ML20129D639

Rev 2 to Calculation Package for Application of Revised DBA Source Term to Browns Ferry Nuclear Power Plant

ML20129D639
Person / Time
Site:	Browns Ferry
Issue date:	09/25/1996
From:	Metcalf J POLESTAR APPLIED TECHNOLOGY, INC.
To:
Shared Package
ML19353D888	List: ML20129D562 ML20129D639 ML20129D737 ML20129D760 ML20129D786 ML20129D810 ML20129D819 ML20129D844 ML20129D874 ML20129D883 ML20129D891 ML20129D898
References
PSAT-040004.04-R02, PSAT-04000U.04, PSAT-40004.04-R2, PSAT-4000U.4, NUDOCS 9610250094
Download: ML20129D639 (11)

Text

. .

. Polostar NON-PROPRIETARY PSAT 04000U.04 Page 1 of 11 Revision: 2 1

CALCULATION PACKAGE FOR APPLICATION OF THE REVISED DBA SOURCE TERM TO THE BROWNS FERRY NUCLEAR POWER PLANT prepared by:

POLESTAR APPLIED TECHNOLOGY,INC.

for the ELECTRIC POWERRESEARCHINSTITUTE Revision Reason Project Manager Reviewer Number for Revision (print / sign /date) (print /sian/date) 0 InitialIssue am Dave Leaver etcalf 4 $()9 f m 8 5 L 7/c5/fr 1 Revision 1 of Attachment 3 es calf Dave Leaver bE W zirin 14 2 Modified discussion of two leak paths: es calf Dave Leaver MSIVleak MF L c[zthc Main condenserleak 42 #

Clarified main steamline/ condenser depo 'tionjn list ofTVADOSE features Revised " Key Assumption 3" to reflect de ' tion in main condenser Changed Conservatism 9 Removed Conservatism 11 Revised title and mmmary of Calc 04002H.08 to include main condenser Revision 2 ofAttachment 2 Revision 2 of At+=^hment 3 Revision 1 of Attachment 5 Revision 1 of Attaclunent 6 Revision 1 of Attachment 7 Revision 1 of Attachment 10 Polestar NON-PROPRIETARY 9610250094 961018 PDR ADOCK 05000259 P PDR

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! PSAT 04000U.04 Page 2 of 11 3

Revision: 2 l

Table of Contents

} Sections Eggs 4

Purpose 3 Methodology 3 Key Assumptions /Conservatisms 5 References 9 Summary ofCalculations 9 Results 11 Conclusions .

11 Attachments Revision Title and Date 1 0 ABB-CE Calculation 1066-S&T95-C-002, " Calculation of

- Containment Leakage Doses for Browns Ferry", September 29, ,

1995 which includes a letter dated September 20,1995 from James Metcalf(Project Manager, PSAT) to Raymond Schneider (ABB-CE Technical Contact) modifying the original contract workscope 2 2 PSAT Project Data Base 04000U.03, June 25,1996 3 2 PSAT Calculation 04011R01, " Volumetric Flowrate as a Function of Tirne from Drywell to Torus (and Return)", June 13,1996 4 0 PSAT Calculation 04001R02, " Aerosol Decay Rates (Lambda) in Drywell", September 1,1995 5 1 PSAT Calculation 04011R03, " Maximum Iodine Decontamination Factors", June 13,1996 6 1 PSAT Calculation 04011R04, " Suppression Pool Scrubbing Ef5ciency (Including Pool Bypass)", June 13,1996 7 1 PSAT Calculation 04011R05, " Additional Radionuclide Data",

June 13,1996

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Revision: 2 l l

8 0 PSAT Calculation 04011R06, " Source Term for Use on Browns Ferry Application of NUREG-1465", September 28,1995 9 0 PSAT Calculation 04011R07, "Drywell Leakage Rate Direct to Environment Mimicking Case 2 Early Bypass of SGTS", September 28,1995 10 1 PSAT Calculation 04002R08, " Aerosol Remova! Efficiency in Main Steam Lines and Main Condenser", May 31,1996 11 0 PSAT Calculation 04002R09, " Elemental Iodine Filter Efficiency in Main Steam Lines", September 28,1995 12 N/A Fax dated September 1,1995 from Don McCamy (TVA Technical Contact) to James Metcalf(PSAT Project Manager) providing a reference for Item 3.28 of the Project Data Base 13 N/A Notes of Telecon dated September 13,1995 between James Metcalf(PSAT Project Manager) and Don McCamy (TVA Technical Contact) providing concurrence for time-shift of fbmigation X/Qs (Item 5.1 ofProject Data Base)

Purpose The purpose of this calculation package is to compile a set of Safety-Related calculations which, together, constitute the application of the revised DBA source term from Reference 1 to the Browns Ferry Nuclear Power Station for the purpose of demon h.Gg compliance with 10CFR100 and with 10CFR50, Appendix A, GDC-19. Two cases are calculated: Case I which includes three trains of SGTS operating and Case 2 which includes two trains of SGTS operating.

Methodology The approach makes use of two computer models and manual calculations documented in A"=eha=ats 1 through 11. A*=nh-t 2 is not a calculation, itself, but rather a plant-specific data base which has been used to control and coordinate the preparation of the calculations in three locations (the two Polestar Applied Technology offices in Los Altos, CA and Portsmouth, NH and the ABB-Combustion Engineering office, a subcontractor to Polestar, in Wm' dsor, CT) as well as to facilitate TVA-Browns Ferry review ofplant-specific input.

The two computer models include STARNAUA (Reference 2), an aerosol physics code

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i PSAT 04000U.04 Page 4 of 11 i

Revision: 2 proprietary to Polestar, and TVADOSE (Reference 3), a slightly modified version of the

proprietary LDOSE dose calcult. tion code belonging to ABB-CE.

The overall approach has been to treat the containment as a two-contiof volume model (drywell and toms airspace) up until the time that the core debris from the recowrd core damage accident (see Key Assumption 1 and Attachment 3) has been quenched. Beyoil that time, a high mixing

! rate is used between the two control volumes to effectively create a s!ngle-control volume l containment similar to current practice. Prior to the debris quench, vent flow and suppression 1

pool bypass from the drywell to the torus is considered. The source term (the activity release

from the reactor to the drywell) is based on Reference 1 (see Attachment 8).

i

Release paths that are included are as follows (see Attachment 1):

Enth Comment Drywellleak Equal to two percent of the drywell volume per day - leak is to the reactor building Torus airspace leak Equal to two percent of the torus airspace volume per day - leak is to the reactor building MSIVleak Equal to 250 scfh from the drywell (total for four stammth) - leak ,

is to the main condenser via the drain line pathway and to the high pressure (HP) turbine via the stop valve leakage pathway (only the former pathway is included in the dose model - see " Leak from main condenser" below)

CAD operation Equal to 8340 cfh from the torus airspace intermittently over 30 days (containment is well-mixed by the time of CAD operation so a drywell source would be equivalent) - all of the flow is through the 4 SGTS filters,99.97% of the flow is to the stack,0.03% to the stack l

room (i.e., stack bypass)  :

Torus vent leak Equal to 10 cfh from the torus airspace - leak is to the stack SGTS operation Equal to 1.32E6 cfh for Case 1 and 0.9E6 cfh for Case 2 - flow is from the reactor building to the stack (except for 300 cfh which goes to the stack room as bypass) through the SGTS filters SGTS bypass Equal to 3.lE-3 cfh directly from drywell for Case 2 only (leakrate is two percent per day with reduction factor of approximately 40000 to account for reactor building hold-up)-leak is to the

PSAT 04000U.04 Page 5 of 11 Revision:* 2 i

j environment i

]E Stack room leak Equal to 300 cfh (assumed to be same as inleakage, see "SGTS

operation", above) - leak is to the environment

' Main condenser leak Equal to 250 cfh (assumed to be same as leakage out of drywell as scfh, see "MSIV leak", above) - leak is to the environment (leak is actually to turbine building but turbine building hold-up is neglected

- leak also includes 0.5% which goes to HP turbine, but since HP turbine hold-up is equivalent to main condenser hold-up, only the main condenser pathway is modeled for dose assessment)

The TVADOSE code integrates the release through the various pathways; and, using Reference 4-based dose conversion factors, calculates doses at the EAB m (' tegrated over the first two hours of the accident) and at the LPZ and in the control room (integrated over 30 days). TVADOSE also includes the following:

o Deposition in the drywell and, after core debris quench, in the torus airspace e Suppression pool scrubbing e Filtration in the main steamlines representing main steamline/ condenser deposition 4 o Active filtration by the SGTS and CREVS, but without credit for charcoal adsorbers These effects are included in TVADOSE via filtration ef5ciencies and removal " lambdas" provided by Polestar. These matters are discussed further under Summary of Calculations.

Key Assumptions /Conservatisms Key Assumption 1: The core damage which leads to the DBA source term of Reference l is arrested by the restoration of core cooling at about two hours after the start of the accident. Diann== inn This is an extension to operating plants of a position presented in Reference 5 for advanced LWRs and is discussed fully in Attachman* 3.

Key Assumption 2: The source terms of Reference I can be applied to Browns Ferry without regard for fuel burn-up limitations. Discussion This issue derives from a caveat in Section 3.6 ofReference 1 and is being pursued separately by NEI with NRC. Since the Reference I source term is specified in terms of fractions of core inventory and since core inventories are calculated for this

1 i 4 j PSAT 04000U.04 Page 6 of 11  !

Revision: 2 application to Browns Ferry using an appropriate burn-up, the caveat is not j

related to core inventory. As noted in Reference 1, the focus of the caveat j

is the gap activity release; and because of the nature of this application to Browns Ferry, the results would not be greatly sensitive to the exact gap l release timing or magnitude in any case.

Key Assumption 3: The MSIV leakage release is entirely through the drainline and main  ;

condenser; there is no release included via the high pressure (HP) turbine.

Diann== ion: From Reference 6 it is observed that the flow split between the l drainline/ main condenser flowpath and the HP turbine flowpath is about  !

200:1. Since deposition in the steamlines applies to either flowpath, the  !

only potential difference between the two flowpaths is differential hold-up l and retention in the main condenser vs. the HP turbine. HP turbine retention is assumed to be zero in the calculation of the main steamline/ main condenser overall aerosol and elemental radioiodine removal efBciencies (see Ansehments 10 and 11); and, therefore, the only other mechanism that could create a difference in the calculated relative dose between the two pathways is unequal delay due to hold-up. The hold-up effect is estimated to be of the order of a factor of two to three in dose reduction based on the main condenser (non-negligible); but since Reference 6 gives a hold-up volume about a factor of 200 less for the HP turbine (which is the same as the flow split), nimilar hold-up would be s expected for the HP turbine. Therefore, the HP turbine flowpath can be modeled as a simple bypass of the main condenser in the calculation of the overall main steamline/ main condenser " filter" efEciency (which is what is done in Attachment 10), and the HP turbine volume, itself, can be neglected in the dose model; i.e., all of the flow can be ===nmad to go through the main condenser (which is what is done in Anechmaat 1).

Key Assumption 4: A pH value of at least 6.0 will be maintained in the containment water (m' particular, the suppression pool) for at least 30 days (3.7 half-lives ofI-131) after the start of the accident. Discussion: A non-Safety Related scoping study was conducted as part of this project to aupport this assumption. This non-Safety Related study is being followed-up with a Safety-Related calculation being performed under a separate contract.

Key Assumption 5: The results of these analysis are sufEciently conservative to constitute a basis for demonstrating compliance with the requirements of 10CFR100 and with 10CWO, Appendix A, GDC-19. Discussion: The source terms ofReference 1 are comparable in conservatism to the DBA source tenns previously used on Browns Ferry as based on 10CFR100 (and Reference 7) and subsequent regulatory guidance. The noble gas and iodine release

PSAT 04000U.04 Page 7 of 11 Revision: 2 i

fractions (which are the main determinants of the whole body and thyroid dose evaluations specified in 10CFR100) are about the same. The Reference 1 timing and chemical form, while different from the previous source terms, are nonetheless conservative compared to what is expected I under actual accident conditions (e.g., the 1979 accident at Three Mile l

Island) and provide a more physically correct representation of activity  !

release to the containment. Moreover, in terms of activity transport within and through the containment system and release to the environment, there are many other conservatisms included in the calculations of At+=chments 3 through 11. These are as follows:

1 e Conservatism 1: Only gamma energy is considered in cale"1=+iag the core power f used to determine vent flow from the drywell to the torus during l' core degradation and the associated debris quench. Core debris sensible heat during the core degradation (and the formation of a debris bed that would enhance heat transfer to the overlaying water), metal-water reactions, and beta heating are neglected. See Attachment 3.

i e Conservatism 2: A conservatively small sedimentation area has been specified for the drywell. Moreover, the -limaa+=+ ion removallambdas calculated for the drywell are applied to the torus airspace after the debris ,

quench. (No sedimentation is credited in the torus airspace prior to  !

the debris quench). Due to the effects ofpool bypass (see Conservatism 5) the mass airborne in the torus airspace at the end  ;

of the quench is greater than that airborne in the drywell, and the  !

torus airspace has a smaller volume and a greater medimentatinu  !

area. Therefore, to apply drywell lambdas in the torus after debris quench is a significant conservatism. See At+=chment 4.

e Conservatism 3: Hygroscopicity is neglected in the determination of sedimentation lambdas. See Attachment 4.

o Conservatism 4: The maximum iodine DF is based on the maximum pool temperature and does not consider the long-term iodate reaction which will tend to suppress iodine re-evolution. Both of these  ;

effects will tend to limit the long-term potential for iodine re-  !

evolution even if the containment pH falls below the value of 6.0 assumed in the analysis.' See Attachment 5.

e Conservatism 5: The pool bypass flow area used to assess the impact of pool bypass on pool scrubbing efficiency is ten times the value used as the basis j i

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for the surveillance test acceptance limit. Moreover, a review of surveillance test data indicates that the actual measured value is, on  !

['

, average, substantially below the test acceptance value. See j Attachment 6.

• Conservatism 6: Te-132 is treated as elemental I-132 except for half-life which i i

corresponds to Te-132. This is to account for the fact that Te-132 j {

(which may have been removed as particulate and subsequently held J l up on filters and/or in main steam piping) may re-evolve as

elemental I-132. By treating the Te-132 as elemental I-132 from  !

the beginning (with the Te-132 half-life), the same amount ofI-132  !

l activity is released as would be the case in a mechanistic model of-

! tlie process described, but the release occurs much more rapidly.

i This means that more adverse X/Q values, breathing rates, and j control room occupancy factors are used in calanta+ia: the thyroid dose contribution ofTe-132 than would be the case with a l

mechanistic model. See Attachment 7.

!

• Conservatism 7: In general, the PWR gap release is expected to occur much more j rapidly than the BWR gap release (refer to discussion in Reference )

1). However, this application has used a gap release start time of i 30 seconds (appropriate for a PWR) to represent the Browns Ferry ,

BWR. See At+=rhmaa* 8. -

1 j e Conservatism 8: The impact of non-noble gas an'd non-radioiodine components of

the release on the 10CFR100 and GDC-19 dose calculations has l been assessed in two ways

(1) the important isotopes of

, radiocesium and Te-132 have been included explicitly, and (2) the j "other" radionuclides have been approximated using a one percent j release to the containment atmosphere as described in Reference 7 j (with the exception that these radionuclides are subsequently j treated as aerosol and released to the environment accordingly). By l doing so, the impact of the "Other" has been overstated by about a l factor of ten as compared to rigorous application ofReference 1.

See At+=chmants 7 and 8.

j e Conservatism 9: Aerosol retention in the stammline portions inboard of the outboard MSIVs has been neglected. See Attachment 10.

l i e Conservatism 10: The conversion of deposited elemental iodine to re-evolved organic

! iodine is assumed to be instantaneous as opposed to requiring

several days. See Attachment t1.

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N PSAT 04000U.04 Page 9 of 11

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References

' Reference 1: Soffer, L., et al., " Accident Source Terms for Light-Water Nuclear Power Plants",

NUREG-1465, February 1995 h

l Reference 2: "STARNAUA, A Code for Evaluating Severe Accident Aerosol Behaviorin l

Nuclear Power Plant Containments: Code Description and Validation and Verification Report", PSAT C101.02, Revision 0, May 1995 -

L Reference 3: ABB-Combustion Engineering Calculation "TVADOSE: A Computer Program for the Calculation ofBrowns Ferry Advanced Source Term (ABB-Combustion Engineering Proprietary)",1066-S&T95-C-001, September 1995 Reference 4: International Commission on Radiological Protection, " Limits for Intake of Radionuclides by Workers", ICRP Publication 30,1979 Reference 5: Taylor, J., " Proposed Issuance ofFinal NUREG-1465, ' Accident Source Terms for Light-Water Nuclear Power Plants'", SECY-94-300, December 15,1994 Reference 6: General Electric Nuclear Energy, " Browns Ferry Nuclear Plant, Calculation of LOCA Doses to the Control Room from MSIV T >=k=ge", DRF A00-04146, Section C, Attached to Letter ALJ92049 from A.L. Jenkins (GE Nuclear Energy) to J.L. Kamphouse (TVA) dated August 28,1992  !

l Reference 7: DINunno, J. J., et al., " Calculation of Distance Factors for Power and Test Reactor

{

Sites", TID-14844, March 1%2 I Summary of Calculations Polestar calculations provide the following:

e PSAT 04011H.01 (Attachment 3): ,

i This calculation establishes the overall thermal-hydraulic behavior of the containment and calculates the exchange rate between the drywell and the torus airspace (Items 3.10 and 3.11 of Attachment 2).

o PSAT 04001H.02 (Attachment 4):

This calculation establishes the removal rate of aerosol and elemental iodine from the I containment atmosphere (Items 4.3 and 4.4 of Attachment 2) and the volumetric leakrate I

4 PSAT 04000U.04 Page 10 of 11 Revision: 2 i

for MSIV leakage Gtem 3.23 of Attachment 2).

• PSAT 04011H.03 (Attachment 5):

This calculation establishes the maximum iodine decontamination factor (ratio of total j

iodine in containment to that airborne)(Items 4.5 and 4.6 of Attachment 2). 1 L

i e PSAT 0401IH.04 (Attachment 6):

This calculation establishes the scrubbing efficiency of the suppression pool (Item 4.2 of

} Attachment 2).

l

• PSAT 0401IH.05 (Attachment 7):

l l This calculation provides radionuclide data not available in the TACTS Users Manual 1

(NUREG/CR-5106) for Item 1 of Attachment 2.

i j e PSAT 04011H.06 (Attachment 8):

l This calculrfm defines the source term for Browns Ferry based on Reference 1 Otems 2.1 l

through 2.3 of Attachment 2).

j e PSAT 0401IH.07 (Attachment 9):

j This calculation provides the leakrate directly from the dywell to the environment for the early part of the Case 2 dose assessment (when the reactor building internal pressure is not l sub-atmospheric)(Item 3.20 of Attachment 2).

j e PSAT 04002H.08 (Anechment 10):

! This calculation provides the removal efficiency for aerosol in the main steamline/ main i condenser (Item 4.7 of A stachment 2).  !

i

)

• PSAT 04002H.09 (Attachment 11):

4 l . This calculation provides the removal efficiency for elemental iodine and tellurium in the l

! main steamline (Item 4.7 of AMechment 2).

i

\ In addition to the above Polestar calculations, the ABB-CE dose calculation 1066-S&T95-C-002 i provides the EAB, LPZ, and control room thyroid, whole body, and (for the control room) skin i

doses. This calculationis Attachment 1.

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PSAT 04000U.04 Page 11 of 11 Revision: 2 Results The results are presented in the following table. Case 1 (three trains of SGTS operating) produces the highest doses and is, therefore, the limiting case.

Dose Magnitude - rem Location and Duration Dose Tyne Case 1 Cug,,2 Control Room - 30 day Thyroid 17.9 17.4 Whole Body 0.046 0.045 Skin 1.79 1.78 4

EAB - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Thyroid 3.16 2.74 Whole Body 0.075 0.059 LPZ - 30 day Thyroid 5.79 5.55 Whole Body 0.282 0.269 i Limiting case contributions are as follows: I e I-131 to control room thyroid dose = 16.64 rem I-133 to control room thyroid dose = 1.13 rem .

Te-132 to control room thyroid dose = 0.06 rem Other contributors = 0.07 rem o Of the 16.64 rem I-131 thyroid dose in control room:

12.51 rem organic 3.21 rem elemental 0.92 rem particulate o Non-noble gas, radioiodine, radiocesium, Te-132 (i.e., the "Other") whole body dose is 0.7% of the 2-hour EAB total and 0.12% of the 30-day LPZ total and is, therefore, confirmed to be negligible.

Conclusions These doses, which have been conservatively calculated, are well within the limits of 10CFR100 and 10CFR50, Appendix A, GDC-19.

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PSAT 04000U.04 ' i t

Attachment 1 ABB-CE Calculation 1066-S&T95-C-002 - I 1

4

" Calculation of Containment Leakage Doses for Browns Ferry" '

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{Ib Ybff '

i Document

Title:

CALCULATION OF CONTAINMENT LEAKAGE DOSES FOR BROWNS FERRY l

1 e-4

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1066-S&T95-C-002 0 I

i Calculation Number Rev' COMBUSTION ENGINEERING, INC. l 2 of 46 i SYSTEMS & TRANSIENTS l Page Number 1 i

i 1

RECORD OF REVISIONS j i

i ~= j

.' rc N .- w ead on ww> e age N6.4 u*#Aisthors 5H de5eriewer ud I 9# INN e?A$proverse- e!eDated h 0 Initial Docuanent n/a R. E. Schneider M. Michonski R. E. [kuith 9/29/95 l

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ABB Combustion Engineering Nuclear Operations i J

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1066-S&*IW-C-002 0 COMBUSTION ENGINEERING, INC.

3 of 46 SYSTEMS & TRANSIENTS Page Number CHECKLIST M2 REVIEW OF DESIGN Af4ALYSIS

, 1.

Is the material presented sufficiently detailed as to purpose, method, assumptions, design input, references, and units?

[ dyes O N/A 2.

Were the inputs correctly selected and incorporated into the analysis? E[Yes O N/A

^

3.

Have the assumptions necessary to perform the analysis been adequately documented and justified?

E[Yes O N/A 4.

Are applicable codes, standards, and regulatory requirements, including issue and addenda, employed in the analysis properly identified, and were their requirements met?

EIYes O N/A

5. Have interface requirements been satisfied?

4 EIYes O N/A

6. Have the adjustment factors, uncertainties, and empirical correlations used in the analysis been correctly applied?

i E[Yes O N/A

7. Was an appropriate analysis or calculation method used?

EIY2 O N/A 8.

Have the versions of the computer codes employed in the analysis been certified for application? If not, has sefScient information been provided to enable

, verification of the program and results?

6Yes O N/A

9. Is the purpose sufficiently clear, and are the results and conclusions reasonable when compared to inputs?

EIYes O N/A

10. Has an appropriate title page similar to Exhibit 3.4-1 been used?

'Yes O N/A 11.

Are all pages sequentially numbered and marked with the analysis number? dyes O N/A 12.

Where necessary, are the assumptions identified for subsequent reverifications when the detailed design activities are completed? EfYes O N/A

13. Is the presentation legible and reproducible?

TYes O N/A 14.

Have all cross-outs or overstrikes in the documentation been initialed and dated by the Author?

E[Yes O N/A Y2 '

T 9[

Reviewer Signature ~ Date ABB Combustion Engineering Nuclear Operations

1 l

Al COMBUSTION INGINEERING, INC.

1066-S&'I95-C-002 0 Calculation Number Rev, l

4 of 46 SYSTEMS & TRANSIENTS Page Number REVIEWER'S COMMENTS 1

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No significant errors were discovered. The results of the cases are reasonable considering the inputs which were used.

4 4

4 1

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1 ABB Combustion Engineering Nuclear Operations

ABB COMEUSTION ENGINEERING, INC.

1066-S&T95-C-002 0 5 of 46 SYSTEMS & TRANSIENTS Pa;c Number TABLE OF CONTENTS SECTION DESCRIPTION PAGE NO.

RECORD OF REVISIONS . . . . . . . . . . . . . . . . . . . . . 2 CHECKLIST NO. 2 ......................... 3 REVIEWER'S COMMENTS . . . . . . . . . . . . . . . . . . . . 4 TABLE OF CONTENTS ... .................. 5 I.

INTRODUCTION / PURPOSE . . . . . . . . . . . . . . . . . . . 6 II.

DISCUSSION and METHOD OF ANALYSIS . . . . . . . . . 7 III.

ASSUMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 IV. CALCULATION ..........................29 V. CODE USED / UPDATES ....................41 VI. COMPUTER RUN

SUMMARY

. . . . . . . . . . . . . . . . . 42 VII. RESULTS AND CONCLUSIONS . . . . . . . . . . . . . . . . 43 VIII. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 IX.

APPENDICES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 A. CASE 1 INPUT / OUTPUT . . . . . . . . . . . . . . Al B. CASE 2 INPUT / OUTPUT . . . . . . . . . . . . . . El C. BROWNS FERRY DESIGN DATA BASE ... C1 D. WORKSCOPE (REFERENCES 4 AND 6) ...........D1 ABB Combustion Engineering Nuclear Operations

1 1

1066-S&T95-C-002 0 Calcuhtion Number - Rev.'

COMBUSTION ENGINEERING, INC.

6 of 46 SYSTEMS & TRANSIENTS Page Number I. INTRODUCTION / PURPOSE The purpose of this calculation is to provide containment leakage dose assessments for the TVA Browns Ferry Nuclear Unit for a design basis maximum hypothetical accident using source term input based on the revised source term as defined in NUREG-1465 (Reference 7). This information will be combined with other calculations to be performed by Polestar Applied Technology, Inc.(PSAT) to establish the total radiological dose following the 10CFR100 maximum hypothetical accident.

4 Specifically, this report provides two calculations for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 30 day doses in the control room and at the exclusion area boundary (EAB) and low population zone (LPZ): l l

Case 1: 'Ihree SGTS Fans no SGTS/CREVS charcoal filters SGTS flow of 22,000 CFM l 4

Case 2: Two SGTS Fans no SGTS/CREVS charcoal filters SGTS flow of 15,000 CFM -

Main steam line leakage is assumed to be initially 120 CFH and increase after 7230 seconds to 4 177.5 CFH and remain at that value for the remainder of the 30 day calculation. l Calculations will be performed using TVADOSE (See Reference 2) (version TVD 92395). The equations governing TVADOSE are presented herein. The validation of TVADOSE for application to Browns Ferry is contained in Reference 2.

Results of this calculation show that the leakage contribution to the 30 day control room dose is under 18 Rem. Case 1 was predicted to be the more limiting case due to the higher outflow from the Reactor Building. Additional details, including the EAB and LPZ thyroid, and whole body are presented in Section VII.

3 ABB Combustion Engineering Nuclear Operations

AIy 1066-S&T95-C-002 0 Calculation Number Rev.

COMBUSTION ENGINEERING, INC.

7 of 46 i SYSTEMS & TRANSIENTS Page Number H. DISCUSSION and METHOD OF ANALYSIS H.1 Introduction l

The methodology used for this calculation is defined below. The equations definb in this section are  !

implemented for the Browns Ferry Nuclear Unit. The resultant program (TVADOSE) has evolved from an Combustion Engineering's "in house" LDOSE computer code (Reference 1 ). The modified code version is fully described and specifically qualified for use in the Browns Ferry calculation in Reference

2. Modifications made to LDOSE to create TVADOSE are based on the workscope outlined in 1 Reference The model and equations implemented in TVADOSE are based on standard engineering methodology for the calculation cf activity transport; doses calculations are based on a Dose Conversion Factors (DCFs) Methodology with DCFs provided to ABB via Polestar Applied Technology , Inc. in Reference

3. All relevant co;eations are presented in this document.

H.2 Overview of Model H.2.1 Calculation of Area Activity The computer model to be used for TVADOSE consists of 7 nodes, with eight identified regions (See Figure II-1). These regions are:

1. Atmosphere

2. Drywell

3. Wetwell

4. Control Room

5. Reactor Building

6. Stack Base '

7. Main Steam Line Piping

8. Condenser The model follows the guidelines of Reference 2. The regions are defined as follows:

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8 of 46 SYSTEMS & TRANSIENTS Page Nusr.bcr FIGURE II.1. NODAL ARRANGEMENT FOR TVADOSE I

PROPRIETARY 1

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FIGURE II.2. SOURCES OF ACTIVITY TO THE CONTROL ROOM 2 1

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Page Number FIGURE II.3 SOURCES OF ACTIVITY TO THE EXCLUSION AREA BOUNDARY 1

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SYSTEMS & TRANSIENTS Page Number MGURE II.4 SOURCES OF ACTIVITY TO THE LOW POPULATION ZONE 1

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12 of 46 SYSTEMS & TRANSIEN"rS Page Number 4

Region 1: Atmosphere Region 1 is not ud in the activity calculations. Their impact on dose is evaluab based on the methodology defined in section II.2.2.

Region 2: Drywell i

Region 2 is the drywell. All source releases from the RCS fuel are directed into region 2.

Radionuclide removal in this region is allowed via natural and active removal mechanisms. This is accomplished by providing a radionuclide removal time constant. This time constant is user specified.

1 Region 3: Wetwell Region 3 is the BWR wetwell (torus). All flow from the drywell that passes into the wetwell must '

pass through vent pipes. When the vacuum breakers on the vent pipes are closed, the gas mixture driven from the drywell into the wetwell will exit at a submerged elevation within the suppression pool. Fission products traversing this pass will be scrubbed (decontaminated) prior to entering the wetwell air space.

When the vent pipe vacuum breakers are open, the gas space of the wetwell and drywell communicate directly, without further scrubbing.

Fission product removal in the wetwell air space is considered via a user input table of radionuclide removal time constant.

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13 of 46 SYSTEMS & TRANSIENTS Page Number Region 4: Control Room Region 4 is the Control Room. The control room receives air intake from the enviroc.rnent. Both filtered and unfiltered air enters the control room.

The activity entering the control room originates from:

reactor building and wetwell releases through the stack stack room releases drywell releases main condenser releases due to leakages through the main steam isolation valve Region 5: Reactor Building Region 5 is the reactor building. In this model the reactor building accepts containment leakage from the drywell and wetwell air space. ESF leakage is not modeled in the TVADOSE model.

This is consistent with the guidelines of Reference 4 '

Region 6: Stack Room Region 6 is room at the base of the stack. Leakage may enter this room via filtered leakage originating in the wetwell and the reactor building. Leakage from the stack room is not filtered.

Region 7 and 8: Main Steam Line and Main Condenser Node 7 and 8 are only weakly coupled to the remainder of the model. Flow leaving the Drywell (Region 2) through the MSIVs enters the main steam line volume (Region 7). The activity of this flow is decremented by the decontamination factor associated with pipe settling, plateout and natural deposition processes.

Region 8 is the Main Condenser. Fission products enter the main condenser at a low rate and are diluted by the large air volume of the condenser. Settling of the fission products within the condenser is modeled with a user input radionuclide removal time constant.

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26 of 46 SYSTEMS & TRANS!ENTS Page Number Introduction of the source into the containment drywell is based on the Revised Source Term and is provided as input into this calculation via Reference 3. This information is summarized below.

Source Term (S(I,i,k))

(fraction of initial inventory released to the drywell) (Ref. 3)

TIME (SEC) 0 TO 1830 1830-7230 'i !$[EI"i% $$$$2fddd$$

Noble Gases .05 0.95 $@Nd$ $E$$isMNkNb Iodm.e .05 0.25

en uwn;w.

..a:p 's msg %s e g 4;weg Sc m s e@timd e m a+

Cesium .05 0.20 - ' CM $ INN @ :@M!kh$k$$$

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27 of 46 SYSTEMS & TRANSIENTS Page Number IIL ASSUMPTIONS / INPUT A list of assumptions follows:

1. ESF leakage is not considered Per scope of project as defined by Polestar Technology , Inc. and TVA (Reference 4).

2. Source term composition and release characteristics are based on NUREG-1465 Revised Source Term. This information is provided in Reference 4.

This is consistent with the intent of the calculations. Releases into containment are as linear over the appropriate time interval.

All releases are delayed until 30 seconds.

I

3. Dose Conversion Factors baspd on specifications supplied in the Polestar data base.

(Reference 4) 4..

Tellurium is considered to behave t.s elemental iodine for purposes of scrubbing. Tellurium doses are based on I-132 DCFs (See workscope, Reference 4))

5.

Atmospheric dispersion from regions 3 and 6 are assumed equal. No impact on results since region 3 releases do not contribute to dose calculations 6.

Kr-90 contribution is neglected due to its short halflife (see References 3 and 6 ). This is accomplished by setting DCFs for Kr-90 equal to zero.

7. All filters neglect removal due to charcoal filters. Elemental and organic Iodine removal in t

HEPA filters assumed to have a zero efficiency (References 3 and 4)

8. No fission product removal due to settling or plateout is assumed in the Main condenser.

This is a conservative assumption in that increased airborne activities implies increased leakage.

9.

Fission product removal in the main steam line allowed for elemental iodine and particulates (See References 3 and 4) 10.

radionuclides are assumed to instantaneously mix with volume atmosphere. This assumption it consistent with standard review plan methodology and the cuncut worktenne (Reference 4)

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SYSTEMS & TRANSIENTS Page Number 11.

Fumigation time interval is selected at the worst one half hour period over the first two hours. This occurs in the 1.5 to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time frame. Including this effect later in the event acknowledges the impact of the later release of radionuclides. Doses will be maximized wit this assumption since the later interval has the higher atmospheric releases.

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29 Of 46 SYSTEMS & TRANSIENTS Page Number IV. CALCULATION IV.1 RadlOnuclide and Dose Data Data for RadiOnuclide Activity, decay constants and Dose Conversion Factors are.Obtained from Reference 3, Section 1. ~

ISOTOPE LDOSE AcnVITY D E C A Y WB DCF BETA DCF THYROID NAME LAMBDA DCF (skin)

CI DIS /SEC Rem m'/(Ci-S) Rem.m'/(Ci-S) 104/ Cia IS ISNAH(IS) AOREF(IS) YD(IS) DCFWB2(IS DCFSK2(IS DCFTH2(IS

) ) )

10 Kr-83m KRYPTO 1.127E+07 1.04E-04 1.27E-05 0 0 11 Kr-85m KRYPTO 2.351E+07 4.39E-05 2.30E-02 4.97E-02 0 12 Kr-85 KRYPTO 1.360E+06 2.04E-09 3.31E-04 4.84E-02 0 13 Kr-87 KRYPTO 4.481E+07 1.52E-04 1.33E-01 3.36E-01 0 14 Kr-88 KRYPTO 6.303E+07 6.89E-05 3.38E-01 7.76E-02 0 15 Kr-89 KRYPTO 7.653E+07 3.63E-03 3.03E-01 3.47E-01 0 16 Kr-90** KRYPTO 7.554E+07 .215E-1 0.0 0.0 0  %

17 Xe-131m XENON 1.050E+06 6.68E-07 1.25E-03 1.33E-02 0 18 Xe-133m XENON 5.960E+06 3.49E-06 4.29E-03 2.96E-02 0 19 xe-133 XENON 1.847E+08 1.52E-06 4.96E-03 9.67E-03 0 20 Xe-135m XENON 3.761E+07 7.40E-04 6.37E-02 2.14E-02 0 21 Xe-135 XENON 6.610E+07 2.09E-05 3.59E-02 6.32E-02 0 22 Xe-137 KENON 1.655E+08 2.96E-03 2.83E-02 4.59E-01 O d

23 Xe-138 XENON 1.552E+08 6.80E-04 1.87E-01 1.47E-01 0 1 1-131 IODINE 9.378E+07 9.96E-07 5.59E-02 3.07E-02 110 2 I-132 IODINE 1.355E+08 8.27E-05 3.55E-01 1.10E-01 0.63 3 I-133 IODINE 1.898E+0B 9.22E-06 9.11E-02 8.90E-02 18 4 I-134 IODINE 2.081E+08 2.23E-04 4.11E-01 1.42E-01 0.11 5 I-135 IODINE 1.778E+08 2.86E-05 2.49E-01 7.86E-02 3.1 6 Cs-134 CESIUM 2.508E+07 9.55E-09 2.58E-01 1.15E-01 0 7 Cs-137 CESIUM 1.503E+07 7.29E-10 9.30E-02 1.27E-01 8 Te-132 0 TELLUR 1.333E+08 2.51E-06 3.55E-01 1.10E-01 9 Other 0.63 ,

OTHER 4.967E+9 7.05E-5 .168 0 Oj

• INHALED

• • DCF set = 0.00 per ref. 6 The following factors are not used in the TVADOSE calculation: DCFTH1, DCFSK1, DCFWB1, EB and EO TVADOSE. these parameters appear in trie database but are not used in ABB Combustion Engineering Nuclear Operations

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l LOCATOR CONSTANTS:

I131 "IS" SUBCRIPT FOR ENTRY FOR I-131 1 NID - TOTAL NUMBER OF IODINE ISOTOPES: 5 KRN -TOTAL NUMBER OF KRYPTON ISOTOPES: 7 IST -TOTAL NUMBER OF ISOTOPES: 23 --

NOI-TOTAL NUMBER OF OTHER ISOTOPES: 1 NCS-TOTAL NUMBER OF CESIUM ISOTOPES 2 NXE-TOTLA NUMBER OF XENON ISOTOPES :7 1 l

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31 of 46 SYSTEMS & TRANSIENTS Page Number IV.2 Revised Source Term Release Profile Data obtained from Reference 3 items 2.1, 2.2, 2.3.

n) Fractional Releases (REVISED SOURCE TERM)

Fractional Releases into Containment in time Interval Time Noble Iodine Cesium Tellurium Other Interval Gases (sec)

O to 30 0 0 0 0 0 30 to 0.05 0.05 0.05 0 1830 1830-7230 0.95 0.25 0.20 0.05 0.01 7230 to 0.0 0.0 0.0 0.0 0.0

! End Note that the

+ organic Iodine contribution includes aerosols (CsI) + elemental iodine From Reference 5 item 2.2 and 2.3 total iodine between 30 to 1830 sec = .0024 + .000075 + .0475

=.049975 (round up to .05)

Ratio of CSI/ Total = .0475/.05= 0.95 Ratio of I2(elemental) / Total = 0.048 Ratio of Organic I/ Total = .000075/.05 =.0015 Note the ratios do not add to 1. Therefore, 4

• Fraction CsI = 0.95 Fraction I2(gas)=.0485 Fraction Organic =.0015 This composition applies to both early and late releases ABB Combdstion Engineering Nuclear Operations

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32 of 46 SYSTEMS & TRANSIENTS Page Number b) Iodine Composition From (a) above FRCTK:

Particulate Iodine: 0.95 Elemental Iodine  : 0.0485 Organic Iodine  : 0.0015 The code assumes the following:

All aerosols are particulate in nature and can be filtered via particulate filters s All noble gases cannot be filtered a Tellurium is an aerosol that is assumed to not be filtered via filter flovpaths (treated like elemental iodine)  ;

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SYSTEMS & TRANSIENTS Page Number IV.3 Browns Ferry Model System Description n) Volumes (Data taken from Reference 3) {

VOLUME )

NODE TVADOSE ITEM DESCRIPTION CUBIC FT COMMENT NO var. N O l Ref. 3 1

ENVIRONMENT 1.00E+08 ARBITRARY W(1) NOT l USED 2 W(2) 3.1 DRYWELL 159000 ITEM 3.1 3 W(3) 3.2 WETWELL 124000 ITEM 3.2 4 W(4) 3.6 CONTROL ROOM 210000 ITEM 3.6 5 W(5) 3.4 REACTOR BLDG 1.93E+06 ITEM 3.4 6 W(6) 3.5 STACK ROOM 34560 ITEM 3.5 7 W(7) 3.7 MS PIPE 692 ITEM 3.7 8 W(8) 3.8 MAIN CONDENSER 122400 ITEM 3.8 t

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1 34' of 46 SYSTEMS & TRANSIENTS Page Number IV.4 REGICN FLOARATES 4

TVADOSE ALLONS FIXED AND VARIABLE FLOARATES. 'IEE FLOARATES 'IHAT ARE FIXED ARE:

L25, L35 L31U, L81,L51, L61, L56, L78, L14U, L14F --

FLOARATES 'IRAT VARY W1'IH TIhE INCLUDE:

L21, L23, L31F, L32, L27 ALL Il3AKAGES ARE INPUT 'IO 'IBE OIE AS A P/JtAhETER WDH CG RE1YACDG L.

SIM%RY OF FIXED FLOARATES VARIABLE NAME DESCRIPTION VALUE (CFH) REF 3, ITIM OG14U CR UNFILTEKED. INFIDV 2.23E+5 3.22 OG14F CR FILreKED INFIDV 1.BE+5 3.21 GG25 IaV TO RB LEAKAGE 132.5 3.12 OG31U WV TO STAG 'IEROUGH 10 3.16 ,

HARDENED VENT OG61 STAG ROCM 'IO 300 3.19 ENVIRO MENT OG35 WEIVFII LEAKAGE 'IO RB 103.3 3.13 OG51 CASE 1 RB FIDV TO SGTS FILTER 1.32E+6 3.14 TO STAG OG51 CASE 2 RB FLOV T3 SGTS FILTER 0.9E+6 3,14 TO STAG OG56 FIDV FRCM SO13 FILTER 300 3.15 TO STAG ROCM 0G78 MAIN STEAM LINE 'IO 475 3.24 01 DENSER OG81 LEAKAGE FRCM MAIN 250 3.25 OCNDENSER -

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35 of 46 SYSTEMS & TRANSIENTS Page Number IV.4 (CCNT'D)

TBE VARYIN3 FIDNS:

FIDN OG21 :

Ref 3) FLOV FRCM DRYAELL TO ENVIRCN&NT: UNFILTERED 41 Tai 3.20 OF OG21 = 0 for all time for case 1.

TBE(s ec) FLCMRATE (CFH) 0.0 3.1E-3 (case 2) 105 _

0.0 3.E+6 0.0-i FLOV OG23 AND OG32 MIXIN3 AND TRANSPORT FIDNS WI'IHIN THE OCNTAINENT  !

(ITEG 3.10 AND 3.11) l TBE (SEC) OG23 :DN-W OG32 :W-DN CFH CFH 0.0 0.0 0.0 '

1830. 1.6E+5 0.0 7230 1.2E+6 0.0 7890 1.2E+6 1.2E+6 3.E+6 1.2E+6 1.2E+6 N3rE 30 DAYS = 2.592E+6 SECCNDS (CASE RUNS TO 30 DAYS)

OG27: LEAKAGE FRCM DRYAELL TO MAIN STEAM LINE (3.23)

TBE(SEC) OG27 (CFH) 1 0.0 120 l 7230 177.5 3.E+6 177.5 ABB Combustion Engineering Nuclear Operations

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36 of 46 SYSTEMS & TRANSIENTS Page Number OG31F:

FILTERED FLOE FROd VW THROUGH CAD (ITSO .17)

TIME (DAYS) OG31F (CFH) 0 0.0 10 8340 .

11 0.0

_20 8340.

21 0.0 29 8340 30 0.0 SET FLOV BEYCND 30 DAYS =0. (ICT USED IN CALC)

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' 38 of 46 SYSTEMS & TRANSIENTS Page Number B) REOVAL COEFFICIENTS:

Removal coefficients are provided for the drywell, wetwell gas space and the main condenser gas space. See i tems 4.3 and 4.4 of Reference

3. -- I TIhE (SEC) IAhEDA DN TIhE (SEC) LAMBDA VW (PER HOUR) REF. ITBd (PER HOUR) REF ITEM 0 0 4.3 0 44 0

30 .35 7890 .95 2400 .45 8570 .85 3200 .55 9840 .75 l 4000 .65 11760 .65 4885 .75 14530 .55 6300 .85 18650 .45 7360 .95 24980 .35 8570 .85 35570 .25 9840 .75 57220 .162 11760 .65 100000 0 14530 .55 3.E+6 0 18650 .45 1 24980 .35 i

35570 .25 57220 .162 10000 0.0

3. E + 6

_ 0.0 l

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39 of 46 l SYSTEMS & TRANSIENTS Page Number IV.6 Atmospheric Dispersion I

CHI /Q Values for Various Regions (SEC/M3) '

node 5 @MFV9&P Midd$ TACK 1RM FASE F A *

-m sM XQS.. ITad T I M E EAB XQ5EB LPZ 5.1 INr(XQ5T) G XQ5G XQ5LP (hrs) 1 0 TO 1. 5 9.70E-07 8.00E-07 5.91E-15 2 1.5 TO 2 2.40E-05 1.300E-05 3.31E- 5 3 2 TO 8 8.00E-07 3.80E-15 4 8 TO 24 4.00E-07 3.00E-15 5 24 TO 96 2.00E-07 1.90E-15 6 96 TO 720 6.50E-08 9.60E-16 node 6 MSgn&%s@NR%@ M >

1. MSTACKAROCM:RM FASE XQ6.. XQ6T(hrs) EAB- XQ6EB LPZ XQ6LP ITEM - G XQ6G 5.2 '

1 0 TO 2 1.22E-04 5.65E-05 8.89E-04 2 2 'ID 8 5.65E-05 7.30E-04 3 8 TO 24 2.24E-05 6.60E-04 4 24 TO 96 7.94E-06 5.40E-04 5 96 TO 720 1.71E-06 4.00E-04 node 8 GRC%mhs@@N m.

MAINiCODENSER9 BM FASE XQ8.. ITBd end time interval XQ8T(hrs) EAB XQ8EB LPZ XQ8LP G XQ8G ITEM 5.3 1 0 'IO 2 2.70E-04 1.32E-04 1.74E-04 2 2 TO 8 6.02E-05 1.47E-04 3 81024 4.07E-05 1.27E-04 4 24 'ID 96 1.73E-05 1.01E-04 5 96 10 720 5.10E-06 7.20E-05 node 2 W D #iP % M2sW "iDRWF MPM FASE XQ2.. 5.4 1 2 0 0 1.12E-03 XQ2 (NLY IMPACTS RELEASE FOR DURAT12 OF OG21 RELEASE ABB Combustion Engineering Nuclear Operations

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40 of 46 SYSTEMS & TRANSIENTS Page Number IV.7 Breathing Rates and Occupancy Factors (Data From Ref. 3) a breathing ITai .5. 5 rates -

end rate time (hr) m3/seci 8 3 . 4 7E- C '," '

24  !.~I5E-04 720 3.32E-04j b

occupancy ITai 5. 6 factors end factor time (hrs) 24 1

96 0.6 720 0.4 ABB Combustion Engineering Nuclear Operations

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41 of 46 SYSTEMS & TRANSIENTS Page Number

. V. CIE USED / UPIRTES The ve r scalculation employs the TVADOSE computer code (See Reference 2) i on 'IVD92395.

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1066-S&'I95-C-002 0 42 of 46 SYSTEMS & TRANSIENTS Page Number VI. COdPUTER RIN SWhBRY Case 1 refers to the base analysis as defined in Section IV. Case 2 is similar to case 1 except that per Reference 3:

and OG21 = 3.1E-3 CFH for 105 seconds ,

OG51 = 0.9E+6 CFH l

1 COdPUTER OUTPUT LIST I Case # JOB ID Run Date/ Time Description or Name Cd Rom Date j 1 or see below Run date 9/23/95 HICE RB FLCW; BFCASE1 Run time 11:22:57 gyp.

Cd Rom date 9/25/95 BFNE1 M 2 or see below Run date 9/23/95 LO,V RB FLOV BFCASE2 Run time 11:25:40 MUr-Cd Rom date 9/25/95 BFCASE2. CUT 1

l Filename: Job Id:

Description:

bfcasel. inp Ou5ha26c.cdf Input file bfcasel.out Ou5hab88.cdf Standard output file bfcasel.act Ou5h980o.cdf Regional activity file bfcasel.dpr Ou5h9s4g.cdf Detailed debug edits bfcase2.inp Ou5hedho.cdf Input file b fcas e2.ou t Ou5hesj k. cd f Standard output file bfcase2.act Ou5hb8c0.cdf Regional activity file bfcase2.dpr Ou5hclf s.cdf Detailed debug edits Computer output stored in directory / misc /6602r00/out on t he CD-RCM volume identified on the cover sheet. Output files are also included in Appendices. ,

l

• No t e , in the file which contains the activity information, the following heading is used:

ACTIVITIES (CI) -- REGION NUMBERS IN ( )

DW (2) WW (3) RR (4) RB (5) SR (6) MSL (7) MC (4)

In this heading "RR (4)" should be interpreted to mean the Control Room (Region 4) and "hC (4)" should ac tually l be ennee indicating Main Condenser (Region 8). This code anomaly should not nny eigniricant nrnhieme ABB Combustion Engineering Nuclear Operations

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43 of 46 SYSTEMS & TRANSIENTS Page Number VII . RESULTS AbD OCNIUSICNS The criteria for control room habitability given in NUREG-0800, '

Standard Review Plan 6.4, Rev. 2, 1981 is as follows:

limit for 30 day dose accumulation: --

5 Rem whole bod 30 Rem Thyroid (ytodine inhalation) 30 Rem skin The NRC allowable offsite doses are given in 10CFR100.11 to be:

25 Rem total whole body 300 Rem total to the thyroid due to iodine exposure.

These limits apply to EAB 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and LPZ thirty day doses.

The maximumresults of thedose thyroid two cases areissurrmarized in the CR under 18 rem. in the table below: The Dose sumnaries can be found on pages A-56 thru A-58 and B-56 thru B-58 of Appendices.

s I.OCATICN DOSE TYPE CASE 1 CASE 2 2 HR DOSE 30 DAY 2 HR DOSE 30 DAY (Raf) DOSE (RHM) DOSE (RBf) (PIM)

CR 'IRYROID 17.9 I

17.41 CR SKIN 1.794 1.782 CR  % BOLE 0.046 BCDY 0.045 EAB 'IHYROID 3.159 2.738 EAB SKIN 0.05658 0.0441 EAB  % HOLE 0.0750 0.059 BCDY LPZ 'IHYROID 5.786 5.552 LPZ SKIN 0.4928 0.4826 LPZ  % HOLE 0.2823 BCDY 0.269 ABB Combustion Engineering Nuclear Operations

1066-S&T95-C-002 0 COMBUSTION ENGINEERING, INC.

44 of 46 SYSTEMS & TRANSIENTS Page Number The contribution of "other" isotopes to the whole body LPZ and EAB doses are as follows:

Case 1: EAB 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> : other contribution = .000527 Rem Case 1: LPZ 30 day: other contribution = .00034 Rem Case 2: EAB 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> : other contribution = .00050 Rem Case 2: LPZ 30 day: other contribution = .00033 Rem For case 1 Iodine -131 contributes 16.64 rem to the control room thyroid dose (30 day), distributed among the three iodine forms as follows:

1 elemental : 3.21 Rem organic: 12.51 Rem Particulate:0.9232 Rem  !

Additional details associated with the radionuclide contributions to l the dose can be found in the computer output files (See Appendices A and B). Also included in the appendices are the predicted plant s activities for various times into the event. l I

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i I

1066-S&'I95-C-002 0 kB I COMBUSTION ENGINEERING, INC.

Calculation Number Rev.

l 45 of 46 SYSTEMS & TRANSIENTS Page Number

. VIII. AtwtKBrCES

1. 'IML- 9 0 - 12 0, " Analytical Models for LOCA Radiological Dose Consequences (Basis for LDOSE Program)", S. Rosen, November, 1, 1990.(Combustion Engineering Proprietary) 2.

1066-S&T95-C-001 TVADOSE:OCMPUTER PROGRAM FOR 'IEE C5i.CULATICN OP l BROANS FERRY ADVANGD SOURCE TEmi LEAKAGE DOSES, M. Michonski, I September 29, 1995, (Combustion Engineering Proprietary)

3. PSAT-04000U.03,Rev.'1, " Design Data Base for Application of the Revised DBA Source Term to the TVA Browns Ferry Nuclear Power Plant", J. Metcalf, September 22, 1995

4. Attachment A "Workscope" to Letter L. Brown-Herzl (Polstar Applled Technology, Inc) to Raymond Schneider (Combustion Engineering, Inc.) August 31, 1995.

5. PVN3S UPDATED FSAR (APPENDIX 15B)

6. Letter from J. Metcalf (PSAT) to Ray Schneider (ABB), dated -

September 20, 1995 '

7. NUREG-1465, Accident Source Terms for Light Water Power Plants, January 1995, USNRC ABB Combustion Engineerinc Nuclear Operations

1066-S&'I95-C-002 0 h Calculation Number Rev.

COMBUSTION ENGINEERING, INC.

46 of 46 '

SYSTEMS & TRANSIENTS Page Number IX. APPEDICES 4

NOTICE

This Non-Proprietary Version of Calculation Number i 1066-S&T95-C-002 Rev. 0 (Attachment 1 to Non-Proprietary PSAT 04000U.04) does not include anyof the four appendices identified in the Table of Contents.

Appendices A, B, and D are proprietary, and Appendix C contains information identical to that of Attachment 2 of PSAT 04000U.04.

I ABB Combustion Engineering Nuclear Operations j j