage from the drywell is at the rate of 1.5% per day.(Note: The contribution of ESF component leakage wasdetermined to be negligible in comparison to the drywellleakage.)
(d) Noble gases and halogens leaking from the drywell mixuniformly with the RB atmosphere (2.60E+06 ft3) and areexhausted to the atmosphere via the Standby Gas TreatmentSystem (SGTS) and the main stack at a flow rate of 6000scfm.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 33 -OF __
PROJECT: JAF PRELM [ I PREPARED BY Mt DATE JI/L3/9FINAL [XI CHECKED BY 26,- DATE q//3/g 7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 2.6
JAF Power Uprate project - Dose Rates (rad/hr) at VariousDistances from the RB East Wall due to Post-LOCA Shine
from Airborne Radioactivity in RB El. 272'
Time (hr) Contact 3 ft 6 ft 9 ft
0 .00.51. 02.0
4.08.0
12. 018.0
24. 036.048.072. 0
96.0168.0336.0744. 0
0. OOOE+007.247E-021.073E-011.442E-01
1.542E-011.083E-016.709E-023.366E-02
1.859E-027.247E-033.608E-031.462E-03
8.277E-043 .427E-041.519E-042. 735E-05
0. OOOE+006. 815E-021. 009E-011.357E-01
1.4="E-011. 019E-016 .3:9E-023. 172E-02
1.754E-026.843E-033.410E-031.383E-03
7. 834E-043.245E-041.438E-042. 590E-05
0. OOOE+006.422E-029.505E-021.278E-01
1.367E-019. 609E-025. 960E-022. 994E-02
1.656E-026.468E-033.225E-031.310E-03
7.421E-043.077E-041.364E-042.456E-05
0. OOOE+006. 079E-028. 998E-021.210E-01
1.294E-019. 101E-025. 647E-022.838E-02
1. 571E-026.138E-033. 062E-031.244E-03
7. 053E-042. 925E-041.297E-042.335E-05
Time (hr) 12 ft 15 ft 18 ft 21 ft
0. 00.51. 02.0
4.08.0
12. 018. 0
24. 036.048.072. 0
96.0168.0336 .0744. 0
0.OOOE+005. 773E-028.545E-021.150E-01
1.229E-018. 648E-025.367E-022.699E-02
1.494E-025.844E-032. 917E-031.186E-03
6.723E-042.790E-041.237E-042.228E-05
0. OOOE+005.491E-028.129E-021.094E-01
1.170E-018.231E-025.111E-022.572E-02
1.425E-025. 575E-032.785E-031.133E-03
6.425E-042.667E-041.183E-042.130E-05
0. OOOE+005.229E-027.741E-021.041E-01
1. 114E-017. 843E-024.872E-022.453E-02
1.359E-025.325E-032.661E-031.084E-03
6.148E-042.554E-041. 133E-042.039E-05
0. OOOE+004.981E-027.374E-029. 922E-02
1.062E-017.476E-024. 647E-022.341E-02
1.298E-025. 090E-032.546E-031.038E-03
5. 888E-042.448E-041.086E-041.955E-05
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 3 9 OF
PROJECT: JAF PRELM [ ] PREPARED BY DATE 1JZ1L?FINAL [XI CHECKED BY Ileg DATE _1/__I__
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
Fig. 2.1 Receptor locations for the dose rates in Table 2.5from airborne radioactivity accumulatir. in the RBrefueling level.
- 0 0 0 8
NYPA - CALC.# JF-CALC-RAD-00048 REV 1 PAGE 3• OF
PROJECT: JAF PRELM t I PREPARED BY At DATE I
FINAL [XI CHECKED BY A2G DATE /I//3/g7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Fig. 2.2 Direct Shine Dose Rates due to Post-LOCA Activityin the RB Refueling .evel at 4 hours after a LOCA(Receptor Locati-ns jf1 - #16 from-Table 2.5,arranged as a function of distance from the RB)
D-
LEGEND0 - 4 Hours of ter occxdent
L
c\0L
a)
0C
0
0CD
C) -
'o 2 I I I I I
10 3 103
Distonce from RB CenterLxne (ft)
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 39 OF
PROJECT: JAF PRELM E I PREPARED BY /W6 DATE 1113FINAL [X] CHECKED BY A DATE I11/ 9 7
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
3. METHODS OF ANALYSIS
Post-accident radiation exposures at the locations of interest
were computed using the following:
a) The methodology and assumptions in the regulatory
guides (Refs. 14 through 19) and the pertinent sections
of the Standard Review Plan (Ref. 10),
b) Appropriate source terms, release pathways,
decontamination factors and other assumptions,
c) Post-accident atmospheric dispersion factors based on
8-years' worth of hourly meteorological data collected
on site by Niagara Mohawk, from JAF-CALC-RAD-00007,
Rev. 2 (Ref. 2), and
d) The following CRE Computer Codes:
DORITA-2 (Ref. 6) Computation of radiation exposures,
and definition of gamma spectra
associated with post-LOCA airborne
radioactivity within the RB.
QAD-CGGP (Ref. 7) Determination of the relative gamma
fluxes at the locations of interest
(in terms of MeV/sec-cm2 per
MeV/sec emitted by a source, as a
function of gamma energy), for
gamma radiation emanating from the
RB refueling level and from El.
272'.
MATILDA (Ref. 8) Computation of dose rates (and
cumulative doses) at the receptors
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 37a OF 3tPROJECT: JAF PRELM 1 1 PREPARED BY f DATE 111tj'
FINAL [X] CHECKED BY 4&-- DATE It//a/3 7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
of interest as a function ot post-
accidert time, using the gaL.Lna
spectra generated by DORITA-2 and
the relative gamma fluxes produced
by QAD-CGGP.
Refer to Secs. 4 through 8 for further details.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 3 8 -OF T?PROJECT: JAF PRELM [ ] PREPARED BY t DATE 4LftL/9-
FINAL [X] CHECKED BY 126, DATE //3/97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
4. RADIATION EXZOSURES FROM A LOSS OF COOLANT ACCIDENT
Rele:ase pathwv.ays and contributing radiation s-1rces which
are typically addressed in the analysis of a LOCA are the
following:
(a) Drywell leakage and ESF component leakage, followed by
atmospheric releases and cloud exposures,
(b) Direct gamma radiation from airborne radioactivity
accumulating on the refueling floor of the reactor
building, and
(c) MSIV leakage.
This part of the calculation addresses the immersion
exposures at onsite and offsite outdoor receptors due to drywell
and ESF component leakage. Shine from the RB [Item (b) above] is
addressed in Sec. 8 of this calculation. The MSIV-leakage
pathway is not applicable at JAF since the plant is equipped with
a Main Steam Leakage Collection System (MSLCS) whose safety
objective is to collect and process leakage past the MSIVs
following a LOCA; see Ref. 5 for more information.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 31- OF r8PROJECT: JAF PRELM 1 ] PREPARED BY PAZ. DATE if/
FINAL [XI CHECKED BY X6; DATE fr///y9-7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
4.1 Drywell Leakage
4.1.1 Basic Data and Assumptions
The following data and assumptions were used in the computation
of immersion exposures at outdoor receptors as a result of post-
LOCA drywell leakage:
(a) A LOCA takes place at full power (2535.8 MWt + 2%
uncertainty, i.e., 2586.5 MWt) [Ref. 11 and Reg. Guide 1.49
(Ref. 17)].
(b) The full-power core inventory for the radionuclides of
interest is shown in the table which follows (based on
information from Ref. 12):
Nuclide Activ.(Ci) Nuclide Activ.(Ci)
Br 83 8.078E+06* Kr 83m 8.114E+06Br 84 1.432E+07 Kr 85m 1.742E+07Br 85 1.717E+07 Kr 85 7.798E+05
Kr 87 3.342E+07I 129 2.254E+00 Kr 88 4.733E+07I 130 2.705E+06 Kr 89 5.887E+07I 131 6.805E+07I 132 9.945E+07 Xe 131m 4.092E+05I 133 1.423E+08 Xe 133m 5.962E+06I 134 1.566E+08 Xe 133 1.430E+08I 135 1.344E+08 Xe 135m 2.695E+07I 136 6.479E+07 Xe 135 1.847E+07
Xe 137 1.255E+08Xe 138 1.192E+08
* 3.123E+03 (Ci/MWt from Ref. 12) x 2586.5 (MWt)(c) 100% of the noble gases and 25% of the halogens present in
the core are released instantaneously to the drywell where
they are available as an aerosol for leakage to the
secondary containment [Reg. Guide 1.3 (Ref. 14)].
(d) The halogen co...Vsition airborne within the drywell is as
follows: 91% elemental, 4% organic and 5k particulate [Reg.
Guide 1.3 (Ref. 14)].
(e) Leakage from the drywell is at the rate of 1.5% per day
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 4c OFPROJECT: JAF PRELM [ ] PREPARED BY fizL DATE 11L/3/Lf
FINAL [X] CHECKED BY Anti DATE 0/13/97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
(UFSAR, Rev. 0, 7/82, Secs. 14.8.1.5 and 14.8-22). This
leak ra&e accounts for brth dryvell containment leakage and
MSIV leakage and is assumed to be constant for the accident
duration. The design leak rate is 0.5% per day of
containment volume (Technical Specifications Sec. 4.7.A.2.8,
and UFSAR, Rev. 0, 7/82, Secs. 11.5.3.10 and 14.6.1.3.5).
Use of the 1.5% per day value is conservative.
(f) All the noble gases and halogens leaking from the drywell
are instantaneously exhausted to the atmosphere via the
Standby Gas Treatment System (SGTS) and the main stack
without mixing in the reactor building.
(g) The SGTS filter efficiency for the removal of halogens is
90k for all halogen species (Ref. 19). Although an
efficiency of 95 percent for halogen removal could have been
employed for the SGTS filters, the present revision
conservatively employs an efficiency of 90% to provide some
relief in testing.
(h) The atmospheric dispersion factors associated with the
transport of released radioactivity to the onsite and
offsite outdoor receptors of interest are as follows (from
revision 2 of Ref. 2):Time Dispersion Parameter (sec/m3 )
Receptor IntervalLocation (hrs) Conc. X/Q Gamma X/Q
SB 0 - 2 5.24E-5 4.75E-5
LPZ 0 - 4 2.04E-5 1.90E-54 - 8 2.17E-6 3.91E-68 - 24 9.53E-7 1.52E-6
24 - 96 3.90E-7 5.68E-796 744 1.08E-7 1.38E-7
Onsite 0 - 8 9.26E-7 3.24E-68 - 24 6.75E-7 2.45E-6
24 - 96 3.39E-7 1.34E-696 - 744 1.26E-7 5.60E-7
NYPA - CALC.# JF-CALC-RAD-00048 REV 1 PAGE 41 OF tPROJECT: JAF PRELM ( I PREPARED BY /1 DATE
FINAL [X] CHECKED BY /I&- DATE 11A3191TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Note the following:
1. The concentration (X/Q)s are for computing the
inhalation exposures and the beta component of the skin
dose.
2. The gamma (X/Q)s are for computing the whole body doses
due to exposure to finite radioactive clouds above.
3. The dispersion parameters at the SB and during the
first 4 hours at the LPZ are due to the prescribed
assumption of fumigation conditions prevailing at the
site at the time of the accident.
4. The dispersion parameters listed above for onsite
outdoor receptors are for the CR air intake (from Ref.
2). They were conservatively assumed to apply to
onsite outdoor receptors at grade elevation, in the
general vicinity of the old administration building.
Fumigation conditions at these locations are not
applicable; see Ref. 2 for details.
(i) The breathing rates at the various receptors of interest
were as follows (Ref. 14 for the SB and LPZ, and
conservative high breathing rate for the onsite outdoor
receptors):
Time BreathingReceptor Interval RateLocation (hrs) (m 3/sec)
SB 0 - 2 3.47E-4
LPZ 0 - 8 3.47E-48 - 24 1.75E-4
24 - 744 2.32E-4
Onsite 0 - 744 3.47E-4
(j) Thyroid exposures were based on the dose conversion factors
in ICRP-30 (Ref. 45).
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 42. OF TePROJECT: JAP PRELM E I PREPARED BY q1 DATE
FINAL [X] CHECKED BY ,2&- DATE II//,g7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
4.1.2 Results
Radiation exposures at the onsite and offsite outdoor receptors
of interest due to drywell leakage following a design-basis LOCA
were calculated using the DORITA-2 computer code and the data and
assumptions listed above. Copies of the DORITA-2 outputs appear
in Attachment B to this calculation (Computer Run Cases #1, #2
and #3).
Table 4.1 which follows presents the time-dependent thyroid,
whole body and skin doses at the receptors of interest due to
post-LOCA drywell leakage. Refer to Tables 2.1, 2.2 and 2.4 for
a summary of the exposures, and to Table 2.3 for the time-
dependent dose rates at onsite outdoor receptor locations in the
general vicinity of the old administration building.
NYPA - CALC.# JAF-C.LC-RAD-00048 REV 1 PAGE 43 -OF ' g
PROJECT: JAF PRELM t I PREPARED BY OI DATE 3 9-FINAL EX] CHECKED BY A/6 DATE I1/L3197
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
Table 4.1
JAF Power Uprate Project - Poit-LOCA Time-DependentDoses at Outdoor Receptors Due to Drywell Leakage
Time(hours)
SB:o.OOOE+002. OOOE+00
LPZ:0. OOOE+002. OOOE+004. OOOE+008. OOOE+00
2 .400E+019. 600E+017.440E+02
ThyroidDose (rem)
o.OOOE+005. 824E+01
0. OOOE+002.267E+014.462E+014. 910E+01
5.263E+015. 890E+016.317E+01
Whole BodyDose (rem)
o.OOOE+002.321E+00
o.OO0E+009.283E-011.470E+001.635E+00
1.784E+001.842E+001.856E+00
SkinDose (rem)
0. OOOE+004 . 059E+00
0. OOOE+001. 609E+002. 508E+002. 748E+00
2. 984E+003. 092E+003 .124E+00
Onsite:0. OOE+005. OOOE-011. OOOE+002. OOOE+00
0 . 0OOE+002. 607E-015.190E-011.029E+00
0. OOOE+006. 142E-029.912E-021.583E-01
0 . 0OOE+008.248E-021.31SE-012. 085E-01
4. OOOE+008. OOOE+001.200E+011. 800E+01
2.400E+013. 600E+014.800E+017. 200E+01
9. 600E+011. 680E+023 .360E+027.440E+02
2. 026E+003.935E+005. 257E+007. 132E+00
8.897E+001. 054E+011.203E+011.470E+01
1. 705E+011. 914E+012. 212E+012.449E+01
2.507E-013.872E-014.664E-015.587E-01
6.277E-016.770E-017. 059E-017.399E-01
7.635E-017. 846E-018.094E-018.23OE-01
3.290E-015.098E-016.165E-017.433E-01
8 .397E-019. 094E-019.517E-011.004E+00
1. 041E+001. 074E+001. 113E+001. 134E+00
XYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE -4 OFPROJECT: JAF PRELM [ I PREPARED BY K DATE J/3/98
FINAL [X] CHECKED BY o.2 - DATETITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
4.2 ESF Component Leakage
4.2.1 Basic Data and Assumptions
The following data and assumptions were used to calculate the
post-LOCA dose contribution from ESF component leakage:
(a) A LOCA takes place at full power (2586.5 MWt).
(b) The core inventory for the radionuclides of interest
(halogens in this case) is as shown in Sec. 4.1.1, Item (b).
(c) 50% of the total halogen activity present in the core mixes
uniformly with the coolant in the RHR system, which has a
total fluid mass of 3.21 x 109 grams (Ref. 26) . This is
equal to approximately 113,400 ft3, consisting of (431190
lbs / 62.4 lbs/ft3) = 6,900 ft3 of cold RCS coolant (from
JAF Drawing 5.01-lOlA), 105,600 f3t of torus water (from
UFSAR, Rev, 0, 7/82, Table 5.2-1), and 900 ft3 of water from
other sources.
(d) Total ESF component leakage rate into the RB is 5 gpm (Tech.
Specifications, Sec. 3.6.D, for unidentified leakage inside
the containment, and UFSAR, Rev. 1, 7/83, Sec. 4.10.3.2, for
maximum allowable leakage rate from unidentified sources in
the reactor coolant pressure boundary [both inside and
outside the primary containment and systems essential to
safe plant shutdown, i.e., ECCS]); it corresponds to a
fractional rate from the recirculating water system of
0.00849 vol/day.
(e) The ESF component leakage of 5 gpm is assumed to be constant
from the start of the LOCA through the duration of the
accident.
(f) An additional leakage contribution due to a gross failure of
a passive component with an assumed leak rate of 50 gpm is
included in the model (Ref. 10, SRP, Sec. 15.6.5, Appendix
B). This leakage is assumed to begin at the time of LOCA
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE .S3 OF
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Offsite Outdoor Receptors Following Design-Basis Accidents
onset and lasts for a period of 30 minutes. (This
assumption is more conservative than the SRP model wLich
assumes that the additional leakage begins at 24 hours after
the LOCA).
(g) It is further assumed that 10% of the halogens contained in
the water from ESF component leakage become airborne within
the Reactor Building (Ref. 10, SRP, Sec. 15.6.5, Appendix
B), and mix uniformly with the RB atmosphere.
(h) Release from the reactor building is through the SGTS and
the main stack at the rate of 3.3 air changes per day [based
on an SGTS flow of 6000 scfm with one fan operating (UFSAR,
Rev. 0, 7/82, Sec. 5.3.3.4)].
(i) The SGTS filter efficiency for the removal of halogens is
90% for all halogen species (Ref. 19); see discussion in
previous section for additional details.
(j) The atmospheric dispersion factors associated with the
transport of released radioactivity to the onsite and
offsite outdoor receptors of interest, and other exposure-
related parameters are described under Items (h), (i) and
(j) in Sec. 4.1.1.
4.2.2 Results
Radiation exposures at the onsite and offsite outdoor receptors
of interest due to ESF component leakage following a design-basis
LOCA were calculated using the DORITA-2 computer code and the
data and assumptions listed above. Copies of the DORITA-2
outputs appear in Attachment B to this calculation (Computer Run
Cases #1, #2 and #3).
Table 4.2 which follows presents the time-dependent thyroid,
whole body and skin doses at the receptors of interest due to
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE -OFPROJECT: JAF PRELK [ ] PREPARED BY g DATE
FINAL [XI CHECKED BY H DATE I//7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
post-LOCA ESF component leakage. Refer to Tables 2.1, 2.2 and
2.4 for n. summary of the exposures, and to Table 2 3 for the
time-dependent dose rates at onsite outdoor receptor locations in
the general vicinity of the old administration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE -OFPROJECT: JAF PRELM [ I PREPARED BY AL DATE /11/37f6
FINAL EX] CHECKED BY 1-2 DATE _1//3/97
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 4.2
JAF Power Uprate Project - Post-LOCA Time-DependentDoses at Outdoor Receptors Due to ESF Component Leakage
Time(hours)
SB:
0. OOOE+005. 0OOE-012. OOOE+00
ThyroidDose (rem)
0. OOOE+005. 600E-013 . 988E+00
Whole BodyDose (rem)
o.OOOE+003.151E-032.203E-02
SkinDose (rem)
0. 0OOE+004. 797E-033.400E-02
LPZ
0. OOOE+005. 0OOE-012. OOE+004. OOOE+00
8. OOOE+002.400E+019. GOOE+017.440E+02
0. OOOE+002.180E-011. 553E+003.445E+00
3. 871E+004. 253E+004. 973E+005.496E+00
0. OOOE+001.260E-038.812E-032. 025E-02
2.585E-023 .382E-023.617E-023.648E-02
0. OOOE+001. 907E-031.351E-023. 150E-02
3.935E-025. 088E-025.451E-025.505E-02
Onsite0. OOOE+005.000E-011. OOOE+002. OOOE+00
0. 000E+009. 896E-032 .965E-027. 048E-02
o.OOOE+002.149E-046.331E-041.503E-03
0. OOOE+002.725E-048.044E-041.917E-03
4. OOOE+008. OOOE+001.200E+011. 800E+01
2.400E+013. 600E+014. 800E+017. 200E+01
9. GOOE+011. 680E+023 .360E+027.440E+02
1.564E-013. 382E-014.745E-016.777E-01
8.746E-011. 061E+001. 232E+001.539E+00
1. 811E+002. 057E+002.418E+002.724E+00
3 .453E-038. 097E-031.184E-021.694E-02
2.094E-022.364E-022.495E-022. 6OOE-02
2.648E-022.686E-022.735E-022.773E-02
4 .431E-031.046E-021.532E-022.197E-02
2.720E-023.073E-023.246E-023 .388E-02
3.456E-023.507E-023.574E-023. 626E-02
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE if OF.
PROJECT: JAF PRELM [ ] PREPARED BY PC- DATE 1FINAL [XI CHECKED BY A26- DATE _//1 3g 7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
4.3 Total LOCA Dose
The tot-1 LOCA radiation doses (Rue to both drywell and ESF
component leakage are shown in Table 4.3. The table was prepared
by summing the results in Tables 4.1 and 4.2. Note that, for the
31-day exposures at the LPZ, drywell leakage contributes 92% of
the total thyroid dose and 98.4% of the total whole body dose.
NYPA - CALC.# JXF-CALC-RAD-00048 REV . PAGE 4 OF +8PROJECT: JAF PRELM t ] PREPARED BY 6¢_ DATE /I 9
FINAL EX] CHECKED BY /26- DATE 1/1/g9/7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 4.3
JAF Power Uprate Project - Pc.st-LOCA Time-Dependent Dosesat Outdoor Receptors Due to Drywell and ESF Component Leakage
Time(hours)
ThyroidDose (rem)
Whole BodyDose (rem)
SkinDose (rem)
SB:0. OOE+002. OOE+00
o.OOE+006. 22E+01
0 . OOE+002. 34E+00
o.OOE+004.09E+00
LPZ0. OOE+002.OOE+004.OOE+008. OOE+00
2.40E+019. 60E+017.44E+02
0 . OOE+002.42E+014 . 81E+015.30E+01
5. 69E+016.39E+016. 87E+01
0. OE+009.37E-011.49E+001. 66E+00
1. 82E+001. 88E+001.89E+00
0. OOE+001.62E+002.54E+002. 79E+00
3 .03E+003.15E+003.18E+00
Onsite0. OOE+005. OOE-011.OOE+002. OE+00
0. OOE+002. 71E-015.49E-011. 10E+00
0. OOE+006.16E-029.98E-021.60E-01
0. OOE+008.28E-021.32E-012. 1OE-01
4. OOE+008.OOE+001. 20E+011.80E+01
2.40E+013. 60E+014. 80E+017.20E+01
9. 60E+011. 68E+n23.36E+027.44E+02
2. 18E+004.27E+005. 73E+007.81E+00
9.77E+001. 16E+011. 33E+011. 62E+01
1. 89E+012. 12E+012.45E+012. 72E+01
2.54E-013.95E-014.78E-015.76E-01
6.49E-017. 01E-017.31E-017.66E-01
7. 90E-018.11E-018.37E-018.51E-01
3.33E-015.20E-016.32E-017.65E-01
8.67E-019.40E-019.84E-011. 04E+00
1.08E+001. 11E+001. 15E+001. 17E+00
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE SO OF le
PROJECT: JAF PRELI [ I PREPARED BDATE E_ DATFINAL [XI CHECKED BY Ae-- DATE Jf/3191
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
5. RADIATION EXPOSURES FROM A MAIN STEAM LINE BREAK
5.1 Basic Data and Assumptions
As was the case with all accident analyses documented in this
calculation, the computation of radiation exposures associated
with a postulated MSLB outside containment was based on data and
assumptions consistent with the regulatory guidelines,
specifically, Ref. 15 (Regulatory Guide 1.5), and the Standard
Review Plan (Ref. 10, Sec. 15.6.4). A description of the data
and assumptions (as extracted from Ref. 5) follows.
(a) A main steam line break occurs outside containment during
full power operation.
(b) The main steam isolation valves close in 10.5 seconds after
the break (UFSAR, Rev. 0, 7/82, Sec. 14.6.1.5.1.e, pg 14.6-
29). (Note: Actual closure time is approximately 3 to 5
seconds.)
(c) The accident involves a break in the 16" bypass line leading
to the turbine bypass steam chest; this would release more
reactor coolant into the turbine building than a break in
one of the 24" main steam lines in the steam tunnel.
(d) The release through the break consists of 11,621.5 lb of
steam during the initial steam-phase flow, and 93,675.4 lb
of steam and water during the two-phase flow.
Total steam released through the break = 18,179 lbs
Total liquid released through the break = 87,118 lbs
(e) The ensuing high fuel temperatures do not lead to any fuel
damage.
(f) The noble gas fission product concentrations in the steam
correspond to the design values which would yield the
standard release rate to the atmosphere during normal
operation (i.e., 100,000 jiCi/sec following a 30-min decay).
50t of all noble gases leaving the reactor vessel during the
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE I OF l lPROJECT: JAF PRELM [ ] PREPARED BY A(2- DATE
FINAL EX] CHECKED BY o>- DATE /1/9 7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
10.5-sec MSIV closure time are released via the break.
From Ref. 5 (Sec. 5.1), the total noble gas releases
following an MSLB are as follows:
Nuclide MSLB Release(Ci)
Kr 83m 1.82E-02Kr 85m 3.27E-02Kr 85 1.07E-04Kr 87 1.07E-01Kr 88 1.07E-01Kr 89 6.96E-01
Xe 131m 8.03E-05Xe 133m 1.55E-03Xe 133 4.39E-02Xe 135m 1.39E-01Xe 135 1.18E-01Xe 137 8.03E-01Xe 138 4.77E-01
The halogen inventory in the steam was determined to be
insignificant in comparison to that in the discharged
liquid, and was not considered.
(g) The halogen source term in the discharged liquid was
selected to correspond to the proposed technical
specification limit for the maximum permissible reactor
coolant activity, namely 0.2 .LCi/gm I-131 DE2. This is the
GE Standard Technical Specification limit (Ref. 22). The
specified RCS concentration is assumed to accommodate the
pre-accident iodine spike which would occur as a result of
reactor shutdown or depressurization of the primary system.
Also, it is conservatively assumed that the total two-phase
2 I-131 DE (Dose Equivalent) is that concentration of I-131which alone would produce the same committed thyroid dose as allthe iodines in a given mixture.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE i>. -OFPROJECT: JAF PRELM [ I PREPARED BY RJZD DATE
FINAL EX] CHECKED BY z26- DATE U/X3J97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
flow release through the break (93,675.4 lbs of liquid and
staam) would contain iodines at the conc -trations equal to
those for the liquid phase. Under these conditions, the
total halogen activities discharged into the turbine
building would be as follows (Ref. 5, Sec. 5.1):
Nuclide MSLB Release(Ci)
Br 83 3.199Br 84 5.247Br 85 2.687
I-131 3.455I-132 26.87I-133 23.03I-134 48.63I-135 31.99
Activation products and other particulates in the coolant
were neglected since they would not become airborne.
(h) 100 k of the coolant halogens discharged in the turbine
building are assumed to become airborne and released to the
atmosphere at ground level over a period of 2 hours. The
selected release rate was equivalent to 72 air changes per
day, and the cumulative releases to the atmosphere (ignoring
buildup and decay) as a function of time would be as
follows:
Post MSLB Time Cumulative(min) Release (%)
0 0.05 22.1
10 39.315 52.820 63.230 77.745 89.560 95.090 98.9
120 99.8
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 53 OF
PROJECT: JAF PRELM [ ] PREPARED BY A42. DATE ilLL3IIIFINAL EX] CHECKED BY /26- DATE
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
(i) The atmospheric dispersion factors associated with the
transport of radioactivity at cw7ound level to the outdoor
receptors are as follows (from Ref. 2):
TimeReceptor Interval Dispersion Parameter (sec/m3)Location (hrs) Conc. X/Q Gamma X/Q
SB 0 - 2 1.79E-4 1.32E-4
LPZ 0 - 8 2.OOE-5 1.61E.-58 - 24 1.34E-5 1.06E-5
24 - 96 5.59E-6 4.27E-696 - 720 1.60E-6 1.16E-6
Onsite 0 - 8 3.29E-3 4.06E-48 - 24 2.81E-3 3.48E-4
24 - 96 2.OOE-3 2.49E-496 - 720 1.22E-3 1.54E-4
Note the following:
1. The concentration (X/Q)s are for computing the
inhalation exposures and the beta component of the skin
dose.
2. The gamma (X/Q)s are for computing the whole body doses
due to exposure to finite radioactive clouds above.
3. The dispersion parameters listed above for onsite
receptors are for the CR air intake (from Ref. 2).
They were conservatively assumed to apply to onsite
outdoor receptors at grade elevation, in the general
vicinity of the old administration building.
(j) The breathing rates at the various receptor locations are as
described under Item (i) of Sec. 4.1.1.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE S OF 5t
PROJECT: JAF PRELM E I PREPARED BY K DATE / IFINAL EX] CHECKED BY ,2 6- DATE SI/i7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
5.2 Results
Ra-iation exosures at the onsite and offsite outdoor
receptors of interest following a design-basis MSLB were
calculated using the DORITA-2 computer code and the data and
assumptions listed above. Copies of the DORITA-2 outputs appear
in Attachment B to this calculation (Computer Run Cases #1, #2
and #3).
Table 5.1 which follows presents the time-dependent thyroid,
whole body and skin doses at the receptors of interest. Refer to
Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to
Table 2.3 for the time-dependent dose rates at onsite outdoor
receptor locations in the general vicinity of the old
administration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE S• -OF C lPROJECT: JAF PRELM [ I PREPARED BY (R_ DATE JJ/ft9
FINAL [X] CHECKED BY MGS DATE i//n i&7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 5.1
JAF Power Uprate Project - Time-Dependent Doses at OutdoorReceptors Following a Main Steam Line Break Accident
Time(hours)
ThyroidDose (rem)
0. OOOE+005.573E-01
Whole BodyDose (rem)
0. OOOE+007.056E-03
SkinDose (rem)
0. 000E+001.113E-02
SB:0. 000E+002. 000E+00
LPZ0. OOOE+002. 000E+008 . 000E+002.400E+01
O . OOOE+006. 226E-026.241E-026.241E-02
0. OOOE+008.606E-048.616E-048. 616E-04
O . OOOE+001.328E-031.330E-031.330E-03
OnsiteO . OOOE+005. 000E-011. OOOE+002. 000E+00
O. OOOE+008. 016E+009. 773E+001. 024E+01
O . OOOE+001.786E-022.102E-022 .170E-02
O. OOOE+006.377E-027.500E-027. 746E-02
4. OOOE+008.OOOE+001.200E+011.800E+01
2 .400E+01
1. 027E+011. 027E+011. 027E+011. 027E+01
1. 027E+01
2.173E-022. 173E-022.173E-022.173E-02
2.173E-02
7.756E- 027.756E-027.756E-027.756E-02
7.756E-02
Note: The results for the "onsite" receptor in this tablewere conservatively based on the atmosphericdispersion factors applicable to the control-roomoutside air intake located on the roof of the oldadministration building.
NYPA - CALC.# JAE-CALC-RAD-00048 REV 1 PAGE ___ OF
PROJECT: JAF PRELM [ ] PREPARED BY 4 DATE Iift/flFINAL [X] CHECKED BY ^4r& DATE tI//9/
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 5.1
JAF Powe, Uprate Project - Time-:)ependent Doses at Out-door
Receptors Following a Main Steam Line Break Accident
Time(hours)
SB:0. OOE+002. OOOE+00
LPZ0. OOOE+002. 0O0E+008.OO0E+002.400E+01
ThyroidDose (rem)
0. OOE+005.573E-01
0. 0OOE+006.226E-026.241E-026.241E-02
Whole BodyDose (rem)
o.OO0E+007.056E-03
o.OOOE+008.606E-048.616E-048.616E-04
o.00OE+001.786E-022.102E-022.170E-02
SkinDose (rem)
0. OOOE+001.113E-02
0. 0OOE+001.328E-031.330E-031.330E-03
o.O0OE+006.377E-027.500E-027.746E-02
Onsite0. OOOE+005.000E-011. 0OOE+002. OOOE+00
0. OOOE+008.016E+009.773E+001. Q24E+01
4. O0E+008. 0OOE+001.200E+011. 800E+01
2.400E+01
1. 027E+011. 027E+011. 027E+011. 027E+01
1. 027E+01
2.173E-022.173E-022.173E-022.173E-02
2.173E-02
7.756E-027.756E-027.756E-027.756E-02
7.756E-02
Note: The results for the "onsite" receptor in this tablewere conservatively based on the atmosphericdispersion factors applicable to the control-roomoutside air intake located on the roof of the oldadministration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 5 -OF 19PROJECT: JAF PRELM 1 ] PREPARED BY g DATE
FINAL EX] CHECKED BY A2- DATE t/n/97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
6. RADIATION EXPOSURES FROM A CONTROL ROD DROP ACCIDENT
6.1 Basic Data and Assumptions
All assumptions and data employed in the analysis of a CRDA
are consistent with the guidance in the Standard Review Plan
(Ref. 10, Sec. 15.4.9), applicable portions of Regulatory Guide
1.77 (Ref. 18), the updated UFSAR, and JAF-CALC-RAD-00041 (Ref.
4). They are as follows:
(a) The reactor has been operating at full power until 30
minutes before the CRDA. As described in the JAF UFSAR,
Rev. 0, 7/82, Sec. 14.6.1.2.4, this assumption means that
the reactor was shut down from design power, taken critical,
and brought to the initial temperature and pressure
conditions within 30 minutes of the departure from design
power.
(b) The reactor power was at the level for design-basis accident
analyses (i.e., 2586.5 MWt, from Sec. 4.1.1). The core
inventory for the radionuclides of interest at the end of a
1000-day continuous operation is as shown under Item (b) in
Sec. 4.1.1 of this calculation.
(c) A CRDA takes place and leads to the failure of 850 fuel rods
(Ref. 20, Sec. 6.2.1, and Ref. 28, Sec. 3.7). The total
number of fuel rods in the core is equal to 36472 (Ref. 1)3.
(Note: According to the UFSAR, Rev. 0, 7/82, Sec.
14.6.1.2.4, the total number of fuel rods that fail
following a CRDA is 330.)
3 UFSAR Table 3.2-1 lists the total number of fuel rods forCycle 11 as 35784. This number will be revised to 38708 forCycle 12. The number of fuel rods employed in the currentcalculation (364-,2) was selected for consistency with theoriginal calculation (Ref. 1). As the fuel designs continue tochange over the next few fuel cycles to a 10x10 bundle, thequantity of interest is the ratio of the number of failed fuelrods to the total number of rods in the core. The fuel failureratio for the CRDA employed in this calculation (850/36472) isnot expected to significantly change with the new fuel designs.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE ''9 OF Ads
PROJECT: JAF PRELM [ ] PREPARED BY (Wf DATE IIFINAL EX] CHECKED BY +,g, DATE
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
(d) The failed fuel rods were at a core location with a radial
peaking factor of 1.5 (Ref. 10, SRP Sec. 15.4.9).
(e) All activity within the gaps of the failed fuel rods is
released to the reactor coolant and is instantaneously and
uniformly mixed with the coolant in the pressure vessel at
the time of the accident. The released activity is
conservatively assumed to correspond to 10t of all halogens
and 10t of all noble gases (30% for Kr 85) in each failed
rod (Ref. 18, as recommended in the SRP).
(f) Based on the above information, and without taking credit
for the pre-accident decay time of 30 minutes referred to
under Item (a), the noble gas and halogen inventories which
are released to the coolant are as shown below. They were
computed by applying the following multiplying factors to
the core inventory data given in Sec. 4.1.1 of this
calculation:
Multiplying factor for all noble gases except Kr 85:
1.5 (peaking factor) x (850 failed rods / 36472 rods)
x 0.1 (gap fraction) = 3.496E-03
Multiplying factor for Kr 85:1.5 (peaking factor) x (850/36472) x 0.3 (gap fraction)
= 1.049E-02
Multiplying factor for all halogens:1.5 (peaking factor) x (850/36472) x 0.1 (gap fraction)x 0.1 (fraction reaching turbine/condensers)= 3.496E-04
NYPA - CALC.# JAF-CPLC-RAD-00048 REV 1 PAGE q OF AtPROJECT: JAF PRELM E ] PREPARED BY g DATE ,,/?/9
FINAL [X] CHECKED BY -S6- DATE ll//97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Post-CRDA Falogens and Noble Gases Reaching the Condensers
Activity ActivityNuclide (JC Nuclide (Ci)
Br 83 2.824E+03* Kr 83m 2.836E+04Br 84 5.007E+03 Kr 85m 6.089E+04Br 85 6.001E+03 Kr 85 8.179E+03
Kr 87 1.168E+05I 129 7.881E-04 Kr 88 1.655E+05I 130 9.458E+02 Kr 89 2.058E+05I 131 2.379E+04I 132 3.477E+04 Xe 131m 1.430E+03I 133 4.975E+04 Xe 133m 2.084E+04I 134 5.476E+04 Xe 133 4.998E+05I 135 4.697E+04 Xe 135m 9.422E+04I 136 2.265E+04 Xe 135 6.456E+04
Xe 137 4.387E+05Xe 138 4.168E+05
* 8.078E+06 (from Table 4.1) x 3.496E-04
(g) As a result of elimination of the MSIV-closure and reactor-
shutdown functions of the main steam line radiation monitors
(modification No. F1-93-086) the pathway of post-CRDA
atmospheric releases at JAF has changed. Under the new CRDA
scenario, the MSIVs stay open and the release is to the
offgas system. [See Ref. 4 for the radiological analysis of
the revised CRDA scenario under pre-uprate conditions.]
(h) As a result of plant shutdown following a CRDA, or as a
result of offgas system automatic isolation (following a 15-
minute delay, which is not considered in the analysis) due
to high radiation fields at the offgas monitors, the
released radioactivity is retained within the turbine,
condensers and the ofrgas system. Release to the environs
is due to leakage from the various contaminated systems into
the turbine building. [Note: Without offgas system
isolation, releases would be via the charcoal holdup system
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 60 OF
PROJECT: JAP PRELM C ] PREPARED BY Iht_ DATE
FINAL EX] CHECKED BY AMd- DATE ,/,3/97
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
and the stack and would be significantly less restrictive
than the scenario analyzed. See J.AF-CAL( RAD-00041, Ref. 4,
for a comparison under pre-uprate conditions.]
(i) Plateout and partitioning of the halogens in the turbine,
condensers and other internal surfaces is conservatively
assumed to be equal to 90t [Ref. 10 (SRP Sec. 15.4.9), Ref.
20 (Sec. 6.3.1.1), and Ref. 9].
[Note: The 90% halogen depletion due to plateout and
partitioning was numerically accounted for in the DORITA-2
runs by assuming filtration of the release.]
(j) The leakage rate amounts to 1% per day and lasts for 24
hours (Reg. Guide 1.77, Ref. 18). The release to the
atmosphere is at ground level and there is no holdup within
the turbine building.
(k) Transfer of the released radioactivity to the outdoor
receptors is governed by the atmospheric dispersion factors
listed under Item (i) in Sec. 5.1 of this calculation. The
breathing rates at the various receptor locations are as
described under Item (i) of Sec. 4.1.1.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 61 OFPROJECT: JAP PRELM [ ] PREPARED BY A DATE
FINAL [X] CHECKED BY n L-- DATE 11/13/97TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
6.2 Results
Radiation exposures at the onsite and offsite outdoo-
receptors of interest following a design-basis CRDA were
calculated using the DORITA-2 computer code and the data and
assumptions listed above. Copies of the DORITA-2 outputs appear
in Attachment B to this calculation (Computer Run Cases #1, #2
and #3).
Table 6.1 which follows presents the time-dependent thyroid,
whole body and skin doses at the receptors of interest. Refer to
Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to
Table 2.3 for the time-dependent dose rates at onsite outdoor
receptor locations in the general vicinity of the old
administration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 62- -OF 1tPROJECT: JAF PRELM t ] PREPARED BY DATE 9
FINAL [XI CHECKED BY 116&- DATE J//a/g 7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 6.'
JAF -ower Upra':e Project -Receptors Following a
Time-Dependent Dose- at OutdoorControl Rod Drop Accident
Time(hours)
ThyroidDose (rem)
0. OOOE+001.855E-01
Whole BodyDose (rem)
0.OOOE+001.312E-02
SkinDose (rem)
o.00OE+002.514E-02
SB:0. OOOE+002 . OOOE+00
LPZ0 . OOOE+002. OOOE+008. OOOE+002.400E+01
0. OOOE+002.073E-027. 929E-021.258E-01
o.OOOE+001.600E-033.386E-034.524E-03
0. OOOE+002.966E-036.102E-038.417E-03
Onsite0. 000E+005. 000E-011. 000E+002. OOOE+00
0. OOOE+008. 635E-011. 719E+003.410E+00
0. 000E+001. 667E-022. 621E-024. 035E-02
0. OOOE+001. 079E-011. 574E-012.287E-01
4. OOOE+008. OOOE+001.200E+011. 800E+01
2 .400E+01
6. 712E+001.304E+011. 819E+012. 549E+01
3.237E+01
6.052E-028.539E-029.879E-021.127E-01
1.228E-01
3.266E-014.514E-015.276E-016.178E-01
6.908E-01
Note: The results for the "onsite" receptor in this tablewere conservatively based on the atmosphericdispersion factors applicable to the control-roomoutside air intake located on the roof of the oldadministration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 93 OF AdPROJECT: JAF PRELM [ ] PREPARED BY KjXL DATE
FINAL EX] CHECKED BY M2 DATE }JL/6J37TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
7. RADIATION EXPOSURES FROM A REFUELING ACCIDENT
7.1 Basic Data and Assumptions
The assumptions and data listed below were used in the analysis
of a design-basis refueling accident. All assumptions are
consistent with the guidance in the Standard Review Plan (Ref.
10, Sec. 15.7.4), Regulatory Guide 1.25 (Ref. 16), and the UFSAR.
(a) The reactor has been operating at full power (2586.5 MWt)
for an extended period of time (1000 days).
(b) The core inventory for the radionuclides of interest at the
end of such an operating period is as shown under Item (b)
in Sec. 4.1.1 of this calculation.
(c) The reactor is shutdown, refueling operations are initiated
and a refueling accident takes place at 24 hours after
shutdown (Ref. 10, SRP Sec. 15.7.4).
(d) The accident involves a fuel assembly dropping from the
maximum height allowed by the fuel handling equipment. A
total of 125 fuel rods are ruptured. This is a conservative
number based on information in Ref. 28, Sec. 3.8; also,
according to the UFSAR, Rev. 0, 7/82, Sec. 14.6.1.4.2, the
total number of fuel rods that fail during a refueling
accident is 111. The total number of fuel rods in the core
is equal to 36472 (from Sec. 6.1).
(e) The failed fuel rods were at a core location with a radial
peaking factor of 1.5 (Reg. Guide 1.25, Ref. 16).
(f) All activity within the gaps of the failed fuel rods is
released to the fuel pool water. The released activity is
conservatively assumed to correspond to 10's of all halogens
(except I 129) nLd 10% of all noble gases (except Kr 85) in
each failed rod, and to 30t of I 129 and Kr 85 (Ref. 16).
(g) The noble gas and halogen inventories released to the fuel
pool (prior to adjustment for decay from the time of reactor
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE t_ -OF e38PROJECT: JAF PRELM [ I PREPARED BY g DATE Zl//311
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Offsite Outdoor Receptors Following Design-Basis Accidents
shutdown, which is handled by the DORITA-2 computer code)
are as shown in the table which follows. They were computed
by multiplying the core inventory in Sec. 4.1.1 of this
calculation by the following factors:
Multiplying factor for all noble gases except Kr 85:
1.5 (peaking factor) x (125 failed rods / 36472 rods)
x 0.1 (gap fraction) = 5.141E-04
Multiplying factor for Kr 85:
1.5 (peaking factor) x (125/36472) x 0.3 (gap fraction)
= 1.542E-03
Multiplying factor for all halogens except I 129:
1.5 (peaking factor) x (125/36472) x 0.1 (gap fraction)
= 5.141E-04
Multiplying factor for I 129:
1.5 (peaking factor) x (125/36472) x 0.3 (gap fraction)
= 1.542E-03
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 6A OF $ts
PROJECT: JAF PRELM [ I PREPARED BY A? DATEFINAL [XI CHECKED BY _ DATE 11//3/97
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Pre-Decay Refueling Accident Source Term
Nuclide Activ.(Ci) Nuclide Activ.(Ci)
Br 83 4.153E+03* Kr 83m 4.171E+03Br 84 7.363E+03 Kr 85m 8.954E+03Br 85 8.825E+03 Kr 85 1.203E+03
Kr 87 1.718E+04I 129 3.477E-03 Kr 88 2.433E+04
I 130 1.391E+03 Kr 89 3.026E+04I 131 3.498E+04I 132 5.113E+04 Xe 131m 2.104E+02
I 133 7.316E+04 Xe 133m 3.065E+03I 134 8.053E+04 Xe 133 7.351E+04I 135 6.908E+04 Xe 135m 1.386E+04
I 136 3.331E+04 Xe 135 9.494E+03Xe 137 6.452E+04Xe 138 6.130E+04
* 8.078E+06 (from Sec. 4.1.1) x 5.141E-04
(h) The halogen composition (inorganic, organic and particulate
species) and the pool halogen retention factors are such
that 99% of all released halogens are assumed to be retained
by the water in the fuel pool (Ref. 16). This is equivalent
to an overall decontamination factor (DF) of 100. The
halogen composition of the remaining (airborne) halogens is
equal to 75% inorganic and 25t organic (Ref. 16).
(i) The retention of noble gases by the pool water is negligible
(i.e., noble gas DF = 1).
(j) Radioactive material that escapes the pool is released to
the atmosphere via the SGTS and main stack over a 2 hour
period (Ref. 16). The Reactor Building air exchange rate
was arbitrarily set at the conservative value of 3 air
changes per hour. At this release rate, which is the same
as that used for releases from the turbine building
following an MSLB, 99.8 % of all radioactivity would be
released to the SGTS within the assumed 2 hour period.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 46 OF -t
PROJECT: JAF PRELM [ ] PREPARED BY M?_ DATEFINAL [X] CHECKED BY o2&- DATE I//3,I7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
[Refer to Sec. 5.1, Item (h) for tabulation of the
cumulative release as a function of time.] The actual RB
air exchange rate (at the nominal SGTS flow of 6000 scfm) is
only 3.3 per day.
(k) The halogen-removal filter efficiency of the SGTS is 90% for
all halogen species (Ref. 23).
(1) All releases to the atmosphere are via the main stack.
Transport of the released radioactivity to the outdoor
receptors of interest is controlled by the dispersion
factors listed under Item (h) in Sec. 4.1.1. The presence
of a 4-hour fumigation at the time of the accident is
accounted for.
(m) Other exposure-related parameters are as described in Items
(i) and (j) of Sec. 4.1.1.
7.2 Results
Radiation exposures at the onsite and offsite outdoor receptors
of interest following a design-basis Refueling Accident were
calculated using the DORITA-2 computer code and the data and
assumptions listed above. Copies of the DORITA-2 outputs appear
in Attachment B to this calculation (Computer Run Cases #1, #2
and #3).
Table 7.1 which follows presents the time-dependent thyroid,
whole body and skin doses at the receptors of interest. Refer to
Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to
Table 2.3 for the time-dependent dose rates at onsite outdoor
receptor locations in th_ general vicinity of the old
administration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 4 ) -OF 9 'PROJECT: JAF PRELM t ] PREPARED BY Pe- DATE 1//I3/,d
FINAL [X] CHECKED BY I26,- DATE g///)/g7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Table 7.1
JAF Power Uprate Project - Time-Dependent Doses at OutdoorReceptors Following a Refueling Accident
Time(hours)
ThyroidDose (rem)
0. OOOE+007.376E-01
Whole BodyDose (rem)
o.OOOE+009.131E-02
SkinDose (rem)
0. OOOE+002.108E-01
SB:0. 000E+002 . OOOE+00
LPZ0. 000E+002 . 000E+004. 000E+008. 000E+00
2.400E+01
0. OOE+002.872E-012.879E-012. 879E-01
2. 879E-01
o.OOOE+003.652E-024.074E-024.200E-02
4.277E-02
o.OOOE+008.329E-029.003E-029.179E-02
9.290E-02
Onsite0. 000E+005.OOOE-011. 000E+002. 000E+00
0. 000E+001.016E-021.242E-021.303E-02
0. OOOE+004.448E-035.576E-036.228E-03
0. OOOE+006.825E-038.520E-039.428E-03
4. OOOE+008. OOOE+001. 200E+011.800E+01
2.400E+01
1.307E-021. 307E-021.307E-021. 307E-02
1.307E-02
6. 947E-037.989E-038.510E-038.982E-03
9.238E-03
1. 037E-021.172E-021.240E-021.301E-02
1.335E-02
Note: The results for the "onsite" receptor in this tablewere conservatively based on the atmosphericdispersion factors applicable to the control-roomoutside air intake located on the roof of the oldadministration building.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE _e OF
PROJECT: JAY PRELM 1 I PREPARED BY g DATEFINAL [X] CHECKED BY A?&V- DATE .I l13497
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
8. RADIATION EXPOSURES FROM OTHER POST-LOCA SOURCES
This section addresses the direct shine radiation fields at
various onsite outdoor receptors due to post-LOCA airborne
radioactivity accumulating within the reactor building. Section
8.1 looks at the shine dose rates from activity in the refueling
level, and Sec. 8.2 looks at the dose rates adjacent to the east
side of the reactor building exterior wall.
Airborne radioactivity within the reactor building can result
from either post-LOCA drywell leakage, post-LOCA ESF component
leakage, or from a Refueling Accident. Since the latter two are
relatively insignificant with respect to post-LOCA drywell
leakage (See Table 2.3, for instance), they were not considered
in this part of the calculation.
The radiation dose rates and cumulative doses documented in this
section are to air (rad/hr and rad). They may be conservatively
applied to the whole body (rem/hr and rem).
8.1 Direct Shine from Post-LOCA Airborne Radioactivity in the
RB Refueling Level
8.1.1 Basic Data and Assumptions
Of interest here are the definition of the gamma spectra
associated with the post-LOCA airborne radioactivity within the
refueling level, and the source/receptor geometry. These are
addressed below.
Source Term
The basic data and assumptions for a LOCA are as described in
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 6I OF
PROJECT: JAF PRELM [ I PREPARED BY Re- DATEFINAL [X] CHECKED BY V&6- DATE _IL/3/9 1
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Sec. 4.1 of this calculation. The items of interest in t.e
definition of the airborne radioactivity within the reactor
building and the associated time-dependent gamma spectra are as
follows:
(a) A LOCA takes place at full power (2586.5 MWt).
(b) The core inventory for the radionuclides of interest is
as shown under Item (b) in Sec. 4.1.1 of this
calculation.
(c) 100% of the core-inventory noble gases and 25% of the
halogens become instantly airborne within the drywell
atmosphere and are available for leakage to the
secondary containment.
(d) The halogen composition airborne within the drywell is
as follows: 91t elemental, 4k organic and 5%
particulate.
(e) Leakage from the drywell is at the rate of 1.5% per
day.
(f) As a result of the ventilation system, airborne
radioactivity leaking from the drywell becomes
uniformly distributed within the air volume of the
reactor building (2.6E+06 ft3, Ref. 30).
(g) Release from the reactor building is through the SGTS
and the main stack at the rate of 6000 scfm (or 3.3 air
changes per day).
Source/Receptor Geometry
The source/receptor geometry for use with QAD-CGGP is shown in
Fig. 8.1. The primary component is the reactor building. The
refueling level was represented by a box with no roof and no side
walls; the box dimensions are 125' (W) x 162' (L) x 60' (H) and
the total volume is 1.215E+06 ft3, or 3.439E+04 M3.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 3O -OF !tf
PROJECT: JAF PRELM [ I PREPARED BY A12 DATE i1l3/1IFINAL [X] CHECKED BY Age6- DATE l117
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
The RB volume beneath the refueling level was conservatively
represented as a hox with 2-ft concrete walls all around. The
concrete density was set at 2.35 g/cc and had the following
composition (in weight percent, from Ref. 29, Vol. II, Table
9.1.12-77):
Fe: 1.19 H : 0.85 0 : 50.64Mg: 0.23 Ca: 8.03 Na: 1.66Si: 30.49 Al: 4.44 S : 0.12K : 1.87
The receptor locations were selected to be identical to those
analyzed in Ref. 1, for consistency. They are shown in Fig. 8.2.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE %i OF 5t1
PROJECT: JAF PRELM [ ] PREPARED BY 1 DATE fflf3/1f ?FINAL [X] CHECKED BY /A6- DATE f 1// 7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Fig. 8.1 - QAD-CGGP Source/Receptor Geometri
(Direct Shine from the RB Refueling Floor)
ya xis
l
SOURCE REGION
l
1xis
y=162'
-- x axis
z=429.5'
z=369.5'
-2-ft concrete box(all around)
z=272'-* x axis
SOURCE REGION
!' ' * ' . - '. . ' ...- .. . , . ... _ _, ._ I . -..
I
; & ' _ * , _ .,, .,, _ , t;,_..,,,_-. r- * * * 'a' *'�'' '''t' � {XII
X11
Ln
NYPA - CALC.# JIF-CALC-RAD-00048 REV 1 PAGE :4_ OF jts
PROJECT: JAF PRELM [ I PREPARED BY M DATE /1t39>
FINAL [XI CHECKED BY M 6_ DATE 41yj7315TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Fig. 8.2 - Receptor Locations
(Direct Shine from the RB Refueling Floor)
No. x (cm) y (cm) No. x (cm) y (cm)
1 15003 05 --A^ J
7 -150s)0
9 150011 1000013 1500015 10000
800080002300
10000
-15000-15000-40008000
2 15004 -15006 -150008 -20000
10 -2500012 1500014 2000016 15000
1000060002300
15000
10000-15000-40008000
_ rNz = 8400 cm (or El. 272') at all receptors
(x=0, y=0 at the SE corner of the RB)
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE I3 OF 3PROJECT: JAF PRELM [ ] PREPARED BY /St DATE Jl3(98
FINAL EX] CHECKED BY Al6- DATE l/13/9 7TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Fig. 8.2 - Receptor Locations
(Direct Shine from the RB Refueling Floor)
NO.
1357
9111315
Note:
x (cm)
15000
-10000-15000
1500100001500010000
y (cm)
800080002300
10000
-15000-15000
-40008000
No.
2468
10121416
x (cm)
1500-1500
-15000-20000
-25000150002000015000
y (cm)
1000060002300
15000
10000-15000
-40008000
z = 8400 cm (or El. 272') at all receptors(x=0, y=0 at the SE corner of the RB)
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE -OF -g
PROJECT: JAF PRELM [ 1 PREPARED BY Aid DATE
FINAL [X] CHECKED BY 1M 6- DATE II//?/S 7
TITLE: Power Uprate Project - Radiological Impact at Onsite andOffsite Outdoor Receptors Following Design-Basis Accidents
8.1.2 Results
Post-LOCA direct-shine radiation levels at the various onsite
outdoor locations of interest due to radioactivity accumulating
on the RB refueling level are shown in Table 2.5. These were
extracted from QAD-CGGP and MATILDA Run Case #1 in Attachment B.
The gamma spectra were extracted from DORITA-2 Run Case #4.
Refer to Sec. 2.3 for general remarks, and to Fig. 2.2 for
extrapolation of the worst-case dose rates (at 4 hrs after the
postulated LOCA) to other receptors around the reactor building.
Cumulative doses, if of interest, can be found in the MATILDA
output.
Comparison of the dose rates presented in this subsection with
corresponding results in Ref. 1 (JAF-CALC-RAD-00008) shows that
the former are lower by a factor of approximately 3 to 4. The
reason for this is the conservative RB air exchange rate assumed
in Ref. 1 for this part of the analysis, namely 1 air exchange
per day in lieu of the more appropriate value of 3.3. Note also
that, for the same reason, the dose rates in Ref. 1 peak at a
later time, namely at about 12 hours after the accident instead
of 4.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE -j3 OF 3ftPROJECT: JAP PRELM [ ] PREPARED BY < DATE 1//f-J9
FINAL [X] CHECKED BY &f 6 DATE 1J111Z91
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
8.2 Direct Shine from Post-LOCA Airborne Radioactivity at
El. 272' 'f the RB
8.2.1 Basic Data and Assumptions
Direct shine dose rates from the RB were also calculated at
outdoor receptors adjacent to the E side of the reactor building.
The source term in this case was post-LOCA airborne radioactivity
within El. 272' of the reactor building and is identical to that
described in Sec. 8.1.1 for the refueling level.
The RE area of interest and the source/receptor geometry for use
with QAD-CGGP are shown in Figs. 8.3 and 8.4. Note that the RB
wall thickness in this area is 21" (from Fig. 8.3). The main
components are the source region (with dimensions of 154 ft
length, 50 ft depth, and 27.2 ft high), and the exterior wall
(21" thick). The source volume is equal to 2.09E+05 ft3, or
5.93E+03 M 3. The receptor distances range from contact with the
exterior wall to 21 ft, at 3-ft increments.
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 1t OF %+9PROJECT: JAF PRELM [ I PREPARED BY R DATE ,72~19
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TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
>1-'-
0.1-)
0
0
0)
Fig. 8.3 - RB El. 272' - Plant Drawing 11825-Fm-1E)
(Source area and rece'ptors of tterest)
4'i(2.gja
w.r
4 r04-J0 4J
O a)(C)
tOU)X
"-4
Hi
E-4
d-
I. I
iI
i '
!I_
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE ' OF
PROJECT: JAF PRELM [ I PREPARED BY f DATE
FINAL [X] CHECKED BY CO eG- DATE J/,/97
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Fig. 8.4 - QAD-CGGP Source/Receptor Geometry
(Direct Shine tbzough the RB E Wall)
y axis
SOURCE REGION(27.2' high)
I x axis
8 receptors(contact to 21')(4' above grade)
NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE t -OF ;f-
PROJECT: JAP PRELM 1 1 PREPARED BY ^ DATE &L/9
FINAL EX] CHECKED BY ; DATE _JL4A7
TITLE: Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
8.2.2 Results
Post-LOCA direct-shine radiation levels at the var.i.ous onsite
outdoor locations of interest due to RB shine through the 21' E
wall are shown in Table 2.6. These were extracted from QAD-CGGP
and MATILDA Run Case #2 in Attachment B. The gamma spectra were
extracted from DORITA-2 Run Case #4. Refer to Sec. 2.3 for
general remarks, and to the MATILDA output for cumulative doses,
if of interest.
NYPA Calculation No. JAF-CALC-RAD-00048, Rev. 1
Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Attachment A
Excerpts from References Pertinent to this Calculation
Printed: 10/30/97 ACTS CHANGE FORMI ACTS #:Prined: 0/3097 ATS CANGEFOI 23847
I 9 New Item
- I Change DUE DATE from 11/01/1997
[ Change DEPARTMENT from PRE to
LI Change INDIVIDUAL from GR to
E CLOSURE ----- >> Date Completed:
to (* indic hard date)
>_> Dept Mgr Apvl
_[ ACTS to Startup
____ Startup to ACTS
ACTS Type Code: RNRI Due Date: 11/01/1997 Priority: B
Department Code: PRE Resp Indiv: GARY RE ACTS/Startup: A
Source Document: JLIC-96-220 Search Code:
Descriptn: REVISE CALCULATIONS WHICH USE SBGT EFFIENCY FROM 99% TO 95%.
1.) For CHANGES, provide reason and current status of item.2.) For CLOSURE, describe action taken and attach reference documents.
Status:'EN
RNRI Closure:
[wgiorla ] Initiator: Date:
Department Manager: Date:
Director / Gen Mgr: Date:
QA Manager (RQA Only): Date:
Sr. Sponsor (RBP Only): Date:
Plant Mgr (RNYS & RBP): Date:
PORC/SRC Chmn (RPOR/RSRC): Mtg# Date:
AP-03.08 Rev 8 ACTION AND COMMITMENT TRACKING SYSTEM* ATTACHMENT 1
NYPA Calculation No. JAF-CALC-RAD-00048, Rev. 1
Power Uprate Project - Radiological Impact at Onsite and
Offsite Outdoor Receptors Following Design-Basis Accidents
Attachment B
COPIES OF COMPUTER OUTPUTS
Included in this attachment are copies of the computeroutputs pertinent to this calculation. They appear in thefollowing order:
DORITA-2
Case #1 Radiation fields at the Site boundary for thefollowing design-basis accidents:
(a) Loss of coolant accident (LOCA)(drywell leakage),
(b) Loss of coolant accident (LOCA)(ESF Component leakage),
(c) Main Steam Line Break accident(MSLB),
(d) Control Rod Drop Accident (CRDA),and
(e) Refueling accident (RA)
Case #2 Similar to Case #1, for the Low Population Zone
Case #3 Similar to Case #1, for a receptor located at theCR outside air intake (on top of the oldadministration building) [Note: This locationwas conservatively assumed to apply to outdoorreceptors at ground elevation, in the generalvicinity of the old administration building.]
Case #4 Gamma spectra associated with post-LOCA airborneradioactivity within the Reactor Building (for usewith QAD-CGGP and MATILDA to compute the directshine radiation levels from the refueling leveland from El. 272' of the RB).
Note: DORITA-2 Case #4 employs a SGTS filter efficiency of99% for release from the RB to the outsideatmosphere. Although the efficiency was loweredfrom 99% to 90% in evaluating the radiologicalconsequences, this run was not revised since thechange in filter efficiency does not affect thepost-LOCA Reactor Building spectrum.
QAD-CGGP & MATILDA
Case #1 Direct shine radiation levels from post-LOCAradioactivity accumulating in the RB RefuelingLevel, based on the gamma spectra defined underDORITA-2 Run Case #4 (16 receptors all around theRB)
Case #2 Direct shine radiation levels from post-LOCAradioactivity in El. 272' of the RB based on thegamma spectra defined under DORITA-2 Run Case #1)(8 receptors along the source centerline, at about4 ft above grade).
Note: The title line in the QAD-CGGP runs erroneouslyidentify the plant as IP3. This does not impactthe results.
NYPA/CRE CALCULATION No. JAF-CALC-RAD-00048, Rev. 1
THIS IS THE LAST PAGE OF THIS CALCULATION