ML20135E083

Analyses in Support of Gpu Responses to NRC Request for Addl Info Concerning Compliance W/NUREG-0737,Item II.B.2, Final Rept

ML20135E083
Person / Time
Site:	Oyster Creek
Issue date:	06/30/1985
From:	UNITED ENGINEERS & CONSTRUCTORS, INC.
To:
Shared Package
ML20135E062	List: ML20135E058 ML20135E083
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM NUDOCS 8509160278
Download: ML20135E083 (49)

Text

_ _ _ _

Attachment A ANALYSES IN SUPPORT OF CPUN RESPONSES TO NRC RAIs CONCERNING COMPLIANCE WITH NUREG-0737, ITEM II.B.2 (FINAL REPORT)

Prepared for GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION Prepared by UNITED ENGINEERS & CONSTRUCTORS INC.

June 1985 8509160278 850909 PDR ADOCK 05000219 P PDR 5

% 4 Analyses in Support of CPUN Responses to NRC RAls Concerning Compliance with NUREG-0737, Item II.B.2 CONTENTS i

1. SCOPE OF ANALYSES

2.

SUMMARY

OF RESULTS

3. POST-LOCA RADIOACTIVITY CONCENTRATIONS INSIDE REACTOR BUILDING l

4. DOSE ASSESSMElfr FOR VITAL AESAS DUE TO AIRBORNE RADIOACTIVITY INSIDE REACIOR BUILu1NG l

! 5. OVERALL POST-ACCIDENT DOSE ASSESSMENT POR THE VITAL AREAS I

6. REFERENCES APPENDIX A - POST-ACCIDElfr RADIATION SOURCE TERMS APPENDIX B - POST-ACCIDENT PIPING DOSE BOOK APPENDIX C - INPUI PARAMETERS AND REFERENCE DRAWINGS I .

4 i

l 1

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1

1. SCOPE OF ANALYSES In reviewing the plant design of Oyster Creek Nuclear Generating Station for compliance with Item II.B.2 of NUREG-0737 (Reference 1), the Nuclear Regulatory Commission issued Requests for Additional Information on five items of concern (see Reference 2). The response to Item (a) of the RAls had already been provided by GPU Nuclear Corporation. Analysis in support of the GPUN response to Ites (e) had already been provided by United via a separate study (see Reference 3). The scope of this study is to provide analyses in support of GPUN responses to Items (b), (c), and (d) of the NRC RAIs. Section 2 of this report provides an overall summary of the results of the conclusions from the analyses conducted in this study. Section 3 describes the analysis for Item (b) - Post-LOCA Radioactivity Concentrations Inside Reactor Building. Se ction 4 presents details of the analysis for Item (c) - Dos e Assessment for Vital Areas due to Radioactivity Inside Reactor Building. Section 5 presents details of the analysis for Item (d) -

Overall Post-Accident Dose Assessment for All Vital Areas.

Post-accident dose assessment, as described in Sections 4 and 5, has been conducted for the following vital areas.

Control Room Security Building (Main Gate)

Diesel Generator Building Post-Accident Sampling System (PASS) Room Hot Chemistry Laboratory Technical Support Center

1 i

Stack RAGEMS Room i

Alternate Hot Chemistry Laboratory Turbine Building RAGEMS Room t

f Post-accident dose rates, as well as time-integrated doses (TIDs), bave i been calculated. It is to be noted that, in order to properly address ,

postn ccident accessibility, the TIDs have been calculated on the basis 4

. of when access to a vital area is needed, duration of stay in the vital [

t r

area, route and time taken to reach the vital area, etc.

4 ,

In the overall post-accident dose assessment, two accident-scenarios have .

been considered--the "depressurized" accident scenario and the " pressurized" f accident scenario. Details on the derivation of these source teras are (

presented in Appendix A. The resultant Pos t-Accident Piping Dose Book, I f

along with its associated application-procedures, are presented in Appendix B.

References are listed in Section 6; input parameters, system P& ids, and  ;

General Arrangement Drawings are listed in Appendix C.

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2.

SUMMARY

OF RESULTS Results of this study, as detailed in Sections 3 through 5, indicate that the various post-accident radiation sources will not cause any serious radiation hazard for most of the vital area under consideration. For most of these vital areas, the Depressurized Accident Scenario is the controlling event, with the major dose-contribution from the airborne radioactivity in the upper portion of Reactor Building.

For the Control Room, PASS Room, and Technical Support Center, the post-j accident radiation levels are relatively low throughout the course of the j accident; thus, continuous access / occupancy is afforded for these vital i areas. For the Diesel Generator Building, the post-accident radiaiton levels are extremely low throughout the course of the accident; access to this vital area is permitted at any time during the accident.

For the Security Building (Main Gate), the pos t-accident radiation levels are relatively high; especially during t'.ne first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The TID for a person in this vital area, working 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day for the duration of accident, is about 3.6 reas--still below the allowable limit of 5 reas.  !

r l

, For the Stack RAGEMS Room, the post-accident radiation levels are extremely e i

i high, with a peak dose rate of about 1300 mR/hr during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l However, the overall TID for a technician replenishing the RAGEMS supply of liquid-nitrogen, during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, is only about 0.7 ren. Similarly, F

! I the post-accident radiation levels are also extremely high for the Turbine i

Building RAGEMS Ro .a; however, the overall TID for a technician performing l l

a similar task is about i rem. t l

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~3- [

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1 For the Hot Chemistry Laboratory, the post-accident radiation levels are exceedingly high, with a peak dose rate in excess of 3000 mR/hr during the first hour; thus, only very short duration access may be permitted. For the Alterna:e Hot Chemistry Laboratory, the post-accident radiation levels are relatively lower, with a peak of about 300 mR/hr; the overall TID for a 2-hour operation inside the Alternate Hot Chemistry Laboratory will be limited to 0.7 rem.

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1 b

3. POST-LOCA RADIOACTIVITY CONCENTRATIONS INSIDE REACTOR BUILDING The post-LOCA radioactivity concentrations inside Reactor Building have been calculated, via the United in-house computer code HDOSE (Reference 11), as a function of time. Results of the analysis are detailed in Tables 3-1 through 3-4 Pertinent assumptions and parameters for this analysis are presented in Table 3-5.

The post-LOCA radiation transport has been modeled as a two-compa rtment process, with the leakage from the first compartment going into the second compartment. The Primary Containment is modeled as the first compartment, with no external input source; the initial airborne radioactivity inside the Primary Containment is assumed to be 100% of the core's noble gases and 25% of the core's halogens. The only depletion mechanisms considered, for the Primary Containment, are radioactive decay and leakage; parent-daughter relationships have been properly undeled to account for the buildup of daughter-radionuclides inside Primary Containment.

The Reactor Building is modeled as the second compartment, with a time-dependent input source (leakage from Primary Containment). The depletion mechanisms considered, for the Rasetor Building, include radioactive decay and venting (via Standby Gas Treatment System). Parent-daughter relation-ships have been properly modeled to account for the buildup of daughter-radionuclides inside the Reactor Building.

Two models of Primary Containment leakages have been considered in the analysis. In the " ultra conservative" model, the Primary Containment

i l

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I leakage is assumed to be 1.0% per day for the duration of the accident; it is to be noted that this value is actually twice the design basis leakage j of 0.5% per day (see References 5 and 6). This assumption thus leads to i

4 exceedingly conservative estimates of the radioactivity concentrations i

inside the Reactor Building, especially during the latter stages of the accident. The resultant radioactivity concentrations, based upon this j

i " ultra-conservative" model, are presented in Tables 3-1 and 3-2, as function of time.

i f, In the " variable leakage" model, the pe ,c-LOCA leakage f rom Primary Contain-

ment is evaluated as a function of the post-LOCA pressure inside Primary 4

! Containment. The Primary Coc.:ainment post-LOCA pressure history is based j upon the data presented in OCNGS FSAR, Figure 6.2-3. The relationship i

i between pressure and leakage is based upon the formulations given in ANSI j 56.8-1981 (Reference 12). Results from the 1984 OCNGS Reactor Containment i

Building Integrated Leak Rate Test (Reference 13) have also been utilized.

l The resultant leakage from Primary Containment is presented, as a function

of time, in Table 3-6; it is to be noted that the leakage from Primary i

l Containment has been conservatively set at 0.1% per day for t = 25 minutes l l and beyond, although the Primary Containment pressure is practically l

l atmospheric after the intitial few hours. The resultant radioactivity concentrations inside Reactor Building, based upon the " variable leakage" j model, are presented in Tables 3-3 and 3-4.

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ep et* e= nm e e rom ar ate e e Om em n e.,e. ar.ee e ,n o, e..we o.n n e ass e

e a- -j H . e e

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ee @. e e e e e e e 'e e O. O o. .. e e. e

e. e a e I . e aw W C3 - e e . w ens!w w ww em ww ms == e

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. ' . . . = .e.. ar. es e e. r. .

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n. r*. O. e. O. . . n.ee. - . O. . . ~. m.  %.y.. O. '.T..,.".

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< O < e e- as e > a g e e- , 4tn n re

> 's 0 " eo e Q *e e ele e = - * *eC .e eseeeee-nem e e. e e n --- ---e O .e o eO e>%* %m N I""" O e e. . Jee - e s= ..a a a a e er at ens w ins ensw >

'p x xxx 'x x sg ac g sg sc.es

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M

' e w* - e e . Jim eaeee a mim er .= ==a e= == e ens x e ans e.J.. o > . ,. , e '3.%. <

are u oe ue ' u .u Omor *-" @

W

. e e= e ., x e .

)  : 1 i-y;n ,- ss.: %f" J.9% 9 ), 2

!.e .. x ) . g, M~

De .s.I, ay x(d. . .k L'%s . *

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og? l g

, s* , An. . e

T=LL a-se Lt-LocA LWda CaoAdia M bbhi1L8. 6 RCi/ccl j

l 1%aw L g wMQ O' e

ISOTOPE TIME ::: 0 TIME c.1 ME TIME =:5m* a m T I ME = *.$ ygu*yg T I ME =l. \ .gp TIME =. "1 b TIME :: I4. h CLA55 1

' e .......: .  ; .;r.;W - 1. ; .J*....., ,;+. < ..... .,..... ..-..

? . . Og. A. .p . N

,.4 e ....... > ., ,F e

,TOTALJ s

[0. ' NQ> d(; .h.M@. pJm! .}+ ?Olb ' lp[w90M !i e

., 40. '

L 3 as O.. .' ' O.

cwi

. g: , >

CLA5S 2 a ....... ..... ..... ..... ..... ..... ..... ..-..

" TOTAL O. O. O. O. O. O. O.

• ,. >' c-- * * >

._ . 44 , l' l

',CLA$$1 . . 3L-

• ^

V 4 .,

, o - 'T ...;...i - up sv. wen:sw a... - # ..s..

^

W 4. 1 +

e

' ',: .O .."sl.aa WJ s

' . i . e' . '

G.J1 5.'

I TOTAL O. O. O. O. O. O. O.

I

• I CLA55 4 k{ M g

! 8482M . , ;O.e?f M u g 3.0387E-07 4 11.00902-06,. ' 4.958tE407.... 2.6503E'08pn3:2;2659E-tO- 1.0898E-131

" ' + C y8R82.;w J M OTy Mfj 746097E-07.+ 5.0995E-05

~.k g k gd 5.8640E 05' 1 .

" RR83 ~ /0; :t 3. tO99E;O6 i 3'7.4887E*04'

.2.3526EeO4~ 1;O634E*051:h 's ,i l'1063E-03"';t;8722E-056s

. "1.3994E 3iO933E-05~ 1.364_5E?O3; BR84 O. 9.OO49E-05 3.3772E-04 7.7423E-04 6.4241E-04 3.0953E-04 5.1208E-05 BR85 O. 1.OO94E-04 1.6759E-04 1.8726E-05 5.8053E-08 7.2637E-12 2.1395E-24 I 139 O. 5.7743E-04 2.36ttE-03 e.2620E-03 1.4411E-02 2.4193E-02 4.117tE-02

• ;I .132 . - 4 0.< :. , tBE *04 -:3.26 TIE?O3: . E, ( 1.035fE 02 - 1.5INE *02 ; ,,1,8885E'02- 1.7928E-02 F 'r'!'133+ Q,. ;: 0.'O 'y;w / C 4m s?8.13

.c.

9;4423E-04 *i3.8534E*03 4 1;3354EkO2l . '2.2892E-02? ..:3;73tSE*02 5.988BE*02

!4 134 i#'

=

t O[:b ^ '9W i :1 2057E-03 *4 4.6789Ef03' E 1.2652E'02'

• 114038E 9 1.0956E-02 2 4.2821E-03 I 135 O. 9.6544E-04 3.9212E-03 1.3269E-02 2.1818E-02 3.3ti9E-02 4.6157E-02 I = ....... ..... ..... ..... . ..... ..... ..... .....

I " TOTAL O. 4.7567E-03 1.8819E-02 5.9418E-02 9.OO60E-02 1.2620E-01 1.7089E-01 1

,(. < w. & n , , m . - ~ , . . . . , ,  :+ .. ,a - ,;i i

=

i ' IC L!

< . gag 3g A$'Sk5)"Am-e '<~ : o.%*+W -CpAaA

'2.3378E-04:,,"t!38$lt04dPi2E9936E-03O bikN .Y Nf N.hb. s ,56M (. (.4390E*03- ~C5.6483E-03 5.5633E-03 KR85M O. 5.0962E-04 2.0656E-03 6.8814E-03 1.0993E-02 1.5885E-02 2.OO90E-02 KR85 O. 2.4211E-05 9.6866E-05 3.3832E-04 5.9008E-04 9.9489E-04 1.7050E-03 KR87 O. 8.9944E-04 m.5473E-03 1.0380E-02 1.3220E-02 1.3033E-02 7.8692E-03 kR68 s ,O. ~

y.o ut.3533E-03 ,s . 5.32 ME.03 x .,1.7355E-02.,.....:2.609BE'02~ >v3.4535 F52 3.6757E-02

.KR89 . O. ' dN.0+< BM%1.2844E-03 .- l3 2;2625E'03 V :3.8232E-10 1.3996E-21

= sa- RE 13 iM ' ^OF ' M t'E M 1.3834EeOSs 5.4573E-05 4 Ws1;9813Ea04 2.0994E-04iM-)1;4863E-06' 5^ 23.4601E*04> '5.8219E-04' -9.952SE-04+

XE133M O. 1.5353E-04 6.275tE-04 2.1908E-03 3.8046E-03 6.3407E-03 1.0633E-02 i

XE133 O. 4.4305E-03 1.8940E-02 8.3500E-02 1.1071E-Ot 1.8574E-Of 3.1566E-01 XE135M O. 1.203tE-03 4.5004E-03 7.9852E-03 5.841tE-03 5.6870E-03 7.4565E-03 p7 765BEeO4 m 3;174&Ee03

.p t 5 t456Ef 02 e <2;O273E'02H.n >3.4344E-02; 5.7734E-02.

.e

. s AE135ygt37; ,4: 4u goj o Oynh ~$ .-whm:W i 3.0926E 03 ; "i t.S tBOEaO3i: +1112821E*O3

=  :, : in'W Mfh113477E-05T O.it:1809E-083 4 3.4513E-10 P*

=

AE138

• O.E + ~ 3.3176E-03 - 1, t 173E*02 i < d!1;5521E-02 JX . 5.3089E-03 1 i6.2352E-04 1,2072E-05 i ....... ..... ..... ..... ..... ..... ..... .....

TOTAL O. 1.7324E-02 5.0233E-02 1.4019E-01 2.0165E-Of 3.0338E-01 4.6440E-01 CLASS 1G " . -

. . . . . . . q i , .l, q ..aag g g g;'.. q .es. .

~

=

] .

's-

" ^*

'TOTALi 'Oi * *w A 'O.V "+.^' f;g " Og&.ehr

.' ^ ' Nggp ' ' Hp.4aeg>#

Or N #<,',O.f , [.;. a e ig

• h*c..x f.i. . . ~

' 40; 'O.

TOTAL O. 2.20stE-02 7.705tE-02 1.9961E-01 2.9171E-Of 4.295eE-01 6.3530E-Of

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Table 3-5 POST-LOCA RADIOACTIVITY INSIDE REACTOR BUILDING -

ASSUMPTIONS AND PARAMETERS Core Power Level 1930 MWe Radioactivity per NWt GE Calculation SNUMB = 7007S Radioactivity Releases from Core Noble Gases 100%

Halogens 25%

Airborne Fractions Inside Primsry Containment Noble Gasec 1.0 Halogens 1.0 Leakage from Primary Containment 1.0% per day

" Ultra-Conservative" Model 1.0% per day for duration of accident

" Variable Leakage" Model See Table 3-6 Radioactive Decay Inside Primary Containment Considered Airborne Fractions Inside Reactor Building Noble Gases 1.0 Halogens 1.0 Free Volume of Reactor Building 1.8 x 106 ge3 Venting via Standby Cas Treatment System 2600 cfm Radioactive Decay Inside Reactor Building Considered

Table 3-6

, POST-LOCA LEAKAGES FROM PRIMARY CONTAINMENT -

VARIABLE LEAKAGE MODEL Time Interval Leakage Rate, % per Day 0 - 1 mi m te 0.30 1 - 5 mi m tes 0.23 4

5 - 25 mi m tes 0.16 25 minutes - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 0.10 j 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> - duration of accident 0.10 I

l i

I f

- -., .-, - - - - , . , - .--r , , - n,n_-- n_.. , - - - - . , _ _ ,n.,,- - , ..g_--- -n-- - ,, ,,-.

I 1 .

i

4. DOSE ASSESSMENT FOR VITAL AREAS DUE TO AIRBORNE RADIOACTIVITY  !

I INSIDE REACTOR BUILDING 'l 4

f Based upon the results presented in Section 3, the post-LOCA gama source i

i terms inside Reactor Building are developed. Utilising the United in-house 1

l computer code ACfESP (Reference 9) with the ORNL-RSIC data file DLC-62 ,

i.

j (Reference 8), these time-dependent radioactivity concentrations (in uCi/cc) i i

are then " transformed" into time-dependent multigroup gamma source teras (in l i

MeV/cc-sec). The resultant game source terms are presented in Tables 4-1 I l

l and 4-2. L L

These source terms are then utilised, with the United in-house computer code  ;

QAD-UE (Reference 10), to determine the post-LOCA doses for the following 1

l vital areas.

l

Control Room l I

l Security Building (Main Gate) f i f j Diesel Generator Building i t

[

Post-Accident Sampling System .(PASS) Room  ;

} Hot Chemistry Laboratory [

! i Technical Support Center [

1  !

4 Stack RAGEMS Room i i r i  !

l Alternate Hot Chemistry Laboratory )

1 Turbine Building RAGEMS Room i

I General arrangement drawings of various plant buildings, se listed in i.

, Appendix C, have been utilised in developing the Combinatorial Geometry l

inpute for QAD-UE. The geometric modeling has been realistic and detailed, f 1

! with only minor conservative simplifications. l i i l _,_

{

t

. . . - . - _ . - . . - - - . . . , . . - . . - . - . . . - . - - . - - . - , . - - - . . - - - . - . . -..----.--J

The resultant post-LOCA dose rates, for these nine vital areas, due to the .

airborne radioactivity inside Reactor Building, are presented in F'igures 4-1 through 4-8 as functions of time. Figures 4-1 and 4-3 through 4-8 represent the results based upon the " Ultra Conservative" Model. Figure 4-2 shows the results, based upon the " Variable Leakage" Model, for the Security Building (Main Gate).

From these results, it is observed that the post-LOCA dose rates are quite low for the Control Room, Diesel Generator Building, PASS Room, and Technical Support Center. For the Hot Chemistry Laboratory, the post-LOCA dose rates are relatively high due to its close proximity to the Reactor Building, with only minimal shielding in between. For the Security Building (Main Gate),

Stack RAGEMS Room, Alternate Hot Chemistry Laboratory, and Turbine Building RAGEMS Room, the relatively high post-LOCA dose rates are mainly due to the

" direct shine" through the roofs, from the airborne gaans sources in the upper portion of the Reactor Bui'. ding, with practically no shielding in be tween.

Further discussions are provided in Sections 2 and 5 of this report.

t 0

0

l l

t TABLE 4-1: FOST-IACA CAf98A SOURCE TERM INSIDE REACTOR BUILDING (MeV/cc-sec)

" ULTIMATE CONSERVATIVE" MODEL  !

- Energy (MeV) 0 0.5 Mr 1 Mr 2 Mrs 4 Mrs 8 Mrs 12 Mrs 16 Mrs 24 Mrs 10 Days 20 Days 30 Days  ;

0.0 - 0.2 0 1.3+03 2.4+03 4.1+03 6.8M 3 1.0+04 1.2+04 1.3M4 1.4+04 7.3+03 2.3+03 5.6M2 0.2 - 0.4 0 2.0+03 3.2+03 5.7M3 9.9+03 1.6+04 1.8+04 1.9+04 1.9+04 8.2+03 3.6+03 1.4+03 0.4 - 0.9 0 1.5+04 2.2+04 2.9+04 3.2+04 3.0+04 2.8+04 2.6+04 2.2+04 2.1403 7. 3+02 2.8+02 i

0.9 - 1.35 0 5.9403 9.6+03 1.4+04 1.8404 1.9+04 1.6+04 1.2+04 7.0+03 5.4+01 1.8-01 0 U.

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t

! TABLE 4-2: POST-10CA CAletA SOURCE TERMS INSIDE REACTOR SUILDINC (MeV/cc-sec)

, t VARIABLE LEAEACE MODEL i

Energy (MeV) 0 5 Mins. 25 Mins. 1 Mr. 2 Nrs. 4 Mrs. 8 Hrs. 16 Hrs. 1 Day 4 Days 15 Days 30 Days i 0.0 - 0.2 0 5.9+01 1.9+02 3.1+02 4.7+.02 7.2+02 1.0+03 1. 3+03 1.4+03 1.3+03 5.5+02 1.1+02 0.2 - 0.4 0 1.2+02 3.0+02 4.2+02 6.5+02 1.1+03 1.6+03 2.0+03 1.9403 1.3+03 6.8+02 2.3+02 0.4 - 0.9 0 8.8+02 2.3+03 2.9+03 3.3+03 3.4+03 3.1+03 2.7+03 2.2+03 8.1+02 1.5+02 4.5+01 0.9 - 1.35 0 3.2+02 8.8+02 1.2+03 1.6+03 2.0+03 1.9+03 1.3+03 7.2+02 9.1+01 1.7 0 l

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1

! 5. OVERALL POST-ACCIDENT DOSE ASSESSMENT FOR THE VITAL AREAS Utilizing the Post-Accident Piping Dose Book, as presented in Appendix B of this report, the post-accident doses due to the radioactive piping of various systems are calculated for the nine vital areas. Time-dependent ,

dose rates, as well as time-integrated doses (TIDs), have been calculated.

Two accident-scenarios have been considered-the Depressurized Accident Scenario and the Pressurized Accident Scenario. The relationships between accident scenarios, source terms, and associated systems / piping are detailed in Appendix A of this report.

For the Depressurized Accident Scenario, the corresponding piping doses are combined with the results in Section 4 to yield the overall post-accident doses for the vital areas. For the Pressurized Accident Scenario, since there is negligible airborne radioactivity inside the Reactor Building, the l

l piping doses represent the overall post-accident doses for the vital areas.

The overall post-accident dose rates for the vital areas, as well as the overall time-integrated doses, are presented in Tables 5-1 thru 5-9.

l For the Control Room, the post-accident dose rate peaks at about 60 mR/hr within the first hour, for either accident scenario. However, the TID is quite low, less than one rea for either accident scenario (even with the occupancy factor being conservatively assumed to be 100% for the duration of the accident) . For the Depressurized Accident Scenario, the major dose-contributor is the Core Spray System piping, located at the El. 51' level of the Reactor Building. For the Pressurized Accident Scenario, the major dose-contributor is the Main Steam piping inside the MS/FW Tunnel.

~

For the Security Building (Main Gate), the post-accident doses are relatively high for either accident scenario. Under the Depressurized Accident Scenario, the peak dose rate is about 190 mR/hr; the associated TID, with the occupancy factor assumed to be 33% for the duration of the accident, is about 3.6 rems. The major contribution is due to the " direct shine" from the upper portion of the Reactor Building, through the roof of the Security Building (Main Gate). Under the Pressurized Accident Scenario, the peak dose rate approaches 450 mR/hr, with the associated TID being about 1.5 rems. The major contribution is due to the Reactor Shutdown Cooling System piping located at the El. 38' level of the Reactor Building.

For the Diesel Generator Building, the post-accident doses are very low, with a peak dose rate of only about 3 mR/hr. The only dose-contribution comes from the airborne sources inside the Reactor Building. The piping contribution is negligible, for either accident scenario because of its location (more than 500 ft. from any piping source inside the Reactor Building and with a minimum of three feet of concrete shielding in between).

For the PASS Room, the post-accident doses are quite low for either accident scenarios, with TIDs of about 0.5 rem. Under the Pressurized Accident

, Scenario, the peak dose rate is about 30 mR/hr; the major contribution comes from the Core Spray System piping located at the elevation 51' level of the Reactor Building. Under the Pressurized Accident Scenarin, the  ;

peak dose rate is about 200 mR/hr; the major contribution comes from the Main Steam piping inside the MS/FW tunnel.

l

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For the Hot Chemistry Laboratory, the pos t-accident doses are exceedingly high. Under the Depressurized Accident Scenario, the peak dose rate is in excess of 3000 mR/hr, with the major contribution coming from the Core Spra'y System piping located at the elevation 51' level of the Reactor Building.

Results in Table 5-5 indicate that, under the Depressurized Accident Scenario, I

post-accident accessibility for this area will be severely restricted during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Under the Pressurized Accident Scenario, the peak dose rate is about 300 mR/hr; the major contribution comes from the Main Steam piping inside the MS/FW tunnel.

4 1  !

i For the Technical Support Center, the post-accident doses are very low, with a peak dose rate of less than 20 mR/hr and a TID of about 0.3 rem (even with the occupancy factor being conservatively assumed to be 100% for the duration of the accident). The only dose-contribution comes from the airborne sources in the upper portion of the Reactor Building. The piping contribution is negligible, for either accident scenario, because of its location (more than 400 feet from any piping source inside Reactor Building and with a minima of 3 feet of concrete shielding in between).

For the Stack RAGEMS Room, the post-accident doses are exceedingly high under l the Depressurized Accident Scenario, with a peak dose rate of about 1300 mR/hr. For a technician replenishing the Stack RAGEMS supply of liquid i nitrogen during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, he/she will receive an overall TID of  ;

I about 0.7 rem. The major contribution is due to the " direct shine" from the i upper portion of Reactor Building, with practically no shielding in between.  !

Under the Pressurized Accident Scenario, the corresponding post-accident i

doses are relatively low, as compared with the Depressurized case.  ;

I 1

~

l i .

For the Alternate Hot Chemistry Laboratory, the post-accident doses are relatively high under the Depressurized Accident Scenario, with a peak dose rate of about 300 mR/hr. The TID for a 2-hour operation inside the Alternate Hot Chemistry Laboratory will be limited to 0.7 rem. The major contribution comes form the airborne radioactivity inside Reactor Building. Under the Pressurized Accident Scenario, the corresponding post-accident doses are relatively low, as compared with the Depressurized case.

For the Turbine Building RAGEMS Room, the post-accident doses are exceedingly high under the Depressurized Accident Scenario, with a peak dose rate of about 2300 mR/hr. For a technician replenishing the Turbine Building RAGEMS supply of liquid nitrogen during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, he/she will receive an overall TID of about I res. The major contribution is due to the " direct shine" f rom the upper portion of Reactor Building. Under the Pressurized l Accident Scenario, the corresponding post-accident doses are relatively low, as compared with the Depressurized case. ,

I l

1

[

[

I o

Table 5-1 OVERALL POST-ACCIDENT DOSES FOR CONTROL ROOM i

Post-Accident Dose Rates (aR/hr)

Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days Depressurized Accident Scenario Reactor Bldg. 0 3.3 4.8 5.6 1.1 <1 <1 Piping 39 55 46 7.6 1.0 <1 <1 Total 39 58 51 13 2.1 <1 <1 Pressurized Accident Scenario Reactor Bldg. - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Piping 63 36 28 3.8 <1 <1 <1 Total 63 36 28 3.8 <1 <1 <1 Overall TID (assuming an occupancy factor of 100% for duration of the accident)

Depressurized Accident Scenario 0.6 rea Pressurized Accident Scenario 0.2 ren l

i

..-.g- --,n., w - ,-- -- +-y------

Table 5-2 OVERALL POST-ACCIDENT DOSES FOR SECURITY BUILDING (Main Gate)

Post-Accident Dose Rates (mR/hr)

Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days a

Depressurized Accident Scenario Reactor Bldg. 0 70 79 97 53 12 2.3 Piping 93 120 100 16 4 <1 <1 Total 93 190 180 110 57 12 3 Pressurized Accident Scenario i

Reactor Bldg. - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Piping 360 450 380 78 18 <1 <1 Total 360 450 380 78 18 <1 <1 i

t

. 1 Overall TID (assuming an occupancy factor of 33% for duration of the accident) l l

Depressurized Accident Scenario 3.6 rem ,

Pressurized Accident Scenario 1.5 rem ,

i b

t i

3 t

i

i Table 5-3 OVERALL POST-ACCIDENT DOSES FOR DIESEL GENERATOR BUILDING Post-Accident Dose Rates (mR/hr)

Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days Depressurized Accident Scenario Reactor Bldg. 0 1.7 2.5 3.1 <1 <1 <1 Piping - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - - -

Tot al 0 1.7 2.5 3.1 <1 <1 <1 i

Pressurized Accident Scenario a

Tot al - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - - -

i Overall TID (for a 2-hour operation; including transit dose)

Depressurized Accident Scenario 0.2 rea Pressurized Accident Scenario Negligible l

I l

l l

d

. , - , , . , - - . - - - -.---n, . -

- - - - , ,-e,. _ , . , , -.

Table 5-4 OVERALL POST-ACCIDENT DOSES FOR PASS ROOM Post-Accident Dose Rates (mR/hr)

Radiation 0 1 hr 8 hrs 1 day 10 days 30 days Sources 0.5 hr Depressurized Accident Scenario J

Reactor Bldg. O <1 <1 <1 <1 <1 <1 Piping 22 30 25 5.0 2.4 <1 <1 Total 22 30 25 5.5 2.4 <1 <1 Pressurized Accident Scenario Reactor Bldg. - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Piping 210 110 83 12 <1 <1 <1 i

Total 21 0 110 83 12 <1 <1 <1 Overall TID (for a 2-hour operation; including transit dose)

Depressurized Accident Scenario 0.5 rem Pressurized Accident Scenario 0.1 ren

-2 0-

h t

i l

Table 5-5 i

OVERALL POST-ACCIDENT DOSES FOR HOT CHEMISTRY LABOPATORY l i

Post-Accident Dose Rates (mR/hr) i Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days  ;

i Depressurized Accident Scenario ,

Reactor Bldg. 0 43 63 77 16 <1 <1 I i

Piping 2300 3100 2600 620 150 25 2.2  ;

Tot al 2300 3100 2700 700 170 25 2.2 I Pressurized Accident Scenario  !

Reactor Bldg. - - - - - - - - - - - - - - Neg11ble - - - - - - - - - - - - -

Piping 290 120 90 14 <1 <1 <1 f Total 29 0 120 90 14 <1 <1 <1 ,

I E

L

[

L 1

Table 5-6 OVERALL POST-ACCIDENT DOSES FOR TECHNICAL SUPPORT CENTER Post-Accident Dose Rates (mR/hr)

Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days Depressurized Accident Scenario Reactor Bldg. 0 6.4 9.4 13 3.8 <1 <1 Piping - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Total 0 6.4 9.4 13 3.8 <1 <1 Pressurized Accident Scenario Total - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Overall TID (assuming an occupancy factor of 100% for duration of the accident)

Depressurized Accident Scenario 0.3 rea Pressurized Accident Scenario Negligible i

W 6

__. - __ _ _-- . -- _ = _ = . . _

_ . _ = _ . . ..

Table 5-7 OVERALL POST--ACCIDENT DOSES FOR STACK RAGEMS ROOM Post-Accident Dose Rates (mR/hr)

Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days Depressurized Accident Scenario Reactor Bldg. 0 350 520 830 500 96 14 Piping 490 460 380 110 50 11 3.5 RAGENS Filter 0 110 180 330 210 54 8.7 Total 490 920 1100 1300 760 160 26 j

Pressurized Accident Scenario i

Total - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

1 I

Overall TID (for technician replenishing the RAGEMS supply

( of liquid-nitrogen during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

Depressurized Accident Scenario 0.7 rem Pressurized Accident Scenario Negligible l

I l

O Table 5-8 OVERALL POST-ACCIDENT DOSES FOR ALTERNATE HOT CREMISTRY LABORATORY Post-Accident Dose Rates (mR/hr)

Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days Depressurized Accident Scenario Reactor Bldg. 0 130 190 290 160 30 4.1 Piping 60 57 48 17 6.5 1.4 <1 Tot al 60 190 240 310 170 31 4.6 Pressurized Accident Scenario Total - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Overall TID (for a 2-hour operation; including trans'it dose)

Depressurized Accident Scenario 0.7 res Pressurized Accident Scenario Negli Sible l

4 e

Table 5-9 OVERALL POST-ACCIDENT DOSES FOR TURBINE ROOM RAGEMS ROOM Post-Accident Dose Rates (aR/hr)

Radiation Sources 0 0.5 hr 1 hr 8 hrs 1 day 10 days 30 days Depressurized Accident Scenario Reactor Bldg. 0 940 1400 2300 1500 340 45 Piping - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Total 0 940 1400 2300 1500 340 45

~

Pressurized Accident Scenario Tot al - - - - - - - - - - - - - - Neglible - - - - - - - - - - - - -

Overall TID (for technician replenishing the RAGEMS supply of liquid-nitrogen during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

Depressurized Accident Scenario 1 rem Pressurized Accident Scenario Negligible i

i -

i .

l l

6. REFERENCES

1) NUREG-0737, " Clarification of TMI Action Plan Requirements", U.S.

Nuclear Regulatory Commission. (October 1980) Item II.B.2: Design Review of Plant Shielding and Environmental Qualification of Equip-ment for Spaces / Systems which may be used in Post-Accident Operations.  ;

2) NRC Region I Inspection Report 50-219/83-13 (November 1983)

3) UE&C letter to GPUN (9-5-84), " Evaluation of SGTS Filter for Iodine Loading and Shielding Adequacy".

4

4) " Radiation Source Tera Information for NUREG-0578 Implementation",

General Electric Company'. (November 1979) 4

5) OCNGS Primary Containment Design Report. (September 1967) ,

l

6) OCNGS FDSAR Section V-2. l I

7) " Table of Isotopes", 7th Edition, John Wiley & Sons, Inc. (1978)  !

8) ORNL-RSIC Data File DLC-62. f I

9) ACIESP, "A Computer Code for Generation of Gamma Source Terms fron  ;

Radioisotope Data", Rev.1.0, UE&C (March 1981)  !

I t

10) QAD-UE, "A New Ganana Ray Method for the QAD-CG Shielding Code", [

Rev. 1.0, UE&C. (March 1983)  ;

L i

11) HDOSE, "A Computer Code for Radiological Releases and Population Doses," Rev. 2.0, UE&C (May 1979).

l I

12) ANSI 56.8-1981, " Containment Systes Leakage Testing Requirements." [

13) OCNGS Reactor Containment Integrated Leak Rate Test. (1984) [

h l

l i

Appendix A: Post-Accident Radiation Source Terms A-1 Introduction In this study, the post-accident radiation source terms have been derived from the General Electric calculation SNUMB = 7007S (Reference 4); Refer-ence 4 provides the specific radioactivity in the core, in terms of Ci per MWt.

In accordance with the guidelines in NUREG-0737 (reference 1), two accident scenarios have been considered-the "Depressurized" Accident scenario and the " Pressurized" Accident scenario. Assumptions and parameters for both scenarios are discussed in the following sub-sections of this appendix.

Plant parameters utilized in the radiation transport analysis have been obtained frca References 5 and 6. Radioactive decay data are based upon the information in References 7 and 8. -

e 0

, , _ - , . .,, . -- . ~ , . - - - - --- ,w---I_ -

9 i

l A-2 Radiation Source Terms-Depressurized Accident Scenario For this scenario, a non-isolatable pipe break in the Reactor Coolant System (RCS) is assumed. Various radionuclides are assumed to be released instantaneously from the core to the drywell and the torus. The initial airborne radioactivity in the drywell and the torus air space is assumed to be as follows.

100% of Noble Gases in Core + 25% of Halogens in Core The overall dilution volume for the airborne source terms, under this accident scenario, is calculated as follows:

= 1.8 x 105 ge3 (drywell) + 1.3 x 105 ft3 (torus)

= 3.1 x 105 fc3 The initial liquid source terms in the torus and associated systems are assumed to be as follows:

50% of Halogens in Core + 1% of Other Radionuclides in Core The overall dilution volume for the liquid source terms, under this accident

. scenario, is calculated as follows:

Overall Dilution Volume - Blowdown Liquid volume + Torus Weter Volume

= 7.6 x 103 ft3 (blowdown) + 8.2 x 104 ft3 (torus)

= 9.0 x 104 ft3 The initial radioactivity concentrations, eitbr airborne or liquid, are simply calculated as follows:

Initial Radioactivity Concentrations = (Specific Radioactivity in Core x 1930 MWt) Overall Dilution Volume In deriving the time-dependent radioactivity concentrations, either air-borne or liquid, the only depletion (and buildup) process considered in o

_ .- - _- -, . _= . . . - _ . -. . .__ -__ _ -.- ..- _-

in the analysis is radioactive decay; depletion due to leakages is ignored for deriving these " primary containment" source terms.

Utilising the United in-house computer-code ACTESP (Reference 9) with the ORNL-SRIC data file DLC-63, these time-dependent radios::tivity concentra-tions ' are then " transformed" into a time-dependent multigroup gamma source i terms (in MeV/cc-sec). The resultant gamma source terms, for both airborne and liquid cases, are presented in Table A-1.

I For this depressurised accident scenario, the systems / structures that may contain these post-accident source terms, either airborne or liquid, are tabulated in Table A-2. This list includes structures that may " house" the j radioactivity, systems that any circulate / carry the radioactivity during post-accident operations, and portions of other systems which (although isolated) may contain the radioactivity up to the first isolation valves outside Primary Containment.

The major components, which any contain the "depressurised" source terms outside Primary Containment, belong to the Core Spray System and the Con-i tainment Spray System. These are the systems designed to mitigate the con-sequences under a depressurized accident scenario.

Evaluation of the post-accident radioactivity concentrations inside Reactor Building, for the depressurized accident scenario, is discussed in Section f 3 of the report.

i P

I 5

h I

t

. l t

4 .

A-3 Radiation Source Teras-Pressurized Accident Scenario For this accident scenario, a severe plant transient is assumed, with the 4

RCS pressure boundary remaining intact. Various radionuclides are assumed to be instanteneously released from the core to the RCS. For a BWR, even under pressurized conditions, the RCS will always consist of two regions:

the steam region and the water region. Furthermore, when the pressure approaches the design limits, the reactor will be depressurised via the Automatic Depressurization System; in this case, the accident sequences I

t then progress essentially in the same fashion as a depressurized accident.

i For t'he pressurized accident scenario, the gaseous radionuclides plus  !

portions of the volatile radionuclides will remain in the steam region, l after release from core. The non-gaseous radionuclides plus portions of the volatile radionuclides will remain in the water region, af ter rele'ase from core.

The initial radioactivity in the steam region is assumed to be as follows:

]

100% of Noble Gases in Core + 25% of Halogens in Core

. The overall dilution volume for these steam source terms, under such an 4

accident scenario, is as follows:

Overall Dilution Volume = Reactor Steen Volume

= 5.2 x 103 ft3 The initial radiohetivity in the water region is assumed to be as follows:

! 50% of Halogens in Core + 1% of Other Radionuclides in Core The overall dilution volume for these water source terms, under such an

! accident scenario, is calculated as follows: ,

4

-_.v.- - . , . . - , . ,y- - . _ _ , _ . , , ,. - _, .,.__.y- .,y___.-.,rm. ._.,.._..,%.,,_,,,,.v_,r,+ _,_--.w.,__,, _-

. k i

e .

-Overall Dilution Volume = Reactor Vessel Water Volume + Recirculation i

Water Volume + Pottion of Feedwater Volume (

C

, = 6.2 x 103 ft3 (RV) + 1.0 x 103 ft3 (Recir.) + 3.5 x 102 gt3 (yg) {

= 7.6 x 103 ft3 {

t

[

l Derivation of the time-dependent gamma source terms, for the pressurised  !

i accident' scenario, follows the same procedures as outlined in the previous i

sub-section. The resultant gassa source terms, for both steen and water l regions, as presented in Table A-3. l For this pressurized accident scenario, the systems / components that may i

contain these post-accident source-terms, either steam or water, are tabul- [

d I

{ ated in Table A-4. This list includes components that any " house" the  !

t radioactivity, systems that may circulate / carry the radioactivity during j post-accident operations, and portions of other systems which (although isolated) may contain the radioactivity up to the first isolation valves l outside Primary Containment.  !

The major components, which may contain the " pressurized" source terms i l

outside Primary Containment, belong to the Emergency Condenser System and  !

i the Reactor Shutdown Cooling system, as well as portions of the Main Steam f i

j lines. The Energency Condenser System is a safety feature designed to l 1

l provide emergency core-cooling under such an accident scenario. The Reactor i i

Shutdown Cooling System, although not safety-related, is conservatively  !

I

, included in the analysis; this system, if operable during post-accident  !

i +

l conditions, is a viable option for providing emergency corecooling under [

i the pressurized accident scenario.  !

The Core Spray System and Containment Spray system will not contain these  !

" pressurized" source terms, since both systems take suction directly from j i

the torus. If the torus water contains post-accident source terms, it means that the reactor is already depressurized via the Automatic Depress-urization System.

Under the pressurized accident scenario, the Reactor Building is not expect-ed to contain any significant post-accident source terms.

1 e

t ll l

I t

k r

1 f

1 h

g 4 k

I b

E-1 4

9 Table A-l Radiation Source Terme - Depressurised Accident Scenario Airborne Source Terme (McV/cc-sec)

Energy Croup (MeV) 0 0.5 hr 1.0 hr 8.0 hre 24 hre 10 days 30 days 100 days 1 year 0.0 - 0.2 5.0+07 3.8407 3.5407 2.5+07 2.0+07 6.8+06 5.0+05 2.1+02 0.

0.2 - 0.4 1.1+08 5.6+07 5.0+07 4.4+07 3.2+07 8.2+06 1.5+06 3.5+03 0.

0.4 - 0.9 7.3+08 4.4+08 3.5+08 8.5+07 3.8+07 2.0+06 2.9+05 5.9+03 5.3t03

. 0.9 - 1.35 3.5+08 1.6+08 1.4+08 4.7+07 1.1+07 7.3+04 0. O. O.

1.35 - 1.8 3.8+08 1.5+08 1.1+c8 4.7+07 5.6+06 3.5+04 0. O. O.

1.8 - 2.2 3.2408 1.2+08 7.6+07 1.1+07 5.9+05 1.1+04 0. O. O.

2.2 - 2.6 2.1408 1.6+08 1.3408 2.1+07 7.1+05 1.3+03 0. O. O.

2.6 - 3.0 3.2+07 1.5+06 1.1+06 9.1404 1.5+03 0. O. O. O.

3.0 - 4.0 5.3+07 3.8+05 2.1+06 1.8+04 0. 0. O. O. _ 0.

4.0 - 5.0 2.5+06 6.2+04 2.9+04 0. O. O. O. O.

1.tquid Source Terms (MeV/cc-sec) 0.0 - 0.2 1.9+07 1.5+07 1.3+07 6.4405 5.6+06 2.3+06 9.3+05 3.2+05 1.1+05 0.2 - 0.4 1.8+08 1.7+08 1.7+08 1.4+08 1.3408 5.6+07 1.040 7 3.0+04 3.9+0I 0.4 - 0.9 3.2+09 2.6+09 2.2+09 5.4+08 2.7+08 4.5+07 2.5+07 1.2+07 2.3+06 0.9 - 1.35 1.3+09 1.2409 9.3+08 3.2+08 7.9+07 1.6+06 4.7+05 1.6+05 9.3+04 1.35 - 1.8 7.3+08 9.3+08 9.0+08 2.1+08 6.0+07 1.4+0? 4.6+06 1.7+05 4.0+04 1.8 - 2.2 1.9+08 2.8+08 2.6+08 4.5+07 3.7+06 2.4+05 1.5+05 1.3+05 6.8+04 2.2 - 2.6 1.0406 3.3+05 7.4+03 0.

I, 2.6 - 3.0 0. O. O. O.

3.0 - 4.0 0. 6. O. O.

4.0 - 5.0 0. O. O. O. ,

y m.

i __ _

i l

- TABLE A-2 Systems / Structures Containing Depresstrized Accident Source Terms Airborne Source Terms Drywell -

Torus Air. Space Sampling Lines connected to Drywell/ Torus Air Space Portions of Main Steam piping, up to isolation valves V-1-9 and V-1-10 Liquid Source Terms Torus Liquid Region Core Spray System Containment Spray Systen Sampling Lines connected to RCS/ Core Spray System Portions of Feedwater piping, up to isolation / check valves V-2-71 and V-2-72

~

Portions of Reactor Cleanup System piping, up to isolation valve V-16-14 1

/

I i

l l

~

, ~ . , _, . . . _ _ - - - - . _,_,,r,.m.- y

, , I i

i Table A-3: Radiation Source Terms - Pressurtred Accident Scenario i Steam Source Terms (MeV/cc-sec) 8.0 hre 10 days 30 days 100 days 1 year Eneray Croup (MeV) 0 0,5 hr 1.0 hr 24 're 2.2+09 2.1+09 1.5+09 1.2+09 4.0+08 2.9+07 1.3+04 0, 0 0 - 0.2 2.9+09 2.6+09 ' 1. 9 +09 4.8+08 8.7+07 2.1+05 O.

0.2 - 0.4 6.5+09 3.4+09 2.9+09 2.1+10 5.1+09 2.2+09 1.2+08 1.8+07 3.5+05 3.1+05 0.4 - 0.9 4.4+10 2.6+10 2.8+09 6.6+08 4.4+06 0. O. O.

4 0.9 - 1.35 2.1+10 9.8+09 8.1+09 .

6.5+09 1.9+09 3.4+08 2.1+06 0. O. O.

1.35 - 1.8 2.3+10 8.7+09

• 1.9+10 6.9+09 4.6+09 6.8+08 3.5+07 6.6+05 0. O. O.

1.8 - 2.2 2.2 - 2.6 1.3+10 9.3+09 7.6+09 1.2+09 4.3+07 7.9+04 0. O. O.

6.6+07 5.4+06 9.1+04 0. O. O. O.

2.6 - 3.0 1.9+09 9.0+07 1.3+08 1.1406 0. O. O. O. O.

3.0 - 4.0 3.2+09 2.3+08 1.8+06 0. O. O. O. O. O.

4.0 - 5.0 1.5+08 3.7+06 Water Source Terms (MeV/c :-sec) 1.8+08 1.5+08 7.6+07 6.7+07 2.7+07 1.1+07 3.8+06 1.3+06

. 0.0 - 0.2 2.3+08 2.0+09 2.0+09 1.7+09 1.5+09 6.6+08 1.2+08 2.6+05 4.6+02 0.2 - 0.4 2.1+09 3.1+10 2.6+10 6.4+09 3.2+09 5.3+08 3.0+08 1.4+08 2.7+07 0.4 - 0.9 3.8+10 1.4+10 1.1+10 3.8+09 9.4+08 1.9+07 5.6+06 1.9+06 1.1+06 o.9 - 1.35 1.5+10 1.1+10 9.5+09 2.5+09 7.1+08 1.7+08 5.5+07 2.0+06 4.8+05 1.35 - 1.8 8.7+09 3.1+09 5.4+08 4.4+07 2.8t06 1.8+06 1.5+06 8.1+05 1.8 - 2.2 2.3+09 3.3+09 1.5+09 1.7+08 4.8+07 1.2+07 3.9+06 8.8+04 c.

2.2 - 2.6 8.2+08 1 1+09 O. O.

2.6 - 3.0 1.3+08 1.0+09 8.2+08 5.1+07 9 9tof 0. O.

3.0 - 4.0 5.0+08 3.9+08 2.6+08 1.8+07 3. 2 H15 O. O. O. O.

3.8+07 3.3+07 6.1+06 1.2+05 0. O. O. O.

4.0 - 5.0 9.0+07 I -

me e

TABLE A-4 4 Systems / Structures Containing Pressurized Accident Source Terms Steam Source Terms .

Steam Region of Reactor Vessel

. " Steam" portion of Emergency Condenser System

, Portions of Main Steam piping, up to isolation valves V-1-9 and V-1-10 Sampling Lines connected to Main Steam Liouid Source Terms Liquid Region of Reactor Vessel ,

" Condensate" portion of Energency Condenser Systen Reactor Shutdown Cooling System RCS Recirculation Loops Portions of Feedwater piping, up to isolation / check valves V-2-71 and V-2-72 i Portions of Reactor Cleanup System piping, up to isolation valve V-16-14 s

I i

L i

c 6

L t

. I r

1 h

_ . _ - . - , . . - . . ~ - . - - ,

Appendix B: Post-Accident Piping Dose Book B-1 Post-Accident Piping Doses Utilizing the source terms presented in Appendix A, the post-accident doses due to piping of various sizes are . calculated, via the United in-house computer code QAD-UE (Reference LO). Table B-1 presents the different times af ter accident initiation, for which dose-rates and time-integrated doses are determined.

The geometric source-shield gamma ray transport model is illustrated in ]

Figure B-1. All geometric variables are defined in the figure. Table B-2 ,

provides a list of radial and angular (for axial dependence) detector place-ment data sets. Both detector placement data sets are permutated togetber in the table. j i

Detectors are placed relative to the side and the' end of liquid containing pipes. Table B-3 provides the different liquid containing pipe siza-shield permutation configurations that are given for side detectors. Only the 7 foot long pire permutation configurations are considered for the end detec- i i

tors. Table B-4 provides different air / steam line size-shield permutation [

t configurations. Only side detector data is provided , for the air / steam i lines. {

r i

e s - --

B-2 Interpolation Techniques for using Piping Dose Book Interpolation is performed between tabulated points for shield thickness, time af ter release, detector location, and pipe size. Interpolation rela-tionships are based on physical models. Shield thickness and time after release relationships assume an expcusntial dependence. Radial detector position and pipe radius are assumed to follow an inverse Rn dependence.

Pipe height relations are based on line source uncollided flux models.

All interpolation relations are provided in Table B-5. Geometric parameter definitions are given in Figure B-1.

Interpolating tabulated dose data for parameter X first entails looking up doses at surrounding valuca of X(i.e., Zo, X u). Needed constants are obtained by using the looked up data in the fitting relationship. These constants are used in the interpolation relation, F(S, Xo), to obtain the

. dose at I with:

DOSE (X) = DOSEO (X )

• F(X, XO )

Extrapolation from tabulation from tabulated geometric and time parameters is avoided. Frequently, pipes encountered may be longer than 20 feet.

These pipes are handled by breaking up the pipe into several shorter seg-ments and the dose for each segment is determined. The doses from all segments are combined to obtain the needed data.

l

. ,- ,. -<--y , - , - <<-,_.-,...y__ ,._m_,y_.-. . - - , - , -., ,m-e.-.m ,-n,,_, , , . - . . , - - - , , , , , , , .y_,--w,m..,_m,,,,-e,..-y-

TABLE B-1

. Time After Release Dose Rates Time Integrated Dose O 30 days 0.5 hr. 1 year 1 br.

8 hr.

I day 10 days 30 days 100 days 1 1 year d

I l

I

. \

I i

l O

I

TABLE B-2 Values fo: Detector Location Permutation Distance From Surface Angle *

(Feet) (Degrees)

Contact 0**

1 15 2 30 3 45 4 60 T

5 75 10 30 70 100

• See Figure B-1 for geometric definition l

• • Only angular value considered for and detectors 1

I i

t

- i s

,- -,~ - -,

n . , - ,

TABLE B-3 Values for Permutation Of Liquid Pipe Geometric Variations

- t i

Pipe Pipe Shield l Diameter Length Thickness (INCH) (FT) (PT) k 3/8 3 0 -

1 7 1 ,

i 2 13 2 4 20 3 I

6 4 r

8 i

i 12  :

16 I

f i-l, I

't t

~

l t

e ,

o c ,

TABLE B-4 Values for Permutation of Air Line Geometry

)

Air Line Shield Diameter Length Thickness (INCH) (PT) _

(FT) 1 3 0 2 7 1 24 13 2 20 3 4

l 4

6 1

l l

e e

.---w - - - - ,-, _ - - - . - - . , , , , . . . - ,

~ o -

( ,

TABLE B-W -

NEEDED RELATIONSHIPS l l .

F(XpXO )+

l Idterpolation Fittin N Peremeter Relationship Relatione ip_ Coadment s 6

Time After Release exp (A(t-to)) ,,Ln (g ) 1) Abplied to dose rates only t (tu-to) j i

Shield Thickness esp (gat) g.Ln ( ," ) 1) AT-T-To (Tu-TO) , R= distance traveled from T pipe centerpoint to detect-or throuch shield

2) Due to buildup factors in calculations, use To*0 i

Detector Radial Distance Deo n=La(DO(Dau))

DO(Do o) 1) Relationships also used for p, )n amiel detector distance, Dg De Dq D eu f

i i

g(DO(Oou))

Detector Angular exp (y(1-cos(0,-9,o))) t- M(9so e 1-cos(Gou-Oso)

Dependence Os l

l Fipe Diameter ,Ln(D0 DPu)

)n l Dp Ln(Dub) p l II') 1) Adjust dose for radial de-Pipe Length 1 - B(L-Im) 8*(1-(DO(Im))) tector location before unin,q g Im-lo fitting relationship

2) Only use for L<16 4 3) Bre%k up larger nipes into  !

Dose (X) = Dose """ * * ** ' "*"

• M Do (X) = Dose (X)(Ko)* F(XjXO ) ermine Os for each seament.

e

I r-- ** - -

e .- 1

.- . j l

l ricURE E -l S**  ;

r u, o i i X ,

• es y L u .l D.

JL -

A

. , T ,s

'I DE l

Dp - diameter of pipe T - shield thickness E*

• L - length of pipe

• End detector E - axial distance Dg from and of pipe - radial distance specified

- with angle S '

E

• • Side detector S - radial distance DS from pipe centerline - axial distance t l from midheight of pipe specified with angle es. .

9 I . ,

i

\ '

l i l .

I e m W

o C-3 General Arrangement Drawings Site Plan Drawing #19702, Rev. 2 (1/83)

Reactor Building Drawing #BR 2060, Rev. 4 (4/78)

Drawing #BR 2061, Rev. 5 (3/83)

Drawing fBR 2062, Rev. 2 (4/70)

Drawing fBR 2063, Rev. 4 (4/70)

Drawing #BR 2064, Rev. 4 (4/70)

Security Building (Main Gate) Drawing #KL 8050, Sheets 1-9, Rev. 0 (4/81)

]

Diesel Generator Building Drawing fBR 4020, Rev. 1 (5/70) -

Drawing #BR 4021, Rev. 2 (5/70)

PASS Room and Hot Chemistry Laboratory Drawing fBR M0010, Rev. 3 Drawicg fBR M0031, Rev. 1 (1/83)

Control Room Drawing #BR 4510, Rev. 3 (10/67) l l Drawing fBR 4511, Rev. 7 (6/83)

, Drawing #3E-156-02-001, Rev. 0 (1/84)

Drawing #3E-151-02-008, Rev. 0 (2/84)

Main Steam /Feedwater Tunnel Drawing fBR M550, Rev. 0 (3/73)

Drawing fBR M551, Rev. 0 (3/73)

Drawing fBR M556, Rev. 0 (3/73)

Drawing fBR M557, Rev. 0 (3/73) l Technical Support Center Drawing #S&T 4903, Sheets S-1, A-4, A-5, A-8 (12/83).

Stack RAGEMS Room Drawing #R.C.M-1, Sheets 1 & 2, Rev. O Drawing fBR M0680, Rev. 7 Alternate Hot Chemistry Laboratory Drawing #KL 19400, Sheets 1-3, Rev. O Turbine Building RAGEMS Room Drawing #BR 2052, Rev. 4 i

i Drawing #2054, Rev. 3 i

Drawing #5708, Rev. 0

. C-3 i

I E .

__ _ .u. _ _ _ . - --

, _ . - . - _ - . = . . -. ._

4 .

Appendix C: Input Parmasters and Reference Drawings C-1 Input Parameters for Radiation Source Terms Reactor Power Level 1930 MWe Specific Radioactivity (in Ci per MWt) based upon General Electric calcula-tion SNUMB = 7007S (11/79).

d Release Fractions from Core under various Accident Scenarios based upon J NUREG-0737 guidelines; see Appendix A for details.

Plant Parameters used in evaluating Radiation Transport based upon Pri-aary Containment Design Report (9/67) and FDSAJL Section V-2; see Appendix A for details.

Radioactive Decay Data based upon Table of Isotopes, 7th Edition (1978),

and ORNL-RSIC data file DLC-62.

5 E

k e

I

• 1 4

1

• 1 k

l 1 l

l

A .

C-2 P& ids and Flow Diagrams Core Spray System Drawing #GE 885D781, Rev. 15 (8/83 Containment Spray System Drawing #GE 148F740, Rev. 14,(12/79)

Reactor Shutdown Cooling System Drawing #GE 148F711, Rev.12 (3/84)

Emergency Condenser System Drawing #GE 148F262, Rev. 10 (1/84) 4 I

e f

t t

t t

k

. l a ._ _ . _ _ _ . _ , - _ _ _ _ . . _ . . . _ . - - - , . , . , . .,, , - _ - . . _ . . _ _ _ _ _ _ _ ..,_.____.___,-._.,,__,#.. -- . _ , , _ _ _ .