COMPUTER CALCULATION CONTROL SHEET Page 1 of 78 · 10.5-sec MSIV closure time are released via the...

COMPUTER CALCULATION CONTROL SHEET 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O VOIDEDo SUPERSEDED BY: /S upersedes Revision 0 of this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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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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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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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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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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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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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


7. RADIATION EXPOSURES FROM A REFUELING ACCIDENT ... ..... 63


8. RADIATION EXPOSURES FROM OTHER POST-LOCA SOURCES ..... 68

8.1 Direct Shine from Post-LOCA AirborneRadioactivity in the RB Refueling Level ... ...... 68


8.2 Direct Shine from Post-LOCA AirborneRadioactivity at El. 272' of the RB .... ......... 75


ATTACHMENTS

A. Excerpts from References 'ertinent to this Calculation

B. Copies of Computer Outputs

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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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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

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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.

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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.

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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)

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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.

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(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.

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(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.

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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.

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Table 2.2

JAF Power Uprate Project - Doses &t the Low Population ZoneFollowing Postulated Design-Basis Accidents


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%


75. 01.258E-01

6.04.524E-03 8.417E-03

% of Reg. Limit

RA

0.17% 0.08%


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

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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

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Time(hours)




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

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Time(hours)




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.

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Table 2.4

JAF Power Uprate Project - Integrated Doses (ContinuousOccupancy) at Outdoor Receptors in the General Vicinity



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.

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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.

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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.

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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.

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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

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Fig. 2.1 Receptor locations for the dose rates in Table 2.5from airborne radioactivity accumulatir. in the RBrefueling level.

- 0 0 0 8

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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)


PROJECT: JAF PRELM E I PREPARED BY /W6 DATE 1113FINAL [X] CHECKED BY A DATE I11/ 9 7


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

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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.

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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.

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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

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(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

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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).

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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.

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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

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4.2 ESF Component Leakage


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

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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


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).


whole body and skin doses at the receptors of interest due to

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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.

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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


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

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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.

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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)


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

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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

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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

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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


PROJECT: JAF PRELM [ ] PREPARED BY A42. DATE ilLL3IIIFINAL EX] CHECKED BY /26- DATE



(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.

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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



and #3).


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.

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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


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.

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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


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


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6. RADIATION EXPOSURES FROM A CONTROL ROD DROP ACCIDENT


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.

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(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

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Post-CRDA Falogens and Noble Gases Reaching the Condensers

Activity ActivityNuclide (JC Nuclide (Ci)


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


PROJECT: JAP PRELM C ] PREPARED BY Iht_ DATE

FINAL EX] CHECKED BY AMd- DATE ,/,3/97



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.

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6.2 Results

Radiation exposures at the onsite and offsite outdoo-

receptors of interest following a design-basis CRDA were




and #3).





receptor locations in the general vicinity of the old


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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


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


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FINAL EX] CHECKED BY M2 DATE }JL/6J37TITLE: Power Uprate Project - Radiological Impact at Onsite and


7. RADIATION EXPOSURES FROM A REFUELING ACCIDENT


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

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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:


= 5.141E-04

Multiplying factor for I 129:


= 1.542E-03

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Pre-Decay Refueling Accident Source Term

Nuclide Activ.(Ci) Nuclide Activ.(Ci)


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.

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[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


of interest following a design-basis Refueling Accident were




and #3).





receptor locations in th_ general vicinity of the old


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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


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


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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


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

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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.

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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.

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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

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Fig. 8.2 - Receptor Locations


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)

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Fig. 8.2 - Receptor Locations


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

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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.

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8.2 Direct Shine from Post-LOCA Airborne Radioactivity at

El. 272' 'f the RB


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.

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>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_

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PROJECT: JAF PRELM [ I PREPARED BY f DATE

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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)

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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


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


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

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