ML17297A941

Forwards Amend 6 to FSAR

ML17297A941
Person / Time
Site:	Palo Verde
Issue date:	10/16/1981
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17297A945	List: ML17297A941
References
ANPP-19050-JMA, NUDOCS 8110220476
Download: ML17297A941 (131)

Text

REGULAR'ORY lNFORrMATION DISTRIBUT'ION SYSTEM (R IDS)

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$ g wxa~m msmrmmna mwnwzw STA. P.o. SOX 21666 PHOENIX, ARIZONA 65036 October 16, 1981 ANPP-19050-JMA/WFQ Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission, Washington, D.C. 20555

Subject:

Palo Verde Nuclear Generating Station Units 1, 2 and 3 Docket Nos. STN-50-528/529/530 File: 81-056-026

Dear Sir:

Arizona Public Service Company (APS), as Project Manager and 'Operating Agent for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 & 3, is submitting herewith sixty (60) copies of Amendment 6 to the PVNGS Units 1, 2 & 3 Final Safety Analysis Report.

This amendment provides an update to the Final Safety Analysis Report.

Sincerely, C~Jr.

E. E. Van Brunt, ale ~

APS Vice President Nuclear Projects ANPP Project Director EEVB/WFQ/wp Attachments cc: J. Kerrigan P. L. Hourihan A. C. Gehr Sl f OQPPg7g 81 10i6 PDP @DOCK o DR K

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Director of Nuclear Reactor Regulation ANPP-19050-JMA/WFQ Page Two STATE OF ARIZONA )

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COUNTY OF MARICOPA)

I, Edwin E. Van Brunt, Jr., represent that I am Vice President, Nuclear Projects of Arizona Public Service Company, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority to do so, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.

~EL I Edwin E. Van <33runt, Jr.'

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~s Sworn to before me this~&day of 1981 ~'..'

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DOCKET NUMBERS STN-50-528 STN-50-529 STN-50-530 AMENDMENT NO. 6 INTRODUCTION Amendment No. 6 to the Final Safety Analysis Report for Palo Verde Nuclear Generating Station Units 1, 2 and 3 comprises insert pages that address:

1. AC Power open items contained in enclousre to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated June 17, 1981.

2. NRC Branch 222 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated April 16, 1981.

3. NRC Branch 410 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 10, 1981.

NRC Branch 450 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 14, 1981, July 31, 1981, and August 13, 1981.

5. NRC Branch 460 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 14, 1981.

6. NRC Branch 241 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 14, 1981.

DOCKET NUMBERS STN-50-528 STN-50<<529 STN-50-530

7. NRC Branch 210 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 14, 1981.

8. NRC Branch 490 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 14, 1981.

9. NRC Branch 260 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 14, 1981.

10. NRC Branch 480 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 16, 1981.

11. NRC Branch 230 questions contained in enclosure to the NRC letter from Hr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated July 31, 1981.

12. NRC Branch 231 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Ji . dated August 3, 1981.

13. NRC Branch 281 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated August 13, 1981.

14. NRC Branch 121 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated August 13, 1981.

15. NRC Branch 251 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated August, 13, 1981.

DOCKET NUMBERS STN-50-528 STN-50-529 STN-50-530

16. NRC Branch 640 questions contained in enclosure to the NRC letter from Mr. R. L. Tedesco to Mr. E. E. Van Brunt, Jr. dated August 19, 1981.

17. Miscellaneous text, and figure updates and corrections.

A change bar and number "6" identify new and amended text; shading identifies amended areas of illustrations. "Amend-ment 6" and the date "October 1981" identify pages and figures changed by the amendment.

When an insert page in the Amendment has the same number as an existing page in the FSAR, the existing page has been

.uperseded and should be deleted. Where an insert page does not duplicate an existing page, the insert page is new and should be placed in its proper location in the FSAR.

Those holders of the License Application should insert the transmittal letter, included herein, in the License, Application.

azd PVNGS FSAR CHANGE PAGE LIST FOR AIMNDIWNT 6 Remove Insert iii/iv iii/iv ivA/ivB ivA/ivB ivC/ivD ivC/ivD ivE/ivF vii/viii vii/viii xv/xvi xv/xvi zix/xx zix/xx 1.7-71/-72 through 1.7-71/-72 through 1.7-77/-78 1.7-77/-78 1.8-9/-10 1.8-9/-10 1.8-11/-12 1.8-11/-12 1.8-24A/-24B 1.8-24A/-24B 1.8-24C/-24D 1.8-35/<<36 1.8-35/-36 1.8-36A/-36B 1.8-45/-46 1.8-45/-46 1.8-47/-48 1.8-47/<<48 1.8>>49/-49A 1.8-49/-49A 1.8-49B/-50 1.8-49B/-50 1.8-61/-62 1.8-61/-62 1.8-62A/-62B 1.8-66A/-66B 1.8-66A/-66B 2-iii/-iv 2-iii/-iv 2-xi/-xii 2-xi/-xi'.

2.5-61/-62 2.5-61/-61A 2.5-61B/-62 2.5-97/-98 2.8-97/-98 2.5-99/-100 2.5-99/-99A 2.5-99B/-100 2.5-101/-102 2.5-101/-102 2.5-102A/-102B Figure 2.5>>6 Figure 2.5-6 Figure 2,5-8 Figure 2.5-8 October 1981 Amendment 6

PVNGS FSAR CHANGE PAGE LIST FOR AMENDMENT 6 (Cont.)

Remove Insert Figure 2.5-9 Figure 2.5-9 Figure 2.5-11 Figure 2.5-11 2A-i/Blank 2A-i/Blank 2A-1/-2 2A-1/-2 2A-3/Blank ~p/LP f 2A-3/-4 through P8-X7- <> 2A-61/-62 Figure 2A-1 through Figure 2A-4 3-v/-vi 3-v/-vi

-xi/-xii 3-xi/-xii 3.2-5/-6 3.2-5/>>6 3.2-24A/-24B 3.2-24A/-24B 3.2-37/-38 3.2-37/-38 3.5-7/-8 through 3.5-7/-8 through 3.5-17/-18 3.5-17/-18 3.5-19/-20 3.5-19/-19A 3.5-19B/-20 3.5-33/-34 3.5-33/-34 3.5-39/-40 3.5-39/-40 3.6-25/-26 3.5-25/-25A 3.5-25B/-26 3.7-1/-2 3.7-1/<<2 3.7-15/-16 3.7-15/-16 3.7-17/-18 3.7-17/-18 3.7-18A/-18B 3 . 7-18A/-18B 3.7-27/-28 3/7-27/-28 3.7-29/-30 3.7-29/-30 3.7-31/-32 3.7-31/-32 3.8-11/-12 3.8-11/-12 3.8-21/-22 3.8-21/-22 3.8-45/-46 3.8-45/-46 3.8-91/-92 3.8<<91/-92 3A-i/Blank 3A-i/Blank Amendment 6 October 1981

PVNGS FSAR CHANGE PAGE LIST FOR ANEND1KNT 6 (Cont.)

Remove Insert 3A-15/Blank 3A-15/-16 3A-17/-18 through 3A-31/-32 3A-33/Blank 3B-ii'i/iiiA 3B-iii/-iiiA 3B-v/-vA 3B<<v/-vA 3B-xiii/-xiv 3B-xiii/-xiv 3B.15-1/-2 through 3B.15-5/-6 Figure 3B.15-1 Figure 3B.15-2 3C.3-5/-6 3C.3-5/-6 3F Tab 3F Title Page/Blank 3F-1/-2 through 3F-7/-8 3G Tab 3G Title Page/Blank 3G-1/-2 through 3G-25/-26 Figure 3G-1 through Figure 30-5 4A-i/Blank 4A-1/-2 5-i/-ii 5-i/-ii 5-iii/-iv 5-z.z.a./-iv 5-v/Blank 5-v/-vi 5-vii/-viii 5.2-1/-2 5.2-1/-1A 5.2-1B/-2 5.2-11/-12 5.2-11/-12 5.2-12A/-12B through 5.2-12Y/-12Z October 1981 Amendment, 6

PVNGS FSAR CHANGE PAGE LIST FOR AMENDMENT 6 (Cont.)

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.5.3-1/-2 5.3-1/-1A 5.3-1B/-1C 5.3-1D/-2 5.3-3/-4 through 5.3-3/-4 through 5.3-7/-8 5.3-7/-8 5.3-8A/-8B through 5.3-8E/-8F 5.3-9/-10 through 9.3-9/-10 through 5.3-19/-20 5.3-19/-20 5.3-21/-22 through 5.3-25/-26

5. 3-27/Blank Figure 5.3-1 Figure 5.3-1 Figure 5.3-2 Figure 5.3-2 Figure 5.3-3 Figure 5.3-3 Figure 5.3-3A 5.4-1/-2 5.4-1/-1A 5.4-1B/-1C through 5.4-1X/-1Y 5.4-1Z/-1AA 5.4-1AB/-lAC through 5.4-1AJ/-lAK 5.4-1AL/-2 5A-i/Blank 5A-i/Blank SA-1/Blank 5A-1/-2 through 5A-7/-8 Figure SA-1 6-i/-ii 6-i/-ii Amendment 6 October 1981

PVNGS FSAR CHANGE PAGE LIST FOR AI&ND1'KNT 6 (Cont.)

Remove Insert 6-vii/-viii 6-viz/-vases 6.2.1-96A/-96B 6.2.1-96A/-96B through 6.2.1-96I/-96J 6.2.4-5/-6 6.2.4-5/-6 6.2.4-7/-8 6.2.4-7/-8

.6.2.4-15/-16 6.2.4-15/-16 6,2.4-19/-20 6.2.4-19/-20 6.2.4-25/-26 6.2.4-25/-26 Figure 6.2.4-1 (Sheets 1 Figure 6.2.4-1 (Sheets 1 through 6 of 9) through 6 of 9)

Figure 6.2.4-1 (Sheets 8 Figure 6.2.4-1 (Sheets 8 and 9 of 9) and 9 of 9) 6.2.5-7/-8 6.2.5-7/-8 6.2.5-8A/-8B 6.2.5-9/-10 6.2.5-9/-10 6.2.5-13/-14 6.2.5-13/-14 6.2.5-21/-22 6.2.5-21/>>22 6.2.6-1/-2 6.2.6-1/-2 6.2.6-5/-6 through 6.2.6-5/-6 through 6.2.6-9/-10 6.2.6-9/-10 6.2.6-13/-14 through 6.2.6-13/-14 through 6.2.6-21/-22 6.2.6-21/-22 6A-i/Blank 6A-i/Blank 6A-1/-2 6A-1/-2 6A-2A/-2B 6A-7/Blank 6A-7/-8 through 6A-17/Blank 7.5-7/-8 7.5-7/-8 7A-i/Blank 7A-1/-2 through 7A>>19/-20 8.2-5/-6 8.2-5/-6 8.2-7/-8 8.2-7/-8 October 1981 Amendment 6

PVNGS FSAR CHANGE PAGE LIST FOR AMENDMENT 6 (Cont,. )

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Figure 9.1-6 (Sheet 2 of 2) 9.3-21/-22 through 9.3-21/-22 through 9.3-25/-26 9.3-25/-26 9.3-35/-35A 9.3-35/-35A 9.3-35B/-36 9.3-35B/-36 9.3-39/-40 9.3-39/-40 9.3-47/-48 9.3-47/-48 9.3-48A/-48B 9.3-67/-68 9.3-67/-68 Figure 9.3-5 Figure 9.3-5 Figure 9.3-13 (1 of 5) Figure 9.3-13 (1 of 5) 9.4-3/-4 through 9.4-3/-4 through 9.4-7/-8 9.4-7/-8 9.4-15/-16 9.4-15/-16 9.4-19/-20 9.4-19/-20 9.4-21/-22 9.4-21/-22 Amendment 6 October 1981

PVNGS FSAR CHANGE PAGE IIST FOR AMENDMENT 6 (Cont.)

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Figure 9.4-5 Figure 9.4-5 Figure 9.4-9 Figure 9.4-9 9.5-55/-56 9.5-55/-56 9.5-57/-58 9.5-57/-58

, 9.5-58A/-58B 9.5-58A/-58B 9.5-68A/-68B 9.5-68A/-68B 9.5-91/Blank 9.5-91/-92

'I Figure 9.5-1 (1 of 6) Figure 9.5-1 (1 of 6)

Figure 9.5-1 (2 of 6) Figure 9.5-1 (2 of 6)

Figure 9.5-1 (5 of 6) Figure 9.5-1 (5 of 6)

Figure 9.5-1 (6 of 6) Figure 9.5-1 (6 of 6) 9A-i/-ii 9A-x/-zz.

9A-iii/Blank 9A-7/-8 9A-7/-8 9A-8A/-8B 9A-9/-10 9A-9/-10 9A-10A/-10B 9A-13/-14 9A-13/-14 9A-19/-20 9A-19/-20 9A-20A/-20B 9A-21/-22 9A-21/-22 9A-22A/-22B 9A-27/-28 through 9A-27/-28 thr ough 9A-29/Blank 9A-47/Blank 10-113./-1V 10-z.z.z/-zv 10.4-15/-16 10.4-15/-16 10.4-23/-24 10.4-23/-24 10.4-25/-26 10.4-25/-26 10.4-29/-30 10.4-29/-30 10.4-31/-32 10.4-31/-32 10A-i/Blank 10A-i/Blank 10A-13/-14 10A-13/-14 October 1981 Amendment 6

PVNGS FSAR CHANGE PAGE LIST FOR M4ENDNENT 6 (Cont.)

Remove Insert 10A-15/-16 10A-15/-16 through 10A-19/Blank 11.2-9/-10 11.2-9/-10 11.3-7/-8 11.3-7/-8 11.3-21/-22 11.3-21/-22 11.4-23/-24 11.4-23/-24 Figure 11.5-1 Figure 11.5-1, 11A-i/Blank 11A-i/Blank 11A-1/-2 11A-1/-2 11A-3/-4 11A-3/-3A 11A-3B/-4 11A-7/Blank 11A-7/-8 12.5-13/-13A 12.5-13/-13A 12.5-13B/-14 12.5-13B/-14 13-i/-ii 13-i/-ii 13.5-5/-6 13.5-5/-6 13.5-7/-8 13.5-7/-8 14-i/-ii 14-1/-ll 14.2-11/-12 14.2<<11/-12 14.2-24A/-24B 14.2-24A/-24B 14.2-25/-26 14.2-25/-26 14A-i/Blank 14A-i/-ii 14A-1/Blank 14A-1/-2 through 14A-17/Blank 14B-1/-2 14B-1/-2 14B-3/-4 14B-3/-4 14B-9/-10 through 14B-9/-10 through 14B-13/-14 14B-13/-14 14B-17/>>18 14B-17/-18 14B-19/-20 14B-19/-20 14B-20A/-20B 14B-20A/-20B 14B-20C/-20D 14B-20C/-20D

.14B-21/-22 14B-21/-22 Amendment 6 October 1981

PVNGS FSAR CHANGE PAGE LIST FOR AMENDMENT 6 (Cont.)

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PALO VERDE MUCLEAR GENERATING STATION ge4I

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PVNGS FSAR CHAPTER 12 RADIATION PROTECTION CONTENTS Parte 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE

~ALARA 12.1-1 12.1.1 POLICY CONSIDERATION 12.1-1 12.1.1.1 Desi and Construction Policies 12.1-2 12.1.1.2 0 eration Policies 12.1-3 12.1.2 DESIGN CONSIDERATIONS 12.1-5 12.1.2.1 General Desi Considerations for Shieldin and ALARA Ex osures 12.1-5 12.1.2.2 Facilit La out General Desi Considerations for ALARA 12. 1-7A 12.1.2.3 ALARA General E ui ment Desi n Considerations 12.1-17 12.1.2.4 General Desi Considerations to Kee Post-Accident Ex osures ALARA 12.1-22A 12.1.3 OPERATIONAL CONSIDERATIONS 12.1-23 12.1.3.1 General ALARA Techni es 12.1-24 12.1.3.2 S ecific ALARA Considerations for Steam Generator Re air 12.1-27 12.1.3.3 S ecific ALARA Considerations for Reactor Head Removal and Installation 12.1-28 12.1.3.4 S ecific ALARA Considerations for Inservice Ins ections ISI 12.1-28 12.1.3.5 S ecific ALARA Considerations for Other 0 erations Involvin Radiation Ex osure 12.1-28 12.

1.4 REFERENCES

12.1-29 12.2 RADIATION SOURCES 12.2-1 12.2.1 CONTAINED SOURCES 12.2-1 February 1984 12-i Amendment 12

PVNGS FSAR CONTENTS (cont)

Pacae 12.2.1.1 Containment 12.2-1 12.2;1.2 Auxiliar Buildin 12.2-5 12.2-9 12.2.1.4 Turbine Buildin 12.2-10 12.2.1.5 Radwaste Buildin 12.2-10 12.2.1.6 Sources Resultin from Desi n Basis Accidents 12.2-15 12.2.1.7 Stored Radioactivit 12.2-15 12.2.1.8 Field Run Pi e Routin 12.2-16 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES 12.2-16 12.2.3 SOURCES USED IN NUREG 0737 POST-ACCIDENT SHIELDING REVIEW 12.2-21 12.2.2.1 Model for Calculatin Airborne Concentrations 12.2-20 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 FACILITY DESIGN FEATURES 12.3-1 12.3.1.1 Plant Desi n Descri tion for ALARA 12.3-1 12.3.1.2 Radiation Zonin and Access Control 12.3-11 12.3.1.3 Radiation Zones Post-Accident 12.3-11 12.3.2, SHIELDING 12.3-11A 12.3.2.1 Desi n Ob'ectives 12.3-11A 12.3.2.2 General Shieldin Desi n 12.3-13 12.3.2.3 Shieldin Calculational Methods 12.3-21 12.3.3 VENTILATION 12.3-23 12.3.3.1 Desi n Ob'ectives 12.3-23 12.3.3.2 Desi n Criteria 12.3-23 12.3.3.3 Desi n Guidelines 12.3-24 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIV1TY MONITORING INSTRUMENTATAION 12.3-26 Amendment 17 12-ii May 1987

PVNGS F SAR CONTENTS (cont)

Pa<ac 12.

3.5 REFERENCES

12. 3-27 12.4 DOSE ASSESSMENT 12.4-1 12.4. 1 RADIATION EXPOSURES WITHIN THE PLANT 12.4-1 I

12.4.2 RADIATION EXPOSURE OUTSIDE THE PLANT 12.4-2A 12.4.2.1 Construction Worker Doses 12.4-2A 12.4.2.2 Ex osures at the Site Boundar and'n Uncontrolled'reas 12.4-11 12.

4.3 REFERENCES

12.4-11 12.5 RADIATION PROTECTION PROGRAM 12.5-1 12.5.1 ORGANIZATION 12.5-1 12.5.1.1 Pro ram Administration 12.5-1 12.5.1.2 Pro ram Ob'ectives 12.5-2 12.5.1.3 Radiation Protection Pro ram 12.5-2 12.5.2 E UIPMENT, INSTRUMENTATION AND FACILITIES 12.5-3 12.5.2.1 Radiation Protection E i ment 12.5-3 12.5.2.2 Radiation Protection Instrumentation 12.5-5 12.5.2.3 Facilities Related to Radiation Protection 12.5-12 12.5.3 PROCEDURES 12.5-13A 12.5.3.1 Radiation and Contamination Surve s 12.5-13A 12.5.3.2 Procedures and Methods to Maintain Ex osures ALARA 12.5-14 12.5.3.3 Control of Access and Sta Time in Radiation Areas 12.5-15 12.5.3.4 Contamination Control 12.5-16 12.5.3.5 . Airborne Activit Ex osure Control 12.5-17 12.5.3.6 Personnel Monitorin 12.5-18 12.5.3.7 Handlin of Radioactive Material 12.5-19 12.5.3.8 Radiation Protection Trainin 12.5-20 APPENDIX 12A RESPONSES TO NRC RE UESTS FOR INFORMATION APPENDIX 12B PVNGS REVIEWS TO ACHIEVE AN ALARA DESIGN August 1981 12-iii Amendment 5

PVNGS FSAR CHAPTER 12 TABLES Pacae 12.1-1 RADIATION ZONE CLASSIFICATION 12.1-6 12.2-1 MAIN STEAM SUPPLY SYSTEM MAXIMUM RADIOACTIVITY CONCENTRATIONS (pCi/g) 12.2-3 12.2-2 EXPECTED RADIOACTIVE SOURCE TERMS FOR SHIELDING MAXIMUM CASE 12.2-6

12. 2-3 MAXIMUM RADIOACTIVITY INVENTORIES OF SECONDARY PROCESSING SYSTEMS (Ci) 12.2-7 12.2-4 SPENT FUEI POOL COOLING AND CLEANUP SYSTEM MAXIMUM RADIOACTIVITY INVENTORY (pCi) 12.2-11 12.2-5 MAXIMUM RADIOACTIVITY INVENTORIES OF EQUIPMENT IN THE RADWASTE BUILDING (Ci) 12.2-13 12.2-6 ASSUMPTIONS USED IN DETERMINING AIRBORNE RADIOACTIVITY 12.2-17 12.2-7 NORMAL AIRBORNE RADIOACTIVTY CONCENTRATIONS (pCi/cm 3 ) 12. 2-19

12. 2-8 LOCA SOURCE A CONTAINMENT AIRBORNE 12.2-23 12.2-9 LOCA SOURCE B REACTOR COOLANT 12.2-24 12.2-10 LOCA SOURCE C CONTAINMENT SUMP 12.2-25 12.2-11 SYSTEMS USED IN POST-ACCIDENT 12.2-26 SHIELDING REVIEW 12.4-1 DATA FROM OPERATING PWR PLANTS 12.4-3 12.4-2 YEARLY AVERAGES AND GRAND AVERAGE FOR NUMBER OF PERSONNEL AND MAN-REM DOSES FOR OPERATING PWR PLANTS 12.4-5 12.4-3 DISTRIBUTION OF PERSONNEL AND MAN-REM DOSES FOR VARIOUS FUNCTIONS OF OPERATING LIGHT-WATER REACTORS 12.4-6 12.4-3A MAN-REM EXPOSURES TO CONTRACTOR PERSONNEL, 1979 12.4-6A Amendment 17 12-iv May 1987

PVNGS FSAR TABLES (cont)

Pacae 12 . 4-4 ESTIMATED ANNUAL PVNGS MAN-REM DOSES BY JOB CATEGORY 12.4-6B '

5 12.4-5 ESTIMATED ANNUAL GAMMA DOSE TO PVNGS PERSONNEL (3 UNITS) 12.4-7 12.4-6 DESIGN RADIATION ZONE OCCUPANCY BY JOB 1 12 CLASSIFICATIONS 12.4-8 12.4-7 ANNUAL INDIVIDUAL BODY DOSE AT UNIT 3 FROM UNITS 1 AND 2 12.4-10 12.4-8 ESTIMATED LABOR REQUIREMENTS 12.4-10 12.4-9 MAN-REM DOSES TO CONSTRUCTION PERSONNEL 12.4-11 February 1984 12-v Amendment 12

PVNGS FSAR CHAPTER 12 FIGURES 12.2-1 DECAY CURVE SOURCE A 12.2-2 DECAY CURVE SOURCE B 12.2-3 DECAY CURVE SOURCE C 12.3-1 RADIATION ZONES (OPERATION) BETWEEN EL 40'-0" 100'-0" 12.3-2 RADIATION ZONES (OPERATION) AT EL 100'-0" 12.3-3 RADIATION ZONES (OPERATION) BETWEEN EL 120'-0" 140'-0" 12.3-4 RADIATION ZONES (OPERATION) BETWEEN EL 140'-0" 200'-0" 12.3-5 RADIATlON ZONES (OPERATION) AT ROOF ELEVATION 12.3-6 RADIATION ZONES (OPERATION) SECTION A-A 12.3-7 RADIATION ZONES (OPERATION) SECTION J-J 12.3-8 RADIATION ZONES (OPERATION) SECTIONS D, E, H 12.3-9 RADIATION ZONES (OPERATION) SECTIONS B, C, K 12.3-10 RADIATION ZONES (OPERATION) PLAN AT AUX BLDG EL 88'-0", 6 PLAN AT CONTROL BLDG EL 160'-0",

SECTIONS F AND G 12.3-11 RADIATION ZONES (REFUELING) BETWEEN EL 40'-0" 6 100'-0" 12.3-12 RADIATION ZONES (REFUELING) AT EL 100'-0" 12.3-13 RADIATION ZONES (REFUELING) BETWEEN EL 120'-0" 140'-0" 12.3-14 RADIATION ZONES (REFUELING) BETWEEN EL 140'-0" 200'-0" 12.3-15 RADIATION ZONES (REFUELING) AT ROOF ELEVATION 12.3-16 RADIATION ZONES (REFUELING) SECTION A-A 12.3-17 RADIATION ZONES (REFUELING) SECTION J-J 12.3-18 RADIATION ZONES (REFUELING) SECTIONS D. E, H 12.3-19 RADIATION ZONES (REFUELING) SECTIONS B, C, K Amendment 17 12-vi May 1987

PVNGS FSAR FIGURES (cont) 12.3-20 RADIATION ZONES (REFUELING) PLAN AT AUX BLDG.

EL 88'-0", 6 PLAN AT CONTROL BLDG EL 160'-0" SECTIONS F AND G 12.3-21 REACTOR CAVITY SHIEID 12.3-22 FUEL TRANSFER SHIEI DING (IN CONTAINMENT) 12.3-23 FUEL TRANSFER TUBE INSPECTION FACILITY 12.3-24 VERTICAL DOSE RATE FROM 1 SPENT FUEL ELEMENT 12.3-25 RADIATION ZONES LOCA WITH SUMP RECIRCULATION BETWEEN EL. 40'-0" 6 100'-0" 12.3-26 RADIATION ZONES LOCA WITH SUMP RECIRCULATION AT EL. 100'-0" 12.3-27 RADIATION ZONES LOCA WITH SUMP RECIRCULATION BETWEEN EL. 120'-0" 6 140'-0" 12.3-28 RADIATION ZONES LOCA WITH SUMP RECIRCUIATION BETWEEN EI . 140'-0" 6 200'-0" 12.3-29 RADIATION ZONES LOCA WITH SUMP RECIRCULATION AT ROOF ELEVATION 12.3-30 RADIATION ZONES LOCA WITH SUMP RECIRCULATION SECTION 12.3-31 ZONES LOCA WITH SUMP RECIRCULATION A'ADIATION SECTION J-J 12.3-32 RADIATION ZONES LOCA WITH SUMP RECIRCUIATION SECTIONS D, E, H 12.3-33 RADIATION ZONES LOCA WITH SUMP RECIRCULATION SECTIONS B, C, K 12.3-34 RADIATION ZONES LOCA WITH SUMP RECIRCULATION PLAN AT AUX. 88', CONTROL 160', SECTS. F, G 12.3-35 RADIATION ZONES LOCA OUTSIDE AREAS (DIRECT DOSE FROM BUILDINGS ONLY) 12.3-36 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY BETWEEN EL. 40'-0" 6 100'-0" May 1987 12-vii Amendment 17

PVNGS FSAR FIGURES (cont) 12.3-37 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY AT EL. 100'-0" 12.3-38 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY BETWEEN EI . 120'-0" 6 140'-0" 12.3-39 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY BETWEEN EL. 140'-0" 6 200'-0" 12.3-40 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY AT ROOF ELEVATION 12.3-41 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY SECTION A 12.3-42 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY SECTION J-J 12.3-43 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY SECTIONS D, E, H 12.3-44 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY SECTIONS B, C, K 12.3-45 RADIATION ZONES LOCA DEGRADED CORE-INTACT PRIMARY PLAN AT AUX. 88', CONTROL 160',

SECTS. F, G Amendment 17 12-v111 May 1987

PVNGS FSAR

12. RADIATION PROTECTION This chapter describes the radiation protection measures of the station design and the operating policies to ensure that internal I and external radiation exposures to station personnel, contractors, and the general population due to station conditions, including anticipated operational occur-rences, will be within applicable limits, and furthermore, will be as low as is reasonably achievable (ALARA).

Radiation protection measures include: separation of radio-active components into separately shielded cubicles; use of shielding designed to adequately attenuate radiation emanating from pipes and equipment which are sources of significant ionizing radiation; use of remotely operated valves or hand-wheel extensions; ventilation of areas by systems designed to minimize inhalation and submersion doses; installation of permanent radiation monitoring systems; control of access to the site and to restricted areas; training of personnel'n radiation protection; and development and implementation of administrative policies and procedures to maintain exposures ALARA.

12.1 . ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE 'AS LOW AS IS REASONABLY ACHIEVABLE ALARA 12.1.1 POLICY CONSIDERATIONS It is the policy of the management, of Arizona Public Service Company (APS), operating agent for PVNGS, to keep occupational radiation exposure to personnel AIARA. Administrative programs and procedures, in conjunction with facility design, ensure that the occupational radiation exposures to personnel will be kept ALARA.

PVNGS F SAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE. ACHIEVABLE (ALARA) 12.1.1.1 Desi and Construction Policies The ALARA philosophy was applied during the initial design of the plant and implemented via internal design reviews and documentation. These reviews were conducted and documented consistent with the recommendations of Regulatory Guide 8.8.

A description of these reviews and their scope is provided in appendix 12B. In addition, the design was reviewed for ALARA considerations by the APS Senior Health Physicist, a certified health physicist.

The plant design was reviewed, updated, and modified as necessary during the design and construction phases'ngi-neers reviewed the plant design and integrated the layout, shielding, ventilation, and monitoring designs with traffic control, security, access control, maintenance, inservice-inspection, and radiation protection'spects to ensure that the overall design produced a plant which will achieve exposures that are ALARA.

Piping containing radioactive fluids was routed as part of the engineering design effort. This ensured that lines expected to contain significant radiation sources are adequately shielded and properly routed to minimize exposure to personnel..

Onsite inspections, are also conducted, as necessary during construction, to ensure that the shielding and piping layout meets established criteria. During construction, visual inspec-tions were made to ensure that there were no major defects in the shield walls as they were placed. During initial power operations, radiation surveys will be conducted to ensure that the shielding meets design requirements during normal operation and maintenance of the plant.

12. 1-2

PVNGS, E SAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.1.2 0 eration Policies The Station Manual is one of "'the major means of promulgating the operational ALARA policy. This policy is demonstrated in the radiation protection program, the training program, and station procedures.

Besides describing management's commitment to ALARA, the Radiation Protection Division of the Station Manual designates the station personnel who have the responsibility and author-ity to implement ALARA. The Radiation Protection and Chemistry Manager .has the responsibilities of the onsite Radiation Protection Manager, described in Regulatory Guide 8.8 and 4

section 13.1, and is responsible to implement the radiation protection program and to direct, steps necessary to maintain 1

a 4 exposures ALARA. He reports to the Technical Support Manager.

The Radiological Services Manager reports to the Radiation

't Protection and Chemistry Manager and assists in the day-to-day operati'on of the site radiation protection programs. The l

Radiation Protection Supervisors supervi'se the radiation protection personnel who perform radiation protection functions.

The responsibil'ities and authority of the supervisory positions discussed above and" the qualifications of the personnel who fill them are discussed in sections 13.1.2 and 13.1.3.

It is the responsibilityr of the Training Manager and the Radiation Protection and Chemistry Manager to implement radiation protection training for company employees and contractors commensura'te with the requirements of 10CFR19 and Regulatory Guide 8.8. The Radiation Protection and Chemistry Manager verifies that personnel follow the radiation protection procedures designed to ensure that exposures are maintained November, .1986 12. 1-3 Amendment 16

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW =AS IS REASONABLE ACHIEVABLE (ALARA)

ALARA. To ensure compliance with -this policy, the Radiation Protection and Cheinistry Manager and Radiological Services Manager reporting to him are charged with the responsibility to promptly advise higher management of practices which exceed their authority to correct. In addition, periodic reviews (at least annual) of the ALARA program, including review of radiation exposure records and operating procedures, are conducted by offsite Health Physics personnel.

Personnel requiring access to the restricted area and/or radiological controlled areas will receive training as neces-sary to permit access to these areas. These personnel will be tested to evaluate each worker's knowledge, competency, and understanding relative to the training provided. Additionally, retraining and retesting will be conducted annually. Construc-tion personnel will be instructed in site emergency procedures.

For more details see sections 12.5 and 13.2.

I Prior to startup of the first unit, station procedures to be used for work which involves significant pezsonnel,radiation exposure will be reviewed to verify that the procedures adhere to the ALARA philosophy. Revisions to station procedures involving significant personnel radiation exposure will also receive an ALARA review. System or station modifications affecting personnel radiation exposure will also be reviewed to see that the ALARA concept is applied.

The Radiation Protection and Chemistry Manager will peziodi-cally survey station operations to identify situations in which exposures can be reduced.

Amendment 16 12.1-4 November 1986

PVNGS FSAR ENSURING THAT OCCUPATIONAZ RADIATION EXPOSURES ARE AS LOW AS IS REASONABZE ACHIEVABLE (ALARA) 12.1.2 DESIGN CONSIDERATIONS This section discusses the methods and features by which the policy considerations of section 12.1.1 are applied.

Provisions and designs for maintaining personnel exposures ALARA are presented in detail in sections 12.3.1, 12.3.2, and 12.5.3.

12.1.2.1 General Desi n Considerations for Shieldin and ALARA Ex osures General design considerations, shielding, and methods employed to maintain in-plant radiation exposures ALAI&, con-sistent with the recommendations of NRC Regulatory Guide 8.8, Section C.3, have two objectives:

A. Minimizing the necessity for and amount of personnel time spent in radiation areas.

B. Minimizing radiation levels in routinely occupied plant areas in the vicinity of plant equipment expected to require personnel attention.

Plant operating personnel are protected as necessary by shielding wherever a potential rad'iation hazard may exist.

The shielding performs the following additional functions:

A. Assists in maintaining radiation exposure to plant control room personnel within the limits of 10CFR50, Appendix A, Criterion 19 in the unlikely event of an accident.

'I B. Protects certain components from excessive activation or excessive radiation exposure.

C. Facilitates access for maintenance of components.

In order to maintain exposure ALARA, a design radiation zone classification system has been developed and used. Plant areas have been classified in accordance with table 12.1-1.

February 1984 12.1-5 Amendment 12

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS IOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.2.1.1 Neutrons Neutron shielding design considerations are as follows:

A. Shielding is designed to ensure the area does not become a scattering source, producing excessive doses in other regions.

B. Shielding is designed to ensure that neutron activation does not result in doses exceeding the designed shutdown dose rates in the region.

Table 12.1-1 RADIATION ZONE CLASSIFICATION Zone Dose Rate Designation (mrem/h) Allowed Occupancy (Design)

Less than 0.5 Uncontrolled, unlimited access (plant personnel) 0.5 to 2.5 Controlled, limited access, 40 h/wk to unlimited 2.5 to 15 Controlled, limited access, 6 to 40 h/wk 15 to 100 Controlled, limited access, 1 to 6 h/wk Over 100 Normally inaccessible, access only as permitted by radiation protection personnel 1 h/wk Amendment, 12 12.1-6 February 1984

PVNGS F SAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

C. Neutron radiation damage limits of equipment are not exceeded unless provisions are made for periodic replacement.

12.1.2.1.2 Gamma Radiation Gamma radiation shielding design considerations are as follows:

A. Shielding is designed to reduce gamma dose rates throughout. the plant to levels consistent with expected occupancy during normal operation as specified by the design radiation zones.

's B. a minimum, shielding is designed to reduce gamma dose rates from sources external to a radioactive compartment to levels comparable to dose rates resulting from equipment within that compartment.

This design ensures that radiation levels in a compart-ment undergoing maintenance that are due to operating equipment in adjacent compartments will be the lesser of the zone 3 limit (15 mrem/hr) or the operational dose rate of the equipment under repair.

C. Shielding is provided to attenuate radiation from sources external to equipment compartments so that expected maintenance can be performed without exceed-ing exposure limits.

D. Shielding is designed to reduce gamma radiation after reactor shutdown to levels which allow access. for required maintenance operations.

E. Gamma radiation damage limits for the equipment are not, exceeded unless provisions are made for periodic replacement.

August 1981 12.1-7 Amendment 5

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.2.2 Facilit La out General Desi Considerations for ALARA In order to maintain exposures ALARA, the following design guidelines were considered to the extent practical during plant design.

A. Facility general design considerations The layout for personnel access, routing of piping, and location of components is in a manner to minimize personnel radiation exposure during both operation and maintenance. Access control and traffic patterns have been evaluated to assure that radiation exposures are maintained as low as is reasonably achievable. In the interest. of maintaining doses ALARA, access to a given design radiation zone generally will not require passing through a higher design radiation zone.

Amendment 12 12.1-7A February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

This page intentionally blank Amendment 12 12.1-7B February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (AIdQR,)

2. Radioactive components of the same system are grouped together as practical to minimize radioactive p'iping .runs. Ion exchangers and spent resin collection system components are located as close to the radwaste solidification area as practical. Radioactive wastes are

'stored in shielded enclosures separated from normally accessible areas.

3. Due to the desirability of low background radia-tion levels, the counting room has been located away from highly radioactive components.

Due to frequent access requirements, control panels, readout devices, and transmitters, where possible, are located in low radiation zones (design radiation Zone 2 or less) in order to minimize operator expo'sures.

5. In order to reduce radiation exposures due to sampling operations, sample stations are isolated insofar as practical from other radioactive equipment, and exposed sample piping is minimized.

Primary and secondary system samples taken within the containment except for safety injection tanks are piped to the plant laboratory to minimize the need to access the containment.

I Amendment 12 12..1-8 February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

B. Shielding design guidelines

l. Significantly radioactive components such as tanks, filters, ion exchangers, pumps, and heat exchangers are located in shielded compartments.

2. In those process systems whose components contain major sources of radiation, valves and instrumen-tation are separated by shielding from the components.

3. Although use of permanent shielding is preferred, portable or temporary shielding and convenient means for handling it is provided where shielding is required but fixed shielding is impractical.

Access and the capability of the structure to support such portable or temporary shielding has been evaluated during design review.

4. Access to shielded compartments is generally by means of shielded labyrinth arrangements such that direct exposure to radioactive equipment from normal access areas is eliminated. For highly radioactive passive components such as tanks, ion exchangers, and filters, completely enclosed compartments are provided with access via a shielded hatch, or labyrinth entry way with a locked gate.

5. Where space limitations preclude the use of ordinary concrete for shielding, lead. iron, or high density concrete is used instead.

6. Use of removable concrete shielding blocks for frequent personnel access or equipment removal has been avoided by design when practical.

Febru'ary 1985 12. 1- 9 Amendment "14

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

I

7. Redundant radioactive components are located in separate shielded compartments, where practical, to allow for maintenance on one while the other is in service. If radioactive components are located in the same compartments, space is provided for installation of temporary shielding during maintenance.

8. As a general rule, radioactive equipment, is separated by shielding from nonradioactive equipment. to facilitate maintenance on the latter.

9. When corrugated steel decking is used as forming for concrete floor slabs, the minimum thickness of the resulting slab satisfies the specified shielding requirement.

10. Voids such as those created by embedded HVAC ducts or piping are avoided where practicable in shield walls. Unavoidable large voids in shield walls or slabs are evaluated to determine their effect upon shielding requirements.

ll. Shielding requirements for periodic operations such as filter handling are determined on the basis of ALARA considerations including level of radioactivity and frequency and duration of exposure.

C. Guidelines for the control of radioactive contamination Large tanks containing radioactive fluids are enclosed in water tight compartments or are surrounded by curbs. Threshold berms are Amendment 12 12.1-10 February,1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) provided for other radioactive equipment compart-ments to control the spread of contaminated leakage.

2. Sloped floors and floor drains are provided for radioactive equipment compartments including valve galleries and pipe chases. Radioactive drains are designed to eliminate "backgassing."

3. Radioactive equipment drains are piped directly to one of the drainage systems instead of allow-ing contaminated fluid to flow across the floor to the floor drain. The use of temporary flexible tubing to direct fluids to local floor drains or to portable containers during venting and draining operations may be acceptable when'the frequency of operation and volume of fluid involved do not make the use of" direct connections to a collection system cost effective.

4. Suitable coatings are applied to floors and walls susceptible to radioactive contamination in order to facilitate decontamination operations.

D. Valve gallery design guidelines and considerations Valve galleries are provided for valves serving equipment containing or processing highly radioactive material. Valve galleries provide shielding from the process equipment to personnel operating or servicing valves as follows:

The amount of exposed piping in the valve gallery is minimized.

2. When possible, a shielded pipe chase is provided adjacent to the valve gallery for the'purpose of routing radioactive piping. Such pipe chases are located below valve galleries, if practical, in order to minimize crud deposition in valves.

February 1984 12.1-11 Amendment -12

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

3. If the radiation level within a valve gallery is high, valve stem extensions through the valve gallery wall to an adjacent corridor are supplied for frequently operated valves so that they may be operated from a design radiation Zone 2 area.

4. Sufficient space is provided in valve galleries to facilitate maintenance on valves. I E. Radioactive piping design guidelines

1. Radioactive piping is not field routed.

2. Piping is routed so that it does not exceed applicable design radiation zone level.

3. Radioactive piping routed through design radiation Zone 1 or Zone 2 areas is enclosed in a shielded pipe chase, if required.,

Radioactive piping is routed through the highest design radiation zones practical.

5. Potentially radioactive piping is routed behind components or structures which provide shielding to areas where maintenance is likely to be performed.

6. Radioactive pipes are routed close to floors, ceilings and walls where practical, but are kept away from doors and entrances outside containment.

7. When practical, radioactive piping is separated .

from nonradioactive piping.

8. If practical, valves or instrumentation should not be located within radioactive pipe chases.

Amendment 14 12.1-12 February 1985

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

9. Piping layout provides sufficient space for installing and maintaining special equipment or tools as required.

10. Piping insulation is designed for quick removal in any radiation area where routine maintenance and/or inservice inspection is anticipated.

Care is taken in routing radioactive piping to minimize background radiation interference with area radiation monitoring sensors.

F. Inservice inspection guidelines Provisions are made for inservice inspections to minimize potential personnel exposures. By design, high radiation zones permit prompt ingress and egress and have adequate space to accommodate the work force, special tools, laydown space, removal of internals, temporary shielding, and auxiliary ventilation systems as may be required during inservice inspection.

G. Crud trap design guidelines The length of pipe runs is minimized for radio-active piping when possible.

2. Low points and dead legs in radioactive piping are minimized.

3. Placement of valves at piping low points has been minimized.

Valves have flow characteristics that result in a smooth flow path through the valve to minimize the velocity change and resulting potential for crud trapping.

February 1985 12.1-13 Amendment 14

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

5. Internal surfaces of liquid flow paths are smooth and as free as practical from abrupt discon-tinuities in cross-sections. II

6. Lines, are sized to achieve turbulent flow where practical (Re >10,000).

7. Thermal expansion loops are raised instead of depressed, if possible.

8. Flow paths avoid abrupt changes in direction.

Bends of several pipe diameters or long radius elbows are utilized wherever practical.

9. Butt welds, where practical, are used on radio-active process piping whose nominal pipe diameter is greater than 2 inches. Backing rings are not used.

10. The use of screwed fittings in radioactive piping

,systems is avoided, where practical. 'If such fittings are required, a seal weld is applied.

Flanged joints or suitable rapid disconnect fittings are used in cases where maintenance requirements clearly indicate that such construc-tion is preferable (e.g., for pump maintenance).

12. The origin of branch process lines having little or no flow during normal operation are taken off above the horizontal midplane of the main process pipe, where practical.

13. Strainers are located in process streams immedi-ately downstream of ion exchangers to minimize the introduction of resin fines into process systems and to mitigate the consequences of a resin retention screen break.

Amendment 12 12.1-14 February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 0 In addi tion, the following special considerations are given t o piping which processes spent resins:

14. Deleted

15. Ball, plug, or diaphragm valves are used, depend-ing upon the function. in spent resin lines.

Strainer, check. and Y-valves are not utilized in piping systems which process spent resins.

16. Orifices are not utilized in spent resin'iping systems.

17. Butt welds are employed where practical for spent resin piping regardless of size. Large radius elbows, or large diameter bends, are used where practical in routing all such piping.

18. Ninety-degree tees are not used in spent resin piping systems except to introduce clean services such as nitrogen or water, into such lines. Dead legs are avoided and any necessary flushing connections are taken off above the horizontal centerplane of the resin piping.

19. Spent resin lines are sized to achieve turbulent flow to minimize resin deposits and subsequent buildup.

20. Provisions are made for the ion exchangers as well as the resin lines to be pressurized with nitrogen or water to clear plugged lines. The water or nitrogen is introduced at a tee down-stream of each valve, and the leg of the tee is above the resin line to avoid clogging of the clean service inlet line.

February 1985 . 12.1-15 Amendment 14

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

H. Component. isolation, draining, and flushing considerations:

1. Serviceable radioactive components can be isolated and drained.

2. Major components 'of radwaste systems have redun-dant drainage capability so that, the failure of a single drain will not prevent cleaning or purging of such a system.

3. High-point vent and low-point drain connections are provided where practical in piping systems which process highly radioactive fluids. These connections are placed to facilitate flushing of the associated process equipment.

4. Vent and drain connections are equipped with suitable fittings so that flushing and/or drain system piping can be temporarily connected.

5. Isolation valves for vent and drain connections are located as close to the process piping as practical.

Penetration design considerations.:

Penetrations through the primary shield are designed to minimize neutron streaming and the resultant activation of the steam generators and reactor coolant pumps and piping.

2. When possible, radioactive and nonradioactive piping are separated within the penetration area, with provisions for the utilization of temporary shielding for maintenance purposes.

Amendment 12 12.1-16 February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

3. In general, piping, electrical, and HVAC penetra-tions through radiation shields are designed to minimize radiation streaming into accessible areas. The following considerations apply:

a 0 Cross-sectional areas of penetrations are minimized.

b. Where practical, penetrations are oriented so there is an offset between the radiation source and accessible areas.

c ~ If such an offset orientation is impractical, the penetration is located as far above the floor elevation as possible to minimize direct exposure of personnel.

d. If the above methods are not utilized, alternative means are employed. -For piping penetrations, the alternatives include grouting the annular space of the penetra-tion or utilizing shield collars around the piping at the openings of the penetration to reduce radiation streaming. For ventila-tion and electrical penetrations, use of baffle shields to eliminate streaming into accessible areas is acceptable.

12.1.2.3 ALARA General E i ment Desi Considerations In order to maintain exposures ALARA, the following design guidelines were considered to the extent practical in plant design.

February 1984 12.1-17 Amendment 12

PVNGS F SAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.2.3.1 Radioactive Equipment Design Considerations Radioactive equipment design considerations relative to equip-ment contamination and service time include:

A. Reliability, durability, construction, and design features of equipment, components, and materials to reduce or eliminate the need for repair or preventive maintenance.

B. Servicing convenience of anticipated maintenance or potential repair, including ease of disassembly and modularization of components for replacement or removal to a lower radiation area for repair.

C. Provisions, where practical, to remotely or mechanically operate, repair, service, monitor, or inspect equipment (including inservice inspection in accordance with American Society of Mechanical Engineers (ASME) Sec-tion XI).

D. Redundancy of equipment or components to reduce the need for immediate repair when radiation levels may be high.

12.1.2.3.2 Equipment General Design Considerations Equipment general design considerations directed toward minimizing radiation levels in proximity to equipment or components requiring personnel attention include:

A. Provision for draining, flushing, or if necessary, remotely cleaning equipment and piping containing radioactive material.

B. Design of equipment," piping, and valves to minimize the buildup of radioactive material and to facilitate flushing of crud traps. The use of cobalt containing alloys in systems exposed to primary coolant has been minimized except where very hard erosion resistant surfaces are necessary (e.g., stellite seat materials Amendment 12 12.1-18 February 1984,

PVNGS'SAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS ZOW AS IS REASONABZE ACHIEVABLE (AZARA) on valves). C-E supplied components which will be exposed to primary coolant have less than 0.2% cobalt, nominally. Refer also to CESSAR-FSAR Table 5.2-2.

C. Provisions for minimizing the spread of contamination into equipment service areas, including direct drain connections.

D. Provisions for isolating equipment from radioactive process fluids.

E. Provision for a spent fuel pool cleanup system to maintain the radiation level of the fuel pool area within the design radiation Zone 2 limit. See table 12.1-1 for the description of design radiation zones.

F. Heat, exchangers have been provided with corrosion-resistant, tubes with tube-to-tube sheet joints fabri-cated to minimize leakage. Impact baffles are provided and process fluid velocities are limited as necessary to minimize erosive effects. Provisions are made for removal of the tubes for maintenance.

G. Pumps in radioactive systems have been purchased with mechanical seals to reduce seal servicing time.

Additionally, smaller pumps are provided with flanged connections for ease in removal. Pump casings are provided with drain connections for draining the pump for maintenance.

H. Water is used as the fluidizing medium in tanks from which resin is transported. Resin tanks incorporate integral self-cleaning screens in overflow connections to retain resins within the tank. Overflow connections for radioactive tanks are piped to the liquid radwaste system (LRS).

Filters are supplied with the means to either remotely backflush the filter or to perform cartridge replace-ment with remote tools.

February 1984 12.1-19 Amendment 12

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABIE ACHIEVABLE (ALARA)

Demineralizers are designed to remotely remove spent resins hydraulically and replace new resins from a remote location. Resin strainers have been designed for full system pressure are provided with chemical addition connec-drop.'vaporators K.

tions to allow the use of chemicals for descaling operations.

L. The reactor head laydown area has been designed to substantially reduce both the dose to those changing 0-Ring seals and personnel working in adjacent areas.

An interior shield wall separates the 0-Ring changeout from the radiation field under the hemispherical head.

An exterior shield wall isolates the activated flange and external surfaces from adjacent work activities.

M. Frequently operated valves of highly radioactive systems are designed for remote operation. Motor operators, air operators, and reach rods are provided where necessary. The criteria for selecting valve operators are as follows:

Valves located in design radiation Zones 3, 4 and 5 which are operated frequently (for example, on a weekly basis or more often) are equipped with a remote actuator such as a reach rod, an electric motor actuator, or a pneumatic actuator and position indicator. These valves are controlled from the applicable control station or operating aisle.

2. Valves which are operated occasionally (for example, one to twelve times a year) are classi-fied as follows:

a. Valves which are located in design radiation Zone 4 or less may be manually operated directly.

Amendment 12 12.1-20 February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

b. Valves which are located in design radiation Zone 5 have been equipped with a reach rod, or if a reach rod cannot be installed, an electric motor actuator or pneumatic actuator and a position indicator.

3. Valves which are operated infrequently during normal plant operations (for example, less than once a year) are manually operated directly

'unless their operation results in excessive personnel exposure. In cases where the operation of such a valve may lead to exposures in excess of 0.1 man-rem per year, consideration has been

'iven to fitting the valve either with a simple reach rod if feasible, or a remote actuator and position indicator if necessary. The 0.1 man-rem per year was derived from a judgemental evaluation of the potential exposure savings versus the cost to provide and maintain a remote operator throughout plant life.

A mechanical reach rod assembly is used where an electric motor or pneumatic actuator is not neces-sary. Directional changes and rod lengths are kept to a minimum. This allows easier maintenance if needed, and minimizes radiation exposure.

5. Valves which are operated after an accident are provided with a means of remote operation in cases where the operation of such a valve may lead to exposures in excess of 1.25 rem.

February 1984 12.1-21 Amendment 12

PVNGS F SAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

N. Leakage of radioactive material is minimized by use of appropriate valve gaskets and valve packing. For example, one of the following is provided for radio-active valves 2-1/2 inches and larger:

1. Grafoil or graphite yarn packing.

2. Diaphragm or bellows seal valves are used where minimal leakage is required.

Radiation tolerant materials are used in valves in accordance with their radioactive service. Chemical seals are provided on instrument sensing lines for process piping which may contain highly radioactive solids to reduce the servicing time required to keep the lines free from solids.

0. Primary instrument devices, which for functional reasons are located in high radiation areas, have been designed for uncomplicated removal for calibration or servicing. Some instruments, such as thermocouples are provided in duplicate in highly radioactive areas to reduce access and service time.

P. The sample laboratory is equipped with adequate shielding and a fume hood. The sample laboratory and sample stations are equipped with a sink or funnel arrangement so that sample lines may be purged to the LRS or chemical and volume control system prior to sampling. Also, sample lines incorporate the capa-bility of being flushed.

Q. An automated radwaste solidification system is employed to minimize exposure during radwaste processing.

Remotely operated equipment is provided where practical to minimize operator radiation exposure.

Amendment 14 12.1-22 February 1985

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.2.4 General Desi n Considerations to Kee Post-Accident Ex osures ALARA The facility layout will assist in keeping occupational exposures ALARA even after a design basis accident. While exposures will be significantly higher than during normal operation, required access is provided to vital areas and systems without exceeding 5 rem/hr. Zone maps showing expected dose rates in the event of a LOCA with sump recirculation are provided in figures 12.3-25 through 12.3-35. Zone maps for the hypothetical condition of a LOCA with an intact primary but with a degraded core are provided in figures 12.3-36 through 12.3-45. A discussion of the source terms for these events is provided in section 12.2.3. The dose rates projected for these two sets of drawings do not assume decay beyond that corresponding to the onset of recirculation. Even so, virtu-ally unrestricted access will be permitted within the control and diesel generator buildings, as well as portions of the upper floor of the auxiliary building (such as the area of the operational support center).

To provide sampling capability with exposures kept ALARA, PVNGS will incorporate a post-accident sampling system that meets the requirements of NUREG 0737 and Regulatory Guide 1.97, Revision 2 as described in section 9.3.2 and 18.II.B.3. Refer to section 9.3 for a detailed description of the PVNGS nuclear sampling system.

The only other area where post accident access will be required is to the hydrogen monitors/recombiners. Projected dose rates without the recombiners in operation but at the onset of recirculation are expected to be approximately 10 to 30 rem/hr (sump recirculation). As the recombiners have to be May 1987 12.1-22A Amendment 17

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABIE ACHIEVABLE (ALARA) installed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the DBA, the area will not be accessed until dose rates have dropped due to decay to about 1/10 the doses noted above. Thus, the installation dose rate (assuming sump recirculation) will be less than 5 rem/hr.

While the dose rate would be greater than 5 rem/hr for an intact primary-degraded core event, the recombiners would not need to be installed since an intact primary would not be consistent with hydrogen generation .inside the containment.

If hydrogen generation were postulated, this would necessitate a break or opening in the primary. Consequently, sump recirculation would be available with the concomitant release of noble gases and dilution by the refueling water tank.

These consequences would lead to the doses noted above for the sump recirculation mode of cooli.ng (i.e., dose rates less than 5 rem/hr).

ESF grade filtered ventilation is provided for auxiliary building rooms below elevation 100 feet (refer to section 9.4).

This will reduce airborne sources due to recirculation and/or containment leakage. Non-ESF grade filtered ventilation is available for use to reduce airborne sources above elevation 100 feet in the auxiliary building (refer to section 9.4). The use of non-ESF filtration is acceptable since there are no recirculation components above elevation 100 feet. Thus the only significant source of airborne activity is containment leakage. This leakage has already been accounted for in off-site dose analyses which assumed direct containment, leakage to the atmosphere. Secondly, this filter discharges via the plant vent. The plant vent will be monitored in accordance with NUREG 0737 and Regulatory Guide 1.97, Revision 2 to provide notification of decreased filter efficiency.

Therefore, considering direct and airborne sources, access can be provided to those vital areas necessary for control of the plant and personnel exposures will meet GDC 19 and NUREG 0737 limits.

Amendment,17 12.1-22B May 1987

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.3 OPERATIONAL CONSIDERATIONS In accordance with APS policy and consistent with the recom-mendations of Regulatory Guides 8.8 and 8.10, the radiation exposure of plant personnel will be kept ALARA by means of the radiation protection program discussed in section 12.5. The radiation protection policies and practices contained therein are promulgated through the training program discussed in section 13.2, through the Radiation Protection Division of the Palo Verde. Station Manual discussed in section 12.1.1 and section 12.5.

Based upon the experience of other utilities, the design of PVNGS was reviewed so that this information could be incorpo-rated into the design as previously discussed in section 12.1.1.

The criteria and/or conditions under which various operating, maintenance, and inspection procedures are implemented and some. of the techniques that are used to ensure that occupa-tional radiation exposures are AIARA are discussed below.

From the Atomic Industrial Forum (AIF) National Environmental Studies Project report, it has been determined that the majority of exposure at operating PWRs is received during plant outages from maintenance and inspection activities and not from normal operating activities. This is logical since operators can normally stay outside shield walls to read instruments or operate valves and have to enter cubicles con-taining radioactive equipment for short periods of time only to check equipment, whereas maintenance and inspection person-nel usually must go inside cubicles or behind shield walls and must be in close proximity to the lines, valves, instruments, or other pieces of equipment which are radiation sources in order to perform their job. The major sources of exposure, as defined in the AIF report, are steam generator repairs (88% of exposure at one plant, 27% at the average plant), reactor February 1984 12.1-23 Amendment 12

PVNGS F SAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) vessel head removal (6%) and inservice inspection (6%) account-ing for a total of 40% of the annual exposure at the average PWR Since the above sources cause the major exposures, some of the 12 ALARA techniques that should be considered to reduce these exposures are discussed below.

12.1.3.1 General ALARA Techni es A. Permanent shielding is used, where possible, by having workers stay behind walls or in areas of lower radia-tion level when not actively involved in work in radiation areas. On some jobs temporary shielding is used. Temporary shielding will be used only if the total exposure, which includes the exposure received during installation and removal of shielding, is reduced.

B. Systems and major pieces of equipment which are subject to crud buildup have been equipped with connections which can be used for flushing. Prior to performing maintenance work, consideration will be given to the practicality and effectiveness of flushing and/or chemically decontaminating the system or piece of equipment in order to reduce the crud levels and personnel exposure.

12 C. Work conducted in radiological controlled areas will require a radiation exposure permit, (REP). Refer to section 12.5. The purpose of the REP is to carefully prepare for the job so that it can be performed in a proper and safe manner with minimum personnel radiation exposure.

Amendment .12 12.1-24 February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

D. On complex jobs involving exceptionally high radiation levels, "dry runs" may be made, and in some cases mockups may be used to familiarize the workers with the exact operations they must perform at the jobsite.

This job preplanning will include estimates of the man-rem needed to complete the job. At the completion of the job, a debriefing session will be held with the people who actually performed the work in an effort to determine if the work could have been completed more

'efficiently, resulting in less exposure. This informa-tion, together with the procedures used and actual man-rem expended, will be recorded. The radiation, contamination, and airborne activity levels determined during the work also will be recorded. In addition, if any external body contamination or internal contami-nation was encountered during the job, this information will also be recorded. This information will be used to provide guidance at the preplanning stage of future similar operations. These techniques will assist in improving worker efficiency and thus will minimize the amount of time spent in the radiation field.

E. As much work as practical will be performed outside radiation areas. This includes items such as reading instruction manuals or maintenance procedures, adjust-ing tools or jigs, repairing valve internals, and prefabricating components.

F. For long-tenn repair jobs, consideration will be given to setting up communications systems, such as sound-powered telephones or closed circuit television, so that supervising personnel can check on the progress of work from a lower radiation area.

February 1984 12.1-,25 Amendment 12

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABIE (ALARA)

G. On some jobs, special tools or jigs may be used when their use would permit the job to be performed more efficiently or would prevent errors, thus reducing the time in the radiation field. Special tools may also be used if their use would increase the distance from the source to the worker, thus reducing the exposure

.rate. Unless special tools or jigs are necessary to accomplish the job, special tools or jigs will be used only if the total exposure, which includes that received during installation and removal, is reduced.

H. Entry and exit points will be set up in areas so that personnel are exposed to as low a level of radiation as practical. This will be done because personnel may spend a significant amount of time changing protective clothing and respiratory equipment in these entry-exit areas. These entry and exit points are set up to limit the spread of contamination from the work area.

Protective clothing and respiratory equipment will .be selected on the basis of worker protection and habitability.

Plastic glove boxes, which can be taped around valves or other fixed components,"and plastic bags are used where practical so that personnel can work on equipment without being exposed to the contamination produced during the work, and to limit the spread of contamination.

K. Radiation levels in work areas will be posted at the entrance to the area and/or in the work area so that the areas of highest and lowest radiation level are clearly identifiable. Individuals will be instructed to stay in the lowest radiation area consistent with performing their assigned jobs.

Amendment 12 12.1-26 February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA)

L. Personnel will wear selfreading dosimeters for work in high radiation areas so that they can determine their accumulated exposure at any time during the job. This is in addition to their monthly thermoluminescent dosimeter (TLD), legal TLD, and job TLD.

M. On jobs with exceptionally high radiation levels, a timekeeper, who knows the exposure rate of the radia-tion field, will keep track of the exposure using a timing device. This technique will ensure that personnel are not staying in a high radiation area longer than intended. The timekeeper will remain in the lowest exposure area consistent with performing this task.

12.1.3.2 S ecific ALARA Considerations for Steam Generator

~Re aar The techniques in section 12.1.3.1, except for listing J, are normally used when inspecting and plugging steam generator tubes.

After access to the steam generator primary heads is obtained, covers are installed, if practical, over the hot and cold leg nozzle openings with a layer of material to prevent tools and debris from entering the pipes. If personnel are required to work inside the primary head for a significant. amount of time, consideration is given to adding temporary shielding. Normally, however, this is not worthwhile since at least half the expo-sure is due to shine from crud inside the steam generator tubes which cannot be easily shielded because the tube sheet area must normally be kept clear for inspection or tube plugging.

February 1984 12.1-27 Amendment 12

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ~

ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.3.3 S ecific ALARA Considerations for Reactor Head Removal and Installation Techniques described in section 12.1.3.1, except for listings B, J and M, are normally used when removing and'nstalling the reactor head. Quick disconnect electrical cables and reactor head ventilation ducts which can be quickly removed are used to reduce time spent in radiation areas. Temporary shielding can be installed around the outer control rod drive mechanisms to reduce exposure if crud collects in the control rod drive housing. A multiple stud tensioner is normally used to reduce radiation exposures by expediting removal and installation of the reactor head.

12.1.3.4 S ecific AIARA Considerations for Inservice Ins ectzons ISI Techniques in section 12.1.3.1 are normally used when perform-ing ISI. Remote testing devices will be used in the conduct of the examinations, where applicable. Written and possibly photographic or videotape records will be made of preservice inspection operations that have potential for future signifi-cant radiation exposure to personnel. By training the examiners and alerting them to the specific problems that they can expect to encounter, less time will be spent in radiation areas.

12.1.3.5 S ecific ALARA Considerations for Other 0 erations Involvxn Radar.ation Ex osure Other operations such as refueling, radwaste handling, spent fuel handling, loading and shipping, routine maintenance, sampling, and calibration are discussed in sec'tion 12.5.3.

Amendment 12 12.1-28 February 1984

PVNGS FSAR ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.

1.4 REFERENCES

National Environmental Studies Project, Compilation

/

and Analysis of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants, SAI Services, September 1974.

12.1-29

II k

t f

4

PVNGS FSAR 12.2 RADIATION SOURCES This section discusses and identifies the sources of radiation that form the basis for shield design calculations and the.

sources of- airborne radioactivity used for the design of personnel protecti-:e measures and for dose assessment.

12.2.1 CONTAINED SOURCES The shielding design source terms are based on full-power operation with 1% fuel cladding defects. Sources in the pri-mary coolant include fission products released from fuel clad defects, and activation and corrosion products. The sources in the reactor coolant corresponding to 1% defects are dis-cussed in CESSAR Section 12.2. Throughout most of the reactor coolant system, activation products, principally nitrogen-16 (N-16), are the primary radiation sources for shielding.

design. For all systems transporting radioactive materials, conservative allowance is made for transit decay, while at the same time providing for daughter product formation.

The design sources are presented in this section by Location of the equipment discussed building'ocation and system.

in this section is shown in the general arrangement. drawings of section 1.2.

12.2.1.1 Contain:i>ent 12.2.1.1.1 Reactor Core Refer to CESSAR Section 12.2.1.1.1.

12.2.1.1.2 Reactor Coolant System-Refer to CESSAR Section 12.2.1.1.2.

12.2-1

PVNGS FSAR RADIATION SOURCES 12.2.1.1.3 Main Steam Supply System Radioactivity in the main steam supply system is based'on a steam generator tube leakage rate of 1 gal/min concurrent with 1% failed fuel. These sources are listed in table 12.2-1.

These sources were developed assuming that the condensate polishing demineralizers are not in use. It is assumed that the blowdown rate to the blowdown demineralizers is 1.0% of the main steam rate.

12.2.1.1.4 Spent Fuel Handling and Transfer Refer to CESSAR Section 12.2.1.1.4.

12.2.1.1.5 Processing Systems 12.2.1.1.5.1 Chemical and Volume Control S stem CVCS Radiation sources in the CVCS consist of radionuclides carried in the reactor coolant. Nitrogen-16 (N-16) is the predominant radiation source in the reactor coolant system. The design of the CVCS ensures that most of the N-16 has decayed before the letdown stream leaves the containment by placing a delay mechanism in the letdown flow to obtain an additional 73 sec-onds decay of N-16. All CVCS system heat exchangers other than the regenerative heat exchanger are located in the auxiliary building..

The shielding design is based on the maximum activity in each component.

A. General Design Bases Refer to CESSAR Section 12.2.1.1.5.1 for all CVCS component design bases other than the regenerative heat exchanger.

B. Specific Component Bases

1. Regenerative Heat Exchanger Total tube volume (letdown) 2.0 x 10 cm 5

Total shell volume (charging) 1.59 x 10 cm

12. 2-2

PVNGS F SAR RADIATION SOURCES Table 12.2-1 MAIN STEAM SUPPLY SYSTEM MAXIMUM RADIOACTIVITY CONCENTRATIONS (a) (pCi/g) -(Sheet 1'of- 2)

Primary Activity SG Liquid SG Steam Hotwell Radio-, Concen- Concen- Concen- Concen-nuclide tration tration tration tration Kr-85m 1.40(+00) 5.40(-05) 5.40(-05) 1. 31(-05)

Kr-85 2.60(-02) 1.02(-06) 1.02(-06) 2.49(-07)

Kr-87 8.60(-01) 3.22(-05) 3.22(-05) 7.57(-06)

Kr-88 2.40(+00) 9;19(-05) 9.19(-05) 2.21(-05)

Xe-131m 8.40(-02) 3.28(-06) 3.28(-06) 8.06(-07)

Xe-133 2.70(+01) 1.05(-03) 1.05(-03) 2.59(-04)'.02(-05)

Xe-135 5.30(+00) 2.06(-04) 2.06(-04)

Xe-138 6.10(-01) 1.96(-05) 1.96(-05) 3.92(-06)

Br-84 3.40(-02) 1.88(-05) 1.88(-07) 1.70(-07)

I-129 3.90(-08) 7.87(-11) 7.87(-13) 7.75(-13),

I-131 2.90(+00) 5. 81 (-03) 5.81(-05) 5.72(-05)

I-132 8.10(-01) 1.01(-03) 1.01(-05) 9.77(-06)

I-133 4.30(+00) 8.13(-03) 8.13(-05) 7.99(-05)

I-134 5.70(-01) 4.42(-04) 4.42(-06) 4.13(-06)

I-135 2.40(+00) 4.00(-03) 4.00(-05) 3.90(-05)

Rb-88 2.40(+00) 8.69(-04) 8.69(-07) 8.00(-07)

Rb-89 7.30(-02) 2.36(-05) 2.36('-08) 2.12(-08)

Cs-134 7.60(-02) 1.53(-04) 1.53(-07) 1.74(-07')

Cs-136 9.60(-02) 1.93(-04) 1.93(-07) 2.19(-07)

Cs-137 -3.00(-01) 6.05(-04) 6.05(-07) 6.87(-07)

Cs-138 1.10( ~00) 6.32(-04) 6. 32 (-07) 6.33(-07)

N-16 1.50( i02) 4.04(-04) 4.04(-04) 3.75(-06)

H-3 6. 60 (-01) 1.30(-01) 1.30(-01) 1.30(-01)

Y-90 6.20(-04) 1.21(-06) 1. 21(-09) 1.22(-09)

Y-91 2.10(-02) 4.19(-05) 4. 19 (-08) 4.29(-08)

Mo-99 2.20(+00) 4. 31(-03) 4. 31(-06) 4. 32(-06)

Sr-89 4.00(-03) 7.98(-06) 7.98(-09) 8. 01(-09)

Sr-90 2.00(-04) 4.00(-07) 4.00(-10) 4. 01(-10)

Sr-91 3.80(-03) 6.66(-06) 6.66(-09) 6.64(-09)

Zr-95 5.90(-03) 1.18(-05) 1.18(-08) 1.18(-08)

Ru-103 6.60(-03) 1.32(-05) 1.32(-08) 1.32(-08)

Ru-106 1.60(-03) 3.20(-06) 3.20(-09) 3.21(-09)

Te-129 1.20(-02) 1.12(-05) 1.12(-08) 1.07(-08)

Te-132 3.00(-01) 5.89(-04) 5.89(-07) 5-.91(-07)

Te-134 3.90(-02) 2.68(-05) 2.68(-08) 2.51(-08) a ~ 1 gal/min primary to secondary leakage 12.2-3

PVNGS FSAR RADIATION SOURCES Table 12.2-1 MAIN STEAM SUPPLY SYSTEM MAXIMUM RADIOACTIVITY CONCENTRATIONS (vCi/g) (Sheet 2 of 2)

Primary Activity SG Liquid SG Steam Hotwel 1 Radio- Concen- Concen- Concen- Concen-nuc1 ide tration tration tration tration Ba-140 6.80(-03) 1.35(-05) 1.35(-08) 1.36(-08)

La-140 6.40(-03) 1.24(-05) 1.24(-08) 1.24(-08)

Ce-144 3.70(-03) 7.39(-06) 7.39(-09) 7.41(-09)

Pr-,143 5.40(-03) 1.07(-05) 1.07(-08) 1.08(-08)

Cr-51 1.90(-03) 3.79(-06) 3.79(-09) 3.80(-09)

Mn-54 3.10(-04) 6.19(-07) 6.19(-10) 6. 21(-10)

Fe-55 1.60(-03) 3.20(-06) 3.20(-09) 3.21(-09)

Fe-59 1.00(-03) 2.00(-06) 2.00(-09) 2.00(-09)

Co-58 1.60(-02) 3.19(-05) 3.19(-08) 3.20(-08)

Co-60 2.00(-03) 4.00(-06) 4.00(-09) 4.01(-09) 12.2-4

PVNGS FSAR RADIATION SOURCES Letdown has RCS specific activities Charging has volume control tank (VCT) specific activities Shielding source inventory is in table 12.2-2.

I 12.2.1.1.5.2 Steam Generator Blowdown S stem.

Radiation sources in the steam generator blowdown system shown in table 12.2-2, are based on the primary-to-secondary leakage rate and failed fuel rate stated in section 12.2.1.1.3. How-ever, the liquid is assumed to be processed through the steam generator blowdown system at the maximum rate of 1% of main steam rate. High-conductivity solutions resulting from regeneration of blowdown demineralizers will normally be pro-cessed in the chemical waste neutralizing tanks (see sec-tion 10.4), in the absence of primary-to-secondary leakage.

Radioactive highconductivity solutions will be processed by the LRS (see section 11.2).

12.2.1.1.5.3 Condensate Polishin S stem. .For shielding and dose assessment purposes, the condensate polishing system does not yield substantive doses since radioactivity will be present in the demineralizers only in the event of coincident steam generator tube leaks and condenser tube leaks. Space for shielding has, however, been reserved should dual tube leaks occur. Demineralizer inventory for dual leakage is shown in table 12.2-3.

12.2.1.2 Auxiliar Buildin 12.2.1.2.1 Shutdown Cooling System Refer to CESSAR Section 12.2.1.2.1.

12.2-5

PVNGS FSAR I

RADIATION SOURCES i

Table 12 2-2 EXPECTED RADIOACTIVE SOURCE TERMS FOR SHIELDING MAXIMUM CASE Regenerative Regenerative Heat Heat Nuclide Exchanger Nuclide Exchanger Br-84 Kr-85m 7 '4 3.10

(-4)

(-2)

'I-133

.1-134 9.69 (-2) 1.28 (-2)

Kr-85 6.00 (-4) Cs-134 5.47'-3)

Kr-87 1.95 (-2) I-135 5.36 (-2)

Kr-88 5.49 (-2) Cs-136 6.16 (-3)

Rb-88 5.34 (-2) Cs-137 2.18 (-2)

Sr-89 1.66 (-3) Xe-131m 1.92 (-3)

Sr-90 9.00 (-5) Xe-133 6.12 (-1)

Y-90 4.46 (-6) Xe-135m Y-91 4.78 (-4) Xe-135 1.20 (-1)

Y-91M Xe-138 1. 38 (-2)

Sr-91 8.63 (-5) Ba-140 1.54 (-4)

Zr-95 1.00 (-6) La-140 1.45 (-4)

Mo-99 5.05 (-2) Ce-144 8.43 (-5)

RQ-103 1.51 (-4) Pr-143 1.24 (-4)

RQ-106 3.55 (-5) Co-60 2. 10 (-5)

Te-129 2.75 (-4) Fe-59 1.20 (-6)

I-131 6.70 (-2) Co-58 1.90 (-4)

I-132 1.85 (-2) Mn-54 2.10 (-6)

Te-132 6.82 (-3) Cr-51 1.00 (-4) a ~ Based on 1% failed fuel RCS concentration.

b. ( ) = Denotes power of ten.

c ~ = Considered insignificant.

d. All source terms in curies.

12.;2-,6

PVNGS FSAR RADIATION Table 12.2-3INVENTORIES'OURCES MAXIMUM RADIOACTIVITY OF SECONDARY PROCESSING SYSTEMS (Ci) (Sheet '1 -of 2)

Blowdown Blowdown Polishing Radionuclide Flash. Tank Demineralizer Demineralizer Kr-83m 0.0 0.0 0.0 Kr-85m 1.2 (-03) 0.0 0.0 Kr-85 2.3 (-'05) 0.0 0.0 Kr-87 7.3 (-04) 0.0 0.0 Kr-88 2.1 (-03) 0.0 0.0 Kr-89 0.0 0.0 0.0 Xe-131m 7.5 (-05) 0.0 0.0 Xe-133m 0.0 0.0 0.0 Xe-133 2.4 (-02) 0.0 0.0 Xe-135m 0.0 0.0'.0 0.0 Xe-135 4 ~ 7 (-03) 0.0 Xe-137 0.0 0.0 0.0 Xe-138 4~5 (-04) 0.0 0.0 Br-83 0.0 0.0 0.0 Br-84 4.3 (-03) 8.7 (-04) 1.1 (-04)

Br-85 0.0 0.0 0.0 I-129 1.8 (-09) 3.3 (-05) 3.2 (-06)

I-130 0.0 0.0 0.0 I-131 1.3 (-01) 9.8 (+01) 9.4 I-132 2.3 (-02) 2.0 (-01) 2.2 (-02)

I-133 1.8 (-01) 1.5 (+01) 1.5 I-134 1.0 (-02) 3 4 (-02) 1.8 (-04)

I-135 9.1 (-02) 2.3 2.4 (-01)

Rb-86 0.0 0.0 0.0 Rb-88 2.0 (-02) 2.1 (-02) 1.8 (-04)

Rb-89 5.4 (-04) 5.1 (-04) 4.1 (-06)

Cs-134 3.5 (-03) 5.4 (+01) 5.4 (-01)

Cs-136 4 4 (-03)

~ 5.1 5.1 (-02)

Cs-137 1.4 (-02) 2.4 (+02) 2.4 Cs-138 1.4 (-02) 2.8 (-02) 2.6 (-04)

H-3 3.0 0.0 0.0 Y %90 2.7 (-05) 7.1 (-03) 1.0 (-04)

Y-91m 0.0 0.0 0.0 Y-91 9". 5 (-04) 5.2 7.4 (-02)

Y-93 0.0 0.0 0.0 12.2-7

PVNGS FSA'R RADIATION SOURCES Table 12.2.-3 MAXIMUM'RADIOACTIVITY'NVENTORIES OF SECONDARY PROCESSING SYSTEMS (Ci) (Sheet 2 2)

Condensate Blowdown Blowdown Polishing Radionuclide Flash Tank Demineralizer Demineralizer Mo-99 9.8 (-02) 2.6 (+01) 3.7 (-01)

Sr-89 1.8 (-04) 8.7 (-01) 1.2 (-02)

Sr-90 9.1 (-06) 1.8 (-01) 2.5 (-03)

Sr-91 1.5 (-04) 5.8 (-03) 8.2 (-05)

Zr-91 2.7 (-04) 1.6 2.2 (-02)

Nb-95 0.0 0.0 0.0 Tc-99m 0.0 0.0 0.0 Ru-103 '3.0 (-04) 1.1 1.6 (-02)

Ru-106 7.3 (-05) 1.1 1.5 (-02)

Rh-103m 0.0 0.0 0.0 Rh-106 0.0 0.0 0.0 Te-125m 0.0 0.0 0.0 Te-127m 0.0 0.0 0.0 Te-127 0.0 0.0 0.0 Te-129m 0.0 0.0 0.0 Te-129 2.5 (-04) 1.2 (-03) 1.7 (-05)

Te-131m 0.0 0.0 0.0 Te-131 0.0 0.0 0.0 Te-132 1.4 (-02) 4.2 5.9 (-02)

Te-134 6.1 (-04) 1.7 (-03) 2.4 (-05)

Ba-137m 0.0 0.0 0.0 Ba-140 3.1 (-04) 3.8 .(-01) 5.4 (-03)

La-140 2.8 (-04) 4.6 (-02) 6.4 (-04)

Ce-141 0.0 0.0 0.0 Ce-143 0:. 0 0.0 0.0 Ce-144 1.7 (-04) 2.4 3.3 (-02)

Pr-143 2.4 (-04) 3.2 (-01) 4.6 (-03)

Pr-144 0.0 0.0 0.0 Np-239 0.0 0.0 0.0 Cr-51 8.6 (-05) 2.3 (-01) 3.3 (-03)

Mn-54 1.4 (-05) 2.0 (-01) 2 9 (-03)

Fe-55 7.3 (-05) 1.3 1.8 (-02)

Fe-59 4 5 (-05) 1:9 (-01) 2.7 (-03)

Co-58 7.3 (-04) 4.7 6.6 (-02)

Co-60 9.1 '(-05 ) 1.7 2.4 (-02) 12.2.8

PVNGS FSAR RADIATION SOURCES 12.2.1.2.2 Nuclear Cooling Water System The nuclear cooling water system is normally nonradioactive or of very low level activity due to inleakage. The process radiation monitor (section 11.5) for this system has a sensi-tivity of 1 x 10 -6 pCi/cm 3 of Cs 137 uniformly distributed throughout the nuclear cooling water system. For shielding and dose assessment purposes, the nuclear cooling water system does not yield substantive doses.

~1' d Spent Fuel Storage and Transfer

'2.2.1.3.1 The predominant, radioactivity sources in the spent fuel storage and transfer areas in the fuel building are the spent fuel assemblies. Spent. fuel assembly sources are discussed in CESSAR Section 12.2.1.1.4. For shielding design, the spent fuel pool is assumed to contain the design maximum number of fuel assemblies. Of these, 241 spent fuel assemblies are assumed to be from unloading the full core with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> decay, and 81 assemblies are assumed to be from previous refueling operations with 90 days decay.

12.2.1.3.2 Spent Fuel Pool Cooling and Cleanup System Sources in the spent fuel pool cooling and cleanup (SFPCC) system are a result of transfer of radioactive isotopes from the reactor coolant into the spent fuel pool during refueling operations. The reactor coolant activities for fission, cor-rosion, and activation products are decayed for the amount of time required to remove the reactor vessel head following shutdown, are reduced by operation of the CVCS purification ion exchangers, and are diluted by the total volumes of the water in the reactor vessel, refueling pool, and spent fuel 12.2-9

PVNGS F SAR'ADZATrON SOVRCES pool. This activity then undergoes subsequent decay and accumulation on the SFPCC filters and ion exchangers as dis-cussed in section 11.1.7. These sources are listed in .

table 12.2-4.

12.2.1.4 Turbine Buildin 12.2.1.4.1 Main Steam Supply and Power Conversion Systems

,Potential radioactivity in the main steam supply and power con-version systems is a result of steam generator tube leaks and fuel cladding defects as discussed in section"12.2.1.1.3. This P

radioactivity is sufficiently low so that no radiation'hielding for equipment in secondary systems, other than portions of the steam generator blowdown system (section 12.2.1.1.5), is required in order to meet the design radiation zone requirements.

12.2.1.5 Radwaste Buildin 12.2.1.5.1 Liquid and Solid Radwaste Systems F

Radioactive inputs to the radwaste system sources include fission and activation product radionuclides produced in the core and reactor coolant. The components of the radwaste systems contain varying degrees of activity.

The concentrations of'adionuclides present in the process fluids at various locations in the radwaste systems, such as pipes, tanks, filters, ion exchangers, and evaporators, are discussed in section 11.1. Shielding for each component of the radwaste systems is based on maximum activity conditions as shown in table 12.2-5. Pumps are modeled using the appro-priate geometries and process point, activities.

Amendment 12 12.2-10 February 1984

'PVNGS FSAR RADIATION SOURCES Table 12.2-4 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM MAXIMUM RADIOACTIVITY INVENTORY (vCi) (Sheet 1 of'2)

Radionuclide Ion Exchanger Filter Kr-83m 0.0 0.0 Kr-85m" 0.0 0.0 Kr-85 0.0 0.0 Kr-87 0.0 0.0 Kr-88 0.0 0.0 Kr-89 0.0 0.0 Xe-131m 0.0" 0.0 Xe-133m 0.0 0.'

Xe-133 0.0 0.0 Xe-135m 0.0 0.0 Xe-135 0.0 0.0 Xe-137 0.0 0.0 Xe-138 0.0 0.0 Br-83 0.0 '0. 0 Br-84 0.0 0.0 Br-85 0.0 0.0 I-129 1.3 0.0 I-130 0.0 0.0 I-131 9.1 0;0 I-132 0.0 0.0 I-133 4.9 (+02) 0.0 I-134 0.0 0.0 I-135 1.5 (-08) 0.0 Rb-86 0.03'+06) 0.0 Rb-88 0.0 0.0 Rb-89 0.0 0.0 Cs-134 2.4 (+06) 0.0 Cs-136 4' (+05) 0.0 Cs-137 6.6 (+06) 0.0 Cs-138 0.0 0.0 N-16 0.0 0.0 H-3 0.0 0.0 Y-90 1.0 (+01) 0.0 Y-91m, 0.0 0.0 Y-91 7. (+03) 0.0 12.2-11

PVNGS FSAR RADIATION SOURCES Table. 12 ..2-.4 S PENT FUEL POOL COOLING AND CLEANUP SYSTEM MAXIMUM RADIOACTIVITY'NVENTORY (pCi.)'Sheet 2 of 2)

Radionuclide Ion Exchanger Fi 1te'r Y-93 0.0 0.0 Mo-99 1.5 (+05) 0.0 Sr-89 4.7 (+04) 0.0'.0 Sr-90 4.2 (+03)

Sr-91 1.8 (-06) 0.0 Zr-95 9.6 (+03) 0.0 Nb-95 0.0 0.0 Tc-99m 0.0 0;0 Ru-103 5.0 (+03) 0.0 RQ-106 .4.6 (+03) 0.0 Rh-103m 0.0 0.

Rh-106 0.0 0'.0 Te-125m 0.0 0.0 Te-127m 0.0 0.0 Te-127 0.0 0.0 Te-129m 0.0 0.0 Te-129 0.0 0.0 TQ-131m 0.0 0.0 Te-131 0.0 0.0 Te-132 9.5 (+04) 0.0 Te-'134 0.0 0.0 Ba-137m 0.0 0.0 Ba-140 2.4 (+04) 0.0 La-140 5.3 (+01) 0.0 Ce-141 0.0 0.0 Ce-143 0.0 0.0 Ce-144 1.0 (+04) 0.0 Pr-143 3.3 (+03) 0.0 Pr-144 0.0 0.0 Np-239 0.0 0;0 Cr-51 1.0 (+04) 1.0 (+04)

Mn-54 2.8 (+03) 2.8 (+03)

Fe-55 1.5 (+04) '1. 5 (+04)

Fe-59 6.4 (+03) 6.4, (+03)

Co-58 (+05) 1.1 (+05)

Co-,6 0 2.0 (+04) 2.0 (+04) 12.2-12

Table l2.2-5 MAXIMUM RADIOACTIVITY INVENTORIES OF EQUIPMENT IN THE RADWASTE BUILDING (Ci) (Sheet 1 of 2)

Concentrate High hctiv Low hctiv- LRS Mixed Radionuc1ide High TDS Low TDS Monitor Recycle Monitor Tank LRS ity Spent i.ty Spent Bed Ion LRS hdsorp-Holdup Tank Holdup Tank Tank Evaporator Resin Tank Resin Tank Exchanger tion Bed Kr-85m 1.2 (-1) 4.9 (-6) 0.0 4,6 (-7) 0.0 0.0 0.0 0.0 0.0 Kr-85 1.2 (-1) 9.9 (-7) 0.0 5.6 (-2) 0.0 0.0 0.0 0~0 0.0 Kr-87 1.8 (-2) 8.3 (-7) 0~0 2.2 (-8) 0.0 0.0 0.0 0.0 0.0 Kr-88 1 2 (-1) 5.3 (-6) 0 ' 3.1 (-7) 0.0 0.0 0 0 0.0 0.0 Xe-131la 5.3 (-1) 2 9 (-6) 0.0 5.9 (-2) 0 0 0 0 0.0 0~0 0.0 Xe-133 3.8 (+1) 8.5 (-4) 0~0 1.9 0.0 0.0 0+0 0.0 0.0 Xe-135 5.9 (-1) 3.8 (-5) 0.0 7.1 (-6) 0.0 0.0 0 0 0.0 0.0 Xe-138 2 3 (-3) 9 5 (-8) 0.0 4' (-10) 0.0 0.0 0.0 0.0 0.0 Br-84 2.9 l-4) 9.3 (-7) 0.0 Zoo (-11) 0.0 8 2 l-01) 0.0 1.0 (-8) 1.0 (-8)

Z-129 1 2 ("5) 3,"2 (-6) 8 ' (-5) 2~5 (-9) 5.2 (-5) 8 8 (-03) 0.0 3.2 (-4) 3 2 (-4)

Z-131 3.0 (+1) 8 ' 1 4 (+1) 6.0 (-3) 5.5 (+1) 2 9 (+04) 2 (+Ol) 3 ' l+1) 3.3 (+1)

II-133 132 2.6 (-2) 9.3 (-4) 0~0 4.4 (-8) 0.0 7 1 (+01) 0~0 4.4 l-5) 4.4 l-5) 1.8 5.5 (-1) 8.2 (-5) 2 ' (-4) 3.0 (-3) 3.6 (+03) 0.0 2 4 (-1) 2.4 (-1)

I 134 7 8 (-3) 5 ' (-5) 0.0 1~1 (-9) 0.0 2 3 (+01) 0.0 1 1 (-6) 1.1 (-6)

I-135 2.9 (-1) 3.0 (-2) 1' (-13) 4.2 (-6) 2.0 (-11) 7 6 (+02) 0.0 4 1 (-3) 4.1 (-3)

Rb 88 1.2 (-2) 1.3 (-5) 0~0 3.9 (-9) 0.0 2.4 (+Ol) 0+0 3.9 (-8) 3.9 (-8)

Rb-89 6.1 (-4) 2 6 (-7) 0.0 7.0 (-11) 0.0 1 2 0 0 6.9 (-10) 6.9 (-10)

Cs-134 2.0 (+1) 5.2 1.4 (+2) 2.1 (-1) 8.4 (+1) 1.3 (+04) 2.3 (+02) 2' l+2) 2.3 (-2)

Cs 136 1.5 5 (-1) 1.6 1.7 (-2) 3.9 5.3 (+02) 1.0 1 5 1.5 Cs-137 8 7 (+1) 2 3 (+1) 6.3 (+2) 9. 2 (-1) 3.8 ,(+2) 3.9 (+04) 1.2 (+03) 1.2 (+3) 1.2 (+3)

Cs 138 7.8 3 0 (-5) 0 ' 1.7 (-8) 0.0 1.4 (+01) 0.0 1~7 ("7) 1.7 (-7)

H-3 2 2 (-3)'"4) 1.4 l-1) 3~6 l-1) 5.7 ( 1) 2.2 (-1) 0.0 0+0 0.0 0.0 Y-90 9.1 4.8 (-4) 2.6 (-5) 2.9 (-7) 3.0 (-41 1.2 (-01) 0.0 6.3 (-4) 6.3 (-4)

Y 91 1.7 5.0 (-1) 7.4 F 9 (-4) 6.8 3.8 (+01) 1.3 (+01) 1~4 (+1) 1~4 (+1)

Mo-99 3.8 1.8 1.2 1 1 (-3) 1~4 1.7 (+03) 2.0 F 4 2.4 Sr-89 3.0 (-1) 8 2 (-2) 1.2 6.4 (-5) 1 1 2~2 (+02) 2eO 2oo 2.0 S r-90 6.3 (-2) 1.7 (-2) 4.6 (-1) 1~3 (-5) 2 7 (-1) 3.0 (+01) 1.0 1.7 1 7 Sr-91 1.1 (-3) 1.1 (-4) 2.0 ( 12) 2.2 (-8) 1.6 (-10) 2 8 0.0 2.2 (-5) 2.2 (-5)

Sr-95 Nb 95 5.4 ( 1) 1 2 (-5) 1.5 (-1) 0.0 2.4 3 ' l-5) 1.2 1.0

(-4)

(-12) 2 1 2 (-5) 5.0 (+01) 0.0 0.0 0.0 4.6 4.4 (-9) 4.6 4.4 (-9) aM Ru 103 3.7 (-1) 1.1 (-1) 1.2 8.3 (-5) 1 2.2 (+01) 2.0 2.1 2.1 Ru-106 3.9 (-1) 1.1 (-1) 2.6 8 ' (-5) 1.6 3.0 (+01) 8.0 8.1 8.1 H Te-129 Te-132 1.9 8.4

(-4) 1)

2.8 3.0

(-6)

(-1) 0.0 4.6 (-2) 6 8 1.9

( 11)

(-4) 0.0 5.2 7.4

(-01)

(+02) 0.0 0.0 6.7 (-8) 4.9 (-1) 6.7 (-8) 4.9 (-1) 0 3.2 (-4) (-6) 0.0 4

3.5 (-8)

R Te-134 2.4 3.5 (-11) 0.0 9.5 (-011 0.0 3.5 (-8)

Ba-140 La-140 1 2 4 9

(-1)

(-3) 3.4 (-2) 2.6 (-3) 1 2 (-1) 1.9 (-5) 2.5 1.3

(-5)

(-6) 3.0 3.6

(-1)

(-4) 7 4 3.4 i+01) 0.0 0.0 2.1 2.1

(-1)

(-3) 2 1 ( 1) 2 1 (-3) 0 Ce-144 8 3 ( 1) 2 3.( 1) 5 ' 1.8 (-4) 3.5 7.0 (+01) 1.6 (+01) 1.6 (+1) 1.6 (+1)

Pr-143 9.0 (-2) 2 9 l-2) 1.0 2.1 (-5) 2.4 (-1) 1~0 (+01) 0.0 1.9 (-1) 1.9 l-1) O Cr-51 7 7 (-2) 2.1 l-2) 2' 1.6 (-5) 2.6 (-1) 5.5 0.0 2.9 (-1) 2.9 (-1)

Hn-54 7 2 (-2) 1.9 (-2) 4,7 1.5 (-5) 3.1 4.8 1.0 1 4 1.4 Pe-55 4.6 (-1) 1.2 l-1) 3 2 '(-5) '.8 2.0 3.0 (+01) 1.1 (+01) 1.1 l+1) 1.1 (+1)

Pe-59 6.6 l-2) 1~8 (-2) 2.4 1.4 (-5) 2.5 4.6 0.0 4.0 (-1) 4.0 (-1)

Co-58 1.6 4 5 (-1) 7.6 3.5 (-4) 6.5 1.1 (+02) 1.4 (F01) 1.5 (+1) 1.5 l+1)

Co-60 6.1 (-1) 1.6 (-1) 4.3 1.3 (-4) 2.6 4.0 (+01) 1.5 (+01) 1.5 (+1) 1.5 l+1)

Table 12.2-S MAXIMUM RADIOACTIVITY INVENTORIES OF EQUIPMENT IN THE RADWASTE BUILDING (Ci) (Sheet 2 of 2)

Waste Gas Waste Gas Boric Acid Waste Gas Waste Gas Boric Acid Radionuclide Surge Tank Decay Tank Condensate IX Radionuclide Surge Tank Decay Tank Condensate IX Kr-83m 1.7 {+1) 1.1 0.0 Mo-99 1.2 (-3) 2 ' (-3) 2.3 (-6)

Kr-85m 1;1 (+2) 1.4 (+1) 3 ' (-8) Sr-89 5.2 ("6) 6.7 (-5) 4.8 ("8)

Kr-85 1.7 (+2) 3.3 (+3) 2.9 (-8) Sr-90 1.5 (-7) 3.3 (-6) 1.7 (-9)

Kr-87 4.5 (+1) 2.2 9.2 ("9) Sr"91 9.0 ("6) 2.6 (-6) 4.9 (-9)

Kr-88 1.8 (+2) 1.6 (+1) 5.7 (-8) zr-95 9.0 (-7) 1.3 (-5) 8.7 ("9)

Kr-89 3.8 (-1) 7.1 (-3) 0.0 Nb-95 7.3 8.3 (-6) 0.0 Xe-131m 1.2 (+2) 7.3 <+2) 1.3 (-7) Tc-99m 1.2 (-4) 0.0 Xe-133m 2.5 (+2) 3.2 (+2) 0.0 Ru-103 6.6 (-7) 7.8 (-6) 5.7 (-9)

Xe-133 2.0 (+4) 6.2 (+4) 9.9 (-6) Ru-106 1.5 (-7) 3.0 (-6) 2.3 (-9)

Xe-135m 4.0 9.3 (-2) 0.0 Rh-103m 3.8 (-7) 1.7 (-8) 0.0 Xe-135 3.9 {+2) 9.3 (+1) 2.6 (-7) Rh-106 1.7 (-9) 3.5 (-ll) 0 '

Xe-137 8.7 (-1) 1.6 (-2) 0.0 Te-125m 4.2 (-7) 5.7 (-6) 0.0 Xe-138 1.2 (+1) 2.8 (-11 1.2 (-9) Te-127m 4.2 (-6) 6.7 (-5) 0.0 Br-83 5.4 (-5) 4 ' <-6) 0.0 Te-127 -.1.2 (-5) 3' (-6) 0.0 Br-84 1.6 (-5) ~ "5;7 (-7) 6;7 (-10) Te-129m 2.1 (-5) 2.3 (-4) 0.0 Br-85 2.9 (-7) 6.1 (-9) 0.0 Te<<129 1.5 (-5) 7.8 (-'7) 8.9 (-10)

I"129 0.0 0.0 1.9 (-9) Te-131m 3.7 (-5) 3.2 (-5)

I-130 2.9

'-5) 1.1 (-5) 0.0 Te-131 6.0 (-6) 1.9 (-7) 0.0 I-131 4.1 (-3) 2' (-2) 3.8 (-3) Te-132 4.1 (-4) 9.1 (-4) 5.6 (-5)

I-132 1.1 (-3) 9.6 (-5) 2.4 (-'7) Te-134 0.0 0.0 1.0 (-9) 0.0'-133 5.4 (-3) 3.3 ("3) 8.9 (-5) Ba-137m 1 3 (-5)

~ 2.8 (.-7) 0.0 I-134 3.8 (-4) 1.7 (-5) 3.0 ("8) Ba-140 3.3 (-6) 2.3 ("5) 4.8 <-8)

I-135 2.5 (-3) 5' (-4) 7.1 (-6) I a-140 2.2 (-6) 2.5 (-6) 4.9 (-9)

Rb-86 4.2 (-7) 4.7 (-6) 0.0 Ce-141 1.0 (-6) 1.1 (-5) 0.0 Rb-88 3.0 (-4) 1.1 (-5) 2.9 (-8) Ce-143 5' (-7) 5.5 (-7) 0.0 Rb 89 0.0 0.0 1.5 (-9) Ce-144 4.9 (-'7) 9.4 (-6) 5.5 (-9)

Cs-134 1.2 (-4) 3.4 (-3) 4.5 (-7) Pr-143 7.3 (-7) 5.3 (-6) 6.2 (-9) M Cs-136 6.6 (-5) 6.1 (-4) 1.9 (-7) Pr-144 1.4 (-'7) 4.0 (-9) 0.0 Cs-137 9.0 (-5) 2.6 (-3) 1.2 ("6) Np-239 1.7 (-5) 2.8 (-5) 0.0 Cs-138 0.0 0.0 1.9 (-8) Cr-51 3.2 (-4) 2.9 (-3) 1.9 (-9) M 0

H-3 0.0 0.0 0.0 Mn-54 5.2 (-5) 8.9 (-4) 1.9 (-9)

Y-90 1.8 (-8) 9.5 (-1) 1.6 (-10) Fe-55 2. 7'(-4) 5.0 (-3) 3 ~ 7 (-10) CO Y-91M 2.8 (-6) 1.2 (-7) 0.0 Fe-59 1.7 ("4) 1.8 (-3) 1.1 (-9) O ~

Y-91 9.4 (-7) 1.3 (-5) 7.0 (-9) Co-58 2.7 (-3) 3.4 ("2) 1.8 (-8) C Y-93 4.5 (-7) 1.4 (-7) 0.0 Co-60 3.3 (-4) 6.3 (-3) 2.4 (-9)

A M a ~

PVNGS FSAR RADIATION SOURCES 12.2.1.5.2 Gaseous Radwaste System Radiation sources for each component of the waste gas system are based on operation with the maximum activity conditions as given in sections 11.1 and 11.3. Tabulation of the maximum activities is shown in table 12.2-5.

12.2.1.6 Sources Resultin from Desi Basis Accidents The radiation sources from design basis accidents include the design basis inventory of radioactive isotopes in the reactor coolant, plus postulated fission product releases from the fuel. Accident parameters and sources are discussed and evaluated. in chapter 15.

12.2.1.7 Stored Radioactivit The principal sources of activity not enclosed by plant structures are the refueling water tank (RWT), the holdup tank (HT), the reactor makeup water tank (RMWT), and the con-densate storage tank (CST). The CST is expected to contain concentrations of radionuclides that yield a surface dose rate of 0.25 mrem/h or less. Radionuclide inventories of the RWT, RMWT, and HT are described in CESSAR Section 12.2.1.1.5.1.

No other radioactive fluids are stored outside the plant structures.

Spent fuel is stored in the spent fuel pool until it is placed in the spent fuel shipping cask for transport offsite.

Storage space is allocated in the radwaste building for stor-1 age of spent filter cartridges and solidified spent resins, evaporator bottoms, and chemical wastes. Radioactive wastes stored inside plant structures are .shielded so that there is design radiation Zone I access outside the structure. If radiation levels outside the structure exceed the design radiation zone limit, or it becomes'ecessary to temporarily store radioactive wastes outside plant structures, radiation protection measures are taken by the radiation protection

.staff to assure compliance with 10CFR20 and to be consistent with the recommendations of NRC Regulatory Guide 8.8.

February 1984 12.2-15 Amendment 12

PVNGS FSAR RADIATION SOURCES 12.2.1.8 Field Run Pi e Routin The procedures for routing of radioactive piping are discussed in section 12.1.2.2. Radioactive piping was routed as part of the engineering design effort. Radioactive piping was not field routed.

12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES This section deals with the models, parameters, and sources reguired to evaluate airborne concentrations of radionuclides during plant, operations in various plant radiation areas where personnel occupancy is expected.

Leakage sources are dependent upon the concentrations of radionuclides in the primary system, secondary system, spent fuel pool, and the refueling pool. The assumptions and parameters reguired to evaluate the isotopic airborne concen-trations in the various applicable regions are listed in table 12.2-6. The chemical and volume control system (CVCS) and the spent fuel pool cooling and cleanup system (SFPCCS) were designed to purify reactor coolant through ion exchangers after reactor shutdown and cooldown. This ensures that. the effect, of activity spikes will not significantly contribute to the containment airborne activity during refueling opera-tions. The. contribution to airborne activity due to reactor vessel head removal is considered negligible as the reactor vessel head vent is connected to the gaseous radwaste system.

The detailed listing of the expected airborne isotopic concen-trations in the applicable regions is presented in table 12.2-7.

The final design of the plant ensures that the expected airborne isotopic concentrations in the applicable regions are well below the max'imum permissible concentration for the critical organ for the appropriate isotope for occupational workers, as adjusted on the basis of expected weekly occupancy in the regions.

12.2-16

PVNGS FSAR RADIATION SOURCES Table 12.2-6 ASSUMPTIONS USED IN DETERMINING AIRBORNE RADIOACTIVITY (Sheet 1 of 2)

Item Value Reference Leakage NUREG-0017 Primary to secondary, lb/d 100 Auxiliary and radwaste bldg (with respect to primary coolant), lb/d 160 Turbine building, lb/h 1,700 Charging pump room, gal/h 1 CESSAR Table 11.1.6-1 Auxiliary bldg IX valve gallery, gal/h 0.15 Radwaste bldg conc tank valve gallery, gal/h 0.05 Radwaste bldg LRS pumps valve gallery, gal/h 0.22 Iodine partition factors Steam generators 0.01 CESSAR Table 11.1.8-1 All buildings except containment 0.0075 NUREG-0017 Containment (fraction of RCS iodine released to building atmosphere per day) 0.0000 NUREG-0017 Bldg(area vent flowrates, ft /min Containment Refueling purge exhaust (high volume) 33,000 Normal purge exhaust (low volume) 2,200 Fuel building exhaust 43,500 Auxiliary building exhaust 60,000 February 1984 12.2-17 Amendment. 12

PVNGS FSAR RADIATION SOURCES Table 12.2-6 ASSUMPTIONS USED IN DETERMINING AIRBORNE RADIOACTIVITY (Sheet 2 of 2)

Item Value Reference Bldg area vent flowrates, ft /min (cont)

Radwaste building exhaust 51,000 Turbine building exhaust 385,000 Charging pump room 1,100 Auxiliary bldg IX valve gallery 400 Radwaste bldg conc tank valve gallery 250 Radwaste bldg LRS pumps valve gallery 500 Building/area free volumes, ft3 6

Containment 2.6 x 10 Fuel building 5 7.5 x 10 Turbine building 6.9 x 10 6 Auxiliary building 1.2 x 10 Charging pump room 8.5 x 1033 IX valve gallery 4.0 x 10 5

Radwaste building 4.6 x 10 Concentrate tanks valve gallery 1.5 x 1033 LRS pump valve gallery 4.4 x 10 Amendment 12 12.2-18 February 1984

Table 12.2-7 NORMAL AIRBORNE RADIOACTIVITY CONCENTRATIONS 3

()ICi/Cm )

Containment Bldg Auxiliary Building Radwaste Building Turbine Bldg HPC Air Conc Tk LRS Pumps (vCi/cm3) Power Fuel Charging ZX Valve Valve Valve Operating Radionuclide (40 h/wk) Access Refueling Bldg Corridor Pump Ro Gallery Corridor Gallery Gallery Deck Kr-83m (-6) 7.6 (-8) 5.3 (-10) 4. 0 (-8) 3.4 (-9) 5.3 {-10) 1. 2 (-14)

Kr 85m (-6) 8.0 (-7) 3.1 (-9) 2.2 (-7) 1.9 (-8) 3.1 (-9) 6.8 (-14)

Kr-85 (-5) . 4. 6 (-6) 4.5 (-9) 3. 0 (-7) 2.6 (-8) 4.5 (-9) 3. 0 (-9) 9.9 (-14)

Kr 87 (-6) 1.5 '(-7) 1.4 (-9) 1. 1 (-7) 9.6 (-9) 1.4 (-9) 3.4 (-14)

Kr-88 (-6) 1. 0 (-6) 5.3 (-9) 3. 9 (-7) 3.4 (-8) 5.3 ("9) 1. 1 (-15) 1.2 (-13)

Kr-89 (-6) 5.5 (-)0) 2.1 (-11) 3.7 (-9) 2.7 (-10) 2.1 (-11) 1. 4 (-15)

Xe-131m (-5) 3.4 (-6) 3' (-9) 2. 2 (-7) 1.9 (-8) 3.2 (-9) 8. 7 (-10) 5.3 (-14)

Xe-133m (-5) 5.5 (-6) 6.4 (-9) 4. 4 (-7) 3.7 (-8) 6.4 ("9) 4.5 (-11) 1.4 (-13)

Xe-133 (-5) 5.0 (-4) 5.3 (-'7) 3.6 (-5) 3.1 (-6) 5.3 (-7) 4. 5 (-8) 1.2 (-11)

Xe-135m t-6) 7.1 (-9) 1.7 (-10) 1. 9 (-8) 1.6 (-9) 1.7 (-10) 5.7 (-15)

Xe-135 (-6) 4.2 (-6) 1.0 (-8) 7. 0 (-7) 6.0 (-8) 1.0 ("8) 6. 2 (-15) 2. 3 (-13)

Xe-137 (-6) 1.2 (-9) 4.3 {-11) 7. 6 (-9) 5.6 (-10) 4.3 (-11) 2.7, {-15)

Xe-138 (-6) 2.2 ("8) 5.5 (-10) 6.4 (-8) 5.1 (-9) 5.5 (-10) 1.9 (-14)

Br-83 (-9) 2.4 (-11) 4.5 (-12) 7.0 (-12) 6.0 (-12) 4.5 (-12) 1.5 (-15)

Br-84 (-6) 2.9 (-12) 8.1 (-13) 3.4 (-12) 2.8 (-12) 8.1 (-13)

Br-85 (-6) 3.1 (-14) 9 ' (-15) 1.6 (-13) 1.1 (-131 9.7 (-15)

Z-130 (-9) 5.0 (-11) 3.6 (-12) 3.2 (-12) F 7 (-12) 3.6 (-12) 1.7 (-15)

Z-131 (-9) 4.6 (-8) 5.9 (-10) 4.1 (-10) 3.5 (-10) 5.9 (-10) 8.0 (-14) 1.0 (-11) 3. 2 (-13)

I-132 (-7) 4.6 ("10) 9.2 (-11) 1.5 (-10) 1.2 (-10) 9.2 (-11) 3. 0 (-14)

Z-133 ("8) 1.5 (-8) 7.3 (-10) 5. 8 (-10) 4.9 ("10) 7.3 (-10) 1.1 ( 14) 3. 6 (-13)

Z-134 (-7) 8.8 (-11) 2.2 (-11) 6. 5 (-10) 5.4 (-11) 2.2 (-11) , 6. 5 (-15)

I-135 (-7) 2.6 (-9) 2.8 (-10) 2. 8 (-10) 2.4 (-10) 2.8 (-10) 1.2 (-13)

Co-60 ("9) 1.4 (-9) 3.8 (-11) 4. 0 (-14) 3.5 (-14) 3.8 (-11) 2. 6 (-13) 8. 1 (-15)

Co-58 (-8) 3.2 (-9) 8.4 (-11) 3.2 (-13) 2' (-13) 8.4 (-11) 4. 1 (-13) 3. 6 (-14 )

Fe-59 (-8) 3.2 (-10) 8.4 (-12) 2.0 (-14) 1.7 (-14) 8.4 {-12) 2.7 (-14) 1.8 (-3.5)

Hn-54 (-8) 9.2 (-10) 2.5 (-11)

H 6 3 ( 15) 5.4 (-15) 2.5 (-11) 2. 3 (-12) 1.1 (-15)

Cs-137 (-8) 1.6 (-9) 4.2 (-11) 1.8 (-)2) 3.2 (-13) 4.2 (-11) 2.7 (-12) 7.0 (-,14)

Cs-134 (-8) 9.2 (-10)

(-ll) 2.5 (-11) 2. 5 (-12) 4.4 (-13) 2.5 (-11) l. 4 (-15) 9.2 (-14) H Sr-90 Sr-89

(-9)

{-8) 1.3 7.1 (-11) 2.8 1.8

(-13)

{-12) 2.0 (-16) 1.7 (-16) 7.1 (-15) 6.1 (-15) 2.8 (-13) 5. 3 (-)5) 0a 1.8 (-12) 8. 6 (-10)

H-3 (-6) 6.6 (-8) 2. 5 (-6) 1. 1 (-6) 5.9 (-9) 1.2 (-7) 2 ' (-8) 5.9 1. 1 (-14) 1. 2 (-9) 7. 3 (-10) l>>

C-14 (-6) 4.2 (-8) 0 Ar-41 (-6) 1.1 (-6) g Total 5.2 2. 5 (-6) 1 1 (-6) 6.2 (-7) 3.9 (-5) 3.3 ("6) 5.7 8. 7 (-10) 5. 0 (-8) 2.9 (-9)

PVNGS FSAR RADIATION SOURCES 12.2.2.1 Model for Calculatin Airborne Concentrations Plant areas with airborne radioactivity are characterized by a constant leakrate of a radioactive source at a constant source strength with a constant exhaust rate of the contami-nant. This leads to a peak or equilibrium airborne concentra-tion of the radioisotope in the regions as calculated by the following equation:

-AT.t Tz where:

(L R) ~

3.

= leak or evaporation rate of the ,th radioisotope i

in g/s, in the applicable region, and A

i

~ activity concentration of i the .th leaking or evaporating radioisotope in pCi/g partition factor or the fraction of the leaking activity that, is airborne for the i radioisotope total removal rate constant for the

-1 i radioiso-tope in s from the applicable region (Xdi + Xe)

(Xdi and Xe are the removal rate constants in s -1 due to radioactive decay and the exhaust from the applicable region respectively for the i radioisotope) time interval between the start of the leak and the time at which the concentration is evaluated in seconds 12.2-20

PVNGS FSAR RADIATION SOURCES free volume of the region in which the leak occurs in cm 3 C (t) 1 airborne concentration of the i.th radioisotope at time t in yCi/cm 3 in the applicable region From the above equation, it is evident that the peak or equi-librium concentration, CEEq,i. of the i radioisotope in the applicable region will be given by the following expression:

C Eq.i.

= (L R).1 A.1 (P F).

1

/ (V X Tl.) (2)

With high exhaust rates, this peak concentration will be reached within a few hours.

12.2.3 SOURCES USED IN NUREG 0737 POST-ACCIDENT SHIELDING REVIEW The post-accident shielding review described in sec-tion 12.1.2.4 used initial core releases equivalent to those recommended in Regulatory Guides 1.4 and 1.7, and Standard Review Plan 15.6.5, and considered two LOCA events. 'The first was a LOCA with recirculation accomplished via the containment sump. The second was a LOCA with an intact primary with recirculation accomplished via the shutdown cooling system.

The following core releases were used in the review:

A. Source A: Containment airborne: 100% noble gases, 25% iodines see table 12.2-8.

B. Source B: Reactor coolant: 100% noble gases, 50% halogens, 1% solids see table 12.2-9.

C. Source C: Containment sump: 50% halogens, 1% solids see table 12.2-10.

Volumes used for each source were:

A. Source A: Containment free volume of 2.6 x 10 6 ft3 B. Source B: Reactor coolant system (RCS) volume of 3 3 9.15 x 10 May 1987 12.2-21 Amendment 17

PVNQS FSAR RADIATION SOURCES C. Source C: The minimum volume of water, 7.76 x 10 4 ft 3 present at the time of recirculation. (RCS

+ refueling water tank + safety injection tanks.)

A LOCA with sump recirculation is represented by sources A and C. An intact primary-degraded core LOCA is represented by sources A and B (source A was not reduced even though there is no mechanism to assume noble gases in both sources). The systems assumed to be operating for each event are shown in table 12.2-11. The results of the shield review are presented in section 12.3.1.3.

Decay curves, normalized to initial time equals zero, were developed for the sources as an aid in developing post-accident access plans. These curves are presented as fig-ures 12.2-1 (source A), 12.2-2 (source B), and 12.2-3 (source C).

Amendment 17 12.2-22 May 1987

PVNGS. FSAR RADIATION SOURCES TABLE 12.2-10 LOCA SOURCE C - CONTAINMENT SUMP (Curies)

Nuclide Activity Nuclide Activity Nuclide Activity Se-84 2.55(+5) Sn-129 1.08(+5) Cs-140 2.00(+6)

Br-84 1.28(+7') Sb-129 2.08(+5) Ba-140 2.11(+6)

As-85 5. 52 (+4) Te-129m 3.68(+4) La-140 2.11(+6)

Se-85 3.02(+5) Te-129 2.03(+5) Cs-143 4.52(+5)

Br-85 1.61(+7) I-129 8.05(-1) Ba-143 1.47(+6)

Se-87 3.29(+5) Sn-131- 3.50(+5) La-143 1.81(+6)

Br-87 2.87(+7) Sb>>131 9.30(+5) Ce-143 1.83(+6)

Br-88 3.87(+7) Te-131m 1.72(+5) Pr-143 1.84(+6)

Rb-88 8.61(+5) Te-131 1.00(+6) Cs-144 1.25(+5)

Br-89 3.96(+7) I-131 5.65(+7) Ba-144 8.75(+5)

Rb-89 1.12(+6) Sn-132 3.08(+5) La-144 1.61(+6)

Sr-89 1.13(+6) Sb-132 8.24(+5) Ce-144 1.25(+6)

Br-90 3.89(+7) Te-132 1.15(+6) Pr-144 1.25(+6)

Rb-90 1.31(+6) I-132 5.80(+7)

Sr-90 6.60(+4) Sn-133 5.37(+4)

Y-90 6.57(+4) Sb-133 1.10(+6)

Rb-91 1.45(+6) Te 133m 1.42(+6)

Sr-91 1.49(+6) Te-133 1.01(+6)

Y-91m 8.80(+5) I-133 1.16(+8)

Y-91 1.49(+6) Cs-134 1.28(+4)

Sr-95 1.64(+6) Sb-134 5.69(+5)

Y>>95 2.02(+6) Te-134 2.14(+6)

Zr-95 1.99(+6) I-134 1.27(+8)

Nb-95 1.95(+6) Sb-135 1.80(+5)

Zr-99 2.61(+6) Te-135 1.09(+6)

Nb-99 2.61(+6) I-135 1.10(+8)

Mo-99 2.69(+6) Cs-135 2.50(-1)

Tc-99m 3.23(+5) Cs-136 1.32(+4)

Mo-103 1.46(+6) I-137 7.20(+7)

Tc-103 1.53(+6) CS-137 4.84(+4)

Ru-103 1.53(+6) Ba-137m 4.52(+4)

Tc-106 6.72(+5) I-138 4.39(+7)

Ru-106 5.27(+5) Cs-138 2.07(+6)

a. Numbers in parenthesis denote powers of ten August 1981 12.2-25 Amendment 5

Table 12.2-11 SYSTEMS USED IN POST ACCIDENT SHIELDING REVIEW(a)(b)

Source LOCA Degraded Core Type LOCA with Sump Recirculation Intact Primary

~ Containment Air .~ Containment Air

~ Hydrogen Control System ~ Hydrogen Control System B ~ Safety Injection System

~ Containment Spray System

~ Shutdown Cooling System

~ Post-Accident Sampling System

~ Letdown System (c)

~ Safety Injection System

~ Containment Spray System

~ Shutdown Cooling System

~ Post-Accident Sampling System U M

~ Letdown System (c)

M 0

2l

a.

Where redundant systems exist, both are assumed in use. 0

b. Radwaste systems not used post-accident. C

c. Portions up to purification filter inlet. A PJ

(MeV/sec) Vs time,'(days)

(Me V/sec) t=p

.01 <~ t ~< 2.3 10 10-1 10-2 MeV/sec (MeV/sec) t=p Palo Verde Nuclear Generating Station

lirzzr, FSAR DECAY CURVE SOURCE A (Sheet 1 of 2)

Figure 12.2-1 August 1981 Amendment 5

'3 l

'c V

t

(MeV/sec) Vs time (days)

(Me V/sec) t 2.3 <~ t <

~70 10 10~ 10'0 Me V/sec (MeV/sec) t~o Palo Verde Nuclear Generating Station FSAR DECAX CURVE - SOURCE A (Sheet 2 of 2)

Figure 12.2-1 August 1981 Amendment 5

\

(MeV/sec) Vs time (days)

(MeV/sec) t 0.01 <~ t <~ 180.

10-'eV/sec 10-2 10

'MeV/sec) t=o Palo Verde Nuclear Generating Station FSAR DECAY CURVE SOURCE B Figure 12.2-2 August 1981 Amendment 5

(MeV/sec) Vs time (days (MeV/sec) ~

t

.01< t <180.

2 10 10 10 MeV/sec (MeV/sec) t=o Palo Verde Nuclear Generating Station jjjg FSAR DECAY CURVE SOURCE C Figure 12.2-3 August 1981 Amendment 5

4 t'y

PVNGS FSAR 12.3.1 FACILITY DESIGN FEATURES Specific design features to maintain personnel exposures As Low As Reasonably Achievable (ALARA) are discussed in this section.

The design'eatures recommendations given in Regulatory Guide 8.8, Paragraph C.2, are utilized to minimize exposures.to personnel.

Facilities and equipment of a specialized nature for handling special nuclear source and byproduct material are not required except for fuel handling and radioactive waste processing.

Fuel handling and radwaste processing equipment are described in section 9.1 and chapter 11 respectively. Materials handled in the radiochemistry laboratory and low activity sealed sources used for laboratory calibration purposes do not require special handling equipment. Unsealed sources and radioactive samples are handled in conventional hoods which exhaust to the auxiliary building ventilation system, described in section 9.4.

12.3.1.1 Plant Desi n Descri tion for ALARA Equipment and plant design features employed to maintain radia-tion exposures ALARA are based upon the design considerations of section 12.1 and are outlined here for several general classes of equipment (section 12.3.1.1.1) and several typical plant layout situations (section 12.3.1.1.2).

12.3.1.1.1 Common Equipment and Component Designs for ALARA Refer to CESSAR Section 12.3.1.2 for components in CESSAR scope.

February 1984 12. 3-1 Amendment 12

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES 12.3.1.1.1.1 Filters. Filters in the auxiliary building that accumulate radioactive particles are supplied with the means to perform cartridge replacement with remote tools. Cartridge replacement of the blowdown demineralizer filter will utilize long handled tools.

Cartridge filters have'dequate space for removal, cask load-ing, and transport. A filter handling system has been incorporated into PVNGS for all filters except the low activity blowdown demineralizer filter., In use, the handling system is placed over the filter in the space normally occupied by its concrete hatch. The lead base of the system adequately attenuates cartridge radiation. A remote, closed circuit TV and a leaded glass window provide the operator with a complete view of the filter housing and cartridge while he is performing the changeout with remote tools. The cartridge can then be lifted into a shield cask placed on the base. The operator is never exposed to unattenuated radiation from the cartridge. An overhead monorail is used to transport cask and cartridge to the radwaste storage area.

12.3.1.1.1.2 Ion Exchan ers. With the exception of potentially radioactive blowdown processing system ion exchangers, ion exchangers for radioactive systems are designed so that spent resins can be remotely and. hydraulically Amendment 14 12.3-2 February 1985

PVNGS FSAR t

RADIATION PROTECTION DESIGN FEATURES transferred to spent resin tanks prior to solidification. and that fresh resin can be loaded into the ion exchanger remotely.

Underdrains and downstream strainers are designed for full system pressure drop. The ion exchangers and piping are designed with provisions for being flushed with compressed nitrogen or demineralized water.

chemical addition connections to allow the use of chemicals for descaling operations. Space is provided to allow uncomplicated removal of heating tube bundles. The non-radioactive components are separated from those that are radioactive by a shield wall. Instruments and controls necessary for evaporator operation are located on the design radiation Zone 2 side of the shield wall. Frequently operated valves in radioactive lines are capable of being operated from the design radiation Zone 2 side of the shield wall.

12.3.1.1.1.4 ~porn s. Pumps in tadioactive and potentially radioactive systems are provided with mechanical seals to E

reduce seal servicing time. These pumps include those in the nuclear cooling water, essential cooling waters safety injection, containment spray, spent fuel pool cooling, radwaste, and chemical and volume control systems. Pumps and associated piping, are arranged to provide adequate space for access to the pumps for servicing. Pumps in the above systems are provided with flanged connections for ease in removal.

Pump casings are provided with drain connections for draining the pump for maintenance. Plant layout ensures that maintenance can be performed in such a way that exposure to major radiation sources is minimized.

February 1985 12.3-3 Amendment 14

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES 12.3.1.1'.1.5 Tanks. Tanks in radioactive and poten'tially radioactive systems are provided with sloped bottoms and bottom outlet connections whenever practical.

A. Tanks with flat bottoms sloped toward the outlet include:

1. Reactor makeup water tank

2. Radwaste holdup tanks

3. Refueling water tank

4. CVCS holdup tank

5. Liquid radwaste evaporator condensate storage tanks

6. Condensate storage tank m

7. Chemical waste tanks

8. Liquid radwaste monitor tanks These tanks have outlet, connections located on the side of the tank as near to the bottom as possible.

B. Tanks with rounded bottoms and low point include: outlet'onnections

1. Spent resin tanks

2. Deleted

3. Volume control tank

4. Reactor drain tank

5. Chemical drain tank

6. Equipment drain tank Overflow lines are directed to the liquid radwaste system to control contamination within plant structures. Tanks are contained in separate compartments with drains directed to the liquid radwaste system or the chemical and volume control system.

Amendment 14 12.3-4 February 1985

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES 12.3.1.1.1.6 Heat Exchan ers. Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials with tube-to sheet joints welded or expanded to minimize leakage. Impact baffles are provided, and tube side and shell side velocities are limited to minimize erosive effects.

12.3.1.1.1.7 Instruments. Instrument devices are located in low-radiation zones and away from radiation sources whenever practical. Primary instrument devices, which are located in high-radiation areas for functional reasons, are designed for )12 easy removal for calibration. Readout devices are located in I14 design low-radiation areas, such as corridors and the control.

room, for servicing.

Some instruments (such as thermocouples) are provided in dupli-12 cate in high-radiation areas to reduce access and service time.

In the containment most instruments are located outside the secondary shield.

Integral radiation check sources for response verification for airborne radiation monitors and safety-related area radiation monitors are provided. This allows the remote removal of a shielding window to check the response of each detector.

12.3.1.1.1.8 Valves. To minimize personnel exposures due to valve operation, motor-operated. diaphragm, or other remotely actuated valves are used in highly radioactive systems that 12 require frequent valve operation.

Valves are located in valve galleries and are shielded from major system components wherever possible. Long runs of exposed piping are minimized in valve galleries. In areas where manual valves are used in frequently operated process lines, either valve stem extenders or shielding is provided, so that personnel need not enter the high-radiation area for 12 normal valve operation. The criteria for selecting valve operators are detailed in section 12.1.2.3.2.

February 1985 12.3-5 Amendment 14

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES For valves located in radiation areas, provisions are made to drain adjacent radioactive components when maintenance is required.

Valves for clean, nonradioactive systems are separated from radioactive sources and are located in readily accessible areas.

Manually operated valves in the filter and high activity ion exchanger valve compartments required for normal operation and shutdown are equipped with reach rods extending through the valve gallery walls. Personnel do not enter the valve galleries during flushing operations. The valve gallery shield walls are designed for maximum expected filter and ion exchanger. activities.

For most larger valves (2-1/2 inches and larger) in lines carrying radioactive fluids, a grafoil or graphite yarn packing is, used. Diaphragm or bellows seal valves are used where minimal leakage is required.

12.3.1.1. 1.9 ~Pi in . The piping in pipe chases is designed for the life-time of the unit. There are no valves or instrumentation in pipe chases. Wherever radioactive pip-ing is routed through areas where routine maintenance is required, pipe chases are provided to reduce the radiation contribution from these pipes to levels appropriate for the inspection requirements. Piping containing radioactive material is routed to minimize radiation exposure to the unit personnel.

12.3.1.1.1.'10 Floor Drains. Floor drains and properly sloped floors are provided for each room or cubicle containing serviceable components containing radioactive liquids. Local gas traps or porous seals are not used on radwaste floor drains.

Gas traps a'e provided at the common sump or tank.

Amendment 14 12.3-6 February 1985

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES 12.3.1.1.1.11 Sam le Stations. Sample stations for routine sampling of process fluids are located in the local sample laboratory and the radiochemical laboratory in the auxiliary building. Shielding is provided at the sample stations.

Valves in sample lines are provided with reach rods as 12 necessary to minimize personnel exposure during sampling.

12.3.1.1.1.12 Clean Services. Whenever possible, clean services and equipment such as compressed air piping, clean water piping, ventilation ducts, and cable trays are not routed through radioactive pipeways.

12.3.1.1.1.13 Reactor Coolant S stem Leaka e Control. Refer )2 to CESSAR Section 12.3.1.2K. Additionally, the space between )12 the double 0-ring seal" on the reactor vessel closure head is monitored to detect leakage past the inner 0-ring. Leakage is directed to the reactor drain tank, as discussed in Sec-tion 5.2.5.

12.3.1.1.2 Common Facility and Layout Designs for ALARA This section describes the design features utilized for standard type plant processes and layout situations. These features are employed in conjunction with the general equipment designs described in section 12.3.1.1.1, and include the features discussed in the following sections.

12.3.1.1.2.1 Valve Galleries. Valve galleries are provided with shielded entrances for personnel protection. Where practical, the valve galleries are divided so that personnel requiring access are exposed only to valves and piping associated with one component at any given location. Floor drains are provided to control radioactive leakage. To February 1984 12 0 3 7 Amendment 12

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES facilitate decontamination in valve galleries, concrete surfaces are covered with a smooth-surfaced coating, which allows easy decontamination.

12.3.1.1.2.2 ~Pi inc[. Pipes carrying radioactive materials 12 are routed through controlled access areas. Each piping run is individually analyzed to determine the potential radioactivity level and surface dose rate. Where it is necessary that radioactive 2I Amendment 12 12. 3-7A February 1984

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES This page intentionally blank September 1980 12. 3-7B Amendment 2

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES piping be routed through corridors or other low-radiation areas, shielded pipeways are provided. Whenever practical, valves and -instruments are not placed in radioactive pipeways.

Whenever practical, equipment compartments are used as pipeways only for those pipes associated with equipment in the compartment.

When practical, radioactive and nonradioactive piping are sep-

, arated to minimize personnel exposure. Should maintenance be required, provision is made to isolate and drain radioactive piping and associated equipment.

Piping is designed to minimize low points and dead legs.

Drains are provided on piping where low points and dead legs cannot'e eliminated. Long radius'lbows, or bends of several

.12 pipe diameters are utilized whenever practicable for pipes carrying radioactive material.'iping.

carrying resin slurries or evaporator bottoms, is run 14[ vertically as much as passible.

Whenever possible, branch li'nes having little or no flow during normal operation are connected above the horizontal midplane of the main pipe.

12.3.1.1.2.3 Penetrations. To minimize radiation streaming through penetrations, as many penetrations as practicable are located with an offset between the source and the accessible areas. If offsets are not practical, penetrations are located as far as possible above the floor elevation to reduce the exposure to personnel. If these two methods are not used, alternate means are employed, such as baffle shield walls or grouting the area around the penetration.

12.3.1.1.2.4 Contamination Control. Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. Equipment vents and drains from highly radioactive systems are piped directly to the r

Amendment 14 12.3-8 February 1985

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES collection system instead of allowing any radioactive fluid to flow across to the floor drain. All-welded piping systems are employed on radioactive systems to the maximum II extent pz'acticable to reduce system leakage and crud buildup at joints.

Decontamination of potentially contaminated areas and equip-ment within the plant is facilitated by the application of suitable smooth-surface coatings to the concrete floors and walls.

Sloped floors and floor drains are provided in potential'ly contaminated areas of, the plant. In addition, radioactive and potentially radioactive drain systems are separated from non-radioactive drain systems. Rooms e with equipment or tanks that contain radioactive fluids have curbs to contain potential spills or leaks. Large tanks containing radioactive fluids are enclosed in water tight compartments or are surrounded by curbs. 0'n controlled access areas where contamination is expected, radiation monitoring equipment is provided (sections 11.5 and 12.3.4). Those systems that become highly radioactive, such as the spent resin lines in the radwaste system, are provided with flush and drain connections.

12.3.1.1.2.5 E ui ment La out. In systems where process cleanup equipment is a major radiation source, pumps, valves, and instruments are separated from the process component. This allows servicing and maintenance of these items in reduced radiation areas. Control panels are located in low-radiation areas (Design Radiation Zones 1, or 2).

Major components (such as tanks, ion exchangers, and filters) in radioactive systems are isolated in individual shielded compartments. Labyrinth entranceway shields or shielding doors are provided for each compartment from which radiation February 1985 12.3-9 Amendment 14

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES could stream or scatter to access areas and exceed the design radiation zone dose limits for those areas. For potentially high-radiation components (such as ion exchangers and tanks),

I completely enclosed shielded compartments with hatch openings

14) Amendment 14 12.3-9A February 1985

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES This page intentionally blank February 198S 12.3-9B Amendment 14

PVNGS FSAR-RADIATION PROTECTION DESIGN FEATURES

>4( or labyrinth entryways with locked gates are used. For some infrequently serviced components, completely enclosed shielded compartments with removable concrete block walls are used.

Nonradioactive equipment that requires maintenance is located outside radiation areas.

Exposure from routine in-plant inspection is controlled by locating, whenever possible, inspection points in properly shielded low-background radiation areas. Radioactive and nonradioactive systems are separated as far as practicable to limit radiation exposure from routine inspection of non-radioactive systems. For radioactive systems. emphasis is placed on adequate space and ease of motion in a properly shielded inspection area. Where longer times for routine inspection are required, and permanent shielding is not feasible, sufficient space for portable shielding is provided.

For example, a remotely operated device is provided for inservice inspections of the reactor vessel. Access to high-12 I., radiation areas is under the supervision or the radiation protection personnel 12.3.1.1.2.6 Field-Run Pi in . Radioactive process piping 12 design (i.e.. routing or shielding) is not performed in the field.

12.3.1;1.2.7 Packa ed Units. Each package unit is skid mounted with all motors and pumps located on the periphery at the skid for ease of access and for quick removal to low radiation area for maintenance or repair. Package components are provided with provisions for flushing, draining, and 14 chemical cleaning. Heat exchangers are readily accessible for maintenance. As many control elements as possible are mounted remotely from the radioactive components so that the package can be remotely controlled and monitored. Components are designed with a minimum of crevices to reduce the accumulation of radioactive materials.

Amendment 14 12.3-10 February 1985

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES 12.3.1.2 Desi n Radiation Zonin and Access Control Access into the plant structures and plant yard areas is regulated and controlled.

Plant areas are categorized as design radiation zones according to expected maximum radiation levels and anticipated personnel occupancy with consideration given toward maintaining personnel exposures as low as is reasonably achievable and within the standards of 10CFR20. Each design radiation zone defines the radiation level range to which the aggregate of contributing sources must be attenuated by shielding. Each room, corridor.

and pipeway of every plant building is evaluated foi potential radiation sources during normal, shutdown, spent resin transfer, and emergency operations; for maintenance occupancy require-ments; for general access requirements; and for material expo-sure limits to determine appropriate zoning. The design radiation zone categories employed, and their descriptions are given in table 12.1-1. The specific design zoning for each plant area is shown in figures 12.3-1 through 12.3-20. Fre-quently accessed areas, e.g., corridors, are shielded for design radiation Zone 1 or Zone 2 access.

The control of entry or exit of plant operating personnel to controlled access areas, and procedures employed to ensure that radiation levels and allowable working time are within the limits prescribed by 10CFR20 is described in section 12.5.

12.3.1.3 Radiation Zones Post-Accident Radiation zone maps were developed in accordance with NUREG 0737 to review potential access throughout the plant post accident period. The facility layout assists in keeping oc'cupational exposures ALARA even aftei a design basis accident. While exposures will be significantly higher than during normal operation, required access is provided to vital areas and systems without exceeding 5 rem/hr. Zone maps showing expected dose rates in the event of a LOCA with sump May 1987 12.3-11 Amendment 17

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES recirculation are provided as figures 12.3-25 through 12.3-35.

Zone maps for the hypothetical condition of a LOCA with an intact primary but with a degraded core are provided as figures 12.3-36 through 12.3-45. The source terms correspond to those noted in section 12.2.3. The dose rates projected for these two sets of drawings do not assume decay beyond that corresponding to the onset of recirculation. Even so, virtually unrestricted access will be permitted within portions of the upper floor of the auxiliary building (such as the area of the operational support center) and the lower levels of the control building. Continuous occupancy will be permitted in the control room, .Satellite Technical Support Center (STSC),

TSC, diesel generator building and emergency operations facility (EOF) as dose rates will be 15 mrem/hr or less.

Estimated radiation levels in vital areas were based on radiation sources from the post-accident operation of the following systems: containment, safety injection/shutdown cooling/containment spray, chemical and volume control system (up to purification filter inlet), post-accident sampling, and hydrogen recombiners. The gaseous radwaste system will not be used post-accident. Palo Verde does not have a standby gas treatment system or an equivalent.

As a result of this review, piping used for grab sampling in the hot lab sample room area is lead encased to keep operator doses ALARA.

12.3.2 SHIELDING The bases for the nuclear radiation shielding and the shielding configurations are discussed in this section.

12.3.2.1 Desi n Ob ectives The basic objective of the plant radiation shielding, in con-junction with a program of controlled personnel access to, and Amendment 17 12.3-11A May 1987

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES occupancy of, radiation areas, is to reduce personnel and popu-lation exposures to levels that are within the dose regulations of 10CFR20 and 10CFR50 and are as low as is reasonably achiev-able (ALARA). Shielding and equipment layout and design are considered in ensuring flat exposures are kept ALARA during anticipated personnel activities in areas of the plant contain-ing radioactive materials, utilizing the design recommendations given in Regulatory Guide 8.8, Paragraph C.2, where practical.

An analysis of the PVNGS shielding design was performed to determine if TMI level source strengths would inhibit maintenance access or violate 10CFR50, Appendix A, General Design Criterion 19 (GDC 19). The review demonstrated that personnel radiation exposures in vital areas during post-accident activities will meet the criteria of NUREG 0737 and the GDC 19 design basis. The design review of plant shielding discussed fulfills the NRC requirements outlined in NUREGs 0578 and 0737 as well as in Regulatory Guide 1.97, Revision 2.

Four plant conditions are considered in the nuclear radiation shieldi'ng design: normal. full-power operation; shutdown; spent resin transfer: and emergency operations (for required access to safety-related equipment).

The shielding design objectives for the plant during normal operation (including anticipated operational occurrences),

shutdown operations and emergency operations are:

A. To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and proximate site boundary occupants are ALARA and within the limits of 10CFR20.

B. To assure sufficient personnel access and occupancy time to allow normal anticipated maintenance, inspec-tion. and safety-related operations required for each plant equipment and instrumentation area.

May 1987 12.3-11B Amendment 17

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES C. To reduce potential equipment neutron activation and mitigate the passibility of radiation damage to materials.

Amendment 17 12.3-12 May 1987

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES D. To ensure that the control room will be sufficiently shielded so that the direct dose plus the inhalation dose (calculated in chapter 15) will not exceed the limits of 10CFR50, Appendix A, General Design Criterion 19.

12.3.2.2 General Shieldin Desi Shielding is provided to attenuate direct radiation through walls and scattered radiation through penetrations to less than the upper limit of the design radiation zone for each area shown in figures 12.3-1 through 12.3-20. The shielding requirements for plant areas are presented in figures 12.3-1 through 12.3-20. Design criteria for penetrations are consis-tent with the recommendations of Regulatory Guide 8.8, and are discussed in section 12.3.1.1.2.

Should dose rates in excess of design zone criteria occur after PVNGS begins operation, PVNGS will add shielding or revise access to the affected area so as to ensure proper access control.

The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 140 lb/ft . Whenever poured-in-place concrete has been replaced by concrete blocks, design ensures protection on an equivalent shielding basis as determined by the density of the concrete block selected.

Concrete radiation shields are designed following the recommendations of Regulatory Guide 1.69 as discussed in section 1.8. Water is used as the primary shield material for areas above the the spent fuel storage area.

12.3.2.2.1 Containment Shielding Design During reactor operation, the containment protects personnel occupying adjacent plant structures and yard areas from radiation originating in the reactor vessel and primary loop components. The concrete containment. wall, together with the reactor vessel and steam generator 'compartment shield walls, February 1984 12 ~ 3 13 Amendment 12

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES is designed to reduce radiation levels outside the containment from sources within the containment to less than 0.5 mrem/h.

The containment shield is reinforced, prestressed concrete completely surrounding the nuclear steam supply system. The wall is 4 feet thick, and the dome varies from 4 feet at the springline to 3 feet 6 inches at the top.

For design basis accidents, the containment shield together with the control room shielding reduces the plant radiation intensities from fission products inside the containment to acceptable levels, as defined by 10CFR50 Appendix A, General Design Criterion 19, for the control room.

Where personnel and equipment hatches or penetrations pass through the containment wall, additional shielding is provided to attenuate radiation to the required level defined by the outside design radiation zone during normal operation and shut-I down, and to acceptable levels as defined by 10CFR50 during design basis accidents.

12.3.2.2.2 Containment, Interior Shielding Design During reactor operation, many areas inside the containment are design radiation Zone 5 and normally inaccessible. How-ever, shielding is designed to reduce gamma dose rates to approximately 15 mrem/h in the areas that require potential access during power operation. These are the design radiation Zone 3 areas of the containment shown in figures 12.3-1 through 12.3-20.

The main sources of radiation are the reactor vessel and the primary loop components, consisting of the steam generators, pressurizer, reactor coolant pumps, and associated piping. The reactor vessel is shielded by the concrete primary shield, reactor cavity shield, and by the concrete secondary shield which also surrounds all other primary loop components. Air cooling is provided to prevent overheating, dehydration, and degradation of the shielding and structural properties of the primary shield and reactor cavity shield.

Amendment 12 12.3-14 February 1984

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES The primary shield is a large mass of reinforced concrete surrounding the reactor vessel and extending upward from the containment floor to form the walls of the fuel transfer canal. The concrete thickness is 7 feet up to the height of the reactor vessel flange where the thickness is reduced to 4 feet.

The reactor cavity shield, an annular mass of concrete, is below the'eactor vessel nozzles between the vessel and the primary shield as shown in figure 12.3-21. This shield minimizes neutrons streaming from the annulus between the reactor vessel and the primary shield. The bottom of the cavity shield is located at, about. elevation 96 feet. The top of the cavity shield is located at about elevation 100 feet.

Estimates of neutron dose rates vary spatially over the operat-ing level from 50 to 60 mrem/hr at airlock and near pre-access filter unit. to approximately 100 to 500 mrem/hr at locations that can view the CEDM cable structure and up to 1 to 2 rem/hr at locations along the edge of the refueling canal. Neutron dose levels outside the steam generator compartments at eleva-tions below the operating level are expected to be negligible due to shielding by the concrete operating level floor.

The primary shield and reactor cavity shield are designed to meet the following objectives:

A. In conjunction with the secondary shield, to reduce the radiation level from sources within the reactor vessel and reactor coolant system to allow limited access to the containment during normal, full-power operation.

B. After shutdown, limit the radiation level from sources within the reactor vessel, to permit limited access to reactor coolant system equipment.

C. To limit neutron flux activation of component and structural materials.

February 1984 12.3-15 Amendment 12 I

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES The regenerative heat exchanger of the letdown portion of the .

chemical and volume control system is located in a shielded compartment, that is normally design radiation Zone 5.

Shielding is provided for it consistent with its postulated maximum activity (section 12.2.1) and with the access and design zoning requirements of adjacent areas.

After shutdown, the containment is accessible for limited periods of time and all access is controlled. Areas are 5I Amendment 12 12e3 15A February 1984,

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES This page intentionally blank.

August 1981 12.3-15B Amendment 5

PVNGS F SAR RADIATION PROTECTION DESIGN FEATURES surveyed to establish allowable working periods. Dose rates are expected to range from 0.5 to 1000 mrem/h, depending on the location inside the containment (excluding reactor cavity). These dose rates result from residual fission products, neutron-activated materials, and corrosion products in the reactor coolant system.

Spent fuel is the primary source of radiation during refueling.

Because of the extremely high activity of the fission products contained in the spent fuel elements and the proximity of design radiation Zone 2 areas, extensive shielding is provided for areas surrounding the spent fuel pool and the fuel transfer canal to ensure that radiation levels remain below design zone levels specified for adjacent areas. Water provides the shield-ing over the spent fuel assemblies during fuel, handling (refer to figure 12.3-24). Furthermore, substantial structural bar-riers to limit access in the vicinity of the fuel transfer tube during fuel handling operations have been provided. Refer to 5 v figure 12.3-22 for the fuel transfer shielding arrangement.

The secondary shield is a reinforced concrete structure surrounding the reactor coolant equipment, including piping, pumps, and steam generators.

This shield protects personnel from the direct gamma radiation resulting from reactor coolant activation products and fission products carried away from the core by the reactor coolant.

In addition, the secondary shield supplements the primary shield by attenuating neutron and gamma radiation escaping from the primary shield. The secondary shield is sized to allow limited access to the containment during full power operation.

The thickness of secondary shield walls is 4 feet.

12.3.2.2.3 Auxiliary Building Shielding Design During normal operation, the major components in the auxiliary building with potentially high radioactivity are those in the Amendment 12 12.3-16 February 3.984

PVNGS F SAR RADIATION PROTECTION DESIGN FEATURES chemical and volume control system, the shutdown cooling system, the fuel pool cooling and cleanup system and the primary sampling system.

Shielding is provided as necessary around the following equip-ment in the auxiliary building to ensure the design radiation zone and access requirements are met'or surrounding areas.

A. Letdown heat exchangers and piping B. Purification, preholdup, and deborating ion exchangers C. Chemical and volume control tank D. Charging pumps and piping E. Shutdown cooling heat exchangers F. Chemical drain tanks and pumps G. CVCS and radwaste filters H. Spent fuel pool cleanup ion exchangers and filters I. Spent, resin tanks and piping J. Gas stripper K. Seal injection heat exchanger L. Boronometer M. Process radiation monitor N. Seal injection filters O. Deleted Shielding is based upon operation with maximum activity conditions as discussed in sections 11.1, 11.2, 11.3, and 12.2.1.

Depending on the equipment in the compartments, the access varies from design radiation Zones 2 through 5. Corridors are shielded to allow design radiation Zone 2 access. Operator areas. for valve galleries are designed for design radiation Zone 3 access. Frequently operated valves in high radiation February 1985 12.3-17 Amendment 14

PVNGS 'FSAR RADIATION PROTECTION DESIGN FEATURES areas are provided with remote actuators extending to design radiation Zone 2 or Zone 3 areas. (See section 12.1.2.3.2M.)

Removable sections of block shield walls, or concrete hatches with offset gaps to reduce radiation streaming are provided for replacement of ion exchangers, pumps, and heat exchangers.

12.3.2.2.4 Fuel Building Shielding Design Concrete shield walls surrounding the spent fuel cask loading and storage area, fuel transfer and storage pools., and fuel transfer tube between the containment and fuel transfer pool are sufficiently thick to limit radiation levels outside the shield walls in accessible areas to design radiation Zone 2.

Access to the fuel transfer tube through the concrete radiation shield is provided by a heavy concrete hatch through the roof of the shield as shown in figure 12.3-23. The hatch is labeled to caution maintenance personnel that there are potentially lethal radiation fields during fuel transfer.

Water in the spent fuel pool provides shielding above the spent fuel transfer and storage areas. The relationship between dose rate over spent fuel during transfer. and depth 'of cover-ing water is shown in figure 12.3-24. Radiation levels at the fuel handling equipment are not expected to exceed 2.5 mrem/h.

The spent, fuel pool cooling and cleanup (SFPCC) system (section 9.1) shielding is based on the maximum activity discussed in section 12.2 and the access and design zoning requirements of adjacent areas. Equipment in the SFPCC system to be shielded includes the SFPCC heat exchangers, pumps and piping. (SFPCC filters and ion exchangers are located in the auxiliary building.)

12.3.2.2.5 Radwaste Building Shielding Design Radwaste systems are principally located in the radwaste building. Additionally, the boric acid concentrator and the Amendment 14 12.3-18 February 1985

PVNGS FSAR t RADIATION, PROTECTION DESIGN FEATURES boric acid concentrate ion" exchanger are in the radwaste building.

Major components or areas requiring shielding are:

o'ow activity and high activity spent resin tanks o Spent resin transfer pump

~ Spent resin transfer valve gallery e LRS evaporator and piping

~ LRS concentrate tanks and pumps e Boric acid concentrator

~ Boric acid and LRS ion exchangers and their valve gallery a Waste gas compressors and valve galleries

~ Waste gas surge and decay tanks t Shielding is based on maximum expected radioactivity as noted in section 12.2.

Depending on the equipment in .the compartments, the access I

varies from design radiation Zones 2 through 5. Corridors are shielded to allow design radiation Zone 2 access. Operator areas for valve galleries are designed for Zone,3 access.

Frequently operated valves in high radiation areas are provided with remote actuators extending to design radiation Zone 2 or Zone 3 areas. (See section 12.1.2.3.2M.)

Removable sections of block shield walls and concrete plugs with offset gaps to reduce radiation streaming are provided for replacement of ion exchangers, if required.

Partial shield walls are placed between equipment, in compart-ments with more than one piece of equipment to permit mainten-ance access (e.g., evaporator packages).

Turbine Building Shielding Design Radiation shielding is not required for any equipment in the steam and power conversion system located in the turbine February 1984 12. 3<<19 Amendment 12

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES building. All areas in the turbine building are classified as design radiation Zone l.

12.3.2.2.7 Control Room Shielding Design Figure 1.2-7 represents a layout drawing of the control room, showing its relationship to the containment.

The design basis loss of coolant accident (LOCA) dictates the shielding reguirements for the control room.. Consideration is given to shielding provided by the containment structure.

Shielding and adequate radiation protection are provided to permit access and occupancy of the control room under LOCA con-ditions without personnel receiving radiation exposure in excess of 5 rem whole body or its equivalent to other organs from all contributing modes of exposure for the duration of the accident, in accordance with 10CFR50 Appendix A, General Design Criterion 19.

The design basis LOCA is described in section 15.6 and is based on the recommendations of Regulatory Guide 1.4.

to personnel, from airborne fission products'inside the The'ose containment and in the radioactive cloud outside the control room, for the 30-day period following a postulated LOCA are discussed in section 15.6.

The parameters used in the demonstration of control room habitability, in addition to Regulatory Guide 1.4, are listed in section 6.4.

12.3.2.2.8 Diesel Generator Building Shielding Design There are no radiation sources in the diesel generator build-ing, therefore, no shielding is required for the building.

12.3.2.2.9 Miscellaneous Plant Areas and Plant, Yard Areas Sufficient shielding is provided for plant buildings containing radiation sources so that radiation levels at the Amendment 12 12.3-20 February 1984

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES outside surfaces of the buildings are maintained below design radiation Zone 1 levels. The steam generator blowdown ion exchangers, filter, and valve gallery are located inside of concrete shield walls. Plant yard areas that are frequently occupied by plant personnel are fully accessible during normal

.operation and shutdown. These areas are surrounded by a security fence, and closed off from areas accessible to the general public. Access to outside storage tanks that have a design contact dose rate greater than 0.5 mrem/h is restricted by concrete shield wall sufficiently high so that dose rates to personnel in plant yard areas are limited to less than 0.5 mrem/h from sources within those tanks.

12.3.2.3 Shieldin Calculational Methods The shielding thicknesses provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions described in chapter 11 and sec-tion 12.2. The thickness of each shield wall surrounding radioactive equipment, is determined by approximating the actual geometry and physical conditon of the source or sources. The isotopic concentrations are converted to energy group sources using data from standard references.

The geometric model assumed for shielding evaluation of tanks, heat exchangers, filters, ion exchangers, evaporators and piping, and the containment is a finite cylindrical volume source.

The methods and equations of Rockwell's Reactor Shielding Design Manual are used to, calculate dose rates. Buildup is calculated using Taylor coefficients presented in RSIC-10 (7) and Broder's Method of Buildup Determination, presented in the Engineering Compendium on Radiation Shielding, is used for laminated shields.

February 1984 12 3 2 1

~ Amendment 12

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES In addition, three industry accepted computer programs, ANISN QAD (10) and MORSE (11) are used for shielding analysis. ANISN is a multigroup one-dimensional discrete ordinates transport program that solves the one-dimensional Boltzmann transport equation for neutrons and gamma rays in slab, sphere, or cylinder geometry. Using a finite-difference technique, ANISN allows general anisotropic scattering; i.e., an' order Legendre expansion of the scattering cross-sections.

Monte Carlo techniques may be used for more complicated geometries such as penetrations. QAD is a point,-kernel general purpose program for estimating the effects of gamma rays and neutrons that, originate in a volume distributed source.

Description of the three-dimensional geometry of the problem is accomplished by using quadratic equations to define surfaces.

ANISN is used for primary shield design and QAD is used for configurations not conveniently modeled as a cylindrical source with annular shields.

For design of the reactor cavity, a three-dimensional model was used to simulate rad'iation streaming from the reactor surface to the containment using the MORSE Monte Carlo program. The source terms used for the MORSE program were divided into 13 neutron energy groups.

The shielding thicknesses are selected to reduce the aggregate computed radiation level from contributing sources below the upper limit of the design radiation zone specified"for each plant area. Shielding requirements are evaluated at the point of maximum radiation dose through any wall. Therefore, the actual anticipated radiation levels in the greater region of each plant area is less than this maximum dose and therefore less than the. design radiation zone upper limit.

Amendment 12 12.3-22 February 1984

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES Where shielded entryways to compartments containing high-radiation sources are necessary, labyrinths or mazes are designed using a general purpose gamma-ray scattering program (12) or methods summarized in RSIC-21. (13)

~ ~

'he mazes are constructed so that the scattered dose rate, plus the transmitted dose rate through the shield wall from con-tributing sources, is below the upper limit of the design radi-ation zone specified for each plant area.

12.3.3 VENTILATION The plant heating,, ventilating, and air conditioning (HVAC) systems are designed to provide a suitable environment for personnel and equipment. during normal operation and events of moderate frequency or certain infrequent events.

1 Parts of the plant HVAC systems perform safety-related functions.

12.3.3.1 Desi Ob'ectives The plant HVAC systems for normal plant operation and events of moderate frequency or certain infrequent events are designed to meet the requirements of 10CFR20 and 10CFR50.

12.3.3.2 Desi n Criteria Design Criteria for the plant HVAC systems include:

A. During normal operation and events of moderate frequency or certain infrequent events, the average and maximum airborne radioactivity levels to which plant personnel are exposed are as low as is reasonably achievable (ALARA) and within the limits specified in 10CFR20.

February 1984 12.3-23 Amendment 12

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES B. During normal operations and events of moderate frequency or certain infrequent events, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary is ALMA and within the limits specified in 10CFR20 and 10CFR50, Appendix I.

C. The plant site dose limitations of 10CFR100 will be adhered to following those hypothetical accidents described in chapter 15.

D. The dose to control room personnel shall not exceed the limits specified in General Design Criterion 19 of Appendix A to 10CFR50 following those hypothetical accidents described in chapter 15.

12.3.3.3 Desi n Guidelines To accomplish the design objectives, the following guidelines are followed wherever practical.

12.3.3.3.1 Guidelines to Minimize Airborne Radioactivity A. Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination.

B. Equipment vents and drains are piped directly to a collection device connected to the collection system instead of allowing'ny radioactive fluid to flow across the floor to the floor drain.

C. All-welded piping systems are employed on systems containing radioactive fluids to the maximum extent practical to reduce system leakage.

D. Suitable coatings are applied to the concrete floors and walls of potentially contaminated areas to facilitate decontamination.

Amendment 12 12.3-24 February 19S4

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES E. Design of potentially contaminated equipment incorporates features that minimize the potential for airborne radioactivity during maintenance operations. These features may include connections on pump casings for draining and flushing the pump prior to maintenance, or flush connections on piping systems that could become highly radioactive.

12.3.3.3.2 Guidelines to Control Airborne Radioactivity A. The airflow is directed from areas with lesser potential for contamination to areas with greater potential for contamination.

B. In building compartments with a potential for contamina-tion, the exhaust is designed for greater volumetric flow than is supplied to the area to minimize the amount of uncontrolled exfiltration from the area.

C. Air cleaning systems criteria for emergency systems are discussed under Regulatory Guide 1.52 in Section 1.8. System design for both normal and emergency systems is described in section 9.4.

D. Means are provided to isolate the containment and fuel buildings upon indication of radioactive contamination to prevent the discharge of contaminants to the environment and minimize in-plant exposure.

E. Means are provided to isolate the control room to minimize inleakage of contaminated air to the operator.

F. Suitable containment. isolation valves are installed to ensure that the containment integrity is maintained.

See additional discussion in sections 3.1.47, 3.1.49, and 6.2.4.

February l984 12.3-25 Amendment 12

PVNGS FSAR RADIATION PROTECTION DESIGN FEATURES G. Redundancy and Seismic Category I classification features are provided for components of the HVAC systems required for safe shutdown of the reactor plant.

12.3.3.3.3 Guidelines to Minimize Personnel Exposure from HVAC Equipment 4

Access and service of ventilation systems in potentially radioactive areas are provided by component location to minimize operator exposure during maintenance, inspection, and testing as follows:

A. Ventilation equipment rooms for outside air supply units and building exhaust system components are located in design radiation Zone 2 and are accessible to the operators. Ventilation exhaust filtration units for the auxiliary and fuel buildings are located on the roofs of. the respective buildings and are designated as being design radiation Zone 3 areas. Work space is provided around each unit for, anticipated maintenance, testing, and inspection. Filter-adsorber units gener-ally are consistent with the recommendations of Regulatory Guide 1.52,for access and service requirements. (Refer to section 1.8).

B. Local HVAC equipment that services the normal build-ing requirements is located in areas of low contamination when practical.

12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITYMONITORING INSTRUMENTATION Section 11.5 describes the function, operation, design, and locations of the radiation monitoring system (RMS).

The RMS provides 33 fixed area and 14 fixed airborne monitors per unit as noted in section 11.5. Additionally, there are 18 portable monitor connection boxes located through-out each unit that can connect any of the three portable area Amendment 12 12.3-26 February 1984

PVNGS F SAR RADIATION PROTECTION DESIGN FEATURES or one movable airborne monitors available for each unit, for use during extended maintenance. Thus, a portable monitor can be placed where necessary to provide dose rate surveillance.

This will reduce doses to radiation protection personnel, while providing necessary surveillance.

Refer to section 11.5 for a further description of the radiation monitoring system, its sensitivity, and setpoints, as well as a description of how the RMS fulfills the applicable requirements of Regulatory Guides 1.21, 8.2, 8.8, 8.12, and ANSI N13.1.

l2.

3.5 REFERENCES

Martin, J. J. and Blichert-Toft, P. H., Nuclear Data Tables " Radioactive Atmos, Auger Electrons, a, b, q, and X-Ray Data," Academic Press, October 1970.

2. Martin, J. J., Radioactive Atmos Su lement 1, ORNL 4923 g August 1 973

3. Bowman, W. W. and MacMurdo, K. W., Atomic Data and Nuclear Data Tables, "Radioactive Decay q's Ordered by Energy and Nuclide," Academic Press, February 1970.

4, Meek, M. E. and Gilbert, R. S., "Summary of q and b Energy and Intensity Data," NED0-12037, January 1970.

5. Lederer, C. M., et al., Table of Isoto es, Lawrence Radiation Laboratory, University of California, March 1968.

6. Rockwell, T., Reactor Shieldin Desi n Manual, D. Van Nostrand Co., New York, 1956.

7. Trubey, D. K., "A Survey of Empirical Functions Used to Fit Gamma-Ray Buildup Factors," ORNI-RSIC-10, February 1966.

February 1984 12 ' 27 Amendment l2

PVNGS FSAR

. RADIATION PROTECTION DESIGN FEATURES Jaeger, R. G., et al., En ineerin Com endium on Radiation Shieldin , Shielding Fundamentals and Methods I, 1968.

Engle, Jr., W. W., "A User's Manual for ANISN: A One Dimensional Discrete Ordinates Transport Code with Anistropic Scattering,", Union Carbide Corporation, Re ort No. K-1693, 1967.

Malenfant, R. E., QAD, A Series of Point-Kernel General-Purpose Shielding Programs, Los Alamos Scientific Laboratory, LA 3573, October 1966.

RSIC Computer Code Collection CCC-203, MORSE-CG, General Pu ose Monte Carlo Multi-Grou Neutron and Gamma Ra Trans ort Code with Combinatorial Geometr Malenfant, R. E., "G : A General Purpose Gamma-Ray Scattering Program" Los Alamos Scientific Laboratory, LA 5176, June 1973.

Selph, W. E., "Neutron and Gamma Ray Albedos,"

ORNL-RSIC-21, February 1968.

12.3-28