ML20151U169

Rev 2 to DSAR for Humboldt Bay Power Plant,Unit 3

ML20151U169
Person / Time
Site:	Humboldt Bay
Issue date:	08/28/1998
From:	PACIFIC GAS & ELECTRIC CO.
To:
Shared Package
ML20151U125	List: HBL-98-015, Forwards Rev 2 to DSAR for Humboldt Bay Power Plant,Unit 3, to Comply W/Decommissioning Rule.Dsar Was Formerly Known as Safstor Decommissioning Plan.Encl 1 Identifies & Describes Rept Changes ML20151U169
References
NUDOCS 9809100312
Download: ML20151U169 (265)

Text

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1 DEFUELED SAFETY ANALYSIS REPORT FOR THE HUMBOLDT BAY. POWER PLANT, UNIT 3

$R DOC 00 33 l

Rev 2 Augut 1998

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! EXECUTIVE

SUMMARY

In 1984 PG&E submitted the Humboldt Bay Power Plant, Unit 3 (HBPP) SAFSTOR 1 l Decommissioning Plan (SDP) in support of the application to amend the HBPP Operating 1 l License to a Possession-Only License. As a result of the 1996 NRC decommissioning rule, j the SDP was considered to be a Post-Shutdown Activities Report (PSDAR) because it l contained information related to decommissioning activities. It was also considered to be a Final Safety Analysis Report (FSAR) because it contained information such as plant description, site characterization and accident analysis.

! In compliance'with the 1996 NRC decommissioning rule, PG&E submitted a PSDAR in .

l l February 1998 to provide a general overview of proposed decommissioning activities.- As a

. result, the SDP will focus on providing the type of information contained in an FSAR and will

, contain less information related to decommissioning activities. Thus, the SDP has been more appropriately renamed the Defueled Safety Analysis Report (DSAR).

The 1996 NRC decommissioning rule became effective August 28,1996. This rule modified j 10 CFR 50.71 to require licensees of permanently defueled plants to revise their FSARs at least every 24 months. To comply with the decommissioning rule, PG&E is submitting this DSAR prior to August 28,1998.

DESCRIPTION OF THE DSAR- l Section 1 of the DSAR describes the contents of the report, including a listing of the regulations and other documents that were reviewed and considered in placing HBPP in SAFSTOR.

Section 2 provides a brief summary of the Unit 3 licensing and operating history.

Section 3 includes a description of the site and a physicei and radiological characterization of the facility at the time of the commencement of the SAFSTOR period. l Section 4 describes the plant organization, training programs, and a description of activities performed to place HBPP in SAFSTOR.

Section 5 describes the baseline radiation survey that as conducted at the onset of the SAFSTOR period. This section also includes a description of the monitoring and surveillance program and continued care plan in effect during the SAFSTOR period.

Section 6 is a description of the health physics and occupational health and safety

- programs in effect during the SAFSTOR period. l Appendix l provides a discussion of the safety implications during SAFSTOR and a discussion of accidents that could occur and their consequences.

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4 Appendix IA is a description of spent fuel heatup following loss of storage pool water.

Appendix 1B is a criticality analysis for SAFSTOR decommissioning. ,

Appendix IC is the SAFSTOR baseline radiation study. ,

Appendix ll Is a description of the program for training and certification of operators as fuel l handlers during the SAFSTOR period. i

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TABLE OF CONTENTS EXE C UTIVE S U M MAR Y . . . . . . . . . . . . . . . . . .. . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

TA B L E O F C O NTE NTS . .. .. . . .. . .. . . . . . .. . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.0 1 N TR O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.1 D EFU ELED SAFETY ANALYSIS R EPORT ...... .......................................... 1 -1 1.2 CRITERIA AN D G UIDELIN ES R EVIEW ......... .............. ... .. .............. ... ... 1-2 2.0 HUMBOLDT BAY POWER PLANT - UNIT 3 OPERATING HISTORY ....... ........... 2-1 2.1 1 N TR O D U C TI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.2 INITIAL CONSTRUCTION AND LICENSING HISTORY.............................. 2-1 2.3 OPERATING EVENTS WHICH AFFECT DECOMMISSIONING................. 2-2 2.3.1 Fuel Cladding Failures ............. . ........ . ...... ........ ........ .. . . ... . . . .......... ...... 2-2 2.3.2 Spent Fuel Pool Leakage ..................................... ......... .................. 2-2 2.3.3 Spills of Contaminated Water................. ....................................... .. 2-2

2.3.4 Droppe d Fu el As sembly ... ..... . . ... ..... .... ... ... . .... . .... . . . ..... . ......... 2-3 2.4 O P E RATI N G rR ECO R D. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 3.0 C HARACT E R IZATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 l 3.1 S ITE D E S C R I PTIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . 3-1 i

l i 3.1.1 To p o g ra p h y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 3.1.2 S oils an d G e ology. . . . . . . . ... . . . .. ... . . . .. . . . . . . . . . . . .. . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . 3-1 3.1.3 Hydrology ..................................................................................3-2 3.1.3.1 Surface Hydrology ........... ... ..................... .......... ... 3-2 3.1.3.2 Groundwater Hydrology.................................. ...... .... 3-2 l

3.1.3.3 H u m boldt Bay . . . ... ... . . .. . . . . . . . . . . . . . . . . .. . . . . .. .. .. . . . . . . . . . . . .. . . . 3-3 r

3.1.4 Seismology ...................................................................................3-3 3.1.5 Climatology and Meteorology ............................ ............................... 3-3 l

3.2 FACI LITY DES CRIPTION ....... ............................... ....... ... ...... .. . ... ........ . . . .. .. 3-4 3.2.1 General Plant Description .............. ...................... ... ........... .......... 3-4 3.2. 2 Plant S tructure s .................. .. ... .... . ... . ... ..... .. ......... . . . . . .. .... . . . . 3-5 3.2.2.1 Fou nd atio n s . . . . . . . . .. . .. . . . . . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 3.2.2.2 P ower B uildin g . . . . . . . . .. . . . . . .. . . . . . . . . .. . . . .. . . . . . . . . . . . . . . .. . . . . . . 3-5

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TABLE OF CONTENTS (Cont'd) 3.2.2.3 Turbine Pedestal and Shielding ................................ 3-6 3.2.2.4 Refueling Building ...... . .............. ...... ..... ... .... ... . . . . .. . ... 3-6 3.2.2.5 Re actor Calsson . .. ...... . ...... .......... . ... . . . . . . . . .... . . . . ... 3 F, 3.2.2.6 Ventilation Stack .. .. . .. ...... ............. ....... . ... . . . . . . . .. ... .. .. . 3-8 3.2.2.7 Radwaste Treatment Facilities ............... .. ....... ....... 3-8 3.2.2.8 Yard S tru ctu re s . . . . . . ... . . . .. . . .. .. ... . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . .. 3-8 3.2.2.9 Intake and Discharge Structures ....... ..... ... ............. 3-9 3.2.2.10 C eismic Upgrading . . ... ..... ........ .. ...... .. .. ...... . ... .. . .. ... . 3-9 3.2.2.11 Onsite Combustible Fuel Storage ..... ..... . ........ ...... 3-9 3.2.3 Plant S ystems Description ............................... ............... ... . ...... . .. 3-9 3.2J.1 Nuclear Steam Supply System .......... ....... ... ........ 3-10 3.2.3.2 Turtine Plant Systems ............. . . .......................... 3-12 3.2.3.3 Waste Disposal Systems ........ ................................ 3-15 3.2.3.4 : Instrumentation and Control (l&C) Systems ........... 3-19 3.2.3.53 Service Systems ......... .. .... ....... . .. . ...... . . ........ . .. .. . .... 3-19 3.2.3.6 Electrical Systems .............. ................................. ... 3-21 3.2.3.7 Hydrogen and Seal Oil System ............................... 3-23 3.3 RADIOLOGICAL CHARACTERIZATION ...................... ...... ........ ........... 3-23 3.3.1 Radionuclide inventory . ........................ .. .... ... . .. .. ..... . ......... .. 3-23 3.3.2 Refueling Building / Power Building ......... .... ........... .... ..... ..... .... 3-25 3.3. 3 Yard S truetu re s . . . . . . . .. . . . . . . . .. . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 5 3.3.4 Cro ss-Connection s .... ........ .. . . ........................... ... . .. . . . . . . ... . . . . . . 3-25 3.3.5 Environmental Radiological Characteristics .......... .... ................... 3-26 3.3.6 Summary ...........................................................................3-26 4.0 SAFSTOR DECOMMISSIONING ACTIVITIES ................................ .... .. ......... . . 4-1 l 4.1 O B J E CTIVE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4.2 PLANT STAFF ORGANIZATION AND RESPONSIBILITIES ..... ................ 4-1 4.2.1 S AFSTO R Organization .. ... .. ....... ........... .......... .... . . ..... . .. . .. ... . . ... 4-1 4.2.2 Offsite S u pport ... . . . . . ... . . . . . . . . . . . . . . .. . .. . . . . . .. .. . . .. . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . .. 4-2 4.2.3 Staffing During SAFSTOR ..................................... ......... ..... .. .. . . 4-2  :

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• A D M I N ISTRATIO N AN D C O NTR O L ........... . .... ...... ........... .. ... . .. . ... . ...... ....

l 4.3.1 Cost Estimates a nd Financing ... .... ... ...... ... ........ . . . .... .... . . . .. . . . ... .. ... 4-3 4.3.2 Procurement ................................................... .......................4-4 1 l 4. 3. 3 Tra in i n g P rog ra m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4.3.3.1 Training Program Description ............................ ...... 4-5

' 4.3.3.2 General Employee Training ...... .............. . .... ... ..... 4-5 4.3.3.3 Tech nical Training ............................................. ... .. 4-7 4.3.3.4 O th e r Train in g . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-9

[ 4.3.3.5 Training Program Administration and Records ....... 4-10 4.3.4 O uality Assurance Prog ram .... ........ ....................... .... .. ........... 4-11 4.4 D ECOM M I SSIO N I N G ACTiVITI ES .... ............ .......... ............................. 4-11 l 4.4.1 Preparations for SAFSTOR Decommissioning . ............................ 4-12 l 4.4. 2 S y ste m La y u p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 12 4.4.3 O pe ratio n al Systems . .. .. . . . .. . . ... . . . . .. . .. .. . . . .. . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . 4-16 4.4.3.1 U n it 3 S yste m s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 16  !

4.4.3.2 Units 1 and 2 Systems Common to Unit 3 .. . ........ 4-21 4.4.4 S pe n t F u el S to ra g e . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 3 4.4.5 Radioactive Waste Processing and Disposal .............. .. ............ . 4-24 4.4.5.1 Sources of Radioactive Wastes .... . ....................... 4-24 4.4.5.2 Waste Processing and Disposal ...... .... .. ... ......... 4-25 l 4.4.6 D e co n t a min a tio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-26 4.4.6.1 P u rp o s e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -2 6 4.4.6.2 Decontamination Methods ...... ............................... 4-27 4.4.6.3- System Internal Flushing ........... .. . ......... .. ..... .... 4-29 5.0 S AF STO R S ITE CO N D ITI O N S . .. ..... . . . . . . . .. . .. . . . . . . . . . . . . . .. .. . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 BAS E LIN E RADI ATIO N S U RVEY .. ............. .. ............... . . ... .. .. .... .. . ... ......... 5-1 5.2 M ON ITO RI N G AN D SU RVEILLANC E . . ... . ................ .. ... .. .. ...... .. .. ..... .... . 5-2 i

i 5.2.1 I n- P la n t M o n it o rin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 l

[ 5.2.2 Onsite Environmental Monitoring .. ... .. ...... . . ........ . . . . . . . . . . . . . . 5- 3

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5.2.3 Offsite Environmental Monitoring ......... ...................... ....... ........... 5-3 r

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• 5.'3 CONTINUED CARE PLAN, SAFSTOR TO DECON ...... ....... .................. 5-3 l.

l 5.3.1 Operation of Plant Systems ...... ................. ... . .. . ................ ... . .... .. 5-3 5.3.2 Maintenance of Structures, Systems, and Components ........ ... ..... 5-4 HEALTH PHYSICS, OCCUPATIONAL HEALTH AND SAFETY.,............ ........... . 6-1 l 6.0 6.1 ORGANIZATION AND RESPONSIBILITIES .............................................. 6-1 6.2 ALARA P R O G RAM . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 6.3 RADIATIO N P ROTECTION P RO G RAM.............................. . .................. . 6-3 '

6.3.1 P e rson n el M o nito ring . .. . . . . ..... .. . ... .... .. . . .... .. . . . . .. . . . . . . . . . . . . . . . . .. . .. . . . . . .. .. 6-4 6.3.2 Airborne Radioactivity ... ............... ... ................. ............. ......... ...... 6-5 6.3.3 Respiratory Protection P rog ram ....... ............................ .... ....... ........ 6-5 6.3.4 P ro te ctive C lo thi n g . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.3.5 C o n tro l o f Acce s s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.3.6 Fa cilities Mo nito rin g . .. .. . . .. . . . . . . .. ... .. . .. . . . . ... . . ... . ... . .. . . .. . ... . . .. . . . . .. . . .. . 6-7 6.3.6.1 Ai r S a m p l e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 6.3.6.2 Rad latio n S u rveys.. .... . .. . . .. . .. . . . . . . . .. . .. . . .. .. . .. . . .. ... . . .. ... 6-7 6.3.6.3 Wa te r S a m ple s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 6-8 6.3.7 Radiation Protection Equipment and Instrumentation ... . ............... 6-9 l

6.3.7.1 P ortable Instruments..................... ....... ........... .......... 6-9 6.3.7.2 Area Radiation Monitors ..... ........... ............... .... .... 6-9 6.3.7.3 Laboratory instrumentation ............... .......... ....... .... 6-9 6.3.7.4 Maintenance of Radiation Protection Instruments ... 6-10 6.3.8 Radiatio n Protection Records ....................................................... 6-10 6.4 INDUSTRIAL HEALTH AND SAFETY PROGRAM .......... ..... ... ...... ..... 6-10

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APPENDICES I implications of Accidents During SAFSTOR .................... ......... ..... ........... ..... .......l-1 IA Spent Fuel Heat Up Following Loss of Storage Pool .............................. ............. . lA-1 iB Pacific Gas and Electric Company Humboldt Bay Power Plant Unit 3 Criticality Analysis for SAFSTOR Decommissioning... ................. .........................1B-1 IC Humboldt Bay Power Plant Unit 3 SAFSTOR Baseline Radiation Study....... ........lC-1 11 SAFSTOR Operator Training and Certification Program .........................................11-1 l

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LIST OF TABLES 3A Combustible Fuel Storage Facilities .......... ........ ............ ................. ................ . 3-27 l 3-1 Physical Characterization - Summary ....................... ............ . .... .. .................. 3-28 3-2 R e a cto r D a t a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

l 3-3 Survey Results - Reactor Vessel Drain Test ............................ ... ....................... 3-50 3-4 Spent Fuelinventory July 1984...........................................................................3-53 3-5 Spent Fuel Pool Miscellaneous Inventory - 1 9 84 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-54 3-6 Reactor Vessel Inventory of Radienuclides Corrected for Decay for

l. Conditio ns Mid -1984 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3-7 Corrosion Film Radionuclide Inventory Corrected to July 1 9 84 . .. . . . . . . . . . . .. . . . . . . . . . . . . 3-56

! 3-8 Radionuclide inventory Estimates for Humboldt Bay Reactor S y ste m s , J uly 1984 . . . .. . . . .. . . . . . . ... . . . . . . . . . .. . . .. . . . . . . . . . .. . . . . . . . . . . . .. . . . . . .. . . . . . . . . . . . . . .. . ... 3-57 2

3-9 Radionuclide Concentration in Concrete Cores July 1984 (pCi/cm ) .................. 3-59 l 3-10 Radiation Survey-Refueling Building . .. .. . ....... . ... .................... ..... .... ... ....... . 3 63 3-11 Radiation Survey-Power Beilding .......... ..... ............................... ....... ................. 3-65 l

l 3-12 Radiation Survey-Yard Structures ........ ....................... ..... .. ...... .... . ................. 3-67 3-13 Cross-connections Radionuclide Analysis .... ............... ......... .. ....... ................ 3-69 4-1 l&C System Status During SAFSTOR ......... ......... .............. ............................... 4-30 t

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LIST OF FIGURES  !

I 3-1 Humboldt Bay Power Plant Site Plan ... .............................. ..... ......... ................ 3-70 3-2 Humboldt Bay Unit 3 - Operating Floor Plan, El 2 7'- 0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-71 3-3 Humboldt Bay Unit 3 - Ground Floor Plan ............................................. ............... 3-72  ;

3-4 Hot Machine Shop and Calibration Facility ........................................................ .. 3-73  ;

3-5 Equipment Location - Operating Floor Plan, El 2 7'- 0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-74 '

3-6 Equipment Location - Ground Floor Plan, El 1 2'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-7 5  ;

3-7 Equipment Location - Section a-a .... ... ................................................................ 3-76 3-8 Equipment Location - Section b-b ............. .................. ....... ....................... . ....... 3-77 i 3-9 Equipment Location - Section c-c ............................................................... .......... 3-78 l

3-10 Equipment Location - Section d-d ................................ ............. ......................... 3-79  !

3-11 Equipment Location - Section e-e ............................................. .......... ................ 3-80 3-12 Equipment Locations - Plan at El (-) 2'-0" ................................... .......................... 3-81 3-13 Equipment Locations - Plan at El(-) 14'-0".......................................................3-82 i

. 3-14 Equipment Locations - Plan at El(-) 24'-0".........................................................3-83 '

15 Equipment Locations - Plan at El(-) 34'-0".........................................................3-84 3-16 Equipment Locations - Plan at El(-) 44'-0"........................................................3-85 i 3-17 Equipment Locations - Plan at El(-) 54'-0".......................................................3-86 ~

3-18 Equipment Locations - Plan at El(-) 66'-0"............................................................3-87 3-19 Schematic Diagram of Reactor Pressure Vessel and Intemals ........................ .. 3-88 for HBPP Unit 3 I

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1.0 INTRODUCTION

in 1984 PG&E submitted the Humboldt Bay Power Plant, Unit 3 (HBPP) SAFSTOR Decommissioning Plan (SDP)in support of the application to amend the HBPP Operating License to a Possession-Only License. As a result of the 1996 NRC decommissioning rule, the SDP was considered to be a Post-Shutdown Activities Report (PSDAR) because it

. contained information related to decommissioning activities. It was also considered to be a

- Final Safety Analysis Report (FSAR) because it contained information such as plant description, site characterization and accident analysis.

In compliance with the 1996 NRC decommissioning rule, PG&E submitted P. PSDAR in February 1998 to provide a general overview of proposed decommissioning activities. As a result, the SDP will focus on providing the type of information contair.ed in an FSAR and will contain less information r'etated to decommissioning activities. Thus, the SDP has been more appropriately renamed the Defueled Safety Analysis Report (DSAR).

In addition to the DSAR and PSDAR, PG&E has submitted other documents to the NRC in accordance with 10 CFR 50 that constitute the licensing basis for HBPP. These other documents include: (1) License Amendment Application, (2) revised Technical Specifications, (3) Environmental Report, (4) Quality Assurance Plan, (5) Security Plan, (6)

Emergency Plan.

1.1 DEFUELED SAFETY ANALYSIS REPORT l This DSAR (formerly known as the SDP) was originally prepared in support of PG&E's application to amend the Unit 3 operating license to a possession-only license.- The unit was placed in a state of custodial SAFSTOR for up to 30 years, after which it is planned to dismantle the unit, remove all radioactive material from the site, and terminate the license in accordance with NRC requirements. More specific infont.ation pertaining to future decommissioning activities is contained in the PSDAR.

Section 2.0 of this plan summarizes the licensing and opefating history of the plant.

Section 3.0 includes a description of the site and a physical and radiological __

characterization of the facility.' The characterization describes the conditions present at the time of commencing decommissioning activities. Certain activities such as unloading of the reactor core, processing and disposal of radioactive wastes, and decontamination of some structures, systems, and components performed in accordance with the operating license

_ h before the commencement of SAFSTOR activities, are reflected in this section.

Section 4.0 describes the decommissioning activities that were performed to establish the custodial SAFSTOR mode. This section also contains a description of the plant organization, administration, and control. .

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Section 5.0 describes the conditions that will exist during the SAFSTOR period. Also include.d in this section is a description of the monitoring and surveillance programs used during this period.

Section 6.0 describes the health physics and occupational health and safety program.

Appendix l contains a safety and accident analysis for Unit 3 during the SAFSTOR period. l Appendix IA is a description of spent fuel heatup following loss of storage pool water.

Appendix 18 is a criticality analysis for SAFSTOR decommissioning.

Appendix IC is the SAFSTOR baseline radiation study.

Appendix il is a description of the operator training and certification program used during the SAFSTOR period.

l 1.2 CRITERIA AND GUIDELINES REVIEW in preparation of this DSAR, PG&E complied with the guidance provided in Draft Regulatory Guide DG-1067, " Decommissioning of Nuclear Power Reactors," dated July 1997.

In preparation of the original SDP, govemment regulations, regulatory guidelines, and technical reports were reviewed to determine their applicability to HBPP. Other decommissioning plans, reports, and relevant facility experiences were also reviewed. The following list identifies the regulations and related documents that were reviewed.

CODE OF FEDERAL REGULATIONS TITLE 10 - ENERGY Part 20 - Standards for Protection Against Radiation Part 30 - Rulas of General Applicability to Domestic Licensing of Byproduct Material Part 40 - Domestic Licensing of Source Material Part 50 - Domestic Ucensing of Production and Utilization Facilities Part 51 - Licensing and Regulatory Policy and Procedures for Environmental Protection Part 61 - Licensing Requirements for Land Disposal of Radioactive Waste Part 70 - Domestic Licensing of Special Nuclear Material Part 71 - Packaging of Radioactive Material for Transport and Transportation of Radioactive Material Part 73 - Physical Protection of Plants and Materials TITLE 29 - LABOR Part 1910 - Occupational Safety and Health 1-2 l Rev 2 August 1998

TITLE 40 - PROTECTION OF ENVIRONMENT Part 61 - National Emission Standard for Hazardous Air Pollutant Part 122 - National Pollutant Discharge Elimination System Part 141 - National Interim Primary Drinking Water Regulations Part 190 - - Radiation Protection Standards for Nuclear Power Operations TITLE 49 -TRANSPORTATION Parts 171-179 - Hazardous Materials Regulations REGULATORY GUIDELINES Regulatory Guide 1.86 - Termination of Operating Licenses for Nuclear Reactors TECHNICAL REPORTS (1) Technology, Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, NUREG/CR-0672, June 1980

.(2) Technology, Safety and Costs of Decommissicning a Reference Boiling Water Reactor Power Station, NUREG/CR-0672 Addendum 1, July 1983

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L (3) Draft Generic Environmental impact Statement on Decommissioning of Nuclear Facilities, NUREO-0586, January 1981 I

(4) Final Environmental Impact Statement - Decommissioning of the Shippingport Atomic Power Station, DOEEIS-0080, May 1982

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(5) An Engineering Evaluation of Nuclear Power Reactor Decommissioning Altematives, AIF/NESP 009, November 197.6 DECOMMISSIONING PLANS AND EXPERIENCE (1) Dismantling Plan - Plum Brook Reactor Facility, February 1980 (2) Decommissioning Plan and Safety Analysis Report - Peach Bottom Atomic Power Station Unit 1, May 1975 (3) Decommissioning Peach Bottom Unit 1 Final Report, July 1978 (4) Decommissioning Plan for Indian Point Unit 1 (Proposed), October 1980 (5) Discussions with General Electric Company regarding decommissioning experience with the Vallecitos Boiling Water Reactor, ESADA Vallecitos Experimental Superheat Reactor, and General Electnc Test Reactor. l l-3 l Rev 2 August 1998 ,

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2.0 HUMBOLDT BAY POWER PLANT - UNIT 3 OPERATING HISTORY  :

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2.1 INTRODUCTION

j Humboldt Bay Power Plant, Unit 3 was a natural circulation boiling water reactor and associated turbine-generator operated by Pacific Gas and Electric Company (PG&E). In i addition to Unit 3, the Humboldt Bay Power Plant consists of two oil and/or natural gas i fueled units (Unit i rated at 52 MWe and Unit 2 rated at 53 MWe). Two diesel-fueled gas  !

' turbine Mobile Emergency Power Plants (MEPPs), each rated at 15 MWe, are also currently i located at the plant, but may be relocated to other sites either temporarily or permanently. i r ,

i 2.2 INITIAL CONSTRUCTION AND LICENSING HISTORY Unit 3 was granted a construction permit by the Atomic Energy Commission (AEC) on

' October 17,1960, and construction began in November 1960. The AEC issued Provisional l i Operating Ucense No. DPR-7 for Unit 3 in August 1962. Unit 3 achieved initial criticality on February 16,1963, and began commercial operation in August 1963. i To simplify plant design, Unit 3 included certain features that were not typical of nuclear  ;

plants of that era. Natural circulation within the reactor vessel eliminated the need for i recirculation pumps, a direct cycle design eliminated the need for heat transfer loops between the reactor and turbine-generator, and as a joint effort between PG&E and General Electric Company, the pressure suppression containment system was developed to eliminate the need for the large containment structures that had been used at earlier nuclear plants. The pressure suppression containment design permitted the reactor to be

_ located below ground level.

On July 2,1976, Unit 3 was shut down for annual refueling and to conduct seismic modifications. Seismic and geologic studies were in progress. In December.1980 it became apparent that the cost of completing required backfits might have made it uneconomica'i to restart the unit. Work was suspended at that time awaiting further .

guidance regarding backfitting requirements. In 1983, updated economic analyses l' Indicated that Iastarting Unit 3 would probably not be economical, and in June 1983 PG&E l announced its intention to decommission the unit. This DSAR describes the plans to implement that decision.

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i l 2.3 OPERATING EVENTS WHICH AFFECT DECOMMISSIONING l

During the operation of Unit 3, certain events occurred that affected plant conditions and l had to be considered during decommissioning. The following section describes these  !

events and how they related to the decommissioning effort. None of these events caused conditions that would prevent Unit 3 from being decommissioned with current technologies l and wo* practices.

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l 2.3.1 Fuel Cladding Failures i

j When Unit 3 began operation, the fuel utilized stainless steel claddin? !n 1964 and 1965, l fuel cladding failures began to occur and it was determined that the caue of the failures >

was stress corrosion cratking of the stainless steel cladding. In 1965, the stainless steel-clad fuel was replaced with zircaloy-clad fuel.

The early fuel cladding failures resulted in contamination of the reactor vessel, spent fuel storage pool, and plant systems with fission products and transuranic nuclides. All stainless steel-clad fuel was shipped offsite for reprocessing during the years 1969 through 1971.

2.3.2 Spent Fuel Pool Leakage in March 1966, it was discovered that a leak in the spent fuel storage pool liner had developed. Operating procedures were developed to minimize leakage and investigations ,

were conducted to determine the magnitude of any ground contamination that could have occurred. Samples of groundwater from the plant we!!s, the reactor caisson sump, and two of three test wells did not reveal signs of contamination. One test well drilled north of the ,

spent fuel storage pool (between the pool and the bay) revealed evidence of contamination, but the levels were a factor of 100 below allowable drinking water limits. The test wells have been monitored regulariy since that time and results of the surveillance have indicated no increase in activity.

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2.3.3 Spills of Contaminated Water On several occasions during the operation of Unit 3, radioactively contaminated liquids were spilled in certain areas of the facility. Since access to most areas of Unit 3 is controlled for purposes of contamination and radiation exposure control, the corrective action was to clean up the spill and either decontaminate the area or fix the contamination so that exposures required either for decontamination or resulting from the contamination would be consistent with ALARA considerations. During the SAFSTOR period, any residual contamination resulting from these spills will continue to be contained. Final decontamination of these areas to levels acceptable for unrestricted use will be accomplished as part of the final dismantlement program.

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2.3.4 Dropped Fuel Assembly f in 1975, a fuel assembly was dropped into the spent fuel pool cask loading pit, and several fuel rods separated from the assembly. A special container was fabricated to contain the assembly. The assembly and the loose rods have been retrieved and stored in the  ;

container in the spent fuel storage pool fuel storage racks.

2.4 OPERATING RECORD i

buring the period August 1963 to July 1976, Unit 3 generated over 4.7 billion kilowatt-hours of electricity and had a cumulative availability factor of 85.9 percent.

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i 3.0 CHARACTERIZATION The characterization of Unit 3 includes descriptions of the site and the facility, a summary of radiation dose rates and surface contamination levels, and an estimate of the radionuclide inventory at the unit.

l 3.1 SITE DESCRIPTION Humboldt Bay Power Plant is located about four miles true southwest of the city of Eureka, Humboldt County, Califomia, and consists of 142.9 acres of land. The nuclear plant (designated Unit 3) is located alongside two fossil-fueled power plants (designated Units 1 and 2). Two Mobile Emergency Power Plants are also located on the site. These are gas turbine units designated as MEPP No. 2 and

3. The five electric generating-units on this site comprise the Humboldt Bay Power Plant. A site plan is shown in Figure 3-1.

3.1.1 Topography Terrain of the site varies from submerged and low tidalland, protected by dikes and tide gates, to a high precipitous bluff along the southwestem boundary. Elevations range from approximately -3 feet to .

+75 feet based on a datum of the mean lower low water (MLLW) level. The ground floor of the ,

r; fueling building is at elevation +12 feet.

l 3.1.2 Soils and Geology HBPP lies in the Northem Califomia Coast Ranges geomorphic province. This province consists of a i system of longitudinal mountain ranges (2000 to 4000 foot elevations with occasional 6000 feet peaks) and valleys with a trend of N 30 degrees to 40 degrees W.

The immediate vicinity of the site consists of sand and alluvial soil and strata of the Hookton and

( Carlotta sedimentary formations. These formations are primarily consolidated sands, gravels, and

clays and conglomerates with good engineering properties. HBPP buildings have their foundations in l thtse strata.

l l The principal rocks in the area range in age from late Jurassic to eariy Upper Cretaceous. These rocks I tre in two groups:

o Clastic sedimentary rocks, consisting of sandstone, mudstone, and conglomerate o Volcanic and associated rocks, consisting of greenstone, basalt, chert, and minor amounts of limestone l 3-1 Rev 2 Maal 92%

In the site area, younger rocks overlie the volcanic strata. These rocks are dominantly marine sandstone,*mudstone, and conglomerates ranging in age from the late Cretaceous to early Pleistocene. Recent alluvium forms the shallow strata in the valleys and in areas along the coast.

3.1.3 Hydrology 3.1.3.1 Surface Hydrology The surface runoff from the site is directed into drains discharging into the plant cooling water intake canal, through the plant, and into Humboldt Bay - the discharge canal. Outside the area served by ,

the plant drain system, surface runoff drains int' c3uhne Slough, the natural drainage for the area, which drains into Humboldt Bay.

The nearest streams to the site are Salmon Creek and Elk River, which are within a mile south and l

north of the site, respectively, and which discharge into Humboldt Bay. These streams are used for '

watering livestock, but are not used as a potable water supply.

The Mad River flows west approximately 13-15 miles northeast of the site. The Ruth reservoir, the source of the city's water supply, is located on this river.

To the south, the Eel River discharges to the Pacific Ocean 8-10 miles from HBPP. This river is not used for potable water within 25 miles of HBPP.

3.1.3.2 GroundwaterHydrology Groundwater supplies all domestic, industrial, and agricultural needs in Humboldt County except that which is supplied by the Ruth reservoir. A groundwater study made in the area of HBPP prior to Unit 3 construction (Morfiave,1960) identified the following important features of the groundwater system:

o Movement of all groundwater is generally toward the bay.

i o Vertical rates of groundwater movement in the area of the plant are a few inches per day in the light surface alluvium. I o Horizontal movement in aquifers beneath the site range from several feet to hundreds of feet per day. ,

o Groundwater elevation in the area near the bay is similar to sea level and may be somewhat affected by tidal action. This elevation is approximately 12 feet below the plant floor elevation. l Both a groundwater and slight topographic divide appear to exist between HBPP and Elk River. These features reduce the probability ofliquid discharges or leakage from the plant site to this stream either by surface or groundwater flow.

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Southwest of the plant, an area exists which has slight landward groundwater gradients under some conditions. However, this area lies within an area that is affected by tidal action. Negligible inland flow is cstimated to occur.

Any migration of materials of plant origin into the soils beneath or near the plant would move vertically quite slowly until reaching the saturation zone. Migration would then be horizontal, toward the bay.

3.1.3.3 Humboldt Bay Humboldt Bay is a tidal bay receiving and discharging ocean water through its inlet. Very little fresh water discharges into this bay.

1 A study of tidal hydrology in Humboldt Bay has been made (Hazards Assessment Report,1960). The purpose of this study was to determine the flow pattem of tidal currents in Humboldt Bay, dilution of the l cffluent from the plant, and the flushing action of the tides by movement in and out of the bay. The i study concluded that the discharge of effluents into Humboldt Bay would result in a gradual dilution as l th:,y moved into the bay. Dilution of effluents along the shore of the bay entrance is high because of l l

th3 relatively drastic changes in depth for each tidal cycle. The swift moving water in the deeper l channels leading from the North Bay and South Bay causes rapid dilution. The ebb tides carry most of ,

thm discharged water out to sea and bring in water from the sea on the following tide. l Th3 finished grade elevation for the plant was established at +12.0 feet to be above the U.S. Coast end Geodetic Survey estimate of the highest high tide of +9.5 feet.

l 3.1.4 Seismology l

Th:re have been numerous geology and seismology studies conducted for the site with respect to the cff: cts of potential seismic events in the area. These studies are analyzed in Appendix 10.3 to the Environmental Report.

3.1.5 Climatology and Meteorology The climate at HBPP is mesic oceanic, characteristic of the northwestem coast of the continental United States. The area has two distinct seasons differentiated by precipitation rather than temperature. The wet season extends roughly from November through March and yields cpproximately 75 percent of the average annual precipitation. The dry season, extending from May through September, contributes only 10 percent of the average annual precipitation. The transitiona!

months, April and October, contribute the balance. The mean annual precipitation is 39 inches.

The range of air temperatures is minimal, averaging 52*F annually,46*F in winter and 56*F in summer.

The prevailing wind direction is from the north. The wind distribution is 24.3 percent offshore,57 p:rcent onshore, and 18.7 percent light and variable. Average wind speeds are strongest for the north l 3-3 Rev 2

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winds (16 mph) and the southeast winds (12.5 mph) during the wet seasons. These are lower during the dry season. During the rainy seasons, the wind from the south-southwest dominates slightly.

Prevailing winds can be expected to carTy airtome effluents from the plant south and inland 55 percent of the time. Approximately 20 percent of the effluents would be distributed across the bay entrance to the ocean.

Approximately 25 percent of the effluents would be discharged into calm air and distributed randomly.

3.2 FACILITY DESCRIPTION HBPP is comprised of two fossil-fueled units (Unit 1 - 52 MWe and Unit 2 - 53 MWe), a single nuclear  !

unit (Unit 3 - 63 MWe), and two gas turbine-powered mobile emergency power plants (MEPP No. 2 and 1 3 - 15 MWe each). Necessary support structures, equipme% and tanks are also located on the plant site. A site plan is shown in Figure 3-1. The following description is specific to Unit 3 unless otherwise ,

stated, l 3.2.1 General Plant Description Unit 3 consists of a General Electric natural circulation, single cycle boiling water reactor, the associated turbine-generator, and necessary support and auxiliary systems. The reactor vessel is a 10 fret diameter,40.5 feet long pressure vessel that is suspended in a drywell containment vessel.

The reactor primary containment is located entirely below grade and is comprised of the drywell vessel, l which houses the reactor, and a suppression chamber located concentrically around the drywell. The drywell and suppression chamber are located inside a reinforced concrete caisson which, in the vicinity of the reactor, is approximately 60 feelin diameter and extends to an inside depth of 78 feet below  ;

grade. A caisson access shaft extends from the top of the caisson to the space beneath the drywell.

The access shaft contains the reactor auxiliary systems.

The refueling building encloses the space above the caisson. In addition to the' reactor caisson, the refueling building contains the spent fuel storage pool and the new fuel storage vault.

Adjacent to the refueling building is the power building and turbine pedestal. The power building contains the condenser, feedwater and condensate systems, and the steam cycle auxiliary systems.

The control room is also contained in this building. The turbine-generator is located on the turbine pedestal. The turbine is contained in a shielded enclosure while the generator and associated exciter

! are located outside, l

A hot machine shop / calibration facility is located southeast of the power building. The machine shop is used for maintenance of radioactively contaminated equipment. The calibration facility contains sources used to calibrate radiation survey instruments.

Liquid wastes are processed in the radwaste treatment building, located in an excavated portion of an earthen embankment north of the refueling building. This building contains two concentrated liquid l

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radioactive waste storage tanks and the resin disposal tank and is surrounded on three sides by earth.

A steel building encloses the entire liquid radwaste treatment area. North of the radwaste building are l

! three high-level solid radioactive waste storage vaults, a low-level waste storage building, and a low-Isvalwaste handling building.

3.2.2 Plant Structures l Th3 plant structures are shown in Figures 3-2 through 3-18, which provide details of plant layout and equipmentlocations.

3.2.2.1 Foundations l

Ths power building and turbine pedestal are supported on a 3 feet 6 inches thick continuous concrete mat foundation resting on a grid of 30 ton timber pilings penetrating to the sand strata at elevation -24 f::t.

Ths number of pilings along the north edge of the mat was increased to carry the concentrated edge load of the power building.

Ths refueling building is supported on the reactor' caisson by six 100-ton H-piles driven into the same strata which supports the caisson. The reactor caisson also serves as the ventilation stack foundation by providing the projecting stack support bracket. The few equipment foundations and column foundations not resting on the power building structure or mat foundation are supported on spread footings below grade. The radwaste treatment building, hot machine shop / calibration facility, condensate tank, and radwaste storage and handling buildings are supported on grade slabs or footings.

3.2.2.2 Power Building I The power building is a monolithic, reinforced concrete, two-story structure which houses and provides radiation shielding as required for the main condenser and its auxiliary equipment, the makeup d:mineralizer system, the condensate demineralizer system, reactor feed pumps, access control, monitoring, and laundry facilities.

The control room is located on the +27 foot level and is enclosed by the refueling building, the turbine

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shield wall, and a structural steel framing and roof system. An extension north of the control room housing the instrument shop is of concrete block construction with a concrete roof.

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3.2.2.3 Tur.bine Pedestaland Shielding The turbine-generator is supported at the +27 foot level by a massive concrete pedestal composed of longitudinal and transverse rigid frame bents and shielding walls bearing on the foundation mat. The pedestal is separated from the power building by a joint for vibration isolation.

The turbine is enclosed by massive precast concrete shield walls and roof elements which are sized for removal by the powerhouse crane to permit turbine overhauls.

3.2.2.4 Refueling Building The refueling building is 43 feet x 103 feet x 35 feet high, constructed of reinforced concrete walls with a composite roof of precast, prestressed concrete double tee sections and concrete topping. The 12-inch minimum thickness walls and roof were designed to provide containment and shielding.

The building encloses the refueling area consisting of the spent fuel pool, new fuel storage vault, drywell vessel opening, and shipping cask railroad spur. The deactivated spur enters the building through pneumatically operated 10 ton concrete shielding doors, which, when open, provide a 15 feet 0 inches x 15 feet 0 inches clear opening. The building wall pilasters support the structural steel girders for the 75-ton bridge crane required for handlin'g the reactor shield plug and the spent fuel transfer cask, and other heavy loads in the refueling building. Personnel access to the ground flooris through aitiocks entering from the +27 and +12 foot levels. Access to the reactor caisson is through the refueling building.

3.2.2.5 Reactor Caisson i

The reactor caisson consists of a reinforced concrete structure,59 feet 6 inches in diameter,78 feet 0 l Inches inside depth, which houses the reactor vessel and auxiliary equipment, the drywell. vessel, j suppression chamber, and the rectangular portion described below containing the spent fuel storage

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- pool and new fuel storage vault.  !

The drywell vessel is centrally located in the caisson and serves as the primary containment vessel.

Surrounding the drywell vessel is the 300 degree annular suppression chamber (12 feet 6 inches wide 4 by 49 feet 0 inches high with a 4 feet 0 inches exterior wall and a 3 feet 10 inches interior wall).

The suppression chamber is constructed of reinforced concrete and lined with carbon steel plate. Six 40-inches-diameter vent pipes connect the drywell to a common 40-inches equalizing ring header at the top of the suppression chamber. From the header bottom,46 evenly spaced 14 inches pipes extend to a point 6 feet below the normal operating water level. There were baffle plates between the vent pipes and deflector plates in front of the six 40 inches entrance vents. The baffle plates were -

removed during SAFSTOR. An access was created at the -66 foot elevation into the West chamber and between the East and West chambers to facilitate the removal of the baffle plates.

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I l Collectively, the drywell, the suppression chamber, and the vent piping comprised a pressure  ;

suppression containment system surrounding the reactor during plant operations. ,

A 60-degree portion of the circular caisson below elevation -14 feet serves as the access shaft and houses principally the control rod hydraulic system and instrument vaults. Access to the bottom of the j

drywell and the various levels in the access shaft is provided by e. vertical manlift, 2-ton jib hoist, and .

emargency ladder. In addition, a sealed vertical escape shaft with four access doors in the access shaft provides emergency escape to outside the refueling building.

ThD caisson abovs elevation -14 feet and up to grade at elevation +12 feet is rectangular (49 feet 0 inches wide x 75 feet 0 inches long) and serves as the stmetural foundation for the Refueling building i and 'stack projection. The drywell versel and biological shield continue up to grade and serve as a

cIntral support pier for the floors at elevations -14 feet, -2 feet, and + 12 feet.

A spent fuel storage pool is provided for (1) storage of spent fuel assemblies; (2) removal and inspection of fuel assembly channels; and (3) underwater loading of the spent fuel shipping cask. The spent fuel storage pool is approximately 22 feet wide by 28 feet long. The pool depth is 26 feet deep except for the cask loading pit in the southeast comer, which is 36 feet deep. The pool is constructed of reinforced concrete and has a stainless steel liner. A motor-operated movable service platform with i l

L a 500-pound hoist is provided for handling fuel assemblies in the pool. l l

l Storage racks are provided to contain spent fuel. iA total of 89 rack assemblies are provided, and 44 of thsse racks can each contain eight fuel assemblies or two control rod assemblies. The remaining 45 racks can each contain three fuel assemblies. One location cannot be used because of a bolt protruding into the bottom of the location. This gives a total capacity of 486 fuel assemblies. Presently, l thsre are 390 fuel assemblies stored in these racks.

Tha channel stripping machine for removing irradiated channels from irradiated fuel assemblies is-mounted in the pool. An irradiated channelinspection stand is installed on the stripping machine to -

pilow underwater examination of an irradiated channel. ,

7 A new fuel storage vault used for storage of new fuel is located adjacent to the spent fuel storage pool.  ;

- The vault measures 16 feet long,12 feet 6 inches wide, and 12 feet deep, and contains racks for storing 96 fuel assemblies in six rows of 16 each. Also contained in the vault are the fuel storage pool  !

coolers and access hatches to the turbine building drain tank vault below. The top of the new fuel storage vault is at the refueling building floor level, and the vault is normally covered by %-inch thick ch:ckered steel plates that are hinged for access to the vault.

The base of the caisson was sealed with tremie concrete at elevation -66 feet undemeath the drywell and suppression chamber, and at elevation -14 feet and -24 feet under the spent fuel pool. A 6-inch pervious gravel blanket on the tremie and below the 6-inch floor slab was designed to collect seepage and to prevent any pressure buildup.

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3.2.2.6 Ventilation Stack The 250-foot-high ventilation stack is of reinforced concrete construction complete with aircraft waming system, ladder and platforms for servicing the automatic aircraft waming lights, and stack gas monitoring facilities. The base of the stack serves as the enclosure for three floors of gas treatment cquipment.

3.2.2.7 Radwaste Treatment Facilities The radwaste treatment building is recessed into the hill north of the refueling building. It consists of a 37 x 96 foot slab at grade with a rear retaining wall, wing walls, tank and equipment vaults, and an cnclosed control room. All walls and roof slabs are of monolithic reinforced concrete. The slab at grade provides support for eight liquid waste tanks; five are not vaulted but within the LRW enclosure, and the other three are housed in shielded vaults. The solid waste vault is an underground reinforced concrete vault with a capacity of 1,200 cubic feet. The vault is located on top of an earth bank directly north'of the radwaste treatment building. The top of the vault is at ground level. The interior dimensions are 20 x 8 5 x 8 feet deep. Two interior walls are provided that divide the vault into three equal compartments. Three reinforced concrete roof slabs are designed to overiap and interlock with the walls to prevent entry of rainwater.

North of the solid waste storage vaults is the low-level waste storage building. The building is of concrete block construction and is divided into two sections, one for storage of low-level solid radioactive waste awaiting disposal and the other for storage of contaminated reusable tools and cquipment.

North of the low-level waste storage building is the low-level solid waste handling building. The handling building is a prefabricated metal building that consists of a 30 x 40 foot waste handling area and a 30 x 50 foot covered truck loading area. The building provides weather-protected storage for cmpty radioactive waste packages (drums and boxes) and packages awaiting shipment.

3.2.2.8 Yard Structures The yard structures consist of the hot machine shop /cclibration facility and the off-gas treatment vault.

The hot machine shop / calibration facility is constructed of reinforced concrete block masonry bearing w;lls, structural steel framing, and corrugated roof decking. The ground flooris a reinforced concrete slab on 6 inches of rock. The roof decking is welded to the stnJctural steel framing to provide diaphragm action for resisting lateral loads.

The off-gas treatment vault was constructed in 1976 to contain a new off-gas treatment system. It is a monolithic, reinforced concrete structure. After construction, neither the building nor the contained equipment were used since the plant was not retumed to power operation.

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3.2.2.9 Intake and Discharge Structures The intake structure is a compartmented, reinforced concrete box with overall dimensions 53 feet long x 22 feet wide x 24 feet deep, similar to adjacent Units 1 and 2 structures. The top slab of the structure is ct elevation 12 feet 0 inches.

The intake structure is separated by two cross walls, one between the bar racks and traveling screens and the other between the screens and the pumps. These crosswalls have gated openings. The structure is designed to be stable against uplift at high tide with the gates closed and the compartments l dewatered. The walls and floor design also provide for resistance to the pressures existing under l

th:se extreme conditions.

A precast concrete deck is provided between Units 2 and 3 intake structures for access with a 25-ton capacity truck crane for maintenance of the bar racks and traveling screens. The Unit 3 screen wash troughs extend across this dec'k and discharge into the Unit 2 sump pit.

The discharge structure is a gated reinforced concrete box located at the upstream end of the discharge canal and forms the end anchor of the cooling water discharge line. The structure is open at tha top, providing access to the pipe when the gate is closed.

Both the intake structure and discharge structure,were provided with embedded sheetpile sections to cllow future expansion immediately adjacent to the structures.

3.2.2.10 Seismic Upgrading Structures at Unit 3 . vere upgraded to provide additional protection against earthquake damage.

These modifications primarily consisted of the addition of structural steel and piping supports.

3.2.2.11 Onsite Combustible FuelStorage l The description of combustible fuel storage facilities at the HBPP is given in Table 3A.

t 3.2.3 Plant Systems Description l For this DSAR, the plant systems are addressed in six major system groupings that are functionally oriented. These major groupings are the nuclear steam supply system, the turbine plant systems, instrumentation and control systems, service systems, waste disposal systems, and electrical systems.

The systems and major components comprising each of these major system groupings are identified in Tcble 3-1. In addition, a brief indication of the physical status for the SAFSTOR period is provided.

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l 3.2.3.1 Nuclear Steam Supply System 1

l As shown on Table 3-1, the nuclear steam supply system (NSSS) consists of the reactor vessel and l

intemals, control rod system, liquid poison system, reactor cleanup system, reactor shutdown cooling 1 system, emergency condenser system, and the suppression tank cooling and core spray system. '

Reactor Vessel and Intemais. An isometric view of the reactor vesselis shown in Figure 3-19. The reactor vessel is a vertical cylindrical shape,40.5 feet in length with a 10 foot inside diameter. The vesselis high strength carton steel alloy, ASTM A-302, Grade B, fire box quality, with an interior l surface clad with 304 stainless steel deposited by the weld deposit overlay method. The vessel wall  !

thickness (including cladding) ranges between 4 and 5 inches. A bolted closure head is located on top of the vessel and provides access for refueling. Table 3-2 lists miscellaneous reactor data.

The vessel contains and supports the reactor intemal components, coolant, and fuel assemblies. The intemal components are:

o Core support assembly o Cole spray ring o Feedwater sparger o Steam dryer o Control rods o Control rod guide tubes o Lower core shroud o Upper core shroud l o Upper guide l The steam dryers, which are a screen type design, occupy the cross-section of the reactor vessel i closure head. They are inclined to promote drainage and are removed with the reactor head. l All fu~el assemblies were removed from the reactor in early 1984. In addition, the incore strings containing the fission chamber flux monitors were also removed. The vessel was subsequently vacuumed to remove residues from the bottom of the vessel and the closure head was reinstalled.

The nuts on the closure head were installed hand-tight. A test was conducted with the water level in the vessel lowered to the bottom of the vessel to identify conditions which will exist during dry 1: yup. Table 3-3 tabulates the results of this test. The reactor vessel will remain drained during the SAFSTOR period.

Control Rod Hydraulic System. The control rod hydraulic system provided the hydraulic motive i power for control rod positioning. The system consists of control rod drive mechanisms (located exterior to and below the reactor vessel), two supply pumps, accumulators, a scram dump tank,

(. clong with necessary controls, valves, filters, and interconnecting piping. This system is deactivated and no future use is anticipated.

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To reduce radiation dose rates in areas requiring routine access during SAFSTOR, several components of this system have been removed. These components include hydraulic filters at the l -14 foot level, the scram dump tank, the level pots on the scram dump tank, and associated piping at the -66 foot level of the access shaft.

Liould Poison System. The liquid poison system served as an emergency backup to maintain the r: actor subcritical if the control rod drive system failed to insert a sufficient number of rods into the core.

Tha system includes a poison tank that contained approximately 385 gallon of sodium p:ntaborate solution. The tank is constructed of carbon steel and has a corrosion-resistant int:rior lining of a baked phenolic resin. The tank is located beside the north wall of the refueling building at the +12 foot elevation.

Th3 liquid poison tank is drained and the system depressurized.

Reactor Cleanuo System. The reactor cleanup system was provided to remove corrosion products from the reactor vessel water. The system consists of a cleanup pump, two regenerative h:at exchangers, a nonregenerative heat exchanger, a demineralizer, and a resin storage tank.

The reactor cleanup system has been drained and deactivated. No future use is planned. Resins have been removed from the demineralizer.

The cleanup pump and associated piping at the -66 foot level were a significant source of radiation to workers in the area. To minimize radiation exposures for ALARA considerations, the cl:anup pump and portions of the system piping were removed after the system was no longer rcquired to be operational. Caps were installed to contain contamination.

Rnactor Shutdown Coolina System. The reactor shutdown cooling system was provided to remove decay heat following a reactor shutdown when reactor pressure was less than 120 psig and during rcfueling operations. Cooling water to the system heat exchangers was supplied by the closed cooling water system. The reactor shutdown cooling system consists of two hsat exchangers, two pumps and necessary connecting piping, valves, instrumentation, and controls.

The equipment in this system is located in a shielded room at the -14 foot elevation on the west side of the caisson. The system has been drained and secured. Several sections of pipe in the cccess shaft have been flushed and/or removed due to ALARA considerations. The system has be:n isolated macnanically (by appropriate valving) and electrically.

Emeroency Condenser System. An emergency cooling system was gcVided to dissipate reactor d: cay heat following a reactor scram, if normal heat sinks were unavai'able. This system was d signed to maintain the reactor in a safe condition when all normal plant auxiliary power had l be:n lost. The system could be placed in operation either automatically following a reactor scram or manually by remote opening of one motor-operated valve. The arrangement of the system was such that steam generated in the reactor would be condensed in the emergency condenser and l 3 11 Rev 2 August 1998

l the condensate returned by gravity to the reactor. The system consists of an emergency condenser, an emergency make-up pump, and associated piping and valves.

The emergency condenser is located on the north side of the refueling building with a centerline at I clevation +28 feet over the spent fuel storage pool. It is supported in this position by a steel foundation on each end mounted to the east and west sides of the pool. The emergency makeup pump was located in the yard between the condensate storage tank and the refueling building, but has been removed.

Supply and return lines from the reactor have been cut and capped, and the heat exchanger shell end tubes have been drained.

Core Sorav and Sugession Pool Coolina. The core spray system was available to supply cool?ng water to the 'm for core in the event of a loss of coolant accident. The system could take suction from the suppression pool and discharge the water to a core spray ring in the reactor vessel where it would be distributed over the core. A heat exchanger on the discharge side of the pumps was available to cool the suppression chamber water. During normal operation, the heat exchanger was used to maintain suppression chamber water temperature within technical specification limits.

The core spray and suppression pool cooling system consisted of two pumps located at the -66 l foot level, a heat exchanger located at the -2 foot level, and necessary piping, valves, instruments cnd controls. The system, including the suppression chamber itself, has been drained and isolated. The two core spray pumps and associated suction and discharge piping at the -66 foot I:; vel have been removed. No further use of the system is planned.

Core spray system piping sections in the access shaft at the -14 and -66 foot levels were identified as a source of high radiation. Since routine access through these areas will be required for decommissioning activities and during SAFSTOR, this piping has either been flushed or sections have been removed and the ends were capped to contain internal contamination.

3.2.3.2 Turbine Plant Systems The turbine plant systems consist of the turbine system, the lube oil system, the condensate ,

system, the gland seal system, the feedwater system, the closed cooling water system, the main circulating water system (saltwater), the compressed air system, and the demineralized water system.

Turbine System. The turbine-generator unit is a tandem, compound, double flow, condensing turbine, direct connected to a 13,800 volt,3-phase,60-cycle, hydrogen-cooled synchronous generator. The turbine is a non-reheat, condensing machine with three extraction pcints for f:edwater heating and designed for use with a boiling water reactor, it consists of a single flow high pressure section and a double flow, low pressure section with a crossover pipe connecting the two sections. The two units are mounted in tandem. A valved, turbine bypass line was 3-12 l Rev 2 nwaucurem

l l

provided which could dump steam straight to the condenser in lieu of passing first through the turbine. The turbine is in a layup state and a nitrogen blanket was used to fill the turbine internals.

l Lube Oil System. The turbine lube oil system for Unit 3 is a complete and self-contained system  !

consisting of a reservoir, pumps, coolers, bowser filter, and necessary piping.

The main and auxiliary pumps delivered high pressure oil to the turbine hydraulic control system and cooled, low pressure oil to the bearing header which supplies the unit bearings and other parts to be lubricated. Backup protection to the bearing oil system was provided by the tuming g ar oil pump and the DC emergency oil pump. The bearing oil was drained to a common header end was retumed back to the reservoir. A related system, the hydrogen and seal oil system, received its oil supply from this system. The lube oil system has been drained and isolated from other systems. No further use of this system is planned.

~

Main Condenser. The condenser is a single-pass, horizontally divided waterbox, deaerating-type unit with an effective surface of 30,700 square feet and generally of conventional construction.

The condenser is connected to the turbine by an elbow-shaped connecting piece. Standard construction materials were used, including fabricated carbon steel shell and tube support plates end unlined close-grained cast iron waterboxes. Tubes are 7/8-inch outside diameter 18 BWG aluminum brass. Tube sheets are silicon bronze.

The condenser is located alongside the turbine pedestal with tubes parallel to the turbine c:nterline. The low pressure feedwater heater is installed within the condenser connecting piece.

A 6,500-gallon oversized storage-type hotwell was provided to allow decay of short-lived radioactivity. The hotwell is divided by a partition plate parallel to the tubes to facilitate location of tube leaks. /

Tho condenser contained a significant level of internal surface contamination. Internal d: contamination of the condenser has been performed to reduce these levels. In addition, l s:ctions of the 10 inch condensate piping between the condenser and the LP heater which contained significant intemal contamination have undergone decontamination to reduce radiation lavels. The system has been drained and secured.

Condensate Storace Tank. The condensate storage tank is located outside the north side of the refueling building. It is a 34,000 gallon not capacity,15 feet in diameter by approximately 29 feet high, cone-roof, vertical cylindrical storage tank of aluminum construction. An intemal head s:parates the tank into an upper 5,000 gallon section (designated the demineralized water storage tank) for fresh demineralized water and a lower 29,000 gallon section for condensate.

Tha condensate storage tank section has been drained. The demineralized water storage tank l

remains operational.

, Condensate System. The condensate system is composed of two condensate pumps, a l contaminated drain tank and pump, two low pressure feedwater heaters, three half-capacity l

condensate demineralizers, and necessary piping, valves, instruments, and controls. The system l 3-13 Rev 2

_ - _. _ _ _ _ _ _ __.-----------__--------__ A M e l l M - _- ---- _ _

has been drained and flushed. As part of a program to reduce radiation exposures for ALARA considerations, portions of the condensate system have been decontaminated. The system is presently isolated.

Condenser Air Removal System. The condenser air removal system is comprised of a condenser vacuum pump, two sets of air-ejectors, and their associated condensers. This system has been drained and isolated. Due to ALARA considerations, selected sections of piping have been removed to reduce radiation exposures in the area. The resulting open piping has been sealed.

Condensote Demineralizer Resin Receneration System. A system for regenerating condensate demineralizer resins was originally installed and used during the early years of Unit 3 operation.

Use of the system was discontinued and the equipment was abandoned in place, it consists of three tanks and associated piping, valves, and controls located in a shielded area of the condensate demineralizer rogm. One of the tanks was previously converted for use as a laundry waste hold tank, but is no longer used for this purpose and has been isolated. The remainder of the tanks have been drained, flushed, and are presently isolated. No future use of this system is planned.

Gland Seal System. The gland seal system consists of a gland seal condenser and two gland seal exhausters. This equipment is located on the ground floor of the station building in the air-cjector room. Its function was to remove the mixture of air and steam from the turbine seals.

Since no future use of this system is planned, it has been isolated and hot spots have been removed.

Feedwater System. The feedwater system consists of two feedwater pumps (or reactor feed pumps) and necessary piping, valves, instruments, and controls. The feedwater pumps are located in the feed pump room, which is on the ground floor (elevation +12 feet) of the power building. This system has been drained and secured.

Closed Coolina Water System. The closed cooling water system consists of a cooling water return tank, two cooling water pumps, two cooling water heat exchangers, and intarconnecting piping and valves that supply cooling water and seal water to a variety of coolers and pumps.

Water contained in this system was treated with potassium chromate to inh; bit corrosion.

The return tank is located below grade and immediately outside the east side of the station building. The pumps are located above grade on the outside of the building above the tank and on top of the vault. The heat exchangers are located outside the building along the southeast wall of the power building. The system has been drained, flushed, and isolated. No future use is planned.

Main Circulatina Water System (or Salt Water Systemt The main circulating water system provided salt water cooling to the main condenser, the cooling water heat exchangers, and the suppression pool cooler.

3-14 l Rev 2 fr.wrwRM

Tha system contains an intake structure that housed a trash rake, two parallel traveling screens, two screen wash pumps, three slu'ce gates (including operators), two circulating water pumps, and associated piping, valves, and instrumentation. The system also includes intake piping to the main condenser, discharge piping and isolation valves from the main condenser, piping, valves, and instrumentation to the above noted heat exchangers, and a discharge structure located at the discharge canal. This system has been secured. The traveling screens, screen wash pumps, and circulating water pumps have been removed to prevent deterioration.

Comoressed Air System. The compressed air system was designed to provide service air and instrument d.Gisgbnut the plant. The system consists of the instrument air compressor, the eft:rcooler, instrument airieceiver, service air receiver, instrument air dryer, and filters. Headers era provided for both seruce air and instrument air, and these headers are interconnected with equivalent headers in U iits 1 and 2. Normally all compressed air for Unit 3 is cupplied through this interconnacticn; wnh the Unit 3 compressor maintained in a standby mode. The system is located la the reactor feed pump room. The service air and instrument air systems are presently in service and will remain in service during the SAFSTOR period.

D-mineralized Water System. The demineralized water system was originally designed to provide all demineralized water for Unit 3 operations. Except for the demineralized water storage tank and the demineralized water pump, this system has been abandoned in place with demineralized makeup water supplied instead by the Units 1 and 2 evaporators. Makeup water is transferred directly from one of four condensate storage tanks in Units 1 and 2 to the Unit 3 demineralized water storage tank from which it is distributed by the associated pump. The remainder of the system has been drained and isolated.

3.2.3.3 Waste Disposal Systems The waste disposal systems in Unit 3 include the gas treatment system, the liquid waste collection system, the liquid waste treatment and disposal system, and the solid waste facilities. Collectively thsse systems control and dispose of all plant wastes that are normally or potentially ~

contaminated with radioactive materials.

l Gas Treatment System. The gas treatment system (GTS) consists of two exhaust fans, a gas scrubber column, gas scrubber recirculation tank, two recirculation pumps, and associated system piping, valves, filters, instruments, and controls. The system was intended to remove halogen gases from the refueling building in the event of fuel damage and a resulting release of fission product gases. The system components are located on three levels in the base of the main v:ntilation exhaust stack.

Due to the age of the spent fuel stored on site, the inventory of radioactive halogens is sufficiently low to preclude the need for the gas treatment system. The sodium hydroxide solution from the gas scrubber recirculation tank has been drained and the system secured.

l- 3-15 Rev 2 August 1998

1 Off-Gas System. The off-gas system functioned to receive the non-condensable gases from the main condenser air-ejectors and to delay the release of the gases to permit decay of the short-lived radionuclides. The system consists of a buried holdup pipe and a HEPA filter that contained in a below-grade vault near the ventilation exhaust stack. This system has been abandoned and the off-gas filter removed. ,

l A modification to upgrade the off-gas system was under construction when Unit 3 was shut down in 1976. The new system is in the off-gas treatment vauit located north of the ventilation exhaust stack. The system was never used since Unit 3 did not return to operation. Since this equipment l i.s not contaminated, it will be left as is or removed for salvage.

Liould Waste Collection System. The liquid waste collection system consists of the turbine ,

building drain tank, reactor equipment drain tank, reactor caisson sump, the turbine building floor  !

drain pump, two turbine building drain tank pumps, two reactor equipment drain tank pumps, the (

reactor caisson sump pumps ~ the laundry waste tank, the laundry hold tank, the laundry waste l filter, two laundry waste pumps, and a vent separator, The turbine building drain tank (TBDT), turbine building floor drain pump, and TBDT pumps are I located at elevation -14 feet in the reactor caisson in a shielded vault beneath the new fuel storage vault. The vault is accessible via a ladder through a hatch in the new fuel storage vault.

The tank is pumped using the turbine building drain tank pump or can be valved to drain directly to the reactor equipment drain tank via the caisson floor drain system.

The reactor equipment drain tank (REDT) and associated REDT pumps are located at the -66 foot ,

1: vel of the reactor caisson access shaft. The contents of this 500 gallon capacity tank are I pumped automatically to the radwaste treatment system using either of the two REDT pumps.

The reactor caisson sump and its associated reactor caisson sump pumps are located at the -66 l foot level of the access shaft. The sump, which collects groundwater in-leakage, has a capacity of 50 gallons. The pumps normally transfers its contents automatically to the Discharge Canal, but l  ;

may be valved to the radwaste treatment system if groundwater contamination is suspected or ,

detected through routine samples. l The laundry waste tank is a 250-gallon tank located in the power building undemeath the laundry.

It is suspended from the underside of the operating floor slab (elevation +20 feet), and collects potentially contaminated drains from the decontamination area. The laundry waste tank l discharges to the TBDT.

j The laundry waste hold tank is the current designation for the 685-gallon regenerated resin j storage tank which was part of the condensate demineralizer regeneration equipment. This tank was.used to hold laundry waste tank volumes while awaiting sample analysis prior to discharge, but is no longer used for this purpose and has been isolated.

Liould Waste Treatment System. The liquid waste treatment system processes, stores, and provides for disposal of radioactively contaminated liquid wastes and other liquid wastes that are l 3-16 l l

Rev 2 i .

SgsugLgygL

l potentially radioactively contaminated. These wastes are first collected by the radwaste collection system and are then pumped to the radwaste building on the north side of the refueling building.

The system consists of the following major equipment:

• Radwaste Building Sump Tank e Radwaste Building Sump Pump e Radwaste Receiver Tanks (3) e Radwaste Pump e Concentrator Feed Pump e Radwaste Concentrator e Radwaste Concentrator Condenser e Radwaste Demineralizer e Resin Disposal Tank e Concentrated Waste Tanks (2) e Waste Hold Tanks (2)

• Treated Waste Pump e Radwaste Filters (2)

• Spent Fuel Pool Filter ,

l

• Concentrator Drip Receiver Tank and Pump in the radwaste building, wastes are handled on a batch basis with each batch being analyzed cnd handled appropriately in accordance with the analysis. Final disposition consists of storage cwait.ing offsite disposition, or disposal to the discharge canal which flows into Humboldt Bay.

There is no disposal to the ground.

Radwaste Buildino Sump Tank and Pumo. This 250-gallon tank is located beneath the radwaste building floor and receives liquids from drains in the vicinity of the radwaste building. The sump j pump is located on the operating floor of the radwaste building (elevation +12 feet) over the sump l trnk. This pump automatically maintains the level of the tank and discharges to one of the waste r:ceiver tanks.

Radwaste Receiver Tanks and Hold Tanks. Three 7,500-gallon carbon steel radwaste receiver i tanks are for wastes coming from the radwaste collection system. Two 7,500 gallon carbon steel l waste hold tanks are for storing treated wastes for retreatment or disposal. These tanks are located in an extemal section of the radwaste building, but are within the prefabricated steel radwaste enclosure.

Rydwaste Pumo. The radwaste pump is located in the radwaste building and takes suction from cny of the five receiver or hold tanks for the purposes of processing the wastes through various cquipment.

Radwaste Concentrator and Condenser. The radwaste concentrator was designed to concentrate 7,500 gallon per week. The concentrator consists of a vessel about 14 feet high and 24 inches in l 3-17 Rev 2 August 1998

l l

l diameter with a 40 square foot, callandria-type evaporating section near the bottom. Steam from the Unit 1 or Unit 2 auxiliary steam system is fed to the callandria outside of the tubes.

l Evaporation takes place within the tubes. The concentrator is located in a shielded cubicle in the  !

radwaste building.

I Concentrator vapor goes to a condenser which is cooled with water from an independent cooling loop, and the condensate goes to the drip receiver tank for collection for further treatment or l

disposal. The concentrated radwaste is discharged to one of the two concentrated waste storage tanks.

An independent cooling water pump circulates water between the condenser and a second heat exchanger which receives its cooling water from the Unit 2 bearing cooling water system. This I independent cooling water loop provides a radiological barrier between the concentrator and the i Unit 2 system supplying cooling water.

Concentrator Feed Pumos. A concentrator feed pump is located in the radwaste building, draws  :

suction on the receiver or hold tanks, and discharges to the concentrator. As an altemate, the r' adwaste pump can discharge to the concentrator.

Radwaste Demineralizer. The radwaste demineralizer is a single, mixed bed unit with a flow I capacity of 50 gpm. The demineralizer tank is 24 inches in diameter and was designed for 75 psig i in accordance with the ASME Code. There are no provisions for regeneration; spent resins are sluiced to the resin disposal tank. The demineralizer is located in a shielded cubicle in the radwaste building.

Resin Disposal Tank. This 10,000-gallon tank is located in an individual shielded vault within the radwaste building. It is accessed through a hatch in the top of the vault. All spent resins from the various demineralizers on site are routed to this tank.

Concentrated Waste Tanks. Two 5,000-gallon storage tanks are located in a shielded vault in the radwaste building. These tanks receive concentrated wastes from the concentrator. These tanks have no inherent means for draining and must be pumped down through access ports in the top of the tank. ,

Treated Waste Pumo. This pump is also located in the radwaste building and takes suction on the waste hold tanks. After sampling indicates that the contents of these tanks is within specifications, this pump is used to discharge the contents to the discharge canal. Alternate routings from this pump include (1) recirculation to either hold tank, (2) discharge to the condensate storage tank, or (3) recycle to waste receiver tanks for retreatment.

Radwaste Filters. Two radwaste filters are available in the radwaste building. These are cartridge-type filters, 50 gpm capacity, which can remove particles down to 25 microns in diameter.

(

l l 3-18 l l Rev 2 L Au1ust 1998

i Spent Fuel Storaae Pool Filter. An additional cartridge-type filter is located in the radwaste building which is dedicated to processing water from the spent fuel storage pool. This filter is similar in design but smaller than the radwaste filter. It was removed and shipped for disposal.

Concentrator Drip Receiver Tank and Pumo. A concentrator drip receiver tank is provided to collect the condensed vapors from the concentrator. The concentrator drip receiver pump either r3 circulates water in the tank for sample mixing purposes, or it discharges to the treated waste pump discharge header for final disposition.

Solid Radwaste System. The only equipment available for processing solid radwaste is a Compactor that provides higher density loading of 55-gallon drums than can be achieved by hand loading. It is presently located in the liquid radwaste enclosure.

3.2.3.4 Instrumentation and Control (l&C) Systems Unit 3 was provided with numerous I&C Systems to optimize plant performance, protect equipment from damage, protect plant operating personnel, and protect the public and the environment from harm due to accidents of a radiological or non-radiological nature. These systems and their status for the SAFSTOR period are outlined in Table 4-1.

3.2.3.5 Service Systems Domestic Water System. Domestic water to Unit 3 is supplied by the Units 1 and 2 domestic water system.

Fire Protection System. The fire protection system consists of three fire pumps with associated

~ piping, valves, instrumentation, and controls. Carbon dioxide lines to Unit 3 from the CARDOX system located in Unit 2 have been isolated. Removal of oil from the lube oil system eliminated th] need for this system in Unit 3.

Hnatina and Ventilation System. The plant heating and ventilation system consists of two main .

plant exhaust fans that exhaust from the refueling building to the ventilation exhaust stack, the multizone air handling unit, which supplies filtered air to the refueling building and selected areas of the power building, the drywell purge fan which ventilates the reactor-caisson access shaft, and s;veral small air handling units that ventilate selected areas of the plant.

Tha system flow capacity and flow paths are adjusted to optimize air flow for the existing plant condition.

i Spent Fuel Storace Pool Service System. The spent fuel storage pool service system consists of  ;

th) equipment necessary to maintain water level and quality in the pool as well as equipment nnded for handling and movement of spent fuel. Major components include the pool liner gap  ;

pump, two fuel pool circulating water pumps, two fuel pool coolers, channel handling tools, l

l 3-19 Rev 2 l August 1998 1

miscellaneous fuel handling tools, a jib crane, a transfer cask and winch, an extension tank, a"d the service platform.

Soent Fuel Storace Pool Liner Gao Pumo. This pump is located in a sump in the cask area at the bottom of the spent fuel storage pool. It takes suction on the gap between the fuel pool liner and the wall to maintain the water level below the groundwater level outside the building. Discharge is to the TBDT. The net effect is to maintain a head difference between groundwater outside the  ;

building and water in the liner, providing for preferential inflow leakage into the liner gap from outside. This minimizes leakage of radioactive contaminants to the outside of the building. i Fuel' Pool Circulatina Water Pumps. Two pumps are located on the ground floor (elevation +12 f et) in the refueling building adjacent to the hatch into the new fuel storage vault. These pumps circulate water from the spent fuel storage pool through the spent fuel pool demineralizer and strainer. , l Fuel Pool Coolers. The fuel pool coolers are located adjacent to the fuel pool circulating water pumps in the refueling building. Their function was to remove decay heat added to the pool water by the spent fuel. Due to the age of the spent fuel on site, decay heat is low enough that the coolers are no longer required.

Fuel and Channel Handlina Tools. These tools will be required for final removal of the spent fuel l from the spent fuel storage pool. They will be stored for the SAFSTOR period pending their need for this purpose.

Spent Fuel Pool Jib Crane. This crane is used for movement of spent fuel within the Spent Fuel Storage Pool. It is mounted on the Refueling Platform and will be required for final transfer of spent fuel to the shipping cask.

Transfer Cask and Winch. This cask and winch were used to transfer spent fuel from the reactor v;ssel to the spent fuel storage pool. It will be stored during SAFSTOR in the event that a shielded cask of this type is needed during final shipments of the spent fuel or during final l DECON decommissioning of the plant.

Extension Tank. While not directly associated with the spent fuel storage pool, the extension tank l is associated with refueling and is included in this section for accountability purposes. The extension tank was installed on the reactor vessel flange after closure head removal. It was then l filled with water for shielding purposes during spent fuel removal into the transfer cask, it will no longer be required and will remain in its stored position in the refueling building for the SAFSTOR l p:riod. Its use may be required during final DECON decommissioning of the plant at the end of the SAFSTOR period.

Spent Fuel Pool Service Platform. The spent fuel pool service platform moves north and south over the spent fuel storage pool and is used along with the spent fuel pool jib crane to move spent fuel within the pool. it will be required for final transfer of spent fuel to the shipping cask.

3-20 l Rev 2 August 1998

i Manfift. An electric operated manlift provides access between elevations -14 feet and -66 feet  !

within the access shaft of the caisson.

75-Ton Bridae Crane (or Refuelina Buildina Crane). This crane is supported at elevation 35 feet 9 inches in the refueling building. The crane is used to handle the reactor vessel head, fuel transfer cask, shipping casks, the service platform, and other heavy components within the refueling building. The crane bridge, trolley, and trucks are constructed of built-up steel members with w:Ided, riveted, and bolted connections. The bridge consists of two box girder sections spanning 41 feet between rails which are supported on built-up steel girc'ers spanning 20 feet between refueling building columns. A 10-ton capacity auxiliary hook provides additional range, speed, and simplicity for handling smaller loads.

Sanitary System. The existing sanitary system for Unit 3 will be maintained and remain functional during the SAFSTOR period. This system is common with the de sanitary system.

3.2.3.6 Electrical Systems The Electrical Systems at HBPP Unit 3 consist of (1) the main generator and its 13.8 kV system, (2) the 110 kV system (which originates from step-up of 13.8 kV), (3) the auxiliary power system which includes a 2400V system, a 480V system (including an emergency 480V system) and a 120/208V system, (4) the 125V DC system, (5) the preferred 120V AC system, (6) the annunciator system, and (7) the communications system. Since the main generator and reactor will not be operational at any time during the SAFSTOR period, the 13.8 kV and 110 kV systems will not be operational, and the l preferred 120V AC system control and instrumentation. will of These portions not thebe required overall to supply plant electrical regulated system will remain power for nu l

d: energized and electrically isolated and will not be further discussed in this report. The following s:ctions describe the remainder of the electrical systems.

2400V System. The 2400V system, as the higher voltage component of the auxiliary power system, provides power through step down transformers for the remainder of the auxiliary power system.

House Transformer No. 2 is the normal source of power for Unit 3. House Transformer No. 3 is out of s:rvice and electrically isolated. House Transformer No. 2 can be supplied by either operating Unit via the plant 60 kV bus, from the MEPPs or from either of two 115 kV transmission lines from the Cottonwood Substation via the Humboldt Substation and the plant 60 kV bus.

The 2400V system consists of a single distribution bus, with power from the bus supplied to two 2400/480V transformers. One of the transformers supplies power to Load Center (LC) No.10, and the  ;

other supplies power to LC No.11. Plant low voltage distribution originates at these two load centers with feeders to other motor and valve control centers, lighting and heating transformers, and power panels.

C:ntralization of distribution has occurred subsequent to reactor defueling, and additional centralization may occur during the SAFSTOR period. This reduces the number of centers, transformers, and panels that must be maintained, but does not affect the reliability of the power source for equipment important to safety. No changes have been orwill be made that would result in greater loads than have been l 3-21 Rev 2 August 1998

experienced during normal plant operation or which would result in less backup power availability for this equipment than has been available during normal shutdown periods for the plant in the past.

480V System. The 480V system consists of LC Nos.10,11, and 13, and Valve Control Center No.1.

LC Nos.10 and 11 (located in the control room) are supplied power from 2400/480V Transformers No. l 5 and 6, respectively. LC No.10 has a physically and electrically separate emergency section that provides power to Power Panel No.1. LC No.10 also provides power to LC No.13 in the radwaste treatment building LC No.10 has a physically and electrically separate emergency section that 1 provides power to various emergency loads on Power Panel No.1. LC No.10 also provides power to LC No.13 in the radwaste treatmer.t Lading. The emergency section of LC No.10 switches l automatically to LC No. 5 of Unit 1 upon loss of power from LC No.10. Simultaneously with this transfer, the prcpane-fueled emergency engine generator starts, and after the generator attains rated l voltage and frequency, the emergency bus supply is transfen ed to this source. If the emergency engine generator supply should fail, transfer is automatically initiated back to LC No. 5, after a 3-second delay. Retum to normel configuration is accomplished manually.

120/208V System. The 120/208V system is a 3-phase,4-wire distribution system which supplies single-phase,120V power for station lighting, convenience outlets, and fractional horsepower motors.

Power to this system is supplied through 120/208V transformers from LC Nos.10 and 11.

125V DC System. The 125V DC System consists of a 58 cell storage battery (125V DC Battery No. 3) two chargers, and the 125V, DC Distribution Panel No.1. The chargers operate in parallel to carry the  !

continuous load and float charge the battery. The battery leads terminate first in a disconnect link cabinet. From this point they are run and connect directly to 125V DC Distribution Panel No.1. The DC distribution is made from this panel to the various boards, switchgear, and motor starters requiring DC service for control and power. The DC section of Valve Control Center No.1 is powered by the 125V DC Distribution Panel No.1. The 125V DC Battery No. 4 (non-vital), the 125V DC Distribution Panel No. 2, and two associated chargers, (No. 2 and 4) have been removed from service.

Annunciator System. The annunciator system permits efficient operation of the station with a minimum number of operators by announcing (visually and audibly) any change in operating conditions requiring

~

attention. An operator can silence the audible hom or bell but visual annunciation wi!! persist until the trouble has cleared and the initiating alarm device has been reset. Annunciators for systems which have been removed from service have been deenergized and the alarm boxes removed from the annunciator panel.

Communication System. The existing communication system will remain functional during the SAFSTOR period. This system consists of the following:

o Telephones connected to and operated with the plant's Private Automatic Exchange (PAX) o The code call system for locating personnel o The emergency signal system for sounding of emergency alert and code signals (including siren alert) o The voice communication system to provide quick communi:ation between selected areas of the plant 3-22 l Rev 2 N

I o' -Intercom system l

l 3.2.3.7 Hydrogen and Seal Oil System  ;

Hydrogen gas was utilized to cool the main generator during operation and also to provide a passive 1 environment for electricalinsulation. Seal oil was provided to the air-hydrogen seals of the generator to - l cid in maintaining air-hydrogen separation.- 1 Tha seal oil system has been drained and hydrogen bottles have been removed and system openings j sealed. No further action is considered necessary. j i

13,3 RADIOLOGICAL CHARACTERIZATION 3.3.1 Radionuclide Inventory -

l The largest percentage of the onsite radionuclide inventory is contained in the spent fuel, with the  !

' reretor vessel and intemals containing the next largest percentage. Radionuclides are also present in  !

corrosion films within various in-plant systems. . i These radionuclide sources are not readily'dispersible in their present condition and will continue to i d: cay during the SAFSTOR period.

Additional contributions to the radionuclide inventory are those sources extemal to the closed systems i addressed above. These sources include fixed and removable surface contamination, the ..

radionuclides contained in the spent fuel pool water, and associated systems. l t

Soent Fuel. ' During January and February 1984, all fuel and fission chambers were' removed from the )

reactor vessel. Currently there are 390 spent fuel assemblies stored in the spent fuel pool. Incore ,

fission chambers with a total ofless than 1 gram of2 mu are also stored in the pool.;The rninimum d

cay and cooling time for any element stored in the pool is slightly less than 18 years. Characteristics j of the stored spent fuel are summarized in Table 3-4. Miscellaneous items that are stored in the spent I

- futi pool are listed in Table 3-5.

The 140 spent fuel assemblies removed from the reactor in February 1984 have exposures ranging from 5,009 to 15,492 megawatt-days per metric ton of uranium. The 250 fuel assemblies previously stored had exposures ranging from 14,400 to 19,481 megawatt days per metric ton of uranium. The fuel types include GE Types ll and 111-1 through 111-4, as well as Exxon Types ill and IV. There are no  !

L GE Type I, stainless steel-clad assemblies remaining on site. However, cladding failure of the Type I i essemblies during the early years of operation has contaminated the spent fuel pool.

The radioactive deposits on the spent fuel pool walls, racks, and related equipment are described in Table 3-7. The total activity is estimated at 3.2 Cl with "7Cs (2.1 Cl) comprising 70 percent. "Fe is present at 27 percent (690 mci) and gamma-emitting "Co (12C mci) is present at 4 percent.

I 3-23 Rev 2

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - . _ . August 1938

The pool cpntains 110,000 gallon of water with an average concentration of 0.005 Ci/ml. 7Cs (1.76 Ci) comprises 94 percent of the activity, while transuranics (12 pCi) comprise less than 0.007 percent of the spent fuel pool water inventory of 2.0 Cl.

Reactor Vessels and Intemals. Table 3-6 shows the radionuclide inventory calculated by Gibbs and Hill (1982) and PG&E for the reactor vessel, corrected for decay to mid-1984 conditions. The highest nuclide inventory is contained in the chimney guide and chimney with an inventory of 4,900 Cl. The 32 control rod blades and the core support grid structure were estimated to have the next greatest ,

inventory, estimated at 2,200 Ci each. These structures account for 78 percent of the totalinventory.

l Within the reactor vessel and i:'temais, the primary controlling nuclide is gamma-emitting "Co.

Because of the predominance of "Co and "Fe and their relatively short half-lives (5.27y and 2.7y, respectively), the activity of the reactor vessel and intemals will dec. ease to 2,900 Ci over the 30-year SAFSTOR period.

in-Plant Systems. Intemal and extemal contamination of plant systems have been characterized by  ;

samples and surveys taken by Pacific Northwest Laboratory (PNL-4628) and by recent characterization I cfforts, as well as routine PG&E characterization and radiological monitoring surveys.

1 The corrosion film radionuclide inventory estimates are presented in Table 3-7. These estimates have been corrected for decay from values detemlined by PNL's 1981 surveys (PNL-4628). The most i

abundant radionuclides in the corrosion films associated with reactor piping and components are "Fe, l "Co, 7Cs and "NI. Transuranics are estimated to constitute about 0.04 percent of the total residual '

radionuclide inventory, or about 30 mci.

l Within the piping systems, the major radionuclide repository is contained in the shutdown cooling l system, where the radionuclide inventory is about two times greater than in the radwaste or reactor water cleanup system piping. The main steam lines contain an order of magnitude less "Fe than in the I shutdown cooling system and relatively minorinventory resides in the feedwater, condensate, and emergency condenser piping.

In the nonpiping systems, the largest radionuclide inventory is in the main condenser (26 Cl "Fe),

which is about 2.5 times greater than the inventory estimate for the reactor shutdown cooling system cooler. The 2largest transuranic inventory is contained in the reactor shutdown cooling system cooler (4.5 mci of

• Am). Table 3-8 summarizes the radionuclide inventory estimates by system components.

Surface Contamination. Analyses of concrete cores conducted by PNL in 1981 indicate that the radionuclide contamination resides mainly in the top centimeter of concrete (see Table 3-9). The exception to this is where cracks have occurred and the contaminants have migrated into the rancrete.

The primary radionuclides in the concrete surfaces are 7Cs and "Co. The 7Cs appears b have penetrated painted surfaces and migrated into cracks to a greater extent than "Co. The greatest concentrations of '27Cs are in the radwaste treatment building area. "Co concentrations are highest at ,

the -66 foot elevation of the reactor caisson access shaft and in the condensste demineralizer room.

All radionuclide concentrations in the concrete are two-to-three orders of magnitude below the maximum allowable concentrations for Class A waste under 10 CFR 61. These 3-24 l Rev 2 Aupust 1998

i concentrations will be further reduced due to decay during the SAFSTOR period, particularly "Co with a half-life of 5.3 years. The highest total concentrations of transuranics analyzed by PNL were 2.9 nanocuries/g.

The activity in the extemal surface contamination is estimated to be 1.1 Ci assuming the ratio of intemal to extemal contamination reported for NUREG/CR-0672, " Technology, Safety and Cost of D: commissioning a Reference Boiling Water Reactor." Surveys of radiation and contamination levels in Unit 3 and related structures are presented in Table 3-10. An inventory estimate based on core and swipe analyses for radionuclides associated with concrete or other surficial sources has not been r;ttempted. Such an estimate would be unreliable since it would contain sources of extreme variability of radionuclide concentrations and distributions.

Sealed Sources. A 4 Ci"Co source and a 14 Ci"Co source are stored in the calibration facility.

S:veral other small check / calibration sources are also routinely used.

3.3.2 Refueling Building / Power Building D; contamination of the Unit 3 facilities will be an ongoing process, with systems and components drained, flushed, and partially decontaminated as appropriate. These measures and maintenance of th3 integrity of contamination barriers will minimize radiation and contamination levels throughout the SAFSTOR period. Table 3-10 and Table 3-11 describe the radiological conditions existing within the r: fueling building and the power building respectively.

3.3.3 Yard Structures Structures within the yard of Unit 3 are the hot machine shop / calibration facility, radwaste treatment and storage buildings, the stack, and the recombiner vault. Table 3-12 presents radiation and

~

contamination level surveys of various structures in the ya-d.

3.3.4 Cross-Connections Existing cross-connections provide a possible pathway for cross-contamination between Units 1/2 and Unit 3. In response to NRC's IE Bulletin No. 80-10 of May 6,1980, a review of possible cross-connections was undertaken and identified several systems where cross-contamination may occur. A routine sampling analysis program was initiated to ensure that potential problems would be readily id:;ntified and that releases would be minimized and documented.

Table 3-13 presents the radionuclide analysis of systems with cross-contamination potential. In all l

systems, the contamination has been well below limits established by 10 CFR 20.

l l

l l 3-25 Rev 2 August 1998

l 3.3.5 Envir,onmental Radiological Characteristics .

BIckground activity levels are currently found in the soils, canal and slough sediments, bay sediments,

' terrestrial and aquatic plants, and bay mussels (see Environmental Report, Section 5.6) sampled from outside the Unit 3 restricted area.

Within the Unit 3 restricted area, total beta-gamma dose rates range from background to 100 times background, primarjil due to shine from the refueling building and the radwaste treatment building.

Concentrations of Cs, "Co, and otherisotopes are slightly above background in core samples of soils very near the refueling building (see the Environmental Report, Section 5.6).' These concentrations are primarily the result of spills on the soil surface, leakage from buried transfer pipes, and leakage from the spent fuel storage pool. However, the concentrations are low and do not require remediation for the SAFSTOR period. No significant airbome resuspension is associated with the minimal contamination. -

3.3.6. Summary

' As of July 1984, the largest inventory of radionuclides is contained in the spent fuel rods with an estimated activity of 1.2E+6 Cl. Within 30 years, this activity will be less than 5.OE+5 Ci due to the

. decay of 7Cs and "Srwith half-lives of approximately 30 years.

The largest radionuclide inventory outside of the spent fuel pool consists of activation products in the reactor vessel and surrounding structures. The total activity is estimated at 12,000 Cl. The primary radionuclide is "Co with an activity of 7,100 CI (61 percent of the total.) During the SAFSTOR period the "Co activity will decrease to 110 Cl, and "Ni with an activity of 2,700 Ci will be the most abundant rtdionuclide in the reactor vessel system. ' Since the primary source of exposure is currently from "Co, u the exposure rate will continue to decrease during the SAFSTOR period.

~ Other than several sealed sources and surface contamination, the remaining inventory is estimated at l 100 Cl contained in corrosion films in various piping and components. The primary radionuclides are "Fe (81 percent) and "Co (14 percent). After 30 years, the primary radionuclides'will be "Ni and

7 Cs, which will comprise 90 percent of the estimated remaining 2.5 Cl. The transuranic radionuclide  :

inventory is estimated at 32 mci (0.04 percent). The most abundant transuranic nuclide is 2 cam, I which comprises 38 percent of the corrosion film transuranic inventory.

! I l

t l

3-26 l 1 Rev2 l August 1998

_l

Table 3A e Combustible Fuel Storage Facilities MAXIMUM STORAGE LOCATION

• FUEL CAPACITY METHOD (ft.)

(gals.)

Residual fuel oil 5,760,678 Tanks 559 1.

(Number 6 fuel oil or Bunker C)

2. Diesel storage tank 84,940 Tank 473 (Number 2 diesel oil)

Diesel day tanks 19,800 Tanks 401 3.

Propane 2,098 Tank 229 4.

120 Portable tank 321

5. Gasoline -

EPA restrictions limit HBPP to less than one million ga!!ons of petroleum products on site. All of the fu:Is are delivered to the plant site by tank trucks.

• Locations reflect the distance from the center of the reactor to the center of the closest tank.

I 3-27 I Rev 2 August 1998

Teble 3-1 Humboldt Bay Power Plant, Unit 3 Physical Characterization - Summary SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR -

STATUS SAFSTOR.

(NOTE A)

1. Nuclear Steam Supply System (a) Reactor Vessel and Intemals

• Reactor Vessels Drywell 7 A,1,E Vacuumed, Head &

Shield Plug Replaced,  !

Drained, Layup & Secure

• Control Rods (32) ReactorVessel 7 A,1,E Fully inserted t

• Control Rod Guide Tubes (32) Reactor Vess.el 7 A,1,E in Place

• Core Support Assembly Reactor Vessel 7 A,1,E in Place

• Lower Core Shroud ReactorVessel 7 A,1,E in Place  !

i e Upper Core Shroud ReactorVessel 7.A,1,E in Place e Upper Guide Reactor Vessel 7 A,1,E in Place

• Chimney Reactor Vessel 7 A,1,E in Place

• Feedwater Sparger Reactor Vessel 7 A,1,E in Place e Steam Dryer ReactorVessel 7 A,1,E in Place (b) Control Rod Hydraulic System

• System Piping Drywell, Access 3I Drained, Extemal  ;

Shaft & Refueling Decontamination, Layup  !

Building & Secure e Control Rod Drives (32) Drywell 6 A,1,E in Place, Drained 3-28 l Rev 2 l August 1998

LOCATION RAD CONDITION FOR .

SYSTEMS / COMPONENTS STATUS SAFSTOR (NOTE A)

Access Shaft (-54') 51,E Drain, Ext. Decon.,

• Accumulators (16) Layup Refueling Building 31,E Drain, Ext. Decon. &

• Supply Pumps (2) Disconnected, Layup

(+ 12')

Access Shaft (-66') 51.E Removed

• Scram Dump Tank (c) Liquid Poison System Refueling Building 21,E Drain, Ext. Decon.,

• System Piping

(+ 12') Layup ,

Refueling Building 1E Drain, Ext. Decon.,

• Liquid Poison Tank ,

(+ 12') Layup (d) Reactor Cleanup System t

Access Shaft (-66' 51,E Drain, Ext. Decon.,

• System Piping to -2') Layup & Secure .

Access Shaft (-66') 51,E Removed

• Cleanup Pump

• Regenerative Heat Exchangers Cleanup Heat 61.E Drain, Layup & Secure '

• Exch. Room (-2')

• Non-regenerative Heat Exchangers Cleanup Heat 61,E Drain, Layup & Secure Evch. Room (-2')

• Demineralizer Cleanup Heat 31,E Drain, Layup & Secure Exch. Room (-2') ,

• Resin Storage Tank Ref. Building iE Drain, Ext. Decon.,

(+ 12') Layup & Secure ,

l 3-29 Rev 2 a . . . ... < nno

SYSTEMS / COMPONENTS LOCATION RAD . CONDITION FOR

. STATUS SAFSTOR

? (NOTE A) 'O (e) Reactor Shutdown Cooling System -

• System Piping Access Shaft (-14') 5,61,E Drain, Layup & Secure

• Reactor Shutdown Pumps (2) Shutdown Room 31,E Drain, Layup & Secure

(-14')

• Shutdown Heat Exchangers (2) Shutdown Room 3 I,E Drain, Layup & Secure

(-14')

(f) Emergency Condenser System

• Emergency Makeup Pump Yard 1 1,E Drain, Disconnect, Layup

• Emergency Condenser Refueling Building 21,E Drain, Openings Sealed,

(+28') Layup (g) Suppression Tank Cooling and Core Spray System

• Core Spray Pumps (2) Access Shaft (-66') 31,E Removed

• Suppression Chamber Cooler Access Shaft (-2') 21,E Drain, Ext. Decon., Layup

• Low Pressure Core Flooding Supply , Access Shaft (-14') 61E Drain, Ext. Decon. &

Line Ref. Building Flush, Layup (Note C)

(+ 12')

• Suppression Chamber Reactor Caisson 21,E Drain, Openings Sealed, Level Monitor installed, Monitor for Liquid Level

2. Turbine Plant Systems 3-30 l Rev 2 August 1998

E

- SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR - . .

(NOTE 'A)  ;

(a) Turbine Systems

• Turbine Turbine Encl. 1 1,E External Decon.,

(+27') Nitrogen Blanket, Layup

• Main Steam Stop Valve Power Building 31,E External Decon., Layup  ;

Pipe Tunnel (+6')

• Hydrogen Cooler Seal Oil Room 1- External Decon., Layup

(+6')

• Hydrogen Oil Seal Cooler Main Generator 1- Extemal Decon., Layup

• Turbine Bypass Valves (2) Power, Building 31,E Extemal Decon., Layup  ;

Condenser Rm.

(+ 12')  ;

(b) Lube Oil System

• Bowser Retum Pump Reactor Feed 1- Layup & Secure  !

Pump Room j

(+ 12')  ;

• Lube Oil Coolers (2) Reactor Feed 1- Layup & Secure f

, Pump Room

(+12')  ;

,

• Bowser Tank Lube Oil Filters Reactor Feed 1- Layup & Secure

. Pump Room  ;

(+ 12')  :

• Vapor Extractor Pump Reactor Feed 1- Layup & Secure  !

Pump Room j

(+12')

l 3-31 Rev 2  ;

^

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

. . . . _ . . ._ ???"M.109 - t

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR (NOTE A) .

• Oil Driven Booster Pump Reactor Feed 1- Layup & Secure Pump Room

(+ 12')

• Lube Oil Reservoir Reactor Feed 1- Layup & Secure Pump Room

(+ 12')

• Turning Gear Oil Pump Reactor Feed 1- Layup & Secure Pump Room

(+ 12')

• Auxiliary Oil Pump Reactor Feed 1- Layup & Secure Pump Room

(+ 12') ,

• Main Oil Pump Turbine Enclosure 1- Layup & Secure Turbine Shaft

• Reactor Feed Pump Reservoir Reactor Feed 1- Layup & Secure Pump Room

(+ 12')  ;

• Reactor Feed Pump Aux. Lube Pump Reactor Feed 1- Layup & Secure ,

Pump Room

(+ 12')

• Reactor Feed Pump Lube Pump Reactor Feed 1- Layup & Secure Pump Room

(+ 12')

• Reactor Feed Pump Lube Cooler Reactor Feed 1- Layup & Secure Pump Room

(+ 12')

3-32 l Rev 2 ,

August 1998 '

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR ,

(NOTE A)

• Oil Filter Reactor Feed 1- Layup & Secure Pump Room

(+ 12')

• Emergency D.C. Oil Pump Reactor Feed 1- Layup & Secure Pump Room

(+ 12')

(c) Condensate System

. Main Condenser Condenser Rm. 41,E Drain, Decon.

Internal / External, Layup

(+6')

& Secure e Condensate Storage Tank Yard 21 Drain, Decon., Layup &

Secure e Condensate Pumps Condensate Pump 31,E Drain, Decon. Ext., Layup (2)

Room (+ 12') & Secure e Condensate Demineralizer (3) Condensate Demin. 41E Drain, Convert one Room (+ 12') demin. for SFP. Layup arid secure remaining two

' Condenser Rm. 31,E Drain, Layup & Secure e Contaminated Drain Tank

(+6')

• Contaminated Drain Tank Pump Condenser Rm. 31,E Drain, Disconnect, Layup

(+6') & Secure

• Condenser Vacuum Pump Condensate Pump . 21,E Drain, Disconnect, Layup Room & Secure

(+ 12')

l 3-33 Rev 2

- . - _ . . _ - .. . . -.. - -,. - . . - . . . - . . - . - . . . - . - . - . . - . . . - - - - - - . - . - ...- ..-.-.._ .:: :T* """ - - . -

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR '

STATUS SAFSTOR (NOTE A) .

• Air Ejector Air Ejector Room 31,E Drain, Layup & Secure

(+ 12')

• Inter-after Condenser (Air Ejector Air Ejector Room 41,E Drain, Layup & Secure Condenser) (+ 12')

• Cation Regeneration Tank Condensate Demin. 31,E Drain, Decon.

Room (+ 12') Intemal/Extemal, Layup

& Secure ,

• Anion Regeneration Tank Condensate Demin. 31,E Drain, Decon.

Room (+ 12') Intemal/Extemal, Layup f

& Secure i

• Regenerated Resin Storage Tank Condensate Demin. 31,E Drain, Decon.

(Laundry Hold Tank) Room (+ 12') Intemal/Extemal, Layup

& Secure (d) Gland Seal System .

• Gland Seal Condenser Air Ejector Room 31,E Drain, Layup

(+ 12') ,

o Gland Seal Exhausters (2) Air Ejector Room 31,E Drain, Disconnect, Layup "

(+ 12')

(e) Feedwater System

• Feedwater Pumps Reactor Feed 21 Drain, Disconnect, Layup (2)

Pump Room

(+ 12')

• LP Feedwater Heater Pipe Tunnel (+0') 41,E Drain, Layup

• IP Feedwater Heater Pipe Tunnel (+6') 41,E Drain, Layup 3-34 l Rev 2 August 1998

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR ,

(NOTE A)

(f) Closed Cooling Water System

• Retum Tank Yard, below grade -

Drain, Flush, Layup

• Pumps (2) Yard E Drain, Flush, Layup &

Secure

• Heat Exchangers (2) Yard E Drain, Flush, Layup (g) Main Circulating Water System o intake Structure intake Canal 1- In Place, Layup

• Traveling Screens (2) Intake Structure 1- Removed

• Screen Wash Pumps (2) Intake Structure Removed

• Circulating Water Pumps (2) Intake Structure 1- Removed

• Sluice J ates (2) Intake Structure 1- Removed i

l

• Intake Piping Yard, below grade 1- In Place, Layup

• Discharge Piping Yard, below grade 1- In Place, Layup

' intake Structure 1- Removed l

• Sluice Gate Operators (h) Compressed Air System

• Air Compressor Reactor Feed 1- Available for use, Pump Room Maintain (Standby)

(+ 12')

l 3-35 Rev 2 a.... 4nno

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR-(NOTE A) .

• Aftercooler Reactor Feed 1- Available for use, Pump Room Maintain (Standby)

(+ 12')

• Service Air Receiver Reactor Feed 1- In Use, Maintain Pump Room Operational

(+ 12')

• Instrument Air Receiver Reactor Feed 1- In Use, Maintain Pump Room Operational

(+ 12') *

• Instrument Air Dryer Reactor Feed 1- In Use, Maintain -

Pump Room Operational

(+ 12')

• Instrument Air Filters - Reactor Feed 1- In Use, Maintain Pump Room Operational

(+ 12')

(i) Demineralized Water System

• Demineralized Makeup Water Heater Makeup Demin. 1E Removed Area (+27')

• Demineralized Water Storage Tank Yard 1- In Use, Maintain Operational

• Demineralized Water Pump Yard 1E in Use, Maintain l Operational I

• Makeup Demineralizer Filter Makeup Demin. 1E Removed Area (+27')

3-36 l )

Rev 2 August 1998

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR ,

(NOTE A)

Makeup Demin. 1. E Removed

• Makeup Demineralizer Area (+27')

• Concentrated Caustic Storage Tank Yard 1E Removed

• Concentrated Acid Storage Tank Yard 1E Removed

• Acid Pump, Makeup Demin. Makeup Demin. 1E Removed Area (+27')

Makeup Demin. 1E Removed

• Caustic Pump, Makeup Demin.

Area (+27')

Makeup Demin. 1E Removed

• Caustic Dilution Water Heater Area (+27')

Makeup Demin. 1E Removed

• Filter Aid Tank Area (+27')

Makeup Demin. 1E Removed

• Filter Aid Pump Area (+27')

3. Waste Disposal Systems (a) Gas Treatment System Stack (+27') 21,E Decon. Extemal, Maintain

• Exhaust Fans (2) Operational Stack (O') 21,E Drain, Decon. Extemal,

• Gas Scrubber Column Layup Stack (O') 21,E Drain, Decon. External,

• Gas Scrubber Recire. Tank Layup 3-37 l Rev 2 a.... 4nno

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR (NOTE A) .

• Gas Scrubber Recirc. Pumps (2) Stack (O') 21,E Drain, Ext. Decon, Disconnect, Layup

• Absolute Filter (Scrubber Column Stack (+ 12') 21,E in Use, Maintain Exhaust) Operational

• Holdup Piping Yard, below grade 31,E Layup (b) Off-gas Treatment System (New)

• Jet Cornpressors New Off-gas Vault 1- Not Used

• Preheater New Off-gas Vault 1- Not Used

• Recombiners New Off-gas Vault 1- Not Used

• Condenser / Moisture Separator New Off-gas Vault 1- Not Used

• Off-gas Prefilter (HEPA) New Off-gas Vault 1- Not Used

• Refrigerant Air Dryer New Off-gas Vault 1- Not Used

• Desiccant Towers / Dryer Subsys. (2) New Off-gas Vault 1- Not Used

• Carbon Guard Bed , New Off-gas Vault 1- Not Used

• Carbon Absorber Columns New Off-gas Vault 1- Not Used (c) Liquid Waste Collection System

• Turbine Building Drain Tank (TBDT) Below New Fuel 31,E in Use, Maintain Storage Vault (-14') Operational

• Turbine Building Floor Drain Pump Below New Fuel 31,E in Use, Maintain Storage Vault (-14') Operational 3-38 l Rev 2 August 1998

t b

i LOCATION RAD CONDITION FOR .

SYSTEMS / COMPONENTS SAFSTOR  ;

STATUS -

(NOTE A) I Below New Fuel 31,n in Use, Maintain .[

• TBDT Pumps (2) Operational '  !

Storage Vault ,

(-14')

41,E in Use, Maintain Access Shaft (-66')

j

• ~ Reactor Equipment Drain Tank (REDT) Operational 1

Access Shaft (-66') 41,E in Use, Maintain  !

• REDT Pumps (2) Operational .

5 Access Shaft (-66') 3E in Use, Maintain i

• Reactor daisson Sump .

l Operational i t

Access Shaft (-66') 3E in Use, Maintain  ;

• Reactor Caisson Sump Pumps (2)

( Operational l

Yard 1i in Use, Maintain  :

• Yard Drain System Operational  !

Pipe Tunnel 21,E Modified for Decon Sink

• Laundry Waste Tank  !

(+20') & Shower and Maintain Operational l Condensate Demin. 31,E Drain, Decon. & Layup

• Laundry Hold Tank (Formerly Regenerated Resin Storage Tank) Room (+ 12')

Condensate Demin. 21,E Drain.. Decon. & Layup ,

• Laundry Waste Filter [

Room (+ 12') .

Condensate Demin. 21,E Drain, Decon. & Layup i

• Laundry Waste Pumps (2)  ;

Room (+ 12')

i Valve Gallery 31,E Not Used, Layup  ;

e Vent Separator

(+0') l t

l 3-39 Rev 2 a .,m ,,e+ 4 n o a j

SYSTEMS / COMPONENTS LOCATION RAD. CONDITION FOR STATUS SAFSTOR (NOTE A) ,

(d) Liquid Waste Treatment

• Radwaste Bldg. Sump Tank Radwaste Building 41,E in Use, Maintain Operational

• Radwaste Bldg. Sump Pump Radwaste Building 31,E in Use, Maintain Operational

• Radwaste Receiver Tanks (3) Radwaste Building 31,E in Use, Maintain Operational -

• Radwaste Pump Radwaste Building 21,E in Use, Maintain Operational

• Concentrator Feed Pump Radwaste Building 31E Available For Use, Maintain Operational e Radwaste Concentrator Radwaste Building 41.E Available For Use, Maintain Operational e Radwaste Concentrator Condenser Radwaste Building 31,E Available For Use, Maintain Operational

• Radwaste Demineralizer Radwaste Building 5 I,E Available For Use,

, Maintain Operational e Resin Disposal Tank Radwaste Building 61,E in Use, Maintain Operational

• Concentrated Waste Tanks (2) Radwaste Building 51,E In Use, Maintain Operational

• Waste Hold Tanks (2) Radwaste Building 31,E in Use, Maintain Operational 3-40 l Rev 2 August 1998

SYSTEMS / COMPONENTS LOCATION RAD . CONDITION FOR STATUS SAFSTOR .

(NOTE A)

• Treated Waste Pump Radwaste Building 21,E in Use, Maintain Operational

• Radwaste Filters (2) Radwaste Building 21,E in Use, Maintain Operational

• Spent Fuel Pool Filter Radwaste Building 31,E Removed

• Concentrator Drip Receiver Tank Radwaste Building 21,E Available For Use, Maintain Operational

• Resin Addition Tank Radwaste Building 3E Available For Use, Maintain Operational

• Concentrator Flash Pot Radwaste Building 31,E Available For Use, Maintain Operational

• Concentrator Drip Receiver Tank Radwaste Building 21E Available For Use, Maintain Operational

• Absolute Filter (Liquid Waste Tankage Radwaste Building 21,E Removed Vents) Roof

-(e) Solid Radwaste System

. Compactor Liquid Radwaste 21,E in Use, Maintain l Enclosure Operational

4. Control and instrumentation (a) Reactor Protection System Control 1,5 E Not Used, Layup Rm/Drywell (b) Reactor Nuclear Instrumentation Control 1,5 E Not Used, Layup Rm/Drywell l 3-41 Rev 2

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

- - 2 ":': ^ ^"". -

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR (NOTE A) .

(c) In-core Flux Monitoring System Control 1E Not Used, Layup Rm/Drywell (d) Area Radiation Monitoring System Control Room, 1,3 E Selected Areas Plantwide Monitored, Maintain Operational (e) Refueling Bldg. Isolation Monitoring Control Room, 1,3 E Not Used, Secure System Ref. Bldg.,

Access Shaft (f) Stac'K Gas Radiation Monitoring Control Room, 1,2 E in Use, Maintain Stack Base Operational (g) Off-gas Monitoring System Control Room, Air 1,3 E Not Used, Layup Ejector Room (h) Process Radiation Monitoring System Control Room, 1,3 E In Use, Maintain Radwaste Bldg Operational (i) Reactor Vessel Instrumentation Control Room, 1,3 E Not Used, Layup Drywell, Access Shaft (j) Containment Leak Rate Monitoring Access Sh., Reactor 1,2 E Not Used, Layup Feed Pump Room (k) Control Rod Position Indication Control Room, 1,6 E Not Used, Layup Drywell, Access Shaft (I) Control Rod Drive Instrumentation Control Room, Acc. 1,3 E Not Used, Layup Shaft (-44')

3-42 l Rev 2 August 1998

SYSTEMS / COMPONENTS . LOCATION RAD CONDITION FOR STATUS SAFSTOR .

(NOTE A)

(m) Feed water Control System Control Room, 1,2 E Not Used, Layup React. Feed Pump Room

5. Service Systems (a) Fire Protection System

• Fire Pumps (3) Fire Pump House 1- Available For Use, Maintain Operational (b) Heating and Ventilation System  !

• Multizone Air Handling Unit Over Laundry (+ 37') 11 In Use, Maintain Operational ,

• Air Handling Unit #1 Control Room Roof 1E in Use, Maintain

- Operational ,

• Air Handling Unit #2 Yard 1E Removed

• Air Handling Unit #3 Turbine Encl. Roof 1E Removed i

• Reactor Feedpump Rm. Supply Reactor Feed 1E in Use, Maintain

Fan ,

Pump Room Operational

(+ 12') i

• Reactor Feedpump Rm. Exhaust Reactor Feed 1E in Use, Maintain Fan Pump Room Operational

(+ 12')

• Turbine Bldg. Exhaust Plenum Pipe Tunnel 21,E In Use, Maintain (No.1) Operational i

I l 3-43 Rev2

...-....nno ;

SYSTEMS / COMPONENTS LOCATION RAD STATUS CONDITION FOR (NOTE A) SAFSTOR -

• Refueling Bldg. Exhaust Plenum Yard 1 1,E in Use, Maintain -

Operational

• Plant Exhaust Fans (2) Yard 1 I,E in Use, Maintain Operational e Drywell Cooling Unit Lower Drywell 41,E Not Used, Layup

• Drywell Purge Fan Stack (+27') 21,E in Use, Maintain Operational

• Lab Hood Exhaust Fan Hot Lab Roof 1I in Use, Maintain Operational

• Absolute and Roughing Filter (Hot Hot Lab Roof 1I in Use, Maintain Lab) Operational

• Absolute Filter Laundry 11 Removed

• Exciter Cooling Fan Exciter House 1i Not Used, Layup

• Heater and Fan Unit Hot Machine Shop 1 I,E Available For Use, Maintain Operational e Exhaust Fan Hot Machine Shop 1 1,E Available For Use,

, Maintain Operational

• Absolute and Roughing Filter (Hot Hot Machine Shop 11 Available For Use, Machine Shop) Maintain Operational

• Heater and Fan Unit Inst. Repair Room 1- In Use, Maintain Operational e Isokinetic Sampler Stack (Halfway Up) 1I In Use, Maintain Operational 3-44 l Rev 2 August 1998

l l -t t

l SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR ,

(NOTE A)

(c) Spent Fuel Pool Service System

• Liner Gap Pump Refueling Building 21,E in Use, Maintain

(+ 12') Operational

• Fuel Pool Circulating Water Pump (2) Refueling Building 31,E in Use, Maintain

(-24') Operational

!

• Fuel Pool Coolers (2) Refueling Building 31E Not Used, Bypassed,

(+ 12') Layup

• Fuel Pool Skimmer Refueling Building 31,E Not Used

(+ 12')

• Channel Handling Tools Refueling Building 21,E Note B, Store

(+ 12')

• Manual Fuel Handling Tools Refueling Building 21,E Note B,C, Store

(+ 12')

• Spent Fuel Pool Jib Crane (500 lb) Refueling Building 2E Note B, Maintain

(+ 12') (Standby)

• Extension Tank and Refueling Platform Refueling Building 21,E Note C, Layup

, (+ 12')

• Transfer Cask and Winch Refueling Building 31E Not Used, Layup

(+ 12')

• SFP Demineralizer SFP Demin Room 4 l,E in Use, Maintain

(+ 12 ) Operational e SFP Demineralizer Strainer SFP Demin Room 41,E in Use, Maintain

(+ 12 ) Operational (d) Hydrogen and Seal Oil System l 3-45  !

Rev 2 i

. ._ .- -. . .- - - . _ . - - - _ . _ - - . - ,, -- - ^ "?"'* *? " ..

l i

t SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS SAFSTOR (NOTE A) .

• Hydrogen Coolers (4) Generator Housing 1- Not Used

• Hydrogen Dryer Seal Oil Room 1- Not Used, Layup ,

(+6')

• Seal Oil Storage Tank- Seal Oil Room 1- Not Used, Layup _

(+6')

• Main Seal Oil Pump Seal Oil Room 1- Not Used, Layup

(+6')

• Emergency Seal Oil Pump Seal Oil Room '1 - Not Used, Layup

(+6')

• Seal Oil Filters (2) Seal Oil Room 1- Not Used, Layup

(+6')

(e) Reactor Shield Cooling System

• Cooling Coils Drywell 2A Drained, Flushed, Layup (Embedded)  ;

(f) Cranes  :

2E

• 75-Ton Bridge Crane Refueling Building Available For Use,

' Maintain Operational i

• 2-Ton Jib Crane Refueling Building 2E Available For Use,-

Maintain Operational

• 5-Ton Hot Machine Shop Crane Hot Machine Shop 2E Available For Use, Maintain Operational (g) Security System Plantwide 1,2 E in Use, Maintain Operational 346 l Rev 2 August 1998

SYSTEMS / COMPONENTS LOCATION RAD CONDITION FOR STATUS . SAFSTOR ,

(NOTE A) ';

6. Electrical Systems (a) Protective Relay System Control Room, 1,3 E in Use, Maintain Plantwide Operational (b) Annunciator System Control Room, 1,3 E in Use, Maintain Plantwide Operational (c) Communications Systems Plantwide 1,3 - In Use, Maintain Operational (d) Emergency AC Systems Control Room, 1- In Use, Maintain Yard Operational (e) Preferred AC System Reactor Feed 1- Not Used, Layup Pump Room

(+12')

(f) 125 Volt DC System

• Vital Adjacent to 1- Available For Use, Counting Rcom Maintain Operational e Non-vital Control Room, Yard 1- Removed (g) Auxiliary Power Systems

• 480 Volt System Control Room, Yard 1- In Use, Maintain Operational

• 2400 Volt System Control Room, Yard 1- In Use, Maintain Operational e 120/208 Volt Distribution System Plantwide 1,4 - In Use, Maintain Operational l 3-47 Rev 2

_^ ' ""M__*PP 9_ _ _ _ _ __

SYSTEMS / COMPONENTS LOCATION RAD -

CONDITION FOR STATUS SAFSTOR (NOTE A) ,

(h) 13.8 KV System Yard 1- Not Used, Layup NOTES: A. RADIATION STATUS: 1 <2 mrem /hr" 2 2 - 20 mrem /hr" 3 21 - 100 mrem /hr" 4 101 - 300 mrem /hr" 5 301 - 999 mrem /hr" 6 1 - 10 R/hr" 7 >10 R/hr" A ACTIVATED I INTERNALLY CONTAMINATED E EXTERNALLY CONTAMINATED B. May be required for spent fuel shipments.

C. May be required for final decommissioning.

b i

o As of 1984, approximately.

3-48 l 3 Rev 2 August 1998

. Table 3-2 Reactor Data CATEGORY SPECIFICATION i

Reactor Type Single cycle, natural circulation boiling water ,

l Coolant / Moderator Light water ,

License Number DPR-7 Owner Pacific Gas and Electric Company Architect / Engineer Bechtel Supplier General Electric Company lWeights:

ReactorVessel Assembled, Dry 288,146 lb Closure Head (including Steam Dryer ~ 28,068 lb and Cooling Spray Sparger) i Vessel Shell(including Feedwater 231,372 lb Sparger) t Feedwater Sparger 1,242 lb 1,440 lb I Steam Dryer Studs, Nuts, and Washers 3,640 lb i

I l 3-49 L

l Rev2

! August 1998

_ _ - . _ . . _.m. _ . _ . _ _ _ _ _ . . _ . _ _ - _ . _ . _ . _ . . . _ . . . _ . _ _ . . _ . . _ _ _ ,_

l l

Table 3-3 Survey Results - Reactor Vessel Drain Test Before Vessel After Vessel 1 was drained was drained SURVEY LOCATIONS mR/hr mR/hr i

i (1) Refueling Building On top of roof, directly above reactor <1 <1

+ 12 ft elevation, at center of shield plug 0.4 1 l

+12 ft elevation, around edges of shield 0.5 to 1.3 1 plug (2) Access Shaft

, Wall (4 ft above floor) behind the 1.1 2 Suppression Chamber Cooler at -2 ft elevation Wall (4 ft above floor), south of the manlift 0.5 1 at-14 ft elevation Southwest wall (4 ft above-floor) of void 10 10 space at -14 ft elevation (south of Hydraulic System filters)

Wall (4 ft above floor), south of the manlift 0.5 1 at-24 ft elevation Wall (4 ft above floor), south of mantift at - 0.8 1.5 34 ft elevation Center of Suppression Chamber manway 0.7 1 (west side) at the -34 ft elevation Center of Suppression Chamber manway 2.0 1 (east side) at the -34 ft elevation 3-50 l Rev 2 August 1998

Table 3-3 (Continusd)

Survey Results - Reactor Vessel Drain Test

• • Before Vessel After Vessel was drained was drained mR/hr mR/hr SURVEY LOCATIONS (2) Access Shaft (Continued) 6 5 Wall (4 ft above floor), south of manlift at -

44 ft elevation 7 7 Wall (4 ft above floor), south of manlift at -

54 ft elevation

-66 ft 28 21 Center oflower drywell head at elevation 18 in. above floor at -66 ft elevation directly 23 under center of lower drywell head 6 ft above floor at -66 ft elevation directly 26 26 under center oflower drywell head 28 25 Caisson Sump at -66 ft elevation 11 13 Drywell instrument penetration at -66 ft elevation (south side) l 28 25 l Drywell instrument penetration at -66 ft '

l elevation (north side) 18 in. above floor at -66 ft elevation directly 50 60 under edge of drywell (north wall) 6 ft above floor at -66 ft elevation directly 60 65 under edge of drywell (north wall)

(3) Pipe Tunnel (Valve Gallery) 40 33 Instrument penetration at the +2 ft elevation 8 10 Feedwaterline drywell penetration I

3-51 l Rev 2 August 1998

l Tablo 3-3 (Continued) i Survey Results - Reactor Vessel Drain Test Before Vessel After Vessef was drained was drained i SURVEY LOCATIONS mR/hr mR/hr  !

• Main steam line drywell 16 15 l penetration l

• 10 Wall (6 ft above floor at -14 ft 8 elevation) on the south side of drywell

• 7 6.5 Wall (2 ft above floor at -14 ft elevation) on the south side of drywell

• 18 Wall (6 ft above floor at -14 ft 18 elevation) on the east side of I drywell

• Wall (2 ft -tove floor at -14 ft 8 7 l elevation) on the east side of j drywell i

1 l

l' I

I 3-52 l Rev 2 August 1998

Tabla 3-4 Spent Fuel Inventory July 1984*

Isotope. Half-Life Activity l l Isotope Half-life Activity (Years) (Ci) ll 1I (Years) (Ci)

Ii 3.0 E+1 4.5 E+5 l l 125Sb 2.7 E+0 2.3 E+2 137Cs II 2.9 E+1 3.3 E+5 2.1 E+5 6.7 E+1 90Sr l l 99Tc 1.5 E+1 2.9 E+5 242Cm 4.5 E-1 1.8 E+0 241Pu 2.1 E+0 3.4 E+4 1.5 E+1 5.8 E+0 134Cs lIIl 113mCd 1

l 2.7 E+4 1.5 E+6 1.4 E+0 85Kr 1.1 E+1, l ' 933 1.8 E+1 2.2 E+4 9.3 E+1 3.8 E-1 244Cm l151Sm 8.8 E+1 1.3 E+4 1.0 E+5 1.7 E-1 238Pu 126Sn 1.0 E+0 3.8 E+3 l 79 Se 6.5 E+4 1.3 E-1 106Ru 2.6 E+0 3.6 E+3 2.3 E+6 8.4 E-2 147Pm ll135 Cs Il 6.5 E+3 2.0 E+3 1.3 E+1 8.1 E-2 240Pu ll152 Eu 7.8 E-1 1.8 E+3 121mSn 5.0 E+1 3.9 E-2 144Ce l 1.2 E+1 1.7 E+3 l ,l 93 mNb 1.4 E+1 5.0 E-2 3H 8.2 E+0 1.3 E+3 107Pd 6.5 E+6 3.4 E-2 154Eu l

2.4 E+4 1.3 E+3 8.0 E-1 1.2 E-2 239Pu l l 119 mSn 4.3 E+2 7.2 E+2 1.6 E+7 9.4 E-3 241Am 129l 4.8 E+0 5.6 E+2 3.0 E-1 7.5 E-4 155Eu l1il 127mTm

' Activities based on the following: Westinghouse BLWE-1154, April 17,1980, adjusted for bumup fraction and power; General Electric, SEB-80/027, October 31,1980, adjusted for bumup fraction and power; and W. B. Wilson, et al, Extended Bumup Calculations for Operating Reactor Reload Reviews, NUREG/CR-3108, LA-9563-MS, Los Alamos National Laboratory, Febnjary 1983, adjusted for bumup fraction and power.

l 3-53 Rev2 i August 1998

i Table 3-5 Spent Fuel Pool Miscellaneous inventory - 1984 ESTIMATED CURIE CONTENT l

COMPONENT 55p, 60Co 63Ni Other Total Nuclides Curies incore Instrument Strings' 90 98 69 <1 260 l

(< 1 gram 23sU) l

. Stellite Rollers" < 15 8,000 11 -

8,200 Canned Waste - - - - -

-  ?

Sealed Sources

.Sb-Be Operating Sources (2) < 1 E-6 l

Based on Oaks, H.D., Technology, Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, NUREG/CR-0672, Vols.1 and ,

2.

Based on neutron activation calculations and sample analysis l

l l

l l

3-54 l Rev 2 August 1998

Table 3-6 Reactor Vessel Inventory of Radionuclides Corrected for Decay for Conditions Mid-1984 Estimated Curie Content Reactor Other Total Internal Components 55Fe 63N Nuclides' Curies 60Co 3.1 E+3 1.7E+2 1.6E+3 1.7E+1 4.9E+3 Chimney

• Guide and Chimney Core Shroud" 5.3E+2 2.9E+1 2.7E+2 3.0E+0 8.3E+2 Core Support & 1.4 E+3 7.7E+1 7.3E+2 7.0E+0 2.2E+3 Grid" Fuel Support Plates" 9.9E+2 5.4E+1 5.1E+2 5.0E+0 1.6E+3 ,

Control Rod Guide 6.3E+1 3.7E+0 3.3E+1 < 1 E+0 1.0E+2 Tubes

• Control Rod Blades

• 9.4 E+2 1.0E+3 2.3E+2 3.1 E+0 2.2E+3 Reactor Vessel & 6.9E+1 5.0E+1 9.0E+0 3.0E+0 1.3E+2 Clad" Drywell Vessel Wall" < 1 E+0 < 1 E+0 < 1 E+0 < 1 E+0 < 1 E+0 Drywell Concrete & < 1E+0 < 1 E+0 < 1 E+0 < 1 E+0 < 1 E+0 Rebar 7 Totals 7.1 E+3 1.4E+3 3.4E+3 3.8E+1 1.2E+4 Not corrected for decay since identities not reported.

• Gibbs and Hill, Inc. Decommissionina and Decontamination Study Humboldt Bav Unit 3. March l

19.32 Calculated from Oak, H.D., et.al. Technology, Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, NUREG/CR 0672, Vol.2, June 1980 l

l 3-55 Rev 2 August 1998

l I

Table 3-7 Corrosion Film Radionuclide Inventory' <

Corrected to July 1984  !

1 (PNL - 4628)

Radionuclide Half-Life (years) Inventory (curies) i

'55p, 2.7E+0 6.9E+1 l SOCo 5.3E+0 1.2E+1 l 137Cs 3.0E+1 2.1 E+0 l 63Ni 1.0E+2 1.4E+0 90Sr 2.9E+1 1.7E-2 241% 4.3E+2 1.2E-2 238Pu 8.8E+1 6.8E-3 239,240Pu 2.4 E+4 6.1 E-3 244 '

~

Cm 54Mn 8.6E-1 3.3E-3 i 242Cm 4.5E-1 1.1 E-6 I

Excluding the reactor vessel, biological shield, concrete surfaces, and residues in tanks and sumps.

l l

i i

3-56 l ,

Rev 2 M

Ttbla 3-8 Radionuclida Iny:nt:ry Estimit s for Humboldt Biy R ctor Syst:ms, July 1984 (Based on Data from PNL,1983) 54 Mn 60 Co 125 Sb 134 Cs 137Cs 155 Eu 55p , 63 Ni 90 Sr 99Tc 239, 238 Pu 241 Am 244 Cm TOT mCl 240 Pu NUCLEAR STEAM SUPPLY Reactor Cleanuo Piping 1.1E+0 1.2E+3 3.4E+0 3.5E-1 3.1E+0 6.1E-1 3.1 E+3 1.5E+2 1.1E+0 8.5E-2 7.0E-1 6.9E-1 1.3E+0 3.5E-1 4.5E Reg:nerative Heat Ex. 1.4E+ 0 1.6E+3 4.4E+0 3.0E-1 4.0E+0 6.0E-1 4.0E+3 1.9E+2 1.75+0 1.2E-1 9.2E-1 9.2E-1 1.7E+0 4.6E-1 5.8E R:sh; storage Tank 1.65-2 6.0E+0 7.5E-2 3.0E-3 5.3E-1 5.0E-3 2.4E+3 6.3E-1 1.6E-2 1.6E-1 3.1 E-3 3.7E-3 7.6E-3 8.9E-3 2.4E Reactor Shutdown Coolina System

! Piping 2.0E+0 2.3E+3 6.1E+0 6.4E-1 5.7E+0 1.2E+0 5.6E+0 2.7E+2 2.3E+0 1.7E-1 1.3E+0 1.3E+0 2.5E+0 6.6E-1 2.6E R: actor Shutdown Cooler 3.6E+0 4.2E+3 1.1E+1 1.0E+0 1.0E+1 2.0E+0 1.0E+4 5.0E+2 4.2E+0 3.0E-1 2.3E+0 2.4E+0 4.5E+0 1.2E+0 1.5E Em roency Condenser System i l Piping 1.5E-4 8.3E-1 2.0E-3 1.1 E-3 3.7E-3 5.2E-4 7.9E+0 8.0E-2 2.5E-3 9.0E-3 2.0E-4 4.3E-4 3.5E-4 3.5E-4 8.8E l

i Em:rgency Condenser 3.1 E-4 1.7E+0 4.CE-3 1.4E-3 7.0E-3 1.0E-3 1.6E+1 2.0E-1 4.0E-3 2.0t' ?- 4.1 E-4 8.6E-4 7.4 E-4 7.4E-4 1.8E Suppression Tank Coolina and Core Sorav System t j Suppression Chamber 6.7E-3 4.4E+0 3.0E-3 1.4 E-2 6.5E-1 1.0E-3 9.0E+0 3.8E-1 1.0E-1 3.0E-1 1.5E-3 1.9E-3 2.5E-3 1.3E-3 1.5E Suppression Cooler 3.0E-4 1.8E-1 5.0E-4 7.0E-4 2.7E-2 6.0E-4 3.9E-1 1.6E-2 5.0E-3 1.3E-2 6.3E-5 8.0E-5 1.1E-4 5.6E-5 6.3E-TURBINE PLANT SYSTEM Turbin9 System l Piping 3.6E-1 4.9E+1 8.0E-2 1.8E-2 1.3E-1 9.2E-3 4.5E+2 4.3E+1 1.1 E-1 5.4E-1 1.1E-2 2.4 E-2 2.0E-2 2.1E-2 5.4E ll 3-57 Rev2 7 August 1998

Tcble 3-8 (Continued)

Radionuclide Inventory Estimates for Humboldt Bay Reactor Systems, July 1984 j (Based on Data from PNL,1983) 3 54 Mn 60 Co 125 Sb 134 Cs 137Cs 155 Eu 55p , 63 Ni 90 Sr 88 Tc 239, 238Pu 241 g 244 Cm TOT mCl 240 Pu .

Condensate System Piping 2.4E-3 1.2E+1 2.2E-2 1.6E-3 4.5E-2 3.1E-2 1.2E+1 1.0E+1 1.3E-2 1.0E-2 4.3E-2 2.4E-2 1.1 E-1 1.4E-2 3.4E Miln Condenser 2.1E+1 2.7E+3 5.0E+0 1.0E+0 7.0E+0 6.0E-1 2.6E+4 2.7E+2 6.1E+0 3.1E+1 6.8E 1.4E+0 1.2E+0 1.2E+0 2.9E Cond;nsate Demin. . 7.7E-3 3.0E+0 3.8E-2 1.5E-3 2.6E-1 3.0E-3 1.2E+3 3.1E+3 7.0E-3 8.0E-2 1.6E-3 1.9E-3 3.8E-3 4.4E-3 1.2E e

Feedwater System Piping 8.5E-4 5.1E+0 5.8E-2 3.6E-3 6.3E-2 6.5E-3 6.5E-1 6.4E-1 3.1 E-2 8.0E-3 2.5E-2 1.7E-2 4.2E-2 3.6E-3 6.6E WASTE DISPOSAL SYSTEM r

Liauld Waste Treatment -

Piping 2.4E-2 9.1E+0 1.1 E-1 4.4E-3 8.0E-1 7.8E-3 3.7E+3 9.4E-1 2.5E-2 2.4E-1 4.7E-3 5.7E-3 1.1 E-2 1.3E-2 3.7E  !

WLsta Receiver Tank 3.3E-2 1.3E+1 1.6E-1 7.0E-3 1.1E+0 1.0E-2 5.0E+3 1.3E+0 4.0E-2 3.3E-1 6.6E-3 7.8E-3 1.6E-2 1.9E-2 5.0E ,

Conc. Waste Tank 2.1 E-2 8.2E+0 1.T E-1 4.0E-3 7.3E-1 7.1E-3 3.3E+3 8.6E-1 2.1 E-2 2.2E-1 4.3E-3 5.1 E-3 1.0E-2 1.2E-2 3.3E t i'

Wrsts Hold Tank 2.2E-2 8.4E+0 1.1E-1 4.0E 'J 7.4E-1 7.8E-3 3.3E+3 8.7E-1 2.2E-2 2.2E-1 4.4E-3 5.2E-3 1.1E-2 1.2E-2 3.3E SERVICE SYSTEM Spent Fuel Service System i Fuel Basin-Walls 2.1 E-1 4.7E+1 5.0E-1 3.1 E+1 7.8E+2 2.9E-1 3.3E+2 1.0E+1 3.0E-1 2.7E+0 3.8E-2 4.2E-2 2.4 E-1 1.4E-1 1.2E

- Racks 3.0E-1 6.6E+1 6.5E-1 4.3E+1 1.1E+3 4.0E-1 4.5E+2 1.4E+1 4.1 E-1 3.7E+0 5.1E-2 5.8E-2 3.3E-1 1.9E-1 1.7E Fu:I Pool Cooler 4.6E-2 1.1 E+1 9.0E-2 7.0E+0 1.7E+2 6.0E-2 7.0E+1 2.2E+0 6.0E-2 5.9E-1 8.0E-3 9.0E-3 5.2E-2 3.0E-2 2.6E 3-58 l Rev 2 August 1998

i Tabla 3-9 Radionuclide Concentration in Concrete Cores' July 1984 (pCi/cm')*

Depth Sample (cm) 54Mn 55Fe 60Co 125Sb 134Cs 137Cs i 1 0-1 1.3E-2 c 1.8E+0 <9E-2 < 3E-2 5.2E+0  !

1-2 < 4E-3 c 1.2E-1 <6E-2 <3E-2 2.9E-1 2 0-1 5.4E-1 c 1.7E+2 1.8E+0 4.0E+1 2.1 E+3 1-2 < 2E-3 c 3.4E-1 4.7E-2 < 1 E-2 1.8E-1 3 0-1 1.2E-2 c 5.0E+0 < 4E-2 4.4E-2 2.4E+0 '

1-2 < 4E-3 c < SE-2 < 6E-2 < 3E-2 < 6E-2 d

3.6E+3 4 0-1 1.5E-1 2.8E+6 7.7E+1 < 1 E+0 4.0E+1 1-2 1.2E-2 c 6.8E+0 < SE-1 1.0E+0 6.3E+1 5 0-1 < 1 E-2 c 2.4E+0 < 3E+0 9.1 E+0 4.5E+3 1-2 < 3E-2 c 7.1 E-2 < 9E-1 < 2E-2 4.7E-1 f

f 6 0-1 1.5E-1 c 7.9E+1 < 3E+0 2.2E+2 1.0E+4 l 1-2 <4E-3 c 8.1 E-1 < 3E-1 1.9E+0 1.5E+2 7 0-1 <2E-2 c 1.9E+1 < 2E+0 1.1 E+2 6.3E+3 ,

1-2 <6E-3 c 2.0E+0 <1E+0 4.4E+0 5.5E+2  ;

8 0-1 2.8E+0 c 1.5E+3 6.5E+1 1.2E+2 4.0E+3 1-2 < 4 E-3 c 6.3E-1 < 9E-2 < 3E-2 2.3E-1 j 9 0-1 6.3E+0 c .

1.1 E+4 9.8E+1 1.0E+1 1.1 E+3 [

1-2 <7E-3 c 1.3E+0 <9E-2 <7E-2 2.2E-1 l 10 0-1 1.5E+1 c 1.4E+4 1.9E+2 1.0E+1 7.4E+2 l i 1-2 < SE-3 c 1.1 E+0 <9E-2 <3E-2 1.9E-1 l

i 11 0-1 < SE-3 c 1.2E+1 < 1 E-1 1.1 E-1 1.5E+1

! 1-2 < 7E-3 c 1.8E-1 < 9E-2 < 3E-2 9.3E-2

I 12 0-1 <8E-3 c 8.2E+0 <1 E-1 4.0E-1 2.1 E+1 f

1-2 <8E-3 c 6.1 E-1 <9E-2 <3E-2 1.4E-1 13 0-1 1.2E-2 c 1.1 E+1 <3E-1 4.4E+0 1.9E+2  !

1-2 <4E-3 c 1.1 E-1 <9E-2 <3E-2 1.8E-1  !

14 Whole <6E-2 c 2.4E+1 <4E+0 4.2E+2 2.2E+4 15 0-1 <9E-2 c 2.6E+1 <5E-1 2.7E-1 2.1 E+1 1-2 <4E-3 c 1.7E-1 <9E-2 <7E-2 1.1 E-1 l 3-59  ;

I Rev 2  ;

(

l Tabla 3-9 (Continu:d) l Radionuclide Concentration in Concrete Cores' i July 1984 (pCi/cm')*

l l Depth  !

j Sample' (cm) 54Mn 55Fe 60Co 125Sb 134Cs 137Cs {

! 16' 0-1 3.0E+1 2.1 E+8' 1.0E+4 1.7E+1 9.1 E+1 4.0E+3 '

l 1-2 4.0E+0 c 1.8E+3 4.2E+0 3.4E+1 1.4E+3 l

17 0-1 2.2E+0 c 7.1 E+2 1.4E+0 9.5E+0 4.8E+2 i 1-2 <5E-3 c 1.8E+0 <9E-2 <7E-2 8.9E-1 l 18 0-1 <2E-2 c 8.1 E+1 <9E-2 <3E-2 4.1 E+1 1-2 <8E-3 c 2.0E-1 <SE-2 <4E-2 <6E-2  :

1 19 0-1 2.6E-2 c 1.2E+1 <5E-1 9.8E+0 5.0E+2 1-2 ~<4 E-3 c 1.0E+0 <5E-1 <7E-2 2.0E+0 20 0-1 <4E-2 c 6.7E+0 < 1 E-1 5.8E-1 3.5E+1 .

l 1-2 <4E-3 c 1.6E-1 <5E-2 <3E-2 <7E+0 l 1

21 Whole 7.9E-2 c 1.9E+1 <2E-1 1.1 E+0 6.8E+1 22 0-1 <6E-2 c 2.8E+1 <5E-1 4.0E+0 2.6E+2 1-2 <9E-3 c 1.1 E-1 <9E-2 <3E-2 3.3E-1 23 0-1 7.0E-1 c 1.6E+2 <3E-1 4.4E-1 3.9E+1 1-2 <4E-3 c 2.2E-1 <9E-2 <7E-2 1.7E-1 24 0-1 c c c c c c 1-2 <4E-3 c 1.6E-1. <9E-2 <3E-2 <9E-2 l 25 0-1 2.8E+0 c 9.9E+2 1.1 E+0 2.2E+0 _ 1.2E+2 l 0-2 <6E-3 c 6.6E-2 <4E-2 <7E-3 1.7E-1 26 0-1 <4E-3 c 1.9E-1 <9E-2 <4 E-2 1.5E+0 1-2 <4E-3 c 4.7E-2 <5E-2 <3E-2 <6E-2 27 0-1 <4 E-2 c 5.3E+0 <3E-1 <3E-2 8.0E+0 1-2 <4E-3 c 9.4E-1 <9E-2 <3E-2 <5E-2 l a Excerpted from Residual Radionuclide Distribution and Inventory at the Humboldt i Bay Nuclear Power Plant. PNL-4628, May 1983, decay corrected to 7-1-84.

i 3-60 l l

Rev 2 August 1998

Tabla 3-9 (Continund)-

Radionuclide Concentration in Concrete Cores' July 1984 (pCi/cm')*

b To convert to pCilg multiply by 0,406.

c Not analyzed.

d pCi/kg.

e- Core taken over crack in concrete floor.

f Sample Locations

1. Sand Blast Pad

2. Hot Shop Floor Drain

3. Hot Shop Background

4. Radwaste Tank Area

5. Radwaste Tank Area

6. Radwaste Building

7. Radwaste Building

8. Reactor Building-66 ft l

9. Reactor Building-66 ft

10. Reactor Building-66 ft

11. Reactor Building-34 ft

12. Reactor Building-24 ft

13. Concrete Roof Over Conc Waste Tanks

14. Asphalt

15. Condensate Demin Room

16. Condensate Demin Room

17. Condensate Demin Room

18. Condensate Pump Room

19. Turbine Building

20. Turbine Building

21. Turbine Building

'l 3-61 Rev 2 August 1998

Tcbla 3-9 (Continusd)

Radionuclide Concentration in Concrete Cores' l i

July 1984 (pCi/cm')*

22. In Yard Near Stack  :

1

23. Condensate Storage Tank I

24. Air Ejector Room

25. Reactor Building Refuel Level

26. Intake Pumping Platform

27. Access Control Area l l

l 1

l l

l l

i l

I h

l 1

i i

4 3-62 l l

Rev 2 August 1998

Tabla 3-10 Radiation Survey-Refueling Building' Dose Rate

• Contamination Levels ( ci/100cm*)

. mr/h Contact

• Smearable Location Gamma' Beta Alpha Beta-Gamma Alpha Beta-Gamma *

+ 12 ft floor 10 <1 f 3.6E-2 3.9E-6 1.1 E-3 Elevation wall f 9.8E-3 2.2E-6 3.3E-4 floor 78 h f 1.6E-2 7.1 E-6 1.5E-3 Access Shaft

-2 ft El wall f 2.1 E-3 f 2.7E-5

-14 ft El floor 28 0 f 4.2E-3 4.7E-6 2.3E-3 wall f 2.4E-3 2.3E-6 7.6E-4

-24 ft El floor 18 h f 3.1E-3 1.4E-5 2.4E-3 wall f 1.0E-3 f f

-34 ft El floor in h f 2.1 E-3 1.2E-5 3.0E-3 wall f f f f

-44 ft El . floor . 78 1.5 f 8.3E-2 4.5E-6 1.3E-3 wall ,

f 1.0E-2 f 2.7E-5

-54 ft El floor 18 1.1 f 1.2E-1 4.5E-6 1.2E-3 wall f 2.1 E-2 f f

-66 ft El floor 12 0 f 1.4E-1 2.3E-6 6.1 E-4 wall f 6.4E-2 f f Cleanup floor 65 0 f 1.0E-1 2.1 E-5 9.4E-3 HX Room wall f 4.2E-2 f 1.9E-5

-2 ft El Cleanup floor 6 1.5 f 2.1 E-1 1.0E-4 4.2E-2 D9 min Room wall f 2.1 E-3 2.0E-6 3.5E-4

-2 ft El Shutdown floor 55 1.1 f f 3.7E-6 2.8E-3 HX Room wall f 2.1 E-2 2.8E-7 2.0E-5

-14 ft El West Wing floor 110 7.5 f f 1.2E-5 2.7 E-3

-66 ft El wall - f 9.6E- 2 5.6E-7 f Under floor 23 21 1.7E-3 2.0E+0 9.0E-4 3.3E-1 Reactor wall f 3.2E-2 6.5E-5 4.4E-3

-66 ft El 3-63 l Rev 2 August 1998

Table 3-10 (Continued)

Radiation Survey-Refueling Building

• Dose Rate" Contamination Levels fuci/100cm')

mr/h Contact' Smearable Location Gamma' Beta Alpha Beta-Gamma Alpha Beta-Gamma

• New Fuel floor 5 47 3.4E-4 2.3E+0 1.9E-5 5.4E-3 Vault wall f f 1.1 E-6 6.3E-4

+0 ft El TBDT Area floor 23 35 f 1.6E-1 4.2E-6 9.6E-4

-14 ft El wall f 3.4E+0 1.1 E-6 9.1 E-3 a Average values of PG&E Survey conducted May 1984 unless otherwise specified.

b lon Chamber.

c Minimum Sensitivity Alpha: Approx!.T.ctaly 1E-4 pCi/100cm2 Beta: Appraximately SE-3 pCi/100cm2 for Cutie Pie Approximately 2E-6 pCi/100cm' for HP-210 d Based on 137Cs.

e Based on 90Sr (10%),60Co (45%) and 137CS (45%).

f Not detected.

g Previous survey.

h Data not recorded.

3-64 l Rev 2 August 1998

I

, I Tcbla 3-11 Radiation Survey-Power Building' f Dose Rate" Contamination Levels (pcl/100cm')

. mr/h Contact

• Smearable d '

Location Gamma Beta Alpha Beta-Gamma Alpha Beta-Gamma

• Cond. Demin. floor 11 0 f 3.2E-2 8.5E-6 1.4 E-3 Cubicle wall f 3.2E-2 f 9.7E-5 ,

Cond. Demin, floor 14 1.5 2.6E-4 3.5E-2 1.1 E-5 2.7E-3 Regen. Room wall 1.0E-3 7.1 E-2 1.1 E-5 1.5E-3 Cond. Demin. floor 14g h f 3.5E-3 1.4 E-6 1.5E-4 Op. Area wall f 8.8E-3 f 6.1 E-5 Cond. Pump floor 139 h f f 2.0E-6 5.0E-4 Room wall f f f 2.8E-5 Air Ejector floor 55 56 f 5.6E+0 1.7E-6 7.8E-2

. Room wall f f h 1.5E-3 Condenser floor 19 <1 f 6.0E-3 5.7E-7 5.7E-4 f Area wall f f h h Pipe Tunnel floor 15 1.5 f 4.7E-3 1.lE-6 2.9E-4 wall f f 1.4E-7 2.1 E-5 Feed Pump floor < 1g h f 5.2E-4 f 8.4 E-5 l

Room wall h h h h Seal Oil floor 0.005 g h f f f 2.1 E-5 Room wall h h h h ,

t Turbine Enc floor <1,g h f 3.1 E-3 8.5E-7 1.2E-4

+27 ft El wall f 4.2E-3 2.8E-7 f Turbine floor < 1g h f 1.0E-3 1.7E-6 6.1E-5 Washdown Area +27 ft El Hot Lab floor < $g h f 1.2E-2 f 7.3E-5 Laundry / floor < 1g h f 2.6E-3 4.3E-7 7.7E-5 Demin Area t

l

+27 ft El Laundry / floor h h f 1.0E-3 f 2.0E-4 Hot Lab

+34 ft El l 3-65 Rev 2  :

August 1998  ;

Tcbla 3-11 (Continu::d)

Radiation Survey-Power Building' Notes:-

a Average values of PG&E Survey conducted May 1984 unless otherwise specified.

b lon Chamber.

c Minimum Sensitivity Alpha: Approximately 1E-4 pCi/100cm2 Beta: Approximately SE-3 Cl/100cm' for Cutie Pie Approximately 2E-3 Ci/100cm for HP-210 d Based on 137Cs.

e Based on 90Sr (10%),60Co (45%) and 137 Cs (45%).

'f Not detected.

g Previous survey.

l h Data not recorded.

l i

i 3-66 l

Rev 2 August 1998

[

Table 3-12 Radiation Survey- Yard Structures  ;

Dose Rate

• Contamination Levels (pci/100cm')

mr/h Contact

• Smearable '.

d Location Gamma Beta Alpha BothCamma Alpha Beta-Gamma" Hot Shop floor < $g h f 1.3E-2 4.5E-5 6.0E-3 wall f f f f Calibration floor < jg h f 2.5E-3 f f facility l

Stack

-0 ft El floor f 4.3E-2 f 1.1 E-4

+12 ft El - floor 1.8g h f 9.3E-3 5.6E-7 1.9E-5

+26 ft El floor f f 5.6E-7 1.9E-4 Radwaste floor 15 6.8 f 4.9E-1 1.0E-6 4.2E-3 Treatment wall f 6.5E-3 7.0E-7 1.9E-4 Low Level floor 190 7.5 f 3.8E-1 h h Waste Building j Radwaste floor 59 h f f 2.8E-7 1.5E-4 Handling Building a Average values of PG&E Survey conducted May 1984 unless otherwise specified.

b lon Chamber.

c Minimum Sensitivity i l Alpha: Approximately 1E-4 Ci/100Cm2

( Beta: Approximately SE-3 pCi/100cm* for Cutie Pie  ;

Approximately 2E-3 pCi/100cm' for HP-210 d Based on 137Cs t l 3-67 Rev 2 August 1998

Table 3-12

. Radiation Survey-Yard Structures i 1

i 1

e Based on 90Sr (10%),60Co (45%) and 137 Cs (45%) i i

f Not detected -

g Previous survey ]

i h Data not recorded l I

\

l 1

l l

4 1

i

l l

l i

i i-3-68 l Rev 2 August 1998

l Table 3-13 Cross-connections Radionuclide Analysis' IN-PLANT LIQUIDS (uCi/ml) 1982 - 1983 60Co  % MPCb 134Cs  % MPCb 137CS  % MPCb  ;

Boiler No.1 c c c Boiler No. 2 e c c Turbine Lube e c c Oil INSTRUMENT AND SERVICE AIR (uCi/ml) 1982 - 1983 ,.

SOCo  % MPCd 134Cs  % MPCb 137CS  % MPCb Cond. Demin.

Room c c c i

+12 ft Elevation c c c t

instrument Vault c c c j YARD DRAINS (uCi/ml) 1981 - 1983 i

% MPCb ,e 134Cs  % MPCb ,e 137CS  %

l 60Co i

MPCb,e Unrestricted ,

2.8 E-8 9.3 E-2 c 7.4 E-8 3.7 E-1  !

No. Loop Unrestricted  ;

So. Loop 2.2 E-7 7.3 E-1 3.3 E-9 3.7 E-2 1.1 E-7 5.5 E-1 Oily Water' Drain System 2.9 E-5 g 6.0 E-8 g 3.6 E-6 g Restricted No. Loop 1.1 E-6 1.1 E-1 4.1 E 6 1.4 E0 1.5 E-5 3.8 EO Restricted '

So. Loop 1.6E-6 1.6E-1 c 8.4 E-7 2.1 E-1 l a Average above background.  ;

b 10 CFR 20 Appendix B, Table 11, Column 2 (most restrictive value).  ;

c Background.

d 10 CFR 20 Appendix B Table ll, Column 1 (most restrictive value). ,

e 10 CFR 20 Appendix B. Table I, Column 2, (most restrictive value).  !

f Sludge sample from oily water separator ( Ci/g) 1982.

lg Not Applicable.

l 3-69 Rev 2 August 1998 ,

8 "

"c ogi gonroe O^O{,,

O *O

"~~ ^ 4 g

O mp. o u . ) \. -

3 ocricES O B y a m i'

,pa c**'* 4 f

i s2 kM*'D I

WtREFENCE CHAf88 LNHC FT98CE FEtt e- 100 200 300 800 y,,,,

- A- MAD WASTE HANDttNG BUILDMG s- tow teveL stonAios suuma

. c--soup wAsie vAutT HUMBOLDT BAY POWER PIANT SITE PLAN ,

FIGURE 3-1 3-70 l Rev 2 August 1998

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

. \

l 1

i c

c'

' CALLED NOftTH r 1 ,

MAKE UP DEN!NERALIIER 25 TON REFUELING BUILDING CRANE RML j ROON l

[. 3 llIIni!r!v i e up

-- - H ATCH g ,

1 VESYl8ULE O LAUN AIRLDCK -

WASH DOWN t.

\) _

AREA L E i

jj I .I -4 f

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f f TURBINE l

1 E $ j ENCLOSURE B AR R tER

/

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/ ' "*' '

l TomET /' t y LOCKER a .

j '- W A L KwAY ,

O . E 4- HATCH l

! r l r h 7 l E INSTRUMENT CONTROL RCPAIR ROOM ,

FIGURE 3-2 HUMBOLDT B AY UNIT' S OPERATING FLOOR PLAN (EL 27'-0")

1 .

l 3-71 Rev 2 August 1998 ,

9

+0

[ _ CALLED le0RTH ,

CO8eDEktaTE DEWINt4AL1ZEA ROOM E

• I*II" ) gaygg

! . ,M, UP

1. j N

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e _ -m... =L - -

m i l s t. <-i , ."

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/ LADOEM NO.3 t

_-IP =

WASM ARIA ,

,eESCAPS

% UP w en g staTCa4 6 (U

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,3 4

- FIGURE 3-3 HBPP UNIT 3 GROUND FLOOR PLAN 4

3-72 l Rev 2 August 1998 i

l

f

^

N.

O

• CALLED NORTH w,,, o f h NEUTRON TANK , f Call 8AAM

3. FACILITY

? r a

harr.. --173 b HOT MACHINE SHOP TbOL

STORAGE DECONTAMINATION f " 5; f

, AREA .

TOOL CRB

. n. .

N b - ,1a psi .

DE NTAMINATION

% i

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G c11, . .,,,,a FIGURE 3-4. HOT MACHINE SHOP AND CAllBRATipN FACILITY l 3-73 Rev 2 August 1998

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

e DEMtNERAUZER CONTROL PANEL MAKE.UP , STELLAR l DEMiNERALIZER -

FILTER PANEL.

ELECTRIC - FILTER WATER HEATER AID PUMP CALLED NORTH REFUELING BLDG. " FILTE R DOOR OPER. PANEL AID TANK '

i

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CONDENSER p HATCH i

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/ N -EMERGENCY SECTION NCC NO.10 FEEDWATER VALVE STEAM BYPASS SYSTEM CONTROLCTR CONMCL CABINET .

CABIN ET - Hay CONTROL DOARD REACTOR BOARD

{

TURBINE GENERATOR BOARD EL 2T- 0" l

• i ,

FIGURE 3-5 EQUIPMENT LOCATION - O'PERATING FLOOR PLAN I ,3-74 l l Rev 2 l August 1998 l

1

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FIGURE 3-6 EQUtPMENT LOCATION-GROUNG FLOOR Pt.AN l 3-75 Rev 2 August 1998

-e c04MissGFOOM "A"*"

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SECTION B-B FIGURE 3-8 EQUIPMENT LO' CATIONS 3-77 l Rev 2 August 1998 I

_ _ = _ _

q. ___

_ l TOP CRA A/C MA_tL _

l ci. 4 7 .4 - [l ~

hl S DATTERY MACM MCATw0 ANOm VENT UNir TVMetNC DCMiht'RAL.!E EM ,_

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= MCC NO. Et%_

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tt_f7* O*

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cowocusare puun recom %alano sesi. conoCuscn SECTION 0-C FIGURE 3-9 EQUIPMENT LOCATIONS 3-78 l Rev 2 August 1998

8 o

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i DMYCR NO. 3 scacron trea FOMP M00M SECTION 0-0  !

t EQUIPMENT LOCATIONS FIGURE 3-10 t i 3-79 Rev 2 August 1998 i

l i i

t HEAT EXCHANGERS

( ,

}

l s.

l .. - .

. CRANE E REFUELtNG. BLDO. . "! l l x Fa=W R E SIN. -

JIS. CR ANE  ; J EMERGENCY STORAGE. CI POISON TANK N '

p g-o il TANK g 4p ..

z I

lj 'i:'%

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.. E.-.r,- ) (e -<j.

/; l j. j LOADING. PIT ,

i.;

i n

L .?" -

0 pN R%dd 4fgfY46fg' SUPPRESSION  %-

 : ..- -- N CHAMBER  %,

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l FIGURE 3-19 SCHEMATIC DIAGRAM OF REAOTOR PRESSURE VESSEL AND INTERNALS FOR HBPP UNIT 3 3-88 l Rev 2 August 1998

4.0 SAFSTOR ACTIVITIES 4,1 OBJECTIVES The objectives of SAFSTOR activities are:

o To secure non-operating plant systems to prevent deterioration and minimize potential for release of contained radioactivity o To process and dispose of radioactive wastes generated during SAFSTOR t decommissioning activities .

o To decontaminate pl_ ant facilities to the maximum extent practical to minimize the potential for spread of contamination outside of the facility and to minimize the .

requirements for periodic surveys o To reduce general area radiation levels in the vicinity of equipment operated or maintained during the SAFSTOR period to as low as reasonably achievable (ALARA)

'o To maintain plant facilities to support long-term storage of spent fuel, to minimize +

generation of radioactive wastes, and to minimize necessary maintenance and surveillance during SAFSTOR o To establish baseline conditions and a monitoring and surveillance program for the SAFSTOR period 4,2 PLANT STAFF ORGANIZATION AND RESPONSIBILITIES i

, 4.2.1 SAFSTOR Organization l

Th3 following describes the structure and responsibilities of the SAFSTOR organization:

l Senior Vice President and General Manaaer. Nuclear Power Generation (SVP&GM).

Tho SVP&GM has corporate responsibilities for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety.

1 l Vice President. Nuclear Technical Services (VP.NTS). The VP,NTS reports to the l SVP&GM and provides oversight to both the fossil and nuclear activities at HBPP. The

, VP, NTS is the corporate officer designated as the Approving Officer for HBPP Unit 3 SAFSTOR operations and future decommissioning projects.. The VP,NTS serves as chairman of the Nuclear Safety Oversight Committee (NSOC) which, under the Unit 3 l 4-1 ,

t Rev 2 August 1993

i l

i l

I l

I Technical Specifications, provides independent review and audit of SAFSTOR activities.

HBPP Plant Manaaer. The Plant Manager has overall responsibility for operation,  ;

' maintenance, engineering, radiation protection, training, security and quality control for l

SAFSTOR and decommissioning activities. As chairman of the PSRC and the Plant '

ALARA Committee, the Plant Manager is responsible for ensuring that these committees fulfill their responsibilities as described in the HBPP Unit 3 Technical Specifications. The Plant Manager reports to the VP, NTS.

4.2.2 Offsite Support i

Diablo Canyon & General Office Support. Support for HBPP Unit 3 SAFSTOR activities are available from several departments located at the Diablo Canyon Power Plant and PG&E's General Office. Such technical support covers all functions performed at Humboldt.

,Three Diablo organizations actually have matrixed staff present on the Humboldt site.

These include Quality Assurance, Licensing and Budget & Performance Management.

The Nuclear Quality Services (NQS) Department continues to be responsible for the HBPP Unit 3 Quality Assurance Plan and for performing audits to verify that SAFSTOR activities are performed in accordance with the DSAR and the HBPP Unit 3 License and Technical Specifications.

The Regulatory Services (RS) Department provides assistance in preparing license amendment submittals and coordinating responses to NRC correspondence.

The Budget & Performance Management Department provides assistance in the preparation ofjob estimates and budgets during the SAFSTOR period. In addition, the Procurement Services Section of the Site Services Department provides contract ~ .

administration services for contracts used to obtain assistance in performing' activities during the SAFSTOR period.

4.2.3 Staffing During SAFSTOR

~

Key positions in the plant organization during the SAFSTOR period are described in the i HBPP Unit 3 Technical Specifications. During this period, sufficient expertise will be

) maintained to perform the required maintenance, operations, and surveillance activities l

' for the plant. Contractor assistance will continue to be utilized to perform services beyond the capabilities of the plant staff.

The permanent plant staff for operation and maintenance of the fossil-fueled units and gas turbines, and for SAFSTOR maintenance of HBPP Unit 3 during the SAFSTOR period, is estimated to vary between 60 to 80 PG&E employees.

4-2 l Rev 2 l

August 1998

All plant employees currently reside in Humboldt County. The majority reside in Eureka, Arcata, McKinleyville, Fortuna, and Ferndale. The remaining employees live in other~ smaller communities or unincorporated areas.

No traffic congestion or problems of interference with local traffic are expected to occur due to SAFSTOR activities. The staffing levels at the plant will not be significantly increased due to SAFSTOR activities. Waste shipment traffic is expected to be infrequent with periodic increased activity to accommodate certain decommissioning projects.

4.3 ADMINISTRATION AND CONTROL 4.3.1 Cost Estimates and Financing The project to place HBPP Unit 3 into the custodici SAFSTOR mode cost approximately

$14 million (direct costs plus indirect costs and overheads).

.During the SAFSTOR period, funds are required to maintain the unit in accordance with the possession-only license and associated technical specifications. Maintenance and surveillance activities during the SAFSTOR period are similar to activities conducted during operation, but reduced in scope. Funds for SAFSTOR costs are allocated in the Humboldt Bay Power Plant annual operations and maintenance badget.

Pacific Gas and Electric Company Application Number 83-09-049," Authority to increase its Electric Rates to Reflect Retirement and Decommissioning of Humboldt Bay Power Plant Unit 3", was filed with the Public Utilities Commission of the State of California on September 19,1983.

l In Application No. 83-09-049, PG&E requested rate recovery for the costs associated with the retirement of the plant, including ine cost of decommissioning HBPP. Unit 3.

PG&E requested that the unrecovered capital expenditures for HBPP Unit.3 be '

included in the electric rate base. PG&E also requested permission to recover through rates the costs associated with placing HBPP Unit 3 into the SAFSTOR condition.

i The California Public Utilities Commission (CPUC) has authorized that the HBPP Unit 3 operation and maintenance expenses including certain expenditures for SAFSTOR l activities be recovered in rates. The CPUC has granted.the Company's request to collect the remaining SAFSTOR costs from PG&E customers. This authorization is i expected to continue during the SAFSTOR period.

All electric generating facilities are eventually decommissioned. The estimated costs for the removal of the plants are included as part of depreciation and recovered liirough j rates. Accordingly, the costs for SAFSTOR are part of the costs of removal and will be )

amortized together with PG&E's unrecovered capital. PG&E requested that the CPUC authorize recovery of the estimated nonrecurring cost of $10 million that PG&E used to l 4-3 Rev 2 l August 1998

1 i

place HBPP Unit 3 into SAFSTOR. PG&E proposed, however, that only the actual SAFSTOR costs be recovered. The CPUC reviews actual SAFSTOR costs in PG&E's gener'ai rate case and adjusts the amortization rate to incorporate the recorded SAFSTOR costs. An additional $4 million of initial SAFSTOR work was recovered as part of operating and maintenance costs.

The annual, continuing costs of maintaining the unit in SAFSTOR will be included in base rates. These annual costs will be reviewed as part of PG&E's periodic general rate case.

During the SAFSTOR period, plant workers responsible for the maintenance and operation of fossil-fueled Units 1 and 2 will provide the maintenance and surveillance for Unit 3.

The dollar estimate of the taxes attributable to HBPP Unit 3 during the SAFSTOR time period is zero dollars per year. The State Board of Equalization valued HBPP Unit 3 at essentially zero in 1984. The local taxing jurisdiction did not levy a tax on the unit

.beginning in 1984, and is not expected to levy a property tax on the unit during the SAFSTOR time period.

PG&E proposed that the CPUC establish an external fund, which would qualify under the Tax Reform Act of 1984, for the purpose of accruing the required funds for decommissioning. The cost to decommission was accrued over 5 years, starting in 1987, and the trust fund was determined to be fully funded by 1991. The 5 annual payments to cover the decommissioning costs were invested in an external sinking fund (under the Tax Reform Act of 1984) and managed by an outside party. For the eventual complete decommissioning of Unit 3, PG&E contracted TLG Services, Inc., to prepare a site-specific decommissioning cost estimate in 1997. Information pertaining to the cost estimates for complete decommissioning of Unit 3 is contained in the PSDAR.

4.3.2 Procurement For services that may affect quality, contractors are required to have an approved quality assurance program that is reviewed by PG&E. Contracts and contract revisions are reviewed by NQS to verify that the contractor's quality assurance program is adequate.

PG&E standard practices and administrative procedures govern the selection of contractors and the administration of contracts.

I r

i 44 l Rev 2 August 1998

4.3.3 Training Program 4.3.3.'1 Training Program Description PG&E has established general employee training (GET) requirements for PG&E and contractor employees who work in Unit 3. In addition to GET, programs have been designed to assist personnel with technical aspects of their work. Such topics include Hazardous Material (Waste) Program Training and Radiation Protection Technician Training. Additional topics may include such topics as Radioactive Waste Volume Minimization, Contaminated Asbestos Materials, and Decontamination Workers Training.

Personnel who enter Un[t 3 for the purpose of conducting work need to have basic ,

knowledge of HBPP andits procedures. Initial training is given prior to any assignment of work in Unit 3. Training may be accomplished through the use of formalized classroom lecture (s), video / cassette tapes, Computer Based Training, and/or handouts.

.The level of training provided to employees is based upon a review of the information employees will require in order to perform their job duties safely and efficiently.

l Consideration is also given to the employee's past experience and training. The program provides the flexibility for making the decision on a case-by-case basis.

l .

! In addition, special training will be provided as needed when it is deemed necessary or prudent to assist employees involved with unusual or infrequent procedures associated with decommissioning activities. Special training relating to decommissioning activities may include such topics as radioactive waste volume minimization, handling of contaminated materials, and decontamination workers training. Employees actively

involved with such activities will receive special training appropriate to their job duties and responsibilities as necessary and on a timely basis.,

4.3.3.2 General Employee Training The GET Program provides the flexibility to adapt training depth and scope to personnel needs commensurate with job duties and responsibilities, the areas of HBPP l to be entered, and prior experience. Presentations and lessons are supplemented with l videotapes, written handouts, and other techniques designed to improve training efficiency.

Description of Plant and Facilities. Three levels of trainina are available.

Level t " Plant Orientation" training is given to all new employees and visitors depending on the scope of work to be performed. The training provides orientation of site restrictions and rules, and includes an overview of the structures, allowable entry areas, escorted access, security areas, inadvertent entry and alarms, and radiation hazards as appropriate.

l 4-5 Rev 2 August 1998

l l

l ,

i l

l Level 2; " Power Plant Overviews" training is given to all new permanent plant staff in l addition to the Plant Orientation, Level 1 training described above. This training  !

provides more detailed information regarding plant equipment, operation, and organization. Additional activities included in this training are: QA/QC; radiation j protection; emergency plan training; security plan training; and fire brigade introduction.

! I If appropriate, material safety and hazardous materials management are also included l as part of the Power Plant Overview, Level 2 training.

Level 3; " Detailed Description of Plant Systems," training is detailed technical training reserved for workers engaged in operation, maintenance, or testing of systems. This training also includes any annual retraining and requalification requirements.

General Site Rules. This session discusses plant policies on health and safety, vehicle l

operation, firearms, liquor and drugs, cameras and other personal items, as well as any disciplinary actions which thay be accorded to policy violations.

Radiation Protection Proaram. The sophistication of information presented depends l upon the responsibilities of the individual.

Radiation Protection Engineers; Radiation and Process Monitor Foremen; and Radiation and Process Monitors positions are classified as radiation protection

" professionals" for decommissioning activities. l Individuals in these positions are responsible for the implementation and day-to-day work associated with the Radiation Protection Program. Persons in these positions have prior training / experience in the radiation protection field and/or receive extensive training in the radiation protection procedures and work practices at HBPP.

Radiation protection " professionals" are given comprehensive training in radiation protection, covering theory and practical experience with equipment at a technical level beyond GET.

Radiation workers complete a training course that provides classroom instruction and practical demonstrations to permit them to perform their work in an efficient, safe manner. Topics include: radiation types, exposure, and biological effects; health protection problems, ALARA philosophy and program, and NRC rules and regulations; exposure reports, protective devices, routine radiation' work permits (RWPs), and special work permits (SWPs); and high hazard areas, contamination, and decontamination. Training is based upon 32 key questions identified in Regulatory Guide 8.29," Risks From Radiation Exposure." An examination is administered with a minimum passing grade of 70 percent. Non-radiation workers are provided fundamentalinformation on radiation, health effects, and risks, as necessary. l l

l Respiratory Protection. Individuals required to wear respirators are medically qualified, fit tested, and trained in the proper care and use of respirators. When negative 1

4-6 l Rev 2 August 1998

i i

pressure respirators are used, training includes qualitative fit testing. Individuals are acquainted with OSHA, Cal OSHA, NRC, and ANSI requirements for respirators, their I use, snd when they should be used.

Site Emeraency Plans. Basic instruction helps individuals to recognize and respond  !

correctly to emergency or warning signals and how to report fires or injuries. Annual t emergency drills and exercises are conducted to demonstrate proficiency in various aspects of site emergency plans.

Industrial Safety. First Aid and Fire Protection. PG&E's Accident Prevention Program I provides basic instruction in these topics. Persons having unescorted access to Unit 3 must know the significance of barrier tapes and equipment status tags and must be familiar with fire prevention and fire fighting techniques. Individuals team to use water i

and CO2system extingui.shers and fire brigade equipment. A fire brigade training program is provided to fire brigade members for more extensive practical training. ,

Security Proaram. Individuals granted unescorted access to Unit 3 are required to understand the security program including badging, procedures for entering secure ,

' areas, measures to avoid causing security alarms, recognition and reporting of unauthorized personnel, and searching policy.  ;

Quality Assurance / Control Proaram. Individuals who perform quality-related work  !

receive training in the use of procedures, the documentation of work, discrepancy reporting, and the role of Quality Control inspectors.

4.3.3.3 Technical Training l in addition to GET as discussed above, some workers are provided with more extensive technical training appropriate for their work assignments. As an example, hazardous  ;

material and waste handling training is provided when necessary to supply information ,

on waste management, environmental protection, safety standards, and personrjel protection. .

Operator Trainina and Certification Proaram. Under the Unit 3 possession-only license, all fuel movements are required to be performed under the supervision of a Certified Fuel Handler and a Certified Fuel Handler is to be present at the location of each fuel )

movement. .  ;

i During the SAFSTOR period it is not expected that movements of spent reactor fuel will l be made, except for training, special tests, or inspections to monitor the fuel in storage.

At some time during the SAFSTOR period, fuel handling may be performed to transfer the spent fuel assemblies to the DOE for disposal.

l A training and certification program has been implemented to maintain a staff properly trained and qualified to maintain the spent fuel, to perform any fuel movements that may be required, and to maintain Unit 3 in accordance with the possession-only license.

l l 4-7 '

Rev 2 August 1998

l This program provides the training, proficiency testing, and certification of fuel handling

! personnel. A detailed description of the program is provided in Appendix II. i The Operator Training and Certification Program ensures that people trained and qualified to operate Unit 3 will be available during the SAFSTOR period. This program  !

is similar to that required by 10 CFR Part 72, Subpart I for Independent Spent Fuel l Storage Facility personnel. Licensee certification of personnel makes it unnecessary for the NRC to periodically conduct license examinations for persons involved in infrequent activities and prevents delays due to obtaining NRC Fuel Handler Licenses for any evolutions that may require fuel movements.

Radiation Protection Department Trainino Procram. A comprehensive program is presented to HBPP Radiation and Process Monitors (RPMs). The training consists of academic classroom trainirtg, on-the-job training, and retraining to implement changes and improve skills. The course requirements include:

e Nuclear Technoiogy - basic nuclear and radiation protection theory

. Plant Design and Operation - plant layout, system functions, and equipment ,

l

• Chemistry - analyses, calibration, and instrumentation l l

e Radiochemistry - sample preparation, counting, and data reduction. Use and maintenance of instruments

• Emergency Plan and Procedures - emergency responsibilities, surveys, analysis, radiation protection during accident conditions, and environmental monitoring

• Radiation Protection -in-depth training is presented on atomic theory, radioactivity, properties of types, units and dose, biological effects, standards, detection, l dosimetry, instruments, personnel monitoring, air sampling, instrument operation, counting statistics, internal dose calculations, shielding, exposure control,' surveys, respiratory protection, and decontamination methods l

Use of Monitorina Eauioment. RPMs are provided with on-the-job training to enhance skills for operating various types of radiation detection equipment. Information 'l discussed includes the proper use of probes, read-out evaluation, surveys and data to be collected, instrument response time, efficiency, and modes of operation.

Individuals required to use monitoring equipment are required to demonstrate l proficiency with the equipment. The length of training is dependent on the individual's i

ability to perform his duties satisfactorily. As part of the RPM's periodic retraining, proficiency regarding monitoring equipment is verified every 2 years unless new equipment or procedural changes dictate a more frequent proficiency check.

1 4-8 '

l Rev 2 August 1998

RPMs are required to be requalified at least annually by written examination in addition, every 2 years, RPMs are required to demonstrate satisfactory performance of  ;

the practical skills covered in the initial qualification program. j The Radiation Protection Engineers and RP Foremen are qualified by training and f experience, and are not required to be requalified.

Quality Control Deoartment Trainino Proaram. Initial training of Level ill Quality Control j i

(QC) Inspector (Supervisors) and Level ll QC Inspectors (Journeymen) is provided by l

the Nuclear Quality Services Department. This instruction is supplemented by a review of plant procedures.

4 Refresher training for QC Inspectors is provided annually and may include any of the t i

following: -

• Procedure review with emphasis on new or revised procedures

. Classroom training utilizing videotape and/or lectures

• Participation in specialized training offered outside PG&E l

Additionally, Nuclear Quality Services Department provides annual training for Level lli QC Inspectors. ,

l 4.3.3.4 Other Training -

it is anticipated that other technical topics will be presented to personnel on an as-needed basis. Current administrative guidelines will be followed to establish new l procedures and to ensure the training is completed. Suggested topical outlines are i ,

presented for two such courses i

Radioactive Waste Volume Minimization j The course should identify the methods' and techniques of achieving the following [

goals

. Liquid radioactive waste control I

. Control of material entering radioactive materials area l . Contamination control l . Waste segregation program j

. Radiological work planning i

. Decontamination process ,

. Waste packaging and transport (RPMs only) ,

l l 4-9 Rev 2 August 1998

[

) Additional special topics may be added to this list if the need arises to assure safe and timely execution of new work tasks. Training on any such topics will be presented to those* persons who will actively participate in those new work tasks.  ;

j Decontamination Workers l

Decontamination workers should be instructed in the methods, applicability, and techniques for decontamination. The areas of discussion and demonstration include:

. Categories of processes:

l L - Chemical decontamination l

- Manual and mechanical decontamination

- Electropolishing :

- Ultrasonic decontamination l . Decontamination effectiveness

! . Processing requirements

. Solvent / system interactions

. Equipment

. Surface decontamination of concrete structures i 4.3.3.5 Training Program Administration and Records The HBPP Plant Manager is responsible for ensuring that the training requirements and l programs are satisfactorily completed for site personnel. The HBPP Training

!. Coordinator is responsible for the organization and coordination of training programs, ensuring that records are maintained and kept up to date, and assisting in training material preparation and classroom instruction. .

Training that is required to satisfy a regulatory or procedural requirement is documented on a record of training sheet and accompanied by an attendance sheet. The training topic is identified on the record sheet and any applicable training materials is included or referenced as part of the package.

Records of training and qualification are retained for the duration of an individual's assignment to HBPP or for 5 years, whichever is longer. An exception to this are radiation protection training records which are retained for the duration of the facility license.

L Pursuant to 10 CFR 50.120, "each nuclear power plant licensee, [by November 22, l 1993], shall establish, implement, and maintain a training program derived from a

systems approach to training as defined in 10 CFR 55.4." The intent of 10 CFR 50.120 i is to ensure that civilian nuclear power plant operating personnel are trained and i

4-10 l Rev 2 August 1998

i i

I l

qualified to safely operate and maintain the facility commensurate with the safety status ,

of the facility. '

On December 9,1993, the U. S. Nuclear Regulatory Commission granted an exemption  !

from the requirements in 10 CFR 50.120 to establish, implement, and maintain training  !

programs, using the systems approach to training, for the categories of personnel listed in 10 CFR 50.120. .

Exemption from the training rule,10 CFR 50.120, does not relieve HBPP of any other training requirements or commitments which have been established with the NRC. l 4.3.4 Quality Assurance Program Decommissioning and SAFSTOR activities will be performed in accordance with the Humboldt Bay Power Plant Unit 3 SAFSTOR Quality Assurance Program (QA Program). The QA Program is designed to ensure that decommissioning activities and ,

activities during the SAFSTOR period are performed in accordance with the license,

. applicable codes, standards, and regulatory requirements, and that these activities will .

provide adequate protectio.n for the health and safety of the public. Items and activities subject to the QA program include, but are not necessarily limited to:

• Radioactive material licensed shipping containers, and activities which could affect the required function thereof, as required by 10 CFR 71. This applies to shipment of l licensed materialin excess of type A quantities. .

i

• Effluent and environmental monitoring equipment, and the activities that could affect the validity and accuracy of such measurements, as required by USNRC Regulatory Guide 4.15.

• Activities required by the Technical Specifications. .

The QA Program is implemented by quality assurance procedures and HBPP procedures and instructions. l L

4.4 DECOMMISSIONING ACTIVITIES l  !

g The following sections describe the major activities that have been performed as part of l the project to place Unit 3 Into the custodial SAFSTOR mode. For those activities for  ;

which specific tasks can be identified, a general description of those tasks is included.

Detailed descriptions of specific tasks were defined at the procedural!evel at the time j the task was performed. l l 4-11 ,

Rev 2 August 1998

4.4.1 Preparations for SAFSTOR Decommissioning (This " activity is not considered part of decommissioning but is included for completeness.)

Systems and equipment not required by the Unit 3 Operating License for the Cold Shutdown Mode and not required to support SAFSTOR decommissioning activities l were secured in preparation for the decommissioning. Preparations included unloading the reactor core; drainir;g, flushing, and securing systems; deenergizing instruments and controls which are no longer required; and isolating non-operational systems from systems still in operation.

Securing and lay-up of these systems and components were performed in accordance with procedures reviewed and approved by the PSRC. The reviews included a determination that the system and equipment lay-ups did not involve an unreviewed safety question as defined in 10 CFR 50.59(c) or a change to the Unit 3 Technical Specifications.

As a result of ALARA considerations during the performance of system lay-ups and decontamination work, ce.%n piping sections or components were removed. For systems that are to remain secured for the SAFSTOR period, the piping and equipment l was not replaced. Open pipes were sealed to prevent the spread of contamination. l I

Also during the preparations for SAFSTOR decommissioning, some radioactive wastes t on-site were processed and shipped to licensed disposal facilities. These wastes were primarily radioactive wastes that had been generated during the years that Unit 3 had operated and had been stored on-site awaiting final disposal. Additionally, liquid wastes generated as a result of draining and flushing plant systems were processed by 1 the radioactive waste treatment sys;em. I 4.4.2 System Layup l During SAFSTOR decommissioning, systems no longer required by the SAFSTOR l Technical Specifications were secured and isolated. In addition, systems that were required to support SAFSTOR decommissioning activities but which will not be required l ,

. during SAFSTOR were secured upon the completion of those activities. The objectives of the system lay-up are as follows:

• Systems containing fluids were drained to the maximum extent practical.

• Significant sources of radiation in areas that will be routinely accessible during l

SAFSTOR were either removed or shielded.

4-12 l Rev 2 August 1998

- i l

l

• Connections between secured systems and operating systems were sealed by l either using blank flanges or by cutting and capping the lines. This prevents leakage from an operating system from refilling a system that has been drained.

. Motors, valves, instrumentation, and other electrical components associated with secured systems were deenergized.

The following system descriptions describe the SAFSTOR status of systems that will not be operational during the SAFSTOR period.

Reactor Vessel and Intemals. Following the unloading of all fuel assemblies from the reactor vessel, the irradiated incore fission chambers were removed and stored in the spent fuel storage pool. Control rods and all core internals were left in place. The reactor vessel bottom was vacuumed to remove any existing residue and the reactor l vessel head was installed with the stud nuts hand-tight to facilitate their removal at the i . time of final decommissioning. The main steam line spool piece, safety and relief valves, and reactor associated piping were reinstalled. A blank was installed in the flange on the steam line end of the main steam spool piece and a gap was left in the flange at the reactor vessel to permit venting of the vessel during the vessel drain.

s Following reinstallation of the upper drywell head and reactor shield plug and following closure of the lower drywell head, the reactor vessel was drained. The reactor vessel will be left in the drained condition throughout the SAFSTOR period.

Control Rod Hydraulic System. The system has been flushed and is left in a drained condition. Filters at the -14 feet level of the access shaft and level pots on the scram dump tank were high radiation sources located in areas requiring periodic access and were therefore removed. .

Liouid Poison System. This system has been drained and the nitrogen botties have l been removed and decontaminated. Caps have been installed on alllines entering the i refueling building from this system.

Reactor Cleanuo System. This equipment has been deactivated and-isolated. Resins l

in the cleanup system demineralizer have been sluiced to the radwaste treatment i system for disposal. The cleanup pump and system piping at the -66 feet level were a source of radiation to maintenance workers in the vicinity and therefore have been removed. The cleanup heat exchanger and demineralizer room will not be available for routine access during SAFSTOR. Routine access to the room will be prevented by a locked barrier.

Reactor Shutdown Coolino System. This system has been drained and the water processed as radioactive liquid waste. Several sections of pipe in the access shaft l 4-13 Rev 2 August 1998

i have been flushed and/or removed due to ALARA considerations. The system has been isolated mechanically (by appropriate valving) and electrically. Routine access to the shutdown room will not be required during SAFSTOR. Routine access to the room  !

will be prevented by a locked barrier.

Emeraency Condenser System. The emergency condenser has been drained and isolated. Steam and condensate lines from the reactor have been capped.

l Suporession Tank Coolina and Core Sorav System. The waterin the suppression pool has been drained and processed. The remainder of the system has been drained and isolated. Blank flanges are installed in connections to the fire protection system and spent fuel storage pool. A liquid level detection system has been added to the suppression chamber.

~.

Reactor Shield Coolina System. This system has been drained and isolated.

Turbine-Generator. The turbine-generator system has been layed up to prevent

. deterioration. A nitrogen blanket has been used to fill the turbine internals. Dry air or nitrogen is purged through the generator.

Lube Oil System. The lube oil system has been drained of all oil and isolated from l other systems. No further action is planned throughout the SAFSTOR period.

Condensate System. The condenser and condenser hotwell have been flushed, drained, and isolated from the rest of the condensate system and other connected systems. A nitrogen blanket has been provided in the condenser to prevent internal deterioration.

l' The Condensate System has been drained and isolated. Internal system  !

decontamination has been performed to reduce radiation levels in the vicinity of t,he system piping durihg SAFSTOR.

i Resins in the condensate demineralizers have been transferred to the radwaste system l l for processing and disposal. One of the condensate demineralizers was converted to a i spent fuel storage pool demineralizer for the SAFSTOR period. The remaining demineralizers and their associated equipment will remain drained and isolated. Access  ;

to the condensate demineralizer room will be restricted during SAFSTOR by a locked barrier.

The condensate demineralizer resin regeneration system has been flushed, drained, and isolated (with the exception of piping necessary for resin addition and removal associated with the demineralizer used for the spent fuel storage pool).

Elector and Gland Seal Systems. The air ejector and gland seal systems have been I

drained and isolated. The exhausters have been electrically disconnected.

4-14 l Rev 2

. August 1998 l

Feedwater System. The system has been drained and pumps electrically isolated. Due to ALARA considerations, sections of feedwater piping have been internally decontaminated and/or removed to reduce radiation levels in areas requiring periodic access. System openings have been sealed and the system isolated from other connected systems.

Closed Coolina Water System. The chromated water contents of the system have been drained to the suppression chamber and processed. The system has been isolated.

Main Circulatina Water System. The circulating water pumps, screen wash pumps, and traveling screens have been removed. Openings into system internals have been sealed. The tube side of the main condenser was drained. The system has been isolated from connected systems.

Gas Treatment System. The system has been drained, flushed, and isolated. The off-gas equipment located in the new off-gas treatment vault will either remain in its current status or selected components may be salvaged. Radiological safety precautions are

.not required since none of the equipment or the building is contaminated. The sump pump will be maintained in an operational status so that any in-leakage into the building can be removed.

The gas treatment system (GTS) can be used to mitigate accidents involving the release of airborne particulate radioactive materialinto the atmosphere of the refueling building. This system consists of the original gas scrubber column (with the original packing and scrubber solution remover), demister, high efficiency particulate air (HEPA) filter, and fan.

In the event of an accident that results in high airborne particulate radioactive material in the refueling building, the refueling building ventilation system can be isolated and the refueling building air is then exhausted through the gas treatment system HEPA '

filter.

Instrumentation and Control (l&C) Systems. Unit 3 was provided with numerous I&C systems to optimize plant performance, protect equipment from damage, protect plant operating personnel, and protect the public and the environment from harm due to accidents of a radiological or nonradiological nature. Table 4-1 identifies the major I&C systems and their status during the SAFSTOR period. For those systems or parts of systems that were removed from sen/ ice, the following typical actions may be accomplished subsequent to such removal- either in preparation for SAFSTOR or as convenient during the SAFSTOR period:

. Physical removal of instruments or controls from the installed position. If such removal exposes contaminated system internals, openings will be appropriately resealed to prevent the spread of contamination.

l 4-15 Rev 2 August 1998

I Contaminated instruments or controls will be (1) decontaminated for salvage l

, or disposal, (2) disposed of as radwaste, or (3) stored for potential future use  !

l in a manner to minimize deterioration and prevent the spread of i contamination.

. Noncontaminated instruments or controls will be salvaged or stored for future use in a manner to minimize deterioration.

l The general approach for preparation ofinstrumentation and controls for the SAFSTOR

! period was to remove from service all such equipment not required to support

! " continued care" operations and to inspect, perform maintenance, and test as necessary all such equipment which must remain operational to provide reasonable assurance of its continued, reliable performance. For those instrumentation and controls which are to remain in a standby condition or in continuous or intermittent operation, a maintenance and calibration program was instituted to ensure reliable l performance and availability through the " continued care" period of SAFSTOR.

Preferred AC System. This system has been secured.

4.4.3 Operational Systems The following systems and components will remain operational over the SAFSTOR period. They are required (1) to support storage of spent reactor fuel, (2) to maintain environmental conditions for personnel protection, (3) to provide remote monitoring and alarms, and (4) to collect and process liquid or solid radioactive wastes. System descriptions and operating procedures were revised, where necessary, to reflect the SAFSTOR status and operational requirements.

4.4.3.1 Unit 3 Systems Heatina and Ventilation System. The heating and ventilation system will remain ' '

operational to supply filtered air to the refueling building and to exhaust air frorn the refueling building, hot lab, hot machine shop, and radwaste treatment building  :

(enclosure). The system has been adjusted wherever possible to maintain flow from j areas.of low contamination to areas of higher contamination. Ventilation exhaust is i through the ventilation exhaust stack, which is provided with the stack monitoring system to monitor any release. This effluent receives no routine treatment.

Other than the areas discussed above, there are no other locations for the release of airbome radioactive material from buildings.

No controlled ventilation is provided (or needed) for the waste storage vaults, the low-level waste storage building, or the low-level waste handling building. Wastes in these 4 locations will be packaged prior to storage to preclude a potential for release of airbome ,

radioactivity.

4-16 l Rev 2 August 1998

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

As decommissioning and SAFSTOR activities progress, the ventilation system may be i modified to reduce air flow or to secure air flow to unoccupied areas of Unit 3. In ,

additiori, ventilation from the radwaste treatment facility has been tied into the ventilation system.

l i The distance from the stack to the site boundary in each of the 16 compass direction sectors is listed below:

DIRECTION DEGREES DISTANCE Nodh O' 400' l North - Northeast 22 %* 519' i North - East 45' 725'  :

North - West 315* 358' North - Northwest 337 %* 339' East - Northeast 67 %* 1322' East 90* 897' East - Southeast 112 %* 764' South - East 135' 722' South - Southeast 157 %* 914' South - 180* 1372' South - Southwest 202 %* 1485' South - West 225* 1152' West - Southwest 247 %* 1150' t

West 270* 791' l

West - Northwest 292-%" 464' A site plan drawn to scale is shown as Figure 3-1 Spent Fuel Storace Pool. A cover was installed over the spent fuel storage pool t'o eliminate the spread of airbome contamination originating from the spent fuel storage pool and to provide protection against dropping of objects into the spent fuel storage pool. The cover includes the following characteristics: l e a structure designed to prevent dropping objects into the pool, e a contamination control barrier, and i e the airspace between the poollevel and the contamination control barrier has been tied into the ventilation system.

The cover is designed and constructed to be installed and removed in a manner which

! prevents dropping the cover or its component parts into the pool. The cover design and l the attachment design to the building structure are compatible with the current structure design of the pool. i l 4-17

Rev 2 l August 1998

t Soent Fuel Storace Pool and Fuel Handlino Service System.

. Th'e' spent fuel pool liner leakage pump and fuel pool circulating water pumps will be required during the SAFSTOR period, e Since the fuel stored in the spent fuel pool no longer requires cooling, the fuel pool coolers have been drained, flushed, and removed from service.

• The fuel and channel handling tools are stored until they are needed for final removal of the spent fuel from the spent fuel storage pool or for other fuel handling operations and training as required.

e The spent fuel pool service platform and hoist will receive preventive maintenance and steps will be takendo minimize susceptibility 4 deterioration.

• The reactor vessel refueling platform will be stored and steps will be taken to minimize its susceptibility to deterioration.

Domestic Water System. Domestic water for Unit 3 is supplied from the Units 1 and 2 domestic water system. This system will continue to operate during the SAFSTOR period.

Fire Protection System. This system will be required during the SAFSTOR period. No preparatory action other than routine preventive maintenance is required.

Area Radiation Monitorina System. Area monitors in the following locations will remain I

in service:

• Control Room e Refueling Building (+27 feet El.) ~

e Refueling Building (+ 12 feet El.) l e

Caisson Access Shaft (- 66 feet El.) -

l e Radwaste Treatment Facility

. Radwaste Handling Building ,

All other area monitors will be secured.

l Stack Radiation Monitorina System. The stack gas monitoring system consists of an isokinetic sampling device (located approximately halfway up the 250-foot plant stack),

, a quick disconnect particulate filter holder, two shielded gas sample chambers I (connected in series), beta scintillation detectors (situated one in each gas sample chamber), a flow regulator, and carbon vane pump. A 2-cfm sample of stack gas is continuously pulled from the isokinetic sampler through the particulate filter, then through each sample chamber, the flow regulator, and into the pump. The pump

, discharges into plant ventilation ductwork leading back to the stack.

4-18 l Rev 2 August 1998

The particulate filter is replaced weekly and the old filter is analyzed in the plant laboratory to determine particulate activity in stack effluent. Particulate activities down to the' range of 10 Ci released over a calendar quarter may be detected using this approach.

The beta scintillation detectors consist of a thin mylar window and phosphor crystal, photomultiplier tube, and preamplifier (mounted in a lightproof, watertight probe).

Setpoints are based on gaseous effluent discharge limits using the Offsite Dose Calculation Manual (ODCM).

The stack monitors are designed to be sensitive to "Kr in the stack gas for a range from -

approximately 5 x 10 Ci/cc to 2 x 10-2 Ci/cc. This range is intended to detect a small fraction of the 40 CFR 190 limits for routine operation, and estimated maximum anticipated releases follo~ wing an accident that results in damage to spent fuel assemblies.

Process Radiation Monitorina System. The radwaste discharge line monitor uses a gamma-sensitive scintillation detector consisting of a sodium iodide crystal (thallium-activated), and a photomultiplier tube, mounted in a light proof, watertight probe. The detector is mounted in a sample chamber bolted into the liquid radioactive waste discharge line. The detector monitors the activity of the water flowing through the liquid radioactive waste discharge line and is connected to a preamplifier. The preamplifier is i then connected to a rate meter located in the Unit 3 control room. The rate meter l displays the incoming count rate in logarithmic form and has a range of 10.to 10' cpm.

An alarm is provided to alert personnelif elevated levels of radioactivity are being l released into the discharge canal. Radwaste discharge pumps can be turned off from within the control room. Process alarm levels are set to assure that the limitations on l the instantaneous (averaged over a one hour period) concentrations of radioactive material being released to Humboldt Bay conform to ten times the effluent concentration limits of 10 CFR 20, Appendix B, Table 2, column 2.

l The discharge canal sample station is designed to collect a composite, representative l sample of the discharge canal water being released into Humboldt Bay. The sample l station consists of a small electric motor-driven sample pump, a small motor-driven

metering pump, piping for sample collection and system back flush piping from the plant j fire water system. The sample pump continuously draws from the discharge canal with I

water flowing into a sample scupper and back into the canal. The metering pump continuously draws from the scupperinto a 5-gallon sample bottle. The sample is periodically collected and analyzed for radioactivity. This system is intended to provide a final check to assure liquid radioactive effluent limits are not being exceeded.

No other effluent and process monitoring or sampling systems are planned for SAFSTOR. Grab sampling are utilized as required to determine activity levels in other process streams.

l 4-19 Rev 2 August 1998

f Cornaressed Air System. Both service air and instrument air are required for Unit 3 during the SAFSTOR period but at a much reduced volume. Compressed air for Unit 3 is norinally supplied from Units 1 and 2. The Unit 3 air compressor (No. 5 air compressor) will remain available as a standby unit during SAFSTOR. Due to the lay-up of the Unit 3 closed cooling water system, the No. 5 air compressor and the aftercooler have been modified to receive cooling from the Unit 2 bearing cooling water system.

Demineralized Water System. Demineralized water for Unit 3 is supplied from the Units 1 and 2 demineralized water systems. The demineralized water storage tank (top section of the condensate storage tank), the demineralized water pump, and distribution system piping and valves will remain in service during the SAFSTOR period.

Annunciator System. Annunciators for systems removed from service have been deenergized. During SAFSTOR it is not planned to continuously man the Unit 3 control room. A remote annunciator has been installed in Unit 2 to indicate an alarming condition in Unit 3.

Communications System. The existing Communications System will remain functional during the SAFSTOR period.

Auxiliary Power and Emeraency AC Systems. The auxiliary power system (including the 2.4 kV,480V, emergency 480V, and the 240/120V distribution systems) will remain in service during the SAFSTOR period. As system lay-ups are completed, individual loads no longer required are disconnected from these systems. Some loads may be relocated to different load centers to consolidate and equalize loadings, but a loss of redundancy in those systems that require redundant power supplies will not be permitted.

125V DC System.. Due to significantly reduced DC ret tuirements during SAFSTOR, Battery No. 3 has been retained in service at reduced load. During the SAFSTOR -

period, Unit 3 DC load may be transferred to the Unit 2 battery and Battery No. 3 removed for salvage.

75-Ton Refuelina Buildina Crane. While the crane is not anticipated to be routinely required during the SAFSTOR period, it will ultimately be required to support spent fuel shipping operations and final plant dismantlement. The crane will also be available for use during the SAFSTOR period. To prevent deterioration of the crane, PG&E will maintain the crane in an operable, standby status during SAFSTOR.

Liauid Waste Collection System. The turbine building drain tank will continue to be used during the SAFSTOR period along with the associated valves, pumps, and instrumentation and controls.

4-20 l Rev 2 August 1998

The reactor equipment drain tank and its associated pumps will continue to be used throughout the SAFSTOR period. They will be maintained along with associated valves,' instrumentation and controls in an operabic condition.

The reactor caisson sump and its pumps (2) are required throughout the SAFSTOR period. The tank, pumps, valves, instrumentation, and controls will be maintained in an operable condition.

The laundry waste tank, laundry hold tank, and other equipment associated with the i laundry remained in operation until the completion of the SAFSTOR decommissioning activities. The laundry system has been secured. The laundry waste tank remains in ,

service in order to collect drains from the decontamination shower and sink and other miscellaneous drains requiring processing by the radwaste processing system. It is 3 presently planned that during the SAFSTOR period, anti-contamination clothing and materials used will either be disposable or will be shipped off-site for cleaning.

Liould Waste Treatment and Disposal System. This system will remain operational

.throughout the SAFSTOR period. The system consisting of the radwaste receivers, waste hold tanks, resin disposal tank, concentrated waste tanks, the radwaste '

concentrator, and the associated system pumps, valves, filters, and piping will be used l during the SAFSTOR period to process liquid wastes as they are generated. This system will process the wastes collected in the waste collection system and will store

, those wastes until a sufficient quantity has been accumulated for processing and I' disposal.

The radwaste treatment facility has been modified with the construction of a metal building to enclose tne existing liquid radioactive waste treatment building and radioactive waste tankage area. The purpose of this modification is to minimize the potential for the spread of contamination outside of the building and to minimize the generation of potentially contaminated waste requiring processing by eliminating the need to collect rainwater from the building. The building ventilation is connected to the plant ventilation system. L Solid Radwaste Compactor. The compactor will remain operational for processing solid  !

waste during the SAFSTOR period. .

Mantift. The mantift will be used to provide access to the access shaft of the caisson for maintenance and surveillance activities. The manlift will be maintained in an  ;

operational condition.

4.4.3.2 Units I and 2 Systems Common to Unit 3 The foCowing Units 1 and 2 systems supply services to Unit 3, or they are systems shared jointly by these units which will be required during the continued care period of l 4-21 Rev 2 August 1998 l

l' I -

1 I

SAFSTOR. The following descriptions describe the extent of the connections between the units and what services will continue to be provided to Unit 3 during SAFSTOR.

Fire Protection System. A fire water loop around Unit 3 ties to the loop around Units 1 and 2 north and south of the station building. The system will remain in service. The emergency low pressure core flooding line, the emergency makeup line to the dry-well air coolers, and the reactor shield coolers have been disconnected from the fire main and blank flanged. This reduces the possibility of a leak in an abandoned system l affecting the fire system. l Comoressed Air System. Both the service air and instrument air headers of all three units are tied together. It is planned to maintain the Unit 3 air compressor and its aftercooler in standby with compressed air supplied from Unit 2. The remaining l ancillary equipment (receivers, dryers, and filters) will be maintained operational.

l Demineralized Water. Demineralized water services to Unit 3 are supplied by makeup l water from the Units 1 and 2 evaporators. Makeup wateris normally transferred from l

.one of the four Units 1 and 2 demineralized water storage tanks to the Unit 3 demineralized water storage tank for use in Unit 3.

Auxiliary Steam. Three lines provide services to Unit 3. The first provides steam to operate the radwaste concentrator, which will remain in service. The second provides several steam cleaning stations and shutdown cooling system steam mixing station, which will not be used. The third line supplies steam to the heating coils of the Unit 3 control room air handling unit. The radwaste concentrator, air handling unit, and steam i cleaning stations will continue to receive auxiliary steam. The steam supply to the I shutdown cooling system has been secured.

Domestic Water. A line from the Units 1 and 2 domestic water system is piped to Unit

3. This system provides laundry, sanitary, and drinking water in the operating floor -

area. The system will continue to be used in its present configuration. .

Oliv Water Drains. The floor drains serving the Unit 3 turbine lube oil reservoir and the air compressor are routed to the Unit 2 oily water separator. The system will continue to be used in its present configuration.

Yard Drains. Outside ard drains serving Units 1,2, and 3 are interconnected and normally flow to the inlet canal, with interception in a yard drain sump in case of spills.

The yard drain sump can be valved to the oily water separator or the turbine building drain tank through the turbine building drain system. The system continues to be used in its present configuration.

l i Unit 2 Bearino Coolina Water System. Bearing cooling water is routed from Unit 2 to the Radwaste Building for use as a sink for the heat rejected by the radwaste l

4-22 l Rev 2 August 1998

I l

e i

i concentrator vapor condenser. It is also routed to the reactor feed pump room to i provide cooling water for the Number 5 air compressor.

Radwaste Discharae to Units 1 and 2 Discharae Tubes. The radwaste system effluent  !

discharge line to the Units 1 and 2 discharge tubes mixes with the cooling water before l entering the outfall canal; this line will remain operational during the SAFSTOR period.  !

1 Minimum dilution flow can be provided by one of the circulating water pumps supplying either Unit 1 or Unit 2. Each unit has two circulating water pumps, each with a capacity of 12,500 gpm (nominal). The radioactive waste discharge line can be connected to the circulating water discharge line from either unit.

Electrical. The 2400V bus in Unit 3 is supplied from the 2400 kV buses in either Unit 1 )

or Unit 2. House transfogmer No. 2 has been and will continue to be utilized to be the  !

normal supply of Unit 3,2400V power through the SAFSTOR period.  ;

t 4.4.4 Soent Fuel Storace j i

l A total of 390 spent fuel assemblies will remain in the spent fuel storage pool until such ,

I time that they can be shipped offsite for final disposition. PG&E is not presently i

contemplating spent fuel pin consolidation prior to shipment to the Federal high level l l

waste repository. This spent fuelis a potential source of high radiation within the plant i environs.

I Fuel cladding failure might result from mechanical damage due to accidents or from  !

corrosive deterioration. While occurrence of the above-noted adverse conditions is i unlikely, PG&E has taken the following actions during SAFSTOR to further enhance the l margin of safety for the spent fuel:  :

• A redundant level indicating system is provided to alarm low water level in:the spent ,

' fuel storage pool. l

• Water chemistry is maintained to minimize corrosive deterioration.

• Radioactivity levels in the pool water are monitored to provide an indication of fuel -

integrity.  ;

i e A cover has been installed over the spent fuel storage pool. The cover was .

designed to prevent dropping of objects into the pool. The cover also provides I containment of airborne contamination from the pool area. i

• Spent fuelis stored in a configuration to provide maximum separation of assemblies

~

l during the SAFSTOR period. j l

l 4-23 Rev 2 August 1998 j

i l

! e Spent fuel is only handled as necessary for loading and transfer of spent fuel l shipping casks and for special inspections or tests that have the written approval of th Plant Staff Review Committee.

• The handling of heavy loads over the spent fuel storage racks is prohibited and is controlled administratively.

e Two sources of makeup water for the pool are to be maintained.

• Maintenance is performed on structures, systems, and components as required to ensure that the spent fuel will remain in a safe storage condition.

• The refueling building bridge crane will remain stored away from the spent fuel storage pool. '

The above-noted actions, in conjunction with routine surveillance and maintenance activities in effect for the plant, will ensure adequate management of spent fuel for the l SAFSTOR period. l l

4.4.5 Radioactive Waste Processing and Disposal l 1

l 4.4.5.1 Sources of Radioactive Wastes.

Wastes remaining on site at the start of the SAFSTOR period activity included liquids i and sludges stored in several collection tanks located in Unit 3 and in waste receiving and storage tanks located in the radwaste building. Solid wastes consisted of drums and metal boxes located in various waste storage areas of the plant. Other sources of solid waste on site included contaminated tools and equipment, lumber, and soil. I Waste generated during lnitial SAFSTOR activities included liquids and sludges .

resulting from decontamination activities and sludge from final cleanout of tanks'and surface decontamination cleaning solutions. Processing these liquids as well as th'ose remaining on-site generated a variety ofion exchange resins and concentrates that were solidified and transported to a shallow land burial facility. The generation of solid wastes such as contaminated protective clothing and tools, along with other typical dry radicactive wastes, also occurred during initial activities.

During the SAFSTOR period, wastes will continue to be generated. Spent fuel storage pool water, rain and groundwater in-leakage will be collected and processed on a routine basis. Specific operational improvements or maintenance projects in conjunction l with routine maintenance requirements will result in the generation of liquid and solid l waste.

l 4-24 l Rev 2 August 1998

i 4.4.5.2 Waste Processing and Disposal During SAFSTOR activities, radioactive wastes generated will be processed on or off site and shipped to a licensed burial site for disposal. Off-site secondary processors may be used as appropriate to sort, survey, decontaminate, free-release, and consolidate wastes. The radioactive waste treatment facility will remain operational throughout the SAFSTOR period.

Liquid radioactive wastes released from the site may be processed by filtration, and/or demineralization, and/or other appropriate methods when treatment is required.

Samples of liquid wastes are analyzed before release to ensure that they are within the discharge limits specified in 10 CFR Part 20.

The only release point foi quid li radioactive waste is the liquid radioactive waste discharge line that discharges into either the Unit 1 or Unit 2 circulating water discharge line prior to reaching the plant discharge canal.

The expected sources of liquid radwaste from Unit 3 include: spent fuel pool liner

' leakage; spent fuel pool recirculation pump packing leakage; resin sluice water; wastewater from ongoing decontamination efforts; hot lab waste; cassion inleakage; and rainwater runoff from contaminated areas. Prior to release into the plant discharge canal, the liquid radwaste will be diluted with the liquid effluent from the Unit 1 and/or Unit 2 circulating water pumps. The discharge from each of these four pumps is expected to be no less than 12,500 gpm. Since Units 1 and 2 are expected to be in service for the duration of the SAFSTOR period, and normal operations requires at least one unit operating, the typical flow rate (two circulators) will usually be at least 25,000 gpm.

Liquid radioactive wastes that must be treated before discharge may be treated by vendor (contractor) systems on site if filtration or demineralization is not adequate.

Concentrated liquid radioactive waste will be stored in the concentrated waste tanks.

Processing of liquid radioactive wastes and wet solid (sludge) wastes will be iri accordance with the plant or vendor procedures and in accordance with current regulations. Liquid radioactive wastes and wet solid wastes may be shipped to secondary processors for final treatment before disposal.

Chemical and liquid decontamination wastes generated during SAFSTOR will be treated with other liquid radioactive waste.

Spent resins from the radwaste demineralizer and the spent fuel storage pool demineralizer are also accumulated on site in the resin storage tank. When a sufficient l quantity of resins has accumulated, it will either be dewatered and shipped or solidified l and shipped to a licensed burial site in accordance with applicable regulations. An off-site secondary processor may be used for volume reduction or further processing prior to disposal.

l l 4-25 Rev 2 August 1998 l

l l

Spent cartridge-type filters (and filter crud) will be packaged, processed and shipped in accordance with applicable regulations.

Activated components will be handled via current technology in accordance with applicable regulations.

Solid radioactive wastes include: contaminated protective clothing, plastic, rags, piping, surplus equipment, contaminated soil, rubble, etc. Solid wastes are accumulated and stored in appropriate containers. Radioactive waste containers are inventoried, marked, and stored on site until a sufficient quantity has been accumulated to ship the wastes off site for processing and disposal. Waste processing (e.g. sorting, decontamination, compaction) may occur on-site. Off-site secondary processors may be used for processing and volume reduction before disposal.

Low levelliquid and solid v,astes T will be sampled and analyzed as required by regulation when they are packaged or prior to processing for shipment to a secondary processor or disposal site. Analysis will be conducted using a combination of onsite gamma spectrometry, offsite laboratory analysis, and the development of standard

' plant mixtures for similar wastes that can then be ratioed based on a significant isotope or dose rate. The results of the analysis will determine waste classification in accordance with 10 CFR Part 61, the disposal site license, and any other regulatory requirements in effect at the time. Regulatory guidelines, such as NRC Branch l Technical Position, will also be used to characterize wastes.

Records of samples and analysis will be retained to demonstrate the basis for waste classification and stability requirements.

Disposal of processed and packaged radioactive wastes will be accomplished by shipping the wastes to an authorized secondary processor or shallow land burial facility. l Shipments will normally be made by truck in accordance with Department of

' Transportation regulations contained in 49 CFR Parts 171-179. Low-level wastes ,

shipped for land burial disposal will be characterized in accordance with and meet the waste form requirements in 10 CFR Part 61.

4.4.6 Decontamination 4.4.6.1 Purpose During the SAFSTOR activities an ongoing program of facility decontamination will continue. The purposes of this program are:

i e To minimize contamination levels and radiation dose rates in areas of Unit 3 that will be accessible during the SAFSTOR period for routine maintenance or for periodic surveys

• To minimize the requirements for surveys to detect the spread of contamination 4-26 l Rev 2 August 1998

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• To minimize the requirements for protective clothing and the potential for contamination of tools and equipment. This will reduce maintenance costs during SAFSTOR by reducing the amount of radioactive waste generated.

!

• To reduce the potential for contamination to spread outside of controlled areas l Decontamination efforts will be concentrated in areas to which access will be available during SAFSTOR. Areas to which access will not be routinely permitted without special l authorization are sealed to minimize spread of contamination from the areas and l secured with locked gates or physical barriers to prevent access. l 4.4.6.2 Decontamination Methods l . The decontamination program during the SAFSTOR period will be a continuation of the decontamination work that is routinely performed at Unit 3. The decontamination method utilized is dependent on the level of contamination encountered, the type of surface, type of radioactivity involved, and whether or not the contamination is fixed or

. loose.

The methods listed below are considered to be representative. Although others may exist, they are not included because of their similarity to one of the listed methods or because of some characteristic (such as extreme toxicity) that may render them unsuitable.

Hand Wioino. Rags moistened with water or a solvent such as acetone can be an effective decontamination process. This method may not work well if the item is rusty or pitted. It requires access to all surfaces to be cleaned, is a relatively slow procedure  ;

and its hands-on nature can lead to high personnel exposure. On the positive side, wiping can provide a high decontamination factor (DF), generates easily handled decontamination wastes (contaminated rags), requires no special equipment, and can ,

be used selectively on portions of the component. For these reasons, hand wiping will normally not be used for decontamination to reduce levels for shipping where high dose l rates and inaccessible intemal contamination usually exist. For purposes of decontamination to allow the item to be its own shipping container, the generally large component size will restrict hand wiping to small areas that are discovered to be above shipping limits. Wiping can be used extensively and effectively on smaller items with  :

low-to-medium extemal contamination levels and easily accessible internal contamination. -

Steam Cleanino. This may be performed either remotely in a spray booth or directly by decontamination personnel using some type of hand-held wand arrangement. In the former case, only minimal intemal decontamination is possible; however, reasonable extemal cleaning can be accomplished quickly with low exposure expenditures. .

Containment of the generated wastes and protection of personnel from radioactive ,

contamination become more difficult.

l 4-27 Rev 2 August 1998

Abrasive Blastina. This is a highly effective procedure that can effect total decontamination of even rusty or pitted surfaces. As with hand-held steam cleaning, this niethod suffers from internal accessibility problems. It also generates large amounts of solid wastes and, being a dry process, produces significant quantities of airbome radioactivity. Abrasive blasting may be used ifits high effectiveness can be justified after taking the exposure, waste, and accessibility limitations into account.

Hydrolasina. The use of high pressure waterjets for decontamination falls somewhere between steam cleaning and abrasive blasting in effectiveness. Less effective than j abrasive blasting, it has the advantage of producing liquid wastes (that can be i processed) rather than solid wastes. As an external cleaning technique, it offers reduced airbome generation potential although this is offset by the noed to control splashing. The utility of hydrolasing is generally limited to operations where internal accessibility is not required:

Ultrasonic Cleanina. Since this is basically an immersion process and, as such, is I limited to smaller items, it is generally unsuitable for large scale decontamination. In

. addition, although ultrasonic cleaning can be especially effective in removing contamination from crevices, it is doubtful that releasable levels can be reached ,

consistently enough with this technique to make it a viable option. Therefore, this method, if employed, would be used mostly for decontamination of smaller, highly contaminated components that have crevices or poor internal accessibility.

Electropolishina. This is an electrochemical process where the object to be decontaminated serves as the anode in an electrolytic cell and radioactive contamination on the item is removed by anodic dissolution of the surface material.

Although it is a relatively new process and has not yet been used for a full scale decontaminahon operation, it nevertheless requires consideration as a technique on the basis of its saperior effectiveness in cleaning almost any metallic surface to a completely contamination-free state. On the other hand, this process has several drawbacks iacluding the cost of electrolyte and special equipment, the consumption'of '~

considera' ole power, and the production of highly radioactive solutions.

Imoactors. Two surface removal methods used more extensively than the rest are jack hammers and impactors. Jack hammers, powered by compre.ssed air, are readily available and are easily operated by one man. They are used, to chip off the surface material deep enough to remove the contamination. Because they are difficult to position on walls and ceilings, jack hammers are used primarily on floors. Impactors (or hoe rams), similar in operation to jack hammers but much larger, have been used successfully in several decontamination projects. An impactor uses a pick chisel point that is driven into the concrete surface with high-energy impacts several times per second. Impactors are powered either pneumatically or hydraulically, and are held and positioned with linkages typical of those on tractor-mounted backhoes. A medium-size air filtration system is necessary to control the dust produced by both of these surface removal methods.

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t Chemical Decontamination. This technique uses concentrated or dilute solvents in contact with the contaminated item to dissolve either the contamination film covering the bsse metal or the base metalitself. Dissolution of the film is intended to be nondestructive to the base metal, and is generally used for operating facilities.

Dissolution of the base metal should be considered only in a decommissioning program where reuse of the item will never occur.

Chemical flushing is recommended for remote decontamination of intact piping systems. Chemical decontamination has also proven to be effective as an alternative to partial or complete removal in reducing the radioactivity of large surface areas such as floors and walls.

4.4.6.3 System intemal Flushing Secured plant systems that contained liquids have been flushed (where practical) to remove loose contaminants and the systems drained. Systems / components which contribute radiation exposures significantly above the general area dose rates have

.been reviewed to determine action warranted as a result of ALARA considerations.

Systems / components to which rcutine access will not be required during the SAFSTOR period are isolated using locked barriers. In these cases system internal decontamination was not warranted. Systems / components to which or near which routine access will be required during th'e SAFSTOR period have been evaluated on a case-by-case basis. In some cases, a portion of the system piping or component has been removed. Where such action was not practical, installation of shielding or internal -

decontamination of the system or component was performed to reduce radiation levels l to the minimum practical.

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Table 4-1 I&C System Status During SAFSTOR l

REMOVED

( FROM CONTINUED l&C SYSTEM DESCRIPTION SERVICE OPERATION j (1) Reactor Protection System x (2) Reactor Nuclear Instrumentation x l (including Out-of Vessel Neutron

! Monitoring System and in-Core Neutron Monitoring System)

(3) Off-Gas Monitoring System x L

(4) Process Radiation Monitoring System

. Condensate Demineralizer Line x

. Closed Cooling Water Retum Line to x l

Storage Tank

. Liquid Waste System Vent Monitor x

. Radwaste Discharge Line to the outfall x l Canal  ;

. Reactor Water to Cleanup Demineralizer x (5) Reactor Vessel Instrumentation x

! (6) Containment Leak Rate Monitoring System x .

_(7) Control Rod Position Indication System x .

(8) Control Rod Drive Instrumentation x (9) Feedwater Control System x (10) Discharge Canal Sample Station x .

(ii) Meteorological Facility x l (12) . Area Radiation Monitoring System a (13) Refueling Building isolation Monitoring x System l

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Table 4-1 (Continued) 1&C System Status During SAFSTOR REMOVED FROM CONTINUED I&C SYSTEM DESCRIPTION SERVICE OPERATION (14) Stack Gas Radiation Monitoring System b

, (15) Off-Site Environmental Monitoring Stations c (16) Spent Fuel Storage Pool Water Level d Indicating System a The following Are'a Radiation Monitors will remain in service during SAFSTOR:

- Control Room

- Refueling Building (+27 feet Elevation)

- Refueling Building (+ 12 feet Elevation) l

- Caisson Access Shaft (-66 feet Elevation)

- Radwaste Treatment Facility t

- Radwaste Handling Building t b Two channels of stack monitoring will be used. Each channel will be capable of l monitoring for noble gases (asKr). Particulate will be monitored with replaceable l filters.

c The off-site environmental monitoring stations No.1,2,14 and 15 will be

! maintained in operation.

d Two independent channels of Spent Fuel Storage Poollevelindication~are provided.  ;

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5.0 SAFSTOR SITE CONDITIONS At the outset of the SAFSTOR period, conditions were established and activities continued to protect the health and safety of the plant workers and the public.

Previous sections of this document describe the establishment of suitable conditions.

This section describes the ' establishment of (1) a benchmark (baseline) radiation survey to document quantitative initial radiological conditions at the outset of the i SAFSTOR period, (2) a surveillance and monitoring program by which assurance can l be provided that conditions are not deteriorating, and (3) a program of activities that can provide assurance that safe conditions can be maintained for the duration of the i SAFSTOR period.

5.1 BASELINE RADIATION SURVEY

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The Baseline Radiation Survey establishes the activity levels and nuclide concentrations in the plant and its environs at the beginning of SAFSTOR. The survey includes: l

• Area and contact dose rates; beta-gamma and alpha by elevation and room

• Proportion of loose versus fixed contamination by elevation and room

• Surveys of system components I

• Inventory of nuclides associated with each plant system

• Radionuclide concentrations (if any) in systems which have cross-connections to Units 1 and 2

• Onsite waste inventory

• Activity levels in the Unit 3 area

• Activity levels in the Unit 3 area surrounding the Unit 3 restricted area 1

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• Radicnuclide concentrations in:

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- Vegetation

- Soil (surface and cores around the Refueling Building)

- Canal sediments and slough sediments

- Bay sediments l 5-1 Rev 2 August 1998

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- Bay mussels

- Bay algae Baseline conditions for soil, biota, and sediments were established prior to SAFSTOR and will only need to be reestablished prior to DECON if a significant release occurs during SAFSTOR.

Baseline conditions in the plant will be compared with surveillance values obtained from routine quarterly monitoring. Waste inventory will be updated and documented for 10 CFR 61 disposal. The Baseline Radiation Study is included ac Appendix IC.

5.2 MONITORING AND SURVEILLANCE Section 6.2.1 of the Environmental Report estimates annual gaseous emissions and liquid discharges of radioactivity based on average emissions and discharges for the period 1980-1983. A breakdown by radionuclide is contained in the Annual Effluent Reports submitted to the NRC. This breakdown is summarized in the Humboldt Bay Power Plant (HBPP) Environmental Report, Tables 10.4.3 and 10.4.4.

During the SAFSTOR period, the primary activity will be associated with the operation and maintenance of the spent fuel storage pool and the processing of wastes resulting from the spent fuel pool. This activity is expected to account for the majority of the releases during the SAFSTOR period .

5.2.1 In-Plant Monitoring Rautine surveys will be conducted at pre-established locations using portable beta-gamma and alpha dose rate meters. Surveys willinclude area readings and contact readings at several locations such as step-off pads, storage areas, and maintenance work stations.

Samples of the following will be periodically taken and analysis will include total gamma, beta, and alpha activity, and concentrations of indicator nuclides:

• Spent Fuel Storage Pool Water e Radwaste Liquids, including all discharges from the Radwaste Treatment System e Solid waste in compliance with 10 CFR 61.

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5-2 l Rev 2 August 1998

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Filters and resins will be replaced as requirec to minimize concentrations of nuclides l in spent fuel storage pool water and radwaste discharge.

l 5.2.2 Onsite Environmental Monitoring  ;

1 The following monitors will be maintained through the SAFSTOR period:

• Stack radiation monitor e Continuous sampler in discharge canal j

e Fenceline dosimetry stations f

• Groundwater monitoring wells l Annual reports will be submitted in accordance with the Offsite Dose Calculation l Manual (ODCM) requirements.

l In the area of radwaste treatment buildings routine surveys will be conducted to j identify contamination and record dose rates. These surveys will be conducted ,

quarterly or as needed to support waste management operations. l t

5.2.3 Offsite Environmental Monitoring  :

5 I

Four monitors will be maintained offsite, Stations 1,2,14 and 25. These represent a gradient downwind of the prevailing wind direction. These stations will be equipped i with dosimeters and annual reports will include both average and maximum recorded j values j i

No additional offsite monitoring will be required during the SAFSTOR period unless significant releases occur as a result of an accident. f

[

5.3 CONTINUED CARE PLAN, SAFSTOR TO DECON i

( During the SAFSTOR period and until initiation of the final DECON decommissioning  ;

program, necessary plant systems will be operated and maintenance will be  !

performed on structures, systems, and components as required to ensure continued j safe conditions within the plant.  ;

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! 5.3.1 Operation of Plant Systems j Preceding sections of this document identified those systems required to operate during the SAFSTOR period. The following system classifications were included: l I

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Rev 2 August 1998 i

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• Service Systems, including the Refueling Building Ventilation System, the Spent Fuel Pool Service System, the Fire Protection System, and Electrical Systems e Waste Disposal System, including the Liquid Radioactive Waste System and the Solid Radioactive Waste System e Monitoring System, including the Stack Gas Radiation Monitoring System, the Process Water Monitor, Area Monitors and Portable Monitoring Equipment, Offsite Environmental Monitoring Stations, and Spent Fuel Storage Pool Water Level Monitors.

These systems will be operated as required during the SAFSTOR period. Their operation will be in accordance with approved procedures, and the operational schedule will be of sufficient regularity to ensure adherence to the Unit 3 Technical Specifications. This operational schedule may vary over the SAFSTOR period as conditions warrant.

5.3.2 Maintenance of Structures, Systems, and Components The maintenance program established for SAFSTOR is a modified continuation of the '

previous maintenance program at the plant and includes aspects for both preventive and corrective maintenance.

The preventive maintenance aspect provides for a regularly scheduled series of  ;

inspections, tests, and services for structures, systems, and components. The frequency of preventive maintenance was established on the basis of prior experience, ongoing operational use, plant conditions, and where applicable, Technical Specification requirements. The objective of the preventive maintenance program is to ensure continued reliable function of necessary structures, systems, and components equivalent to or better than the reliability existing at the beginning of the SAFSTOR period.

The corrective maintenance aspect of the program provides for analysis and appropriate action to be taken for all conditions where function of structures, systems, and components is determined to be degraded. Adequate stores of spare parts and ,

servicing equipment are maintained to ensure that corrective action can be taken in a timely manner and where required within the timeframe required by the Technical Specifications.

Plant maintenance will implement the requirements of 10 CFR 50.65 as appropriate (Maintenance Rule).

54 l Rev 2 August 1998 l

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6.0 HEALTH PHYSICS, OCCUPATIONAL HEALTH AND SAFETY l j

During the SAFSTOR period, radiation protection and health physics programs will be provided to ensure the health and safety of Unit 3 workers. The programs also j

! provide the necessary monitoring and control of radiological conditions to protect

- the health and safety of the general public and to ensure compliance with Unit 3 l t

' license requirements. In addition, programs will be provided to maintain radiation exposures as low as reasonably achievable (ALARA). j 8.1 ORGANIZATION AND RESPONSIBILITIES i The organization described below is the organization as it exists during the }

SAFSTOR period. The organization may be changed during the SAFSTOR l period as staffing levels or work requirements dictate. Responsibilities assigned ,

t to a position which is deleted will be assigned to another position in order to t maintain continuity.

The HBPP Plant Manager has the overall responsibility for all onsite activities, including assurance that corporate ALARA policies are carried out at the plant.  !

The Plant Manager is the Chairman of the HBPP Plant Staff Review Committee {

(PSRC) which also serves as the ALARA Committee. and PSRC. In addition, l

he is responsible for approving all Unit 3 administrative and quality-related j procedures.

The Senior Radiation Protection Engineer (SRPE) is designated as the on-site l Radiation Protection Manager (RPM) responsible for implementing the radiation protection and ALARA programs. The RPM shall have a bachelor's degree, or .

the equivalent in a science or engineering subject , including some formal

, training in radiation protection. He shall have at least 5 years of professional l experience in applied radiation protection, at least 3 of which should be in i applied radiation protection work in a nuclear facility dealing with problems l similar to those encountered in nuclear power plants. The SRPE reports to the l l Plant Manager. l The SRPE serves as a member of the PSRC (refer to Unit 3 Technical Specifications). He has the authority and responsibility to halt operations he deems to be unsafe and to report the matter to the Plant Manager; and communicate his concerns directly to any level of Nuclear Power Generation Department management, including the SVP&GM, NPG, if he deems it to be appropriate.

Radiation and Process Monitors are the employees who perform chemical and radiological sampling analyses and radiation and contamination surveys. In addition, they implement the personnel radiation monitoring program, maintain 6-1 Rev 2 August 1998 l

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. radiation protection records, and provide monitoring for work in radiologically controlled areas. Three such positions will normally be filled during SAFSTOR.

6.2 ALARA PROGRAM lt is the policy of PG&E to design, operate, and maintain its nuclear power plants l- - in such a manner as to maintain personnel radiation Total Effective Dose Equivalent (TEDE) ALARA. The TEDE ALARA concept is implemented by l  ;

assuring that every effort be made by all HBPP personnel involved in the planning or performance of radiation work to maintain individual exposures to radiation sources or materials as far below the occupational dose limits as is reasonably achievable, taking into account the state of technology, the economics of

improvements in relation to state of technology, the economics of improvements in j relation to benefits to the public health and safety, and other societal and l socioeconomic considerations, and in relation to utilization of nuclear energy and licensed materials in the public interest. The Company's commitment to maintaining TEDE ALARA involves

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• Design - planning, reviews, system, subsystem, and component selection and location; operator usage considerations and maintainability

• Construction - procedures, planning, methods, testing, and scheduling ,

• Operation - procedures, license compliance, techniques, equipment usage, ,

l maintenance, and operating experience feedback from company and industry l l experience

• Personnel - training, management support, motivation, and supervision

• Administration - policy, guidance, controls, licensing position,' and i documentation

• Management - involvement, commitment, supervision and oversight j l Periodically The Radiation Protection Manager or a designated member of the Diablo Canyon Power Plant (DCPP) ALARA Committee, will participate in a HBPP ALARA Committee review of exposures. Additionally, the SRPE at HBPP ,

l and The Radiation Protection Manager at DCPP will participate in a working level ALARA committee to review topics such as specific exposure reduction techniques, new equipment performance, regulatory changes and the l

application of the change, regulatory rule making activities, cross training and loaned assignments, unique radiological problems, NOV performance and lessons learned.

L The Humboldt Bay Power Plant PSRC also functions as the plant ALARA l Committee.

[ 6-2 Rev 2 August 1998

The committee meets quarterly or as called for by the chairman or the SRPE and has the following functions and responsibilities:

• Review radiation exposures associated with routine operations and maintenance and recommend future exposure reduction goals i

Review planned jobs where potential exposures might exceed 500 person-e millirem for the job and establish exposure limits and person-rem goals for l thatjob

~

e Review completed jobs for achievement of goals and future improvements

'e Review plant radiation and contamination levels annually and recommend f future exposure reduction goals  !

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• Review plant design changes and plant procedures for ALARA considerations (when applicable)  ;

Before the ALARA committee review of a proposed job, the individuals planning the job make estimates of the expected radiation exposures. Estimates are ,

based on radiation surveys conducted in the area where the job will be performed and estimates of the time required to perform the job based on prior  ;

experien:',e. These estimates are reviewed by the Radiation Protection l Department. If the established review threshold of 500 person-millirem for the total job is expected to be exceeded, an ALARA review checklist is completed for  ;

review by the ALARA Committee. The purpose of the checklist is to document the consideration of specific actions that may be taken to reduce radiation exposures. .

All radiation workers at HBPP receive as part of their radiation protection training an indoctrination in the principles of A1. ARA radiation exposure control. In this '

training, the responsibility of the individual worker to follow procedures and l safety rules and to maintain his/her own exposure ALARA, are emphasized. The principles of minimizing the duration of exposure (time), maintaining distance from the source (distance) and reducing the source term (shielding) are included l in the training.

6.3 RADIATION PROTECTION PROGRAM  ;

All employees who routinely work in the restricted areas of the plant, and transient woders whose wo* may involve significant radiation exposure, will  ;

! participate in the radiation protection program. Radiation protection training will '

be commensurate with an' individual's work requirements and the areas to which they are permitted access. Nonradiation workers (i.e., persons whose work  !

I rarely, if ever, requires they enter a restricted area and/or who receive minimal 6-3

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{ radiation exposure) shall be provided with fundamental training on radiation,

health effects, and risks as appropriate.

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Members of the Radiation Protection Department and certified fuel handlers are i responsible for implementing the requirements of the Radiation Protection i Program. These individuals, as part of their initial qualification, will receive i additional training in radiological work practices and the use of specialized survey and analysis equipment to the extent necessary to perform their duties.

The radiation protection program that has been implemented for the SAFSTOR

period is an extension of the program that was in effect during operation of Unit 3.

The program is in compliance with the requirements of 10 CFR 20. Detailed

procedures implement the program at the plant level. The following sections

describe the radiation protection program.

l -

6.3.1 Personnel Monitoring i

Personnel entering the Unit 3 Restricted Area (area to which access requires j written authorization for the purposes of radiation protection) wear personnel l monitoring thermo-luminescent dosimeters (TLDs). The TLDs provide a record l j of radiation exposure received and are normally wom on the upper portion of the body unless the nature of the work or the principle source of radiation is such i

that another portion of the whole body is likely to receive a larger dose, in which 4 case the TLD will be wom on that portion of the body. Additional TLDs and/or j finger rings are used to monitor extremity doses for jobs in which significant

] extremity doses are expected.

i

! Direct reading pocket dosimeters or similar devices are used to provide j- estimated _ exposure. TLDs, provide official exposure. The TLDs are evaluated l l quarterly or whenever exposure estimates indicate exposures approaching an 2

exposure limit.

) Other personnel monitoring instruments consist of personnel contamination i monitor (s) and count rate meters. These instruments are used to conduct l l personnel contamination surveys.

4 j- Intemal radiation exposure monitoring is performed through a program of whole body counting on a routine basis or in the event of suspected intemal i contamination. Personnel who are regularly assigned to work in the Unit 3 4

restricted area and who are likely to receive 500 mrem of exposure per year are

routinely given two whole body counts per year. Persons terminating work

[ assignments are given whole body counts if there is reason to suspect i significant intakes of radioactive material. Other types of bioassay testing (urine, fecal, swabs, etc.) may be used if circumstances warrant.

The Humboldt Bay Power Plant Radiation Control Procedures for performing bioassay conform to the requirements of 10 CFR 20.

6-4 Rev 2 August 1998

6.3.2 Airbome Radioactivity The potential for intemal radiation exposure is minimized through the use of engineering controls and through th9 use of a respiratory protection program.

Air samples are taken to monitor airbome radioactivity.

The spread of airbome radioactivity within the facility is minimized by maintaining air flows from areas of low potential airborne radioactivity to areas of higher potential whenever possible. Other engineering controls such es 1 temporary containments, encapsulating contamination, controlled ventilation techniques, and/or HEPA air filters are utilized when practical.

i Areas with a potential for existing airbome radioactivity are evaluated prior to the start of work. Air samples are taken during activities that may produce airbome radioactivity. In addition, the Refueling Building +12 feet elevation is continuously monitored to detect airbome radioactivity. Samples are evaluated to identify types and sources of radioactivity. Airbome Radioactivity Areas are posted to prevent unauthorized entry. Notices to inform personnel of unusual ,

radiation, contamination, or airbome conditions will be posted at Access Control (s). The use of respiratory protection equipment is specified by the radiation protection department.

i 6.3.3 Respiratory Protection Program ,

1 A respiratory protection program will be maintained during the SAFSTOR period.

l l Use of respiratory protection equipment will be based on a TEDE ALARA review.

Whenever practical, engineering controls will be used to maintain airbome concentrations not only below specified limits but as low as reasonably achievable.

Personnel assigned to use respiratory protection equipment receive a physical  :

examination to qualify them to wear respirators. In addition, respirator users receive training in the proper use of respirators and are fitted to ensure that they ]

can achieve a proper respirator seal. l Only NIOSH/MSHA-approved respiratory protective equipment is used. The following equipment is representative of the respiratory protection equipment that is used:

• MSA Ultra-twin: full-facepiece, air purifying

• MSA Pressure Demand Air Mask, air-line supplied

• Constant Flow Air-Line respirator with Custom Comfo Facepiece

• MSA Self-Contained Breathing Apparatus, Model 401, pressure demand i

6-5 j Rev 2 l August 1998

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i l Air-purifying respirators utilize NIOSH/MSHA -approved filters capable of removing particulate radioactivity. Air-supplied respirators or self-contained respirators may also be used when appropriate. Air supplied respirators are supplied with air from the plant service air system after having been passed through a filter and pressure regulator. Plant procedures are provided for the l fitting, issue, and maintenance of respiratory protection equipment.

L

! The respiratory protection program is designed to comply with the provisions of

NRC Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection."

6.3.4 Protective Clothing L Personnel working in contaminated portions of the Unit 3 restricted area are l provided with protective clothing to minimize the potential for personnel contamination. Prot 8ctive clothing requirements for specificjobs are specified j

by the Radiation Protection Department as part of the work authorization for that job. Protective clothing such as coveralls, lab coats, surgeons caps, hoods, shoe covers, gloves, etc., are readily available for routine use in the controlled area.

Additional items of protective clothing such as waterproof clothing, face shields, etc., are issued when required for a specificjob. Contamination control points l (" step-off pads") are established at certain locations in the plant to permit L changing of protective clothing to prevent the spread of contamination. Clothing '

I potentially contaminated 'to levels in excess of that permitted in certain areas are either removed or exchanged for cleaner items at the step-off pads.

During previous activities in preparation for SAFSTOR, the laundry facility was operational to ensure an adequate supply of clean protective clothing.

l Subsequent to the entry into SAFSTOR, the laundry has been secured and l protective clothing used will be either disposable or will be shipped off site for cleaning.

6.3.5 Control of Access To limit radiation exposures, personnel access is controlled in areas where such exposure is possible. This control consists of a system of physical barriers, warning signs and signals, and administrative procedures which govern authorized entries.- Written authorization for all entries into the restricted area is required. This written authorization is either in the form of a Routine Work Permit (RWP) or a Special Work Permit (SWP).

l RWPs are established to authorize work of a routine nature in the restricted area under relatively stable conditions. RWPs are normally effective for an extended

period of time (up to a maximum of 2 years) but are subject to revision at any i time.

64 Rev 2

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SWPs are short-term authorizations for personnel to perform work of a non-routine nature in specific areas. All entries into the restricted area not covered by a valid RWP must be covered by an SWP.

Both RWPs and SWPs contain a description of the radiological conditions existing in the area covered by the permit, general and special instructions to be followed by persons working under the permit, and a list of the protective  ;

equipment requirements. l 6.3.6 Facilities Monitoring -

A program for routine surveys and monitoring will be continued during the SAFSTOR period. Radiological surveys will be used to maintain a record of radiological conditions in Unit 3 and to evaluate radiological trends during decommissioning activities.

i Radiological surveys include measuring general area dose rates (radiation surveys) and the collection and analysis of representative samples of airbome particulates, water, and removable surface contaminants.

6.3.6.1 AirSamples  ;

Airbome radioactivity surveillance is conducted to detect airbome radioactivity to

( which personnel may be exposed, to detect equipment degradation or failure, and to limit airbome releases from the plant to permissible amounts. Routine air  ;

l particulate samples are obtained from a sampler located in the Refueling Building. These samples are counted for alpha and beta-gamma activities. t l

Particulate samples are obtained from the ventilation exhaust stack weekly to  !

determine release rates.

Non-routine air samples to establish protection requirements for maintenance activities or to verify airbome radioactivity conditions during work activities are  ;

obtained and analyzed when routine samples are not sufficient for monitoring  !

plant conditions.

l 6.3.6.2 Radiation Surveys  ;

i Radiation surveys are conducted for the following purposes: l

• Measure and document radiation and contamination levels in areas of  ;

interest  !

e Identify trends in radiation and contamination levels, particularly during work in progress .

6-7 Rev 2 ,

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• Determine appropriate protective measures for personnel working in restricted areas

• Provide information so that workers can maintain their doses TEDE ALARA e identify locations and situations where special dosimetry is required Two types of routine surveys are conducted. An "A" suniey is conducted by Operations Department personnel as part of their routine plant surveillance.

Surveys are conducted on a daily basis to check dose rates at specific locations. l "A" surveys are not intended to be a comprehensive survey but are used to detect changing plant conditions.

l On average, a "B" survey is conducted weekly by Radiation Protection Department personnbl. A "B" survey includes an extensive dose rate survey of an area. Contamination levels are determined using smears and appropriate instrumentation. When alpha contamination is suspected, an alpha scintillation meter is used. Results are recorded on survey maps or SWP/RWP Job history sheets and maintained for reference. Schedules have been established for surveys to ensure that all plant areas are surveyed periodically. The frequency for "B" surveys of a specific area is determined by the level of work activity in that area and the radiation levels existing in the area. Surveys of high radiation areas are performed prior to entry rather than on a routine schedule.

Special radiation surveys of particular items or areas are performed on an "as  ;

needed" basis. Examples of special radiation surveys are the removal of l equipment or materials from a restricted area, leak testing of sealed radioactive l sources, or the shipment or receipt of radioactive material packages.

i 6.3.6.3 Water Samples  !

Samples of water containing radioactivity are collected and analyzed on a routine basis. Spent fuel pool water is analyzed to detect indications of degradation of the fecal stored in the pool. Samples of liquid radicactive wastes and processed wastes are analyzed to ensure levels of radioactivi:y are below the levels permitted for release. Samples are analyzed by Radiation Protection Department personnel and/or offsite Laboratories, (as applicable) in accordance with established procedures.

6.3.7 Radiation Protection Equipment and Instrumentation A vs.riety of equipment and instruments are used as part of the radiation protection program. Equipment and instrumentation are selected to perform a pa;ticular function. Sensitivity, ease of operation and maintenance, and reliability are factors that are considered in the selection uf a particular 6-8 Rev 2 August 1998

?

i instrument. As the technology of radiation detection instrumentation improves, new instruments are obtained to more accurately measure radioactivity and l ensure an effective radiation protection program.

A suitable number of appropriate radiation detection instruments will be available to perform required radiation and contamination surveys to meet the ,

j requirements of the applicable federa! regulations 6.3.7.1 Portable Instruments i Portable instruments are for radiation surveys to measure dose rates for beta and  ;

gamma radiation and to perform contamination surveys.

Portable dose rate survey instruments are source checked prior to use and are

• calibrated at least semi-annually utilizing a source in the Unit 3 calibration i

facility, or alternate offsite calibration service. The "Co source is calibrated using a precision electrometer rate-meter if the "Co source is used as an onsite  ;

calibration standard, instruments which cannot be calibrated onsite are sent

{

l offsite for calibration by a vendor or PG&E at an offsite location. Detailed procedures are available describing the operation of the instruments. Plant .

personnel are trained in their operation as part of the radiation protection l training program.-

6.3.7.2 Area Radiation Monitors The area radiation monitoring system utilizes fixed gamma monitors with a range of 0.01 to 100 mR/hr. These instruments are calibrated at least quarterly. The  ;

instruments are source checked to test their response at least monthly. ,

6.3.7.3 Laboratory Instrumentation  ;

Laboratory instrumentation used to analyze radioactive samples includes'a high purity Germanium detector-computer based counting system and internal proportional counters used for counting smears, planchets (evaporated water samples), and air samples. A quality control program for laboratory counting  ;

equipment is in effect to perform calibrations and calibration checks of the l equipment when it is in use. Sealed sources traceable to NIST standards are l used for calibration references for this equipment. In addition, the HBPP Radiation Protection Department participates in an interlaboratory sample i plitting program with PG&E's Diablo Canyon Power Plant and an offsite l laboratory. The plant has also participated in NRC directed sample splitting programs, 6-9 Rev 2 August 1998

L 6.3.7.4 Maintenance of Radiation Protection Instruments j l Routine maintenance and calibration of radiation protection instruments is j ' performed by technicians in the HBPP Instrument Maintenance Department. In

! the case of instruments which cannot be calibrated by plant personnel, the instruments are sent to an offsite calibration facility Certification of NIST l l

traceable calibration is required for all instruments included in the calibration program.

6.3.8 Radiation Protection Records The following records related to the radiation protection program are maintained:

• Records of radiation exposure for all plant personnel, including all

' contractors and Wsitors to the plant

• Records of plant radiation and contamination surveys l  !

• Records of radioactivity in liquid and gaseous effluents released to the environmer:t i

e Records of training and qualification of personnel associated with the radiation protection program

!

• Records of periodic checks, tests, and/or calibrations of radiation protection instruments and equipment l

6.4 INDUSTRIAL HEALTH AND SAFETY PROGRAM l The Humboldt Bay Power Plant participates in an industrial safety program l

under the direction of the PG&E Safety, Health, and Claims Department. This program includes accident prevention, hazardous materials control, and l

hazardous waste management programs. The PG&E Safety, Health and Claims l Department has overall responsibility for industrial safety programs within

PG&E.

I L

s 6-10 Rev 2 August 1998 l

l

' . i APPENDIX l l t

Impilcations of Accidents During SAFSTOR Tha fuel in the Spent Fuel Pool contains the greatest percentage of the facility inventory of

. radionuclides. The Spent Fuel Pool will remain in service during the SAFSTOR period to t provide a stable, benign environment for storing the fuel. In comparison to the reactor core i during operations, water temperatures and flows are a small fraction of those experienced paviously by the fuel. The large volume of water in the pool provides a heat sink and l

L containment for small amounts of radioactivity transferred from the fuel. Releases of rrdioactive materials will be minimized by protection of the cladding integrity, containment of  :

spent fuel pool water, and removal of radioactive and other contaminants from the water.

! Thi purity of the water will be maintained during SAFSTOR to prevent corrosion and to control radioactive materials transferred from the fuel. Water pH and levels of contaminants will be  ;

m;intained in ranges where chemical attack is minimized to protect the primary (fuel cladding) cnd secondary (pool liner) barriers against release of radioactivity.

i By maintaining radioactive concentrations in the pool water ALARA, radiation levels in the I

- vicinity of the pool will be reduced and increases in normal rates of radionuclide transfer will be rardily detectable. Also, if failure of the lining should occur, releases will be minimized.

Ecrly in the operation of Unit 3, leakage of the Spent Fuel Pool was detected and a stainless l steel liner was installed to alleviate this problem. Approximately 50 liters (12 gallon) of water is j pumped from the liner every 5 to 7 days with leakage from the pool accounting for about 5  !

l p2rcent of this volume. Sampling of the french drain (under the Spent Fuel Pool), is conducted

- on a periodic basis. "Cs and *Cs radionuclide concentrations in the blotter samples are '

cpproximately 1 percent of the concentrations found in the liner. The radionuclide concentrations are below the limits specified in 10 CFR 20. j Tha fuel itself does not greatly contribute to the personnel exposure associated with the spent fu:1 storage. The water depth (approximately 18 feet over the top of the fuel) p'rovides l

! ad:quate shielding of the spent fuel. A study conducted by Pacific Northwest Laboratory at the Morris Spent Fuel Storage Facility (PNL-3065) found that t' a direct gamma radiation cannot be distinguished from the contribution made by radioactive water contaminants (approximately 3 mR/h at fuel depths of 8 feet).

l l Radiation sources in the vicinity of the spent fuel storage pool are due to deposits of radioactive material on the pool wall, particularly at the surface (tub ring effect), pool water, ci:aning lines, pumps, filters, etc.

Accidents during the SAFSTOR period have a low probability of occurrence and are of minor i consequence, especially when compared with accidents assc,ciated with reactor operations.

' Accidents possible during SAFSTOR operations are analyzed in the assessment presented l below.

1-1 Rev 2 August 1998

l l

)

1.1 IDENTIFICATION AND PROBABILITY OF ACCIDENTS DURING SAFSTOR l The following five accidents were identified as credible and/or worthy of assessment for the I SAFSTOR period, j l

• Spent fuel handling accident

• Spent Fuel Storage Pool rupture )

l

• Heavy load drop into the Spent Fuel Storage Pool )

e Uncontrolled release of radioactive liquid radwaste to the environment

• Explosions, delayed ignition of flammable vapor clouds, release of toxic chemicals, or fire 1.1.1 Soent Fuel Handline Accident it is anticipated that the spent fuel assemblies will continue to be stored in racks throughout much of the SAFSTOR period. A protective cover is in place over the poci The only  !

anticipated fuel handling is for inspection, training, or testing or for removal of the fuel for shipment to a permanent repository. The potential for dropping a spent fuel assembly during SAFSTOR has been calculated based on Unit 3 experience and the number of spent fuel assemblies stored in the Spent Fuel Storage Pool. Three assemblies were dropped during the total handling of all assemblies onsite (over 3000 assembly handlings), including those handlings associated with shipping of 180 stainless steel-clad assemblies off site. The calculated probability assumed a lineer relationship between drop rate and the number of assemblies handled and assumed no decrease in probability as a result of experience. The probability thus calculated is unity, i.e., at least one assembly may be dropped during handling.

Transfer of fuel for shipment is considered as the initial operation of DECON and the probability and consequence of an accident during these activities is discussed in Appendix A of the dismantlement plan.

During an inspection, the protective pool cover would be removed. Spent fuel handling will be accomplished in the same manner as during the operations of Unit 3. The implications of the radiological release, as a result of a fuel handling accident, to conditions at the site boundary i I

were assessed using the assumptions and methodology suggested in Regulatory Guide 1.25,

" Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling l Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors."

The assumptions applied to this analysis were: i e The dropped spent fuel assembly contained the greatest radiological inventory of those stored in the pool.

• "Kr is the single gaseous nuclide of significance for transport up the stack and into the environment. The maximum inventory of "Kr within the dropped assembly is calculated to be 98 CI(July 1984).

1-2 Rev 2 August 1998

o All of the fuel gap inventory of cil rods in ths essembyl cro relscstd. Tha gap inventory is assumed to be 30 percent of the volume, or 29.4 Ci Kr. For purposes of conservatism, the ltnalysis was conducted as if the entire assembly inventory of 98 Ci was vented. >

o No "Kr is retained in the pool. The entire 98 Ci of "Kr is released through the 250 feet ,

stack over a 2-hour period.  ;

i o Still air conditions exist at the time of the accident and are maintained throughout the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l venting of"Kr. l o The individual hypothetically exposed is located on the plant boundary or beyond at the point of maximum concentration of"Kr throughout the 2-hour time period. {

r 1.1.2 A Ruoture of the Spent Fuel Pool The spent fuel storage pool and surrounding structure were designed to withstand an OBE of 0.25g and an SSE of 0.50g with the exception of the 75-ton bridge crane rails. The crane will ,

not be stored over the pool and will not be used for lifting heavy loads over the pool.

A potential for a seismic event in the area exists due to the proximity of the Bay entrance and I Little Salmon faults. The maximum event which is credible from either fault is 7.5 (Modified l' Richter Scale), (Woodward-Clyde Consultants,1980). Because of the nature of seismic svents, it is not t soful to attempt calculation of a probability. However, if it is assumed that a '

seismic event of the maximum activity occurs during SAFSTOR, it is possible to compare the ground motion retulting from the event to the design basis of the spent fuel storage pool. Two e results of earthquake activity which could be of structural consequence to Unit 3 are  !

liquefaction and scil-structure interaction. Based on the Woodward-Clyde Consultants (1980)  ;

l studies, liquefaction potential was assessed to be negligible. These studies also conclude that i the only potential soil structure interaction effect resulting from earthquake activity is on the reactor caisson. Bated on the data available, the design basis for the caisson (Bechtel,1980) ,

is conservative with respect to these effects. The resulting implication of a seismic event to the  ;

opent fuel storage pool's integrity are, therefore, negligible.

For the purpose of analysis, however, it was assumed a rupture of unknown origin occurred in the west side of the pool floor where it overhangs the suppression chamber. The water would then drain into the suppression chamber leaving the spent fuel completely exposed to the air.  ;

Further, for the purpose of analysis, it was also assumed that the pool cover was breached by the event.

1.1.3 Heaw Load Droo into the Soent Fuel Pool A protective cover will normally be installed over the spent fuel storage pool. As a precaution  ;

to minimize eny potential fcr increased reactivity due to a heavy load drop, the spent fuel is l configured in the pool to reduce collective reactivity of the stored array and to provide optimum physical separation of the individual assemblies. In addition, the fuel assemblies are  :

individually enclosed in bora\ channels to further reduce the reactivity of each element.

l The building structure, includlng support for the emergency condenser, was designed and built to seismic criteria considered sufficient at the time for an operating plant. Modifications were ,

1-3 Rev 2 August 1998 j I . - - - . _ _- _ - _.

l mid3 to cnhance th3 scismic d: sign:d to mors stringent standards in tha mid 1970s. Th3 structure is considered to be adequately protected against damage from natural events which could result in heavy load drops into the pool.

There are no plans for lifting or transferring heavy loads over the pool during SAFSTOR.

Transfer of spent fuel to a shipping cask is considered an initial operation of DECON. No other operational accident resulting in a heavy load drop into the pool is considered credible based upon the administrative controls and interlocks that will be in effect during SAFSTOR. The administrative controls include (1) a prohibition to store the crane over the pool, (2) a requirement to not handle heavy loads over the spent fuel storage racks, (3) personnel training requirements, and (4) procedural controls for load handling and crane operation. The circuit i breaker for the crane power supply is locked and administratively controlled. Movement of the crane from its stored position is alarmed.

1.1.4 Uncontrolled Discharoe of Radwaste Tankaae During the operating life of the plant, few incidents were encountered in the radwaste treatment system. In 1973, the storage capacity of the concentrated waste tanks was exceeded, resulting in overflow into the vault containing the tanks. No significant off-site release occurred.

In July 1977, overflow of the radwaste sump resulted in an unmonitored release to the discharge canal. The rate was approximately 1 gallon / min, and total discharge was approximately 2,000 gallons. No radiological release limits were exceeded.

Based on these experiences, it was determined that the worst case accident to the radwaste I treatment system that could occur during SAFSTOR would be the loss of radwaste from the two concentrated wasto storage tanks, a total volume of 10, 900 gallons. For the purpose of the analysis it was assumed that the entirety of the spill would be lost to the discharge canal.

Conversely it was assumed that all of the waste was retained in the soil near the tanks, requiring exhumation and subsequent waste handling as LLW.

1.1.5 Explosions. Fires. and Toxic Chemical Release Offsite accidents could occur in Humboldt Bay or on the railroad tracks east of the HBPP resulting in explosions, fires, or releases of toxic chemicals. Based on the industry experience and the very low shipping rate by either rail or tanker in the area of the plant, the probability of these accidents has been established to be 10 per year.

The worst credible accident is the explosion and associated fire in the two large fuel oil storage tanks, assuming both were filled. The fuel stored onsite is combustible but non-explosive.

Studies of industrial experience with similar tanks suggest that the probability of spontaneous explosion is negligible. For purposes of this analysis, it was assumed that the following conditions would occur as a result of this accident:

f e Offices would be structurally destroyed.

• Fencelines would be breached on the south and east sides of the plant near the intake canal.

1-4 Rev 2 August 1998

' o ' M;jor superstructuro drmags would occur to Units 1 cnd 2.

o Rupture of the refueling building containment would occur.

o ' Damage would occur to the ventilation stack. l l

e ~ Fire would surround the radwaste treatment facility.

l The probability of rupture of the refueling building containment is small, even from a massive i

explosion of both oil storage tanks. Administrative controls and emergency procedures are sufficient to maintain surveillance and security of the fuel inventory throughout the emergency conditions. )

1.2 CONSEQUENCES OF POTENTIAL ACCIDENTS DURING SAFSTOR

- While accidents have an extremely small probability of occurrence during SAFSTOR, the l; consequences were analyzed to determine the potential worst case doses resulting from these {

potentialities.

I 1.2.1 Conseauences of a Fuel Handlina Accident I

if a fuel assembly were damaged during handling (see assumptions in Section 1.1.1) such that the entire "Kr inventory were lost to the atmosphere, the maximally exposed individual would f receive 0.13 mrem over the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> release based on "Kr emission. Total iodine release is negligible. The resulting dose would be negligible, if a fuel handling accident resulted in the

- release of the entire "Kr inventory in the stored fuel, the dose the maximally exposed individual j would ' receive would be 5.1 mrem, less than 1% of the 10 CFR Part 100 guideline value of 25 l Rem to the whole body.

Occupational doses would be increased to collect fuel components, repackage and store them. l i

The occupational dose is expected to not significantly impact annual personnel exposures or increase the number of personnel required for operations.  !

j

- Environmentaiguality would be negligibly affected by such an accident. Air quality would not be affected by Kr release and no other nuclides would be released in concentrations of any {

t consequence. Water quality should not be impacted since the pool water cleanup system would remove the additional radioactivity (except for 'H and "Tc) before any water was j

' discharged to the liquid radwaste system. The maximum concentration of 'H and Mc in any  ;

assembly is 5.9 Ci and 0.24 Ci respectively. If the entire 'H and "Tc inventory of the maximum j activity assembly were mobilized and dispersed in the water in the pool (4.17 x 10' liters), the i pool water concentration would be 1.41 x 10 and 5.76 x 10* pCi/ml for *H and "Tc, respectively, if this water was discharged into the discharge canal through the radioactive  ;

waste treatment system at a rate of 57 liters per minute, the dilution flow of the circulating water pumps (5.4 x 10' liters per minute) would result in a "Tc concentration of 6.08 x 10* pCi/ml, and a *H concentration of 1.49 x 10* pCi/ml.

Assuming the total core inventory of "Tc (67 Ci) and 'H (1700 Ci) was released under these l 5  !

same conditions, the resulting effluent concentration of "Tc would be 1.70 x 10 pCi/ml and the l

I-5 l Rev 2 i August 1998 ,

cfflurnt concentrttions of'H would ba 4.30 x 10"pci/ml. Both of th:so valu:s are well b low the values given in 10 CFR. Part 20, Appendix B, Table 2.

The maximum quantity of **Tc in any fuel assembly is 0.24 Ci. Assuming an accident caused release of this activity to the spent fuel pool (110,000 gallon), the resulting pool concentration of "Tc would be 5.76 x 10" pCi/ml. If this water was discharged into the discharge canal through the radioactive waste treatment system at a rate of 57 liters per minute, the dilution flow of the circulating water pumps (5.4 x 10' liter per minute) would result in a "Tc concentration of 6.08 x 104 pCi/ml.

Assuming core inventory of "Tc (67 Ci) was released under these same conditions, the 4

resulting effluent concentration of"Tc would be 1.70 x 10 pCl/ml. Both of these values are well below the values given in 10 CFR, Part 20, Appendix B, Table 2.

The consequences of a fuel handling accident are minimal. Occupational dose would increase due to handling, fuel repackaging, and repairs. Public dose would not be significantly increased. EnvironmentaLquality would not be significantly impacted.

1.2.2 Consecuences of a Rupture of the Spent Fuel Storace Pool The loss of radiation shielding by complete loss of pool water would create high radiation levels (100 mR/hr) in the refueling building. Since the radiation would be directly above the pool interior and the pool is below grade, off-site radiation increase would be negligible. Recovery from the accident would increase external exposure of workers because of scattered radiation above the pool. Refilling of the pool would provide the necessary shielding for repair activities.

Reestablishment of the water shielding would require an estimated 3 person-rems. The ventilation efficiency might be decreased so that the inhalation dose might be slightly increased by approximately 1 person-rem over the assumed month's repair efforts.

A calculation has been performed to determine the dose rate at the exclusion area boundary assuming total loss of spent fuel pool water. The calculated dose rate for this postulated accident was 0.024 mrem /hr, at the exclusion area boundary. With this dose rate, there would be ample time to take mitigative action (e.g., reflooding the pool or otherwise shielding the spent fuel) before the design basis accident dose (5 rem) would be received offsite.

The calculation was based on the following information:

• Since the total activity in the spent fuel is several orders of magnitude greater than the activity in the pool water, the activity of the trapped pool water was disregarded for this calculation.

5

• The total activity in the spent fuel (8.5 x 10 Ci) was obtained using the inventory given in Table 10.4.5 of the Humboldt Bay Power Plant Environmental Report.

i

• The average photon energy of approximately 0.5 MeV, with approximately 1 photon per j disintegration, is based on the inventory given in Table 10.4.5 of the Humboldt Bay Power

! Plant Environmental Report.

l l-6 Rev 2 August 1998

o The distence from tho rcfu; ling building to tho naarcst sito boundary (exclusion arec .

boundary) is 700 feet, as discussed in Section V of the Final Hazards Summary Report. l t

o The spent fuel was assumed to be a point source located 7 feet (the approximate length of a fuel element) above the spent fuel pool floor. This point is 19 feet below ground level and 39 feet below the top of the concrete containment building walls. Due to the physical location of the spent fuel, there is no offsite dose from direct radiation.

o Per the guidance of ANSI /ANS-6.6.1-1976," Calculation and Measurement of Direct and Scattered Gamma Radiation from LWR Nuclear Power Plants," air-scattered radiation is

. only considered if it is shielded by less than 2 feet of concrete. Therefore, the only source of airscattered radiation considered is that penetrating the 1-foot-thick concrete refueling building roof.

o The point source is geometrically modeled as being located behind a shield wall with a l height of 39 feet and an internal radius of 20 feet. Skyshine is calculated at a receptor location 720 fee; from the source.

o The skyshine calculational methodology was taken from Consumers Power's " Advanced Health Physics Training Manual," prepared by NUS Corporation.

o An attenuation calculation was performed (with no correction for buildup) to account for the shielding provided by the refueling building roof. This resulted in a reduction factor of 0.002 i

to the offsite dose rate.

The loss of water from the pool has no effect on criticality potential. Repositioning of the array l configuration might impact criticality potential if some water remained in the pool. This is discussed in Section 1.2.3. The decay heat rate from the fuel stored in Unit 3 is sufficiently low tsy mid-1984 that the cladding is not subject to deterioration by thermal effects.

l l A preliminary analysis to address total loss of spent fuel storage pool water is provided in Appendix 1A," Spent Fuel Heat Up Following Loss of Storage Pool Water."

Any releases of radioactive material should therefore be limited to relatively minor amounts of l

particulates that could become suspended in air from contaminants on pool surfaces when l

desiccated. These would lead to surface contamination in the refueling building but would not be significantly transported into the environment.

l Ground water could be contaminated by the pool water, but the contamination would be very slight. Resulting concentrations of the three significant isotopes in ground water would approximate 1.2E-7 pCi/ml "Cs,7.5E-9 pCi/mi *Cs, and 2.5E-10 pCi/ml

• Co. This estimate is conservative since the water volume released from the pool would reach an equilibrium with the very high water table in the site soil strata, resulting in less release than the entire pool volume.

All data at the site indicate that the net flow of groundwater in the vicinity of the spent fuel storage pool is towards Humboldt Bay. Flowmeter measurements made by Bechtel in 1984 indicated that the groundwater flow direction is affected by the tidal cycles in Humboldt Bay.

I-7 Rev 2 August 1998

l During th3 summ:r months, and during flood tidss, groundwater flow is g:nnrally landward.

During ebb tides, groundwater flow is towards the bay. During winter months groundwater flow l would also be affected by tidal actions but not as much as in the summer. During the winter, the controlling flow would be always toward Humboldt Bay even though, during flood tides there may be a minor temporary flow reversal.

The tidal cycle in Humboldt Bay for a 24-hour period consists of two flood tides and two ebb tides. Therefore, the flow reversals occur four times a day. Notwithstanding this flow reversal l

phenomenon, the controlling gradient of the flow of groundwater is still toward Humboldt Bay.

The tidal cycle from July 8,1979 to July 15,1979 was studied for flow reversal effects near the i spent fuel storage pool. The tide varied from a maximum elevation of 7.4 feet to a minimum of l

-1.6 feet compared to mean lower low water (MLLW). Assuming a constant summer ,

groundwater elevation of 6 feet, it was calculated that a particle released underneath the spent fuel pool would travel a total distance of 5.10 feet within the seven-day period toward Humboldt i Bay, at a calculated average travel speed of 0.70 feet / day.

(a) Onsite Groundwater l

l The maximum volume of radioactive materials that can be released to the groundwater underneath the spent fuel storage pool was assumed to be equal to the volume from the top of the pool (elevation 12 feet above MLLW) to a mean tide level of 3.3 feet. A release of the radionuclides would be affected by the flow reversal pattern underneath the site, and the l radionuclides would remain in that region for a long period of time due to the cyclic effect of  !

the tides.

, 1 l

Two conditions were evaluated in order to determine the estimate radionuclide concentration of onsite groundwater at various depths due to a rupture of the spent fuel pool. Table 1-1a shows a condition during the rainy season when the ground water level ,

l was assumed to be at an elevation of 9 feet. Table 1-1b shows a condition during the j summer dry season when the groundwater level was assumed at an elevation of 6 feet.

l 1 (b) In the Humboldt Bay I As discussed earlier, the gradient of the groundwater flow is generally towards the Humboldt Bay. Even if there is a temporary reversel of flow due to tidal fluctuations, the effect of the reversal is not significant. The retum flow (toward Humboldt Bay) is always greater than the reversal. Based on this concept, estimates of the concentration at various distances away I from the rupture toward Humboldt Bay are shown in Tables 1-2a and 1-2b. Table 1-2a shows a condition during the rainflood season with an assumed water elevation of 9 feet. Table l-2b shows a condition during the summer dry season when the groundwater level could be at an assumed elevation of 6 feet. Both of these tables were developed with average tide level for exit conditions. The concentration shown at a distance of 420 feet represents the exit l point into the bay at which the isotopes are not yet subject to dilution in the bay. The j concentration at Humboldt Bay will be diluted much more due to the flushing action of ocean

water entering and leaving Humboldt Bay due to tidal action.

The only area which could be impacted by liquid discharges of radionuclides from Humboldt Bay Unit 3 is within the Humboldt Bay itself. Whenever possible, the following information I-8 l Rev 2 August 1998

i on ennual harvasts is giv:n in terms of Humboldt Bay. If Bay-sp cific data is not availablo, annual harvests are for the Bay and coastal areas for which data are available.

Information on the annual commercial harvest of some fish and shell fish is available specifically for Humboldt Bay. For fish such as salmon or surf perch, the harvested area for the port of Eureka ranges from the Mendocino coast to the Oregon border.

Annual Commercial Yield Benthic (Estimated yields within Humboldt Bay)

Oyster - 500,000 lbs Crab - 2,000-3,000 lbs Clam - minor

, Pelagic (Northern California yields brought into the port of Eureka)

Salmon - 130,000 lbs (1984 yield)

Perch - 12,000 lbs (1984 yield)

(Estimated yields within Humboldt Bay)

Herring -

40 to 60 tons Anchovy -

6 tons (annual quota)

The available current information on the recreational harvest of marine fish is contained in a marine fisheries report covering the entire northern California coastline from Morro Bay to the Oregon border. Specific information on Humboldt Bay is not available directly from the report.

The following information on recreational yields has been derived from two sources: (1) the results of current species-specific surveys by the California Department of Fish and Game, and (2) Department of Fish and Game Fish Bulletin 130, dated 1965. The information contained in the Fish Bulletin was obtained in 1958 as part of a survey which estimated the annual yield from the major fishing pier within the bay (Lazio pier). The report does not include estimates of the total yield from Humboldt Bay which may be expected to be as great as two to three times that of the pier.

Annual Recreational Yield Benthic (Estimated yields within Humboldt Bay)

Clams - 123,000 ea. (based on random sampling)

! Crab -

data not available Annual Recreational Yield Pelagic (Humboldt Bay and coastline within 25-mile range) l Salmon - 78,000 lbs (estimate of 13,000 landings based on I-9 Rev 2 August 1998

I rcndom s:mpling; avarego wright l estimated to be 6 lbs)

(Humboldt Bay Lazio pier)

Surf / Rock Fish -

10,000 lbs (1958 yields at Lazio pier)

(c) At the Nearest Offsite Potable Water Supoly Location i'

l The two potable water wells nearest the spent fuel storage pool are owned by PG&E.

l Well No.1 is about 650 feet east of the site and Well No. 2 is about 2,980 feet southeast of the site. These wells, which are sampled quarterly for activity, provide j onsite water supplies.

l As noted earlicr, there is no radionuclide transport landward because of the persistent I l

l gradient of the groundwater level toward Humboldt Bay. As demonstrated earlier, even  ;

l the cyclic variations of flow due to tidal effects would not cause a large enough reversal for a sufficient period of time to transport radionuclides eastward of the site.

I Although not considered possible, an analysis was performed with an assumption that l a severe drawdown at a particular well due to severe pumping resulted in a difference in head of 5 feet. Table 1-3 shows the results of the analysis for Well Nos.1 and 2.

l The limit for total dissolved activity in the spent fuel storage pool water is given in l Technical Specifications Table ill-2. The limit is based on the anticipated levels that I would be maintained following upgrading of the spent fuel pool water cleanup prior to  ;

SAFSTOR. '

i The Technical Specifications IV.A.2 states that the spent fuel pool circulation water pumps shall be operated as necessary to maintain spent fuel pool water within technical specification limits.

l 1.2.3 Conseauences of a Heavy Load Droo into the Soent Fuel Storace Pool Based upon the discussion in Section 1.1.3, the potential for a heavy load drop into the spent i fuel storage pool is extremely small. If, however, design and/or administrative controls used for

! protection were to deteriorate and if a heavy load drop were to occur, the following i consequences could result:

l e The pool structure could crack. If the pool structure cracks, the worst crack could occur as l postulated in Section 1.1.2 whereby water from the pool would drain to the Suppression Chamber. The consequences would be similar to those described in Section 1.2.2.

• A single fuel assembly or multiple fuel assemblies could be damaged such that the cladding would be breached. The consequences would be similar to those described in Section 1.2.1 for fuel handling accident.

• Spent fue! could be reconfigured/ crushed such that an increase in reactivity would occur. If sufficiently severe, criticality could occur. Administrative / procedural controls prevent the

( l-10 Rev 2 August 1998

movsm:nt of hnavy loads over tha sp nt fu:1, climinating ths potential for dropping a haavy load onto the spent fuel.

PG&E has evaluated design alternatives that would prevent possible criticality due to seismic and heavy load events. A report on this completed task is presented in Appendix IB " Pacific Gas and Electric Company Humboldt Bay Power Plant Unit 3 Criticality Analysis for SAFSTOR Decommissioning."

1.2.4 Conseauences of an Uncontrolled Release of Radwaste Tankaae in determining the quantities of radioactive materials that could be released in the worst case radwaste treatment system accident during SAFSTOR, it was assumed that due to radioactive

!. decay, decontamination efforts and lower levels of operational activity, the radioactivity being

! added to the waste storage tank would be less than current additions. The concentrations of the four most significant radionuclides were assumed to be the maximum measured when the storage tank was sampled in 1984.

The tank was assumed to contain its maximum capacity (10,000 gallons) at its maximum concentration. The volume of the tank was assumed to be released to the discharge canal over an 11-hour period (15 gpm) through the 2-inch diameter waste discharge line.

l The consequences of release of the 37,800 liters (10,900 gallons) of concentrated waste tank l storage are measurable. If all the waste were discharged via the radwaste line to the canal, the concentrations of *Cs, *Cs, "Co, and "Sr, the four :!gnificant nuclides, would be less than water effluent concentration limits (10 CFR 20, Appendix R. Table 2) at the discharge point. If j all the waste were spilled on the soil, the contaminated area would be approximately 94 m' to a depth of 40 cm, due to soil permeability. The approximate concentrations of the significant nuclides would be: 3.5 Ci/g *Cs,1.2 pCi/g *Cs,1.1 pCi/g "Co, and 0.01 pCilg "Sr. If this contaminated area were exhumed for LLW disposal,185 55-gallon drums would be required.

Occupational dose is negligible from these operations. There is no impact on public dose or i l environmental quality. .

l i

1.2.5 Consecuences of Exolosion. Fire, and Toxic Chemical Release  !

l An explosion and fire of the large fuel storage tanks on site would obviously cause damage to 1 the plant facilities and incapacitate Units 1 and 2. The consequences to Unit 3 would be minor i and could include:

l Conseauences to Securitv. Physical surveillance of any breached fences and gates would be I

required while repairs are completed.

l Rupture of Refuelino Buildina Containment. The working conditions in the refueling building during SAFSTOR will require personnel monitoring but no protective clothing under normal operating conditions. Negligible nuclide suspension to the air is therefore expected even if the building superstructure were entirely vented.

l Damaae to Ventilation Stack. Ventilation systems would be shut down and the suspended I' l particulate dose to workers might increase slightly during repairs, estimated at less than 0.2 I

l-11 l l

Rev 2 August 1998

parson-rem. No public exposure or environmental quality impact would result from radiological hazards.

Fire in the Unit 3 Restricted Area. There are no significant quantities of flammables or pressurized equipment in the area of the radwaste treatment and storage buildings. It is believed that no loss of stored wastes would result from a fire in their vicinity inside the Unit 3 restricted area.

Although a calculation has not been performed to evaluate this particular sequence of events, it l is not considered possible for a seismic event to rupture the spent fuel storage pool and the  !'

onsite fuel oil storage tank which then causes a fuel oil fire in the pool.

Each of the two main fuel oil storage tanks is surrounded by an earthen dike that has been in place for more tnan 20 years. The minimum dike cross-section is 10 feet top x 50 feet bottom x 10 feet high. The banks of the dikes are covered with vegetation and the tops are paved with 1 asphalt. The capacity within each dike area is greater than the maximum available volume of I the associated fuel oil storage tank (volume above the tank elevation which corresponds to the

~

top of the dike). Therefore, even in the unlikely event of a tank rupture, all oil is expected to be contained within the fuel oil dike area.

In the unlikely event of rupture in the east side of the earthen dike, it is not expected that the fuel oil could reach the spent fuel pool since any flow in that direction would be impeded by the administration building and Units 1 and 2. It is more likely that a rupture of the dike in this area would result in flow to the intake canal.

Furthermore, the fuel oil stored in these tanks is extremely viscous, similar to the consistency of l tar, and as such, it is not of a nature to flow freely. A fuel oil dike rupture in any other direction I would result in flow away from Unit 3.

1 i

I L

1 l

l-12 Rev 2 August 1998

TABLE l-1a ESTIMATED RADIONUCLIDE CONCENTRATION l OF ONSITE GROUNDWATER AT VARIOUS DEPTH DUE TO POSTULATED RUPTURE OF SPENT FUEL STORAGE POOL (Rainy Season) l Given: Groundwater Level = 9 feet. Mean Tide Level = El. 3.3 feet. l Postulated Volume of Water Released from Spent Fuel Storage i Pool = 4524 feet'. Hydraulic Conductivity K = 10,400 feet / year.

Distance Depth Peak from below Travel Concentration Amount of Concentration Rupture Rupture time at location radionuclide NRC Limit isotope (Feet) (Feet) (years) (pCi/ml) released (Ci) ( Ci/ml)*

'C s 30 0' 6.95 2.7 X 10" 0.54 2 X 10

10 5.1 X 10

20 3.5 X 10~7 30 3.5 X 10

• Cs 30 0 6.95 3.3 X 10* 0.055 9 X 1 0

10 6.2 X 10~7 20 4.2 X 10*

30 4.2 X 10

4 7

• Sr 30 0 0.58 4.6 X 10 0.00067 3 X 10 10 8.8 X 10'7 20 5.9 X 10*

30 5.9 X 10-t2 4

• Co 30 0 5.25 1.2 X 10 0.0031 5 X 10

10 2.3 X 10'7  ;

4 20 1.6 X 10 30 1.6 X 10

Note: Computations were based on point concentration model presented by Codell and Duguid (reference)

• From 10 CFR 20, App. B, Table 2, Column 2 1-13 Rev 2 August 1998 I

l i

4  ;

TABLE l-1b

{

ESTIMATED RADIONUCLlDE CONCENTRATION  !

OF ONSITE GROUNDWATER AT VARIOUS DEPTH [

- DUE TO POSTULATED RUPTURE OF SPENT FUEL STORAGE POOL  :

Given: Groundwater Level = 6 feet. Mean Tide Level = El. 3.3 feet.  !

Postulated Volume of Water Released fiom Spent Fuel Storage Pool = 4524 feet'. Hydraulic Conductivity K = 10,400 feet / year.

l (Dry Season) l l Distance Depth Peak .

from below Travel Concentration Amount of Concentration '

Rupture Rupture time at location radionuclide NRC Limit  !

Isotope (Feet) (Feet) (years) (pCi/ml) released (Cl) ( Ci/ml)*  ;

7 i CS 30 0 14.7 2.3 X 10* 0.54 2 X 10 4 I 4.3 X 104 10

! 20 2.9 X 10'7 30 2.9 X 10'"

• Cs 30 0 14.7 2.5 X 10'7 0.055 9 X 104

. 10 4.8 X 10*

j 20 3.2 X 10'"  ;

30 3.2 X 10

i "Sr 30 0 1.23 4.6 X 10* 0.00067 3 X 1 0'7 2

10 8.6 X 10-7 20 5.8 X 10*

30 5.8 X 10

4 I "Co 30 0 11.1 5.7 X 10~7 0.0031 5 X 10 10 1.1 X 10

i 20 7.2 X 10'"

30 7.3 X 10

i i Note: Computations were based on point concentration model presented by Codell and Duguid (reference)

*From 10 CFR 20, App. B, Table 2, Column 2 1-14 Rev 2 August 1998

1 TABLE l-2a PEAK CONCENTRATION AND TRAVEL TIME OF RADIONUCLIDES IN GROUNDWATER IN THE SITE AREA (Assumed Rainflood Season)

Given: Volume of water released from spent fuel storage pool = 4524 feet'.

Groundwater Level = El. 9 feet; Mean Tide Level = El. 3.3 feet.

l Hydraulic Conductivity K = 10,400 feet / year.

Distance Peak from Travel Concentration Concentration at Rupture time at Spent Fuel Specified Distance NRC Limit Isotope (years) Pool ( Ci/ml) (pCi/ml) (pCi/ml)*

Source (ft) 4 50

7 Cs 11.6 4.22 X 10 1.1 X 10* 2 X 10 4

'"Cs 11.6 4.27 X 10 3.2 X 10'7 9 X 10*

4 "Sr 0.97 5.2 X 10* 2.1 X 10 3 X 10'7 f

4 4 "Co 8.75 2.4 X 10 3.6 X 10'7 5 X 10 4 4 100

7 Cs 23.1 4.22 X 10 ' 3.1 X 10 2 X 10 4

• Cs 23.1 4.27 X 10 2.6 X 10* 9 X 10*

"Sr 1.94 5.2 X 10' 7.3 X 10'7 3 X 10'7 4 4 "Co 17.5 2.4 X 10 4.0 X 10* 5 X 10 4 4 150

7 Cs 34.7 4.22 X 10 1.3 X 10 2 X 10 4

• Cs 34.7 4.27 X 10 3.1 X 10'" 9 X 10*

4 "Sr 2.92 5.2 X 10 3.9 X 10'7 3 X 10'7 "Co 26.2 2.4 X 10 4

7.0 X 10* 5 X 10 4 4

200

7 Cs 46.3 4.22 X 10 6.4 X 10* 2 X 10 4

• Cs 46.3 4.27 X 10 4.3 X 10 9 X 10*

"Sr 3.9 5.2 X 10* 2.5 X 10~7 3 X 10'7 "Co 35.0 2.4 X 10 4

1.4 X 10* 5 X 104

7 4 250 Cs 57.9 4.22 X 10 3.7 X 10* 2 X 10 4

i *Cs 57.9 4.27 X 10 7.1 X 10 9 X 10*

4 7 "Sr 4.86 5.2 X 10 1.8 X 10^7 3 X 10 "Co 43.7 2.4 X 10 4

3.5 X 10" 5 X 104 l

l l-15 l Rev 2 l August 1998

I TABLE l-2a (Con'd)

PEAK CONCENTRATION AND TRAVEL TIME OF

' RADIONUCLIDES IN GROUNDWATER IN THE SITE AREA (Assumed Rainflood Season) f l Distance Peak i from Travel Concentration Concentration at i Rupture time at Spent Fuel Specified Distance NRC Limit l

Source (ft) Isotope (years) Pool (pCi/ml) (pCi/ml) (pCi/ml)*

i

7 4 300 Cs 69.5 4.22 X 10 2.2 X 10* 2 X 10*

• Cs ,

69.5 4.27 X 10 4

1.2 X 10 9 X 10*

"Sr 5.83 5.2 X 104 1.4 X 10'7 3 X 10 7

"Co 52.5 2.4 X 104 8.6 X 10'" 5 X 10 4

4 4 "420 7 Cs 97.3 4.22 X 10 7.1 X 10 2 X 10 l

• CS 97.3 4.27 X 10* 7.8 X 10 9 X 10*

l "Sr 8.17 5.2 X 10* 8.2 X 10* 3 X 10'7 "Co 73.5 2.4 X 104 3.4 X 10 52 5 X 10 4

l l

• From 10 CFR 20, App. B, Table 2, Column 2

" Distance to the exit point at bay Note: Computations were based on simplified analytical methods for minimum dilution l presented in the reference (Codell and Duguid) l l

i ,

l l-16 Rev 2 August 1998 ww---y n+ -

}

I TABLE l-2b  :

PEAK CONCENTRATION AND TRAVEL TIME OF  !'

RADIONUCLIDES IN GROUNDWATER IN THE SITE AREA (Assumed Rainflood Season) ,

!Given: Volume of water released from spent fuel storage pool = 4524 feet'.

Groundwater Level = El. 6 feet; Mean Tide Level = El.3.3 feet.

l. Hydraulic Conductivity K = 10,400 feet / year.

Distance Peak from Travel Concentration Concentration at t Rupture time at Spent Fuel Specified Distance NRC Umit .

Source (ft) Isotope (years) Pool (pCl/ml) (pCl/ml) (pCi/ml)*

4 4 4 50

7 Cs 24.5 4.22 X 10 8.4 X 10 2 X 10 7

4

• Cs 24.5 4.27 X 10 4.6 X10* 9 X 10* l 4

"Sr .2.06 5.2 X 10* 2.1 X 10 3 X 10  !

4 4 "Co 18.5 2.4 X 10 9.9 X 10* 5 X 10  ;

4 4 4 i 100 C s 49.0 4.22 X 10 1.7 X 10 2 X 10 I

d

• Cs 49.0 4.27 X 10 5.0 X 10 9 X 10*

"Sr 4.1 5.2 X 10* 7.0 X 10 3 X 10'7 4 4 "Co 37.0 2.4 X 10 3.1 X 10* 5 X 10

'87 4 150 Cs 73.4 4.22 X 10 5.2 X 10* 2 X 10 i

• Cs 73.4 4.27 X 10* 8.6 X 10 9 X 10* l 4

"Sr 6.17 5.2 X 10 3.6 X 10-7 3 X 10'7 4 4 "Co 55.5 2.4 X 10 1.5 X 10'" 5 X 10 4

200 'C s 97.9 4.22 X 10 1.9 X 10* 2 X 10 .

• Cs 97.9 4.27 X 10* 1.9 X 10" 9 X 10* {

"Sr 8.22 5.2 X 10 4

2.2 X 10~7 3 X 10'7 I i

4 4 "Co 74.0 2.4 X 10 8.7 X 10-12 5 X 10 ,

1

7 4 250 Cs 122.4 4.22 X 10 8.3 X 10'7 2 X 10 9 X 104 4

• Cs 122.4 4.27 X 10 4.0 X 10d ,

7 "Sr 10.28 5.2 X 10* 1.6 X 10 3 X 10 4 4 "Co 92.5 2.4 X 10 5.9 X 10' 5 X 10 ,

l-17  !

Rev 2 l August 1998

l l

i t

TABLE l-2b (Con'd)

! PEAK CONCENTRATION AND TRAVEL TIME OF l

RADIONUCLIDES IN GROUNDWATER IN THE SITE AREA (Assumed Rainflood Season)

Distance Peak from Travel Concentration Concentration at Rupture time at Spent Fuel Specified Distance NRC Limit Source (ft) Isotope (years) Pool (pCi/ml) (pCi/ml) (pCi/ml)*

"7 300 Cs 146.9 4.22 X 10 3.7 X 10 2 X 104

• CS 146.9 4.27 X 10* 9.7 X 10" 9 X 10*

"Sr 12.33 5.2 X 10* 1.2 X 1 0-7 3 X 10'7 4

"Co 111.0 2.4 X 10 4.1 X 10'" 5 X 10*

"7 "420 Cs 205.6 4.22 X 10

• 5.9 X 10* 2 X 10 4

i l *CS 205.6 4.27 X 10* 2.3 X 10* 9 X 10*

4 4 l

"Sr 17.3 5.2 X 10 6.6 X 10 3 X 10~7 i "Co 155.4 2.4 X 104 7.6 X 1047 5 X 104 l

• From 10 CFR 20, App. B, Table 2 Column 2 l

" Distance to the exit point at bay Note: Computations were based on simplified analytical methods for minimum dilution j presented in the reference (Codell and Duguid) i l

l l

l l

1-18 Rev 2 August 1998

TABLE l-3 ESTIMATED RADIONUCLIDE CONCENTRATION AT PG&E WELLS NO.1 AND 2 FOR POSTULATED SEVERE DRAWDOWN ONLY AT WELLS Assumed Difference in Head = 5 feet Peak NRC Concentration Travel Concentration Concentration at Spent Fuel time ~at W ell Limit Location isotope Pool (pCl/ml) (Year) (pCi/ml) (pcilmi)*

4 PG&E Well

7 Cs 4.22 x 10 256.4 3.25 x 10* 2 x 10

1. 1 ~

d

• Cs . 4.27 x 10" 256.4 2.15 x 10 ' 9 X 10*

(650 ft. East)

"Sr 5.2 x 10* 21.5 1.1 x 10 3 x 10^7 4 5 x 10*

"Co 2.4 x 10 193.7 9 x 10'"

4 PG&E Well

7 Cs 4.22 x 10 1175.3 3.8 x 10* 2 x 10

1. 2 (2980 ft. SE) *Cs 4.27 x 10" 1175.3 0 9 x 10*

4 l "Sr 5.2 x 10 98.7 3.3 x 10* 3 x 10

"Co 2.4 x 10 4

888.2 6.4 x 10*7 5 x 104

• From 10 CFR 20, App. B Table 2, Column 2 NOTE: The following assumptions were used in all calculations for the tables shown in the response.

1. Hydraulic conductivity was based on Bechtel's report, K = 10,400 ftlyear (28.5 l ft/ day)

2. Other parameters used are:

l Effective porosity = 0.25 l Total porosity = 0.40 Longitudinal dispersion = 1.0 ft i Lateral dispersion = 0.5 ft Distribution coefficients

7 Cs and *CS = 20 ml/g "Sr = 1.5 ml/g l

"Co = 15 ml/g l

l-19 Rev 2 August 1998

I APPENDIX IA 5 pent Fuel Heat Up Following Loss of Storage Pool l

l l-20 Rev2 Atint141998

SUMMARY

A parametric comparison was made between Humboldt Bay Power Plant's spent fuel storage configuration and that used in Sandia Laboratories heatup analyses (Ref. 2) for the unlikely event of total loss of spent fuel storage pool water. It was concluded that the Humboldt Bay spent fuel heatup would be much less severe than the cases analyzed by Sandia Labs for which fuel cladding integrity was maintained. Using Sandia's results, it I

can be concluded that the Humboldt Bay spent fuel clad will remain intact during this postulated accident. Therefore, it is recommended that no further plant-specific spent fuel l heatup analyses be performed.

The assessment is based on the assumption that either ventilation inside the pool is available at a rating of about 300 cfm, or that the pool cover being designed is provided l

with a blowout opening.

l l-21 Rev 2 Ayyust 1998

l i

1. INTRODUCTION Hum'boldt Bay's nuclear unit, a 65 MWe boiling water reactor (BWR), has been shut down since July 1976, and its nuclear fuel is now stored in the unit's spent fuel storage pool.

Currently, the decay heat generation is so low that the fuel pool coolers are not used.

During the recent review of Humboldt Bay's Decommissioning Plan, the NRC Staff requested that a plant-specific calculation be performed to demonstrate that the spent fuel {

will not be damaged in the case of loss of water from the storage pool (Ref.1). This report l uses published results of generic spent fuel heatup analyses under this scenario to ,

qualitatively argue that the analyses performed by Sandia Labs (Ref. 2) are bounding and l that the Humboldt Bay spent fuel cladding will remain intact in the unlikely event of loss of water from the storage pool. ,

I The likelihood of a complete drainage of the pool is judged to be extremely low, in the i 4

range of 10 to 10'7 per year (Ref. 3). Humboldt Bay's spent fuel pool is below ground level. The lowest ground water level is about 7 feet above the top of the fuel. Low water level and high radiation alarms are available to warn the operators of any abnormal conditions, and action can be taken early enough to avoid complete loss of water from the pool.

Postulating a complete pool drainage, however, the fuel elements will heat up and will eventually reach a steady-state temperature when the decay heat generation rate is balanced by the heat loss out of the pool, if the decay heat generation rate is sufficiently high relative to the heat loss, the cladding may reach a high enough temperature (850 -

950*C) to cause the clad to rupture as a result of internal pressure or to undergo rapid exothermic oxidation leading to clad melting. l l

A spent fuel heatup analysis during such a scenario has been performed by Sandia Laboratories for various typical spent fuel pools. Section 2 describes their approaches, assumptions, and the results. Section 3 compares Humboldt Bay's fuel and storage facility parameters that affect such heatup analysis with those used in the Sandia analyses, and concludes that damage will not occur during such an accident at Humboldt Bay.

l l

l-22 Rev 2 A m # 1998

l f

i

2. SANDIA LABORATORIES' SPENT FUEL HEATUP ANALYSES J

i Sandia Laboratories performed spent fuel heatup analyses following loss of water for ,

! several storage facilities (Ref. 2). The discussion in this report concentrates on those  !

related to BWR fuel as it applies to Humboldt Bay.

" The analyses were based upon the SFUEL computer code, to assess the effect of decay time, fuel element design, storage rack design, packing density, room ventilation, and other variables on the heatup characteristics of the spent fuel and to predict the conditions  ;

under which clad failure will occur. It was found that the likelihood of clad failure due to rupture or melting is extremely dependent on the spent fuel decay time period and the j_ storage configuration. ,

t o

Two types of BWR spent fuel storage facilities were investigated. One is a typical onsite  !

storage pool, such as Dresden Unit 2 or Brunswick, and the other one is an away-from- l reactor storage facility, such as the G. E. Morris facility. Since a typical BWR spent fuel i storage pool is open to the entire secondary containment, the amount of air available for cooling is approximately 1,800,000 cubic feet; thus, the heatup is much less severe than i for a typical pressurized water reactor (PWR) spent fuel pool.

The two types of storage racks considered in the analysis (directional and cylindrical) are i shown in Figure 1. The fuel assembly data used in the analyses are also shown in Figure [

1.

- A typical full core discharge loading was used in the analyses. The storage pool is assumed to have a capacity of 1.75 cores. The total decay heat generation rate after.30 days is about 5000 kW. .

l The heat removal problem for the drained spent fuel pool is considered in two parts: (1) i the heat transfer problem within the confines of the spent fuel region in the pool, and (2) l the removal of heat from the containment building.

Heat produced by decay within the spent fuel elements and by chemical oxidation of the [

clad is removed, in part, by buoyancy-driven air flows circulating in the open channels. i Calculation of this air flow uses an iterative, finite-difference solution of conservation l equations. Transient conduction equations are solved in the axial direction of the fuel  !

element to determine the heatup of the fuel rod. Radiation heat transfer between structural elements is accounted for, as is transient conduction into the pool's concrete encasement.  :

Three cases were considered for the containment building heatup: (1) the active  :

ventilation system has enough capacity to keep the ambient air temperature at a constant temperature (well-ventilated room); (2) the active ventilation system is functional (at the  !

ANS Standard 57.2 minimum requirement of two complete air changes per hour), but not -

sufficient to remove all the decay heat; and (3) no active ventilation system. In both cases  !

(2) and (3) the building air temperature will increase to affect the natural convection 1-23  !

Rev 2 i

_ _ _ _ _ _ _ _ _ _ _ swarmwn _ _ _

l process in the pool The SFUEL code computes the amount of heat that is removed by a  !

combination of forced ventilation, leakage of 1

l i

i

! l

{

l 6

l

[

l r  ;

I t

l i

l l

I f

i

e. ,

l I l-24 i Rev 2 i a . ... .a i nn.

air through the building structure, heat storage by the structural heat sinks, and radiation /

natural convection from the building exterior to the outside.

The primary assumptions embodied in the Sandia Laboratories analysis are as follows:

o Water drains instantaneously o Fuel assembly and rack geometries remain undistorted o All rods in a particular fuel assembly have the same vertical temperature distribution o No building leakage occurs until the building internal pressure exceeds 2 psi, after which the leakage rate was assumed at the rate required to keep the building at 2 psi o The room air is assumed to be well mixed The results of the analysis for BWR fuel elements can be summarized as follows:

o The peak clad temperature occurs at the top of the active fuel. Figure 2 shows a typical variation of clad and air temperature with vertical distance.

} o Fuel decay time has a dominant effect on the clad heatup. The peak clad f temperature was decreased by a factor of two as the decay time increased from 90 days to 1 year.

i o The larger the baseplate hole size, the lower the peak clad temperature will be.

This is because more air flow can go into the fuel assembly.

o The effect of the storage rack configuration is significant. Heatup is more severe if the fuel channel is not removed. Most of the spent fuel stored have channels removed. However, some does have the channels attached which will reduce heat removal efficiency and result in higher peak clad temperature.

o Figures 3 and 4 show the partitioning of heat losses for a perfectly ventilated room and a nonventilated room, respectively. For a well-ventilated room, most of the decay heat is removed by natural convection of the air. Without ventilation, most of the heat generated initially is transported to the fuel and structures. Only after a long time does the heat loss to outside by radiation and natural convection from the building simcture become dominant.

o Figure 5 demonstrates the effect of ventilation on the spent fuel heatup. It is clear that the peak clad temperature will become much higher without ventilation.

1-25 Rev 2 August 1999

l In conclusion, for most of the cases analyzed by Sandia, the peak clad temperatures are l

below the critical temperature (about 900*C). Only for those worst cases where the decay time is short, the baseplate hole size is small, the channel is attached, and ventilation is not available, the peak clad temperature was calculated to be higher than the critical temperature. In the following section, those Humboldt Bay parameters that affect the heatup analysis will be compared with those used in the Sandia Labs analysis to demonstrate that, if loss of water from the Humboldt Bay storage pool occurs, the heatup will be significantly less than for those cases analyzed by Sandia.

i 4

1-26 Rev 2 August 1998

3. COMPARISON AND ASSESSMENT l

The Humboldt Bay Power Plant has been shut down for more than 8 years. All nuclear i fuel elements have been removed from the reactor vessel and are stored in the spent fuel pool. There are 390 assemblies in the spent fuel pool, of which some 173 are with channels attached.

A sectional view of the plant is shown in Figure 6, and a plant view showing the spent fuel j pool is shown in Figure 7. The fuel racks are at the bottom of the pool. The fuel rack  ;

arrangement is partially shown in Figure 8 and a picture of one fuel rack is shown in Figure l

9. Each rack can hold eight assemblies. The rack is made of aluminum angle irons  !

welded on a baseplate. Cooling fluid can enter the assembly through the space between '

the angle irons (there is no hole on the baseplate). If the channel is attached, the cooling fluid has to enter the assembly at the bottom space below the channel but above the l baseplate, between the angle irons. As demonstrated by Sandia Laboratories, those with l channels would yield higher peak clad temperature. Those without channels have better l cooling and, therefore, the heatup is less severe. Thus, the discussion below concerns l those fuel assemblies with channels attached. Figure 10 shows the loading pattern of the  ;

spent fuel in the pool. Those with channels attached are identified by asterisks.

Currently in the storage pool there are three types of fuel assemblies. They all have  !

Zircaloy cladding. The fuel data are shown in Table 1. j To assess Humboldt Bay spent fuel heatup effects after the complete loss of spent fuel i pool water, utilizing the results in Sandia Laboratories' study, the major parameters affecting the heatup analysis are compared (Table 2). The following is a discussion of the l comparison. The heatup result of the case that is being compared (Ref. 3) is shown in s Figure 5.

a. Decav Heat  ;

t Decay heat generation rate is one of the most dominant factors in the heatup .

analysis. In Sandia Labs' analysis (Figure 5) for a typical BWR spent fuel pool, a j total decay heat of 5,463 kW (average of 4.17 kW/ assembly) was used. This is j based on a 30-day decay time. For Humboldt Bay, the decay heat rate is much  ;

lower because the decay time is at least 8 years. Using Branch Technical Position  !

ASB 9-2 (Ref. 4), the total decay heat is conservatively estimated to be 35 kW, or l 89 W per assembly, which is about 50 times less than what Sandia Labs used in their analysis. The decay heat per active fuel length per assembly is 29 W/in.  ;

(average) in Sandia's analysis, as compared to 1.154 W/in. for Humboldt Bay, a l factor of 25. See Table 2 for the comparison. I

b. Total AirVolume i 1-27 Rev 2

-- - _ ___ _ _ _ __ ____6EME3ES

l The Sandia Labs' analysis utilized a current BWR facility in which the pool is open to the secondary containment (air volume is about 1.8 mcf) In Humboldt Bay, the pool is currently open to the refueling building (air volume is about 200,000 cu. ft.).

However, there are plans to install a cover over the pool to reduce airbome activity within the refueling building. If the cover is airtight, only the air in the pool (about 1,500 cu. ft.)

, will be available for cooling. This is much less than what was used in the Sandia Labs analysis. As shown in Table 2, the decay heat rate per unit air volume available is comparable to what was used in the Sandia Labs analysis if the cover is airtight. Since the heat loss through the pool wall is much less than the heat loss through the sheet metal wall in the containment, as assumed in the Sandia analyses, the heatup will be more severe. However, the pool cover can be designed such that, if the pressure builds up in the pool, the cover can be made to open upon differential pressure so that the air in the refueling building will be available for cooling. As shown in Table 2, the decay heat per unit air volume in this case will be much less than what was used in Sandia's analyses. Pool ventilation, if available, will also keep the air in the pool at near-ambient temperature.

c. Ventilation in Sandia Labs' analysis, several ventilation rates were assumed. The worst case is without ventilation. It was assumed that the building blowout panel set pressure is 2.0 psig. For Humbc!dt Bay, the ventilation rate for the refueling building is 4,500 cfm and the building design pressure is 1.3 psig. This is more adequate than the case in Sandia Labs' analysis. If the pool cover to be installed is airtight and will not be blown open at high internal pressure, it is estimated that a ventilation rate of 300 cfm will keep the ambient air temperature in the pool to be constant at 150*F or lower for the estimated decay heat.

d. Air Flow Path For the Sandia Labs analysis shown in Fig. 5, air was assumed to circulate only through the space between the fuel and the pool wall. In Humboldt Bay's fuel rack arrangement, air can go down to the bottom opening of the fuel assembly through the space between the fuel rack units, thus providing a much better flow path (see Figure 8).

n Sand!a's analysis, it was found that the distance between the baseplate and the floor could make a significant difference in the peak clad temperature. This is because a larger space there will provide a better flow path for the air to enter the l baseplate hole to the fuel assembly. The space available below the baseplate for the Humboldt Bay fuel rack configuration is significantly less (see Table 2).

However, since the flow path is not through the baseplate hole but through an opening above the baseplate, this parameter is insignificant.

It was also found from Sandia Labs' analysis that the baseplate hole size i significantly affected the heatup analysis. The Humboldt Bay fuel rack does not 1-28 Rev 2 Sugus1998

have a hole at the baseplate. If a fuel channel is attached to the fuel assembly, the air enters the assembly through an opening below the bottom of the channel but above the baseplate and between the two adjacent angle irons. It is roughly estimated that this opening is equivalent to a 2.75-inch diameter hole, which is quite comparable to the large hole size (3 inches) used in the Sandia Labs analysis.

l t

I i

n l

l 1-29 Rev 2 August 1998

The flow area of each assembly for Humboldt Bay is smaller than that used in the Sandia Labs analysis, but the fuel rod pitch / diameter ratio is comparable (see Table 2). Since it was concluded in the analysis that 7x7 and 8x8 fuel assemblies did not yield any notable difference in the heatup, it is concluded that this will not make any significant difference.

e. Buildina Structure The building structure used in the Sandia Labs analysis was assumed to be sheet metal, while the pool wall consists of a 1/4-inch thick stainless steel liner over a concrete wall. This is comparable to Humboldt Bay (the liner plate thicknes.s is, however,1/8-inch thick). If an airtight cover were to be used in Humboldt Bay, the heat must dissipate out through the pool wall and the pool cover, which is certainly less efficient than through sheet metal. Unless active ventilation is available to maintain the air at a low temperature, a blowout opening should be designed in the cover to enhance heat dissipation.

Because ^he extremely low decay heat, the Humboldt Bay spent fuel heatup following Los of water will be less severe than the cases analyzed in the Sandia Labs report. Since Sandia Labs predicted relatively low peak clad temperature as shown in Figure 5, Humboldt Bay spent fuel cladding will remain intact under this low probability accident. As a result of this assessment, it is recommended that no further anafasis be performed.

1-30 Rev 2 August 1PI%

4. REFERENCES

1. NRC Letter, J. A. Zwolinski to J. D. Shiffer, dated January 23,1985, " Request for Additional information," Question No. 84, Docket No. 50-133, LS05-85-01-021.

2. " Spent Fuel Heatup Following Loss of Water During Storage," NUREGICR-0649, SAND 77-1271, March 1979.

3. Reactor Safety Study, WASH '9400, Appendix 1,95-99,1975.

4. Branch Technical Position ASE 9-2, " Residual Decay Energy for Light Water

~

Reactors for Long Term Cooling," Revision 2, July 1981.

l l l

l l

l l

l l

l l-31 Rev 2 August 1998

l Table 1 l <

l l HUMBOLDT BAY FUEL ASSEMBLY DATA i

l Type ll Type 111 Type IV - j l

Rod array 7x7 6x6 6x6 l Active fuel height (in.) 79 77.5 77.125 1 Rod pitch (in.) 0.631 0.74 0.74 l Rod OD (in.) 0.486 0.563 0.5625 Pitch /OD ratio 1.298 1.314 1.315 Cladding thickness (in.) 0.033 0.032 0.035 UO weight / assembly (Ibs) 192 191 181 Assemblies per core 172 172 172 Channelinside dimension (in.) 4.542 4.542 4.542 l l

Net flow area per assembly (in.') 11.54 11.67

~

11.68 Number of fuel assemblies currently 88* 179 123 in the spent fuel storage pool L

.

• Ten of the Type ll assemblies are of higher enrichment and, for the criticality study, they are counted as Type 111.

l

! l-32

Rev2 1 _

Degr3 W12 _

Table 2 PARAMETRIC COMPARISON Sandia Labs Analysis Humboldt Bay  :

Cylindrical Rack,30-day decay Total decay heat 5463 kW 35 kW (Based on peak assembly)

Decay heat per assembly 4.17 kW (avg) 0.089 kW (peak) 7.20 kW (peak)

Flow area per assembly 15.89 in.' 11.54/11.67 in.

Total air volume 1,800,000 ft' Pool only: 11,500 ft' l Inc. refueling bldg:

200,000 ft Decay heat / flow area 0.26 kW/in.2 (avg) 0.0077 kW/in.2 ,

0.45 kW/in.2 (peak) ,

Decay heat / air volume 3 W/ft' Pool only: 3 W/ft' inc. refueling 'bidg:

0.18 W/ft  ;

Decay heat / active length per 29 W/in. (avg) 1.154 W/in.

assembly 50 W/in. (peak)

Pitch / diameter ratio 1.3126 1.298/1.314 Ventilation rate 0 - 60,000 cfm 4500 cfm l Baseplate hole diameter 3 in. 2.75 in. (equivalent)

Baseplate from floor 18 in. 6 in.

l l-33 Rev 2 August 1998

e Spent Fuel Storage M s Consid' red in Drained  !

Pool Analysis ' -

1 Mcosilti Design PUperties of Fuel Asseinblies DnMCilomLisWRI. .

- Used in the Analysis g

N l h

1

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! 'j l

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! Rev 2 August 1998

i

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Figure 2. Typical Variation of Clad Temperature With Normalized Di' stance, Measured from Iowest l End of Fuel Rod.

l l

)

l \

i 1-35 Rev 2 August 1998

l 14 . . i i i

- DECAY HEAT GENERATED ' E E dec dec

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1-36 Rev 2 August 1998

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Figure 4. Partitioning of Heat for a Drained Away-From-Reactor Spent Fuel Pool Without Room Ventilation.

1-37 Rev2 August 1998

i i

i i l

1800 i . . . .

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Spent Fuel in Reactor Storage and Away-From-Reactor Storage. ,

l 1

i 1-38 Rev 2 l' August 1998

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August 1998 l

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h THIS PAGE INTENTIONALLY LEFT BLANK

APPENDIX IB Pacific Gas and Electric Company Humboldt Bay Power Plant Unit 3 Criticality Analysis for SAFSTOR Decommissioning i

l 1

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l  !

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l l IB-1 Rev 2 i

( August 1998 i l _ _ _ - _ _ _ _ _ ____ _ _ . _ - _

i l

l CONTENTS l

Section Pace A. Introduction 1 ,

B. Overall Description 2

1. Existing Rack Configuration 2

2. Proposed Modifications 2 C. Material Considerations 3 D. Nuclear Considerations 4

1. Overview 4

2. PGandE Criticality Analysis With CASMO-2E 4

3. General Electric Criticality Analysis with MERIT 12 i

E. Conclusions 14 References 24 Appendix - MERIT Input Listing A-1 l

l l l 1

l l

I I

-I-Rev 2 August 1998

LIST OF FIGURES l l

Title .Pagg o

Humboldt Bay Power Plant Storage Racks 15 i

Fuel Assembly Protective Can (Upper View) 16 i 2

Fuel Assembly Protective Can (Lower View) 17 3

~

4 Fuel Assembly Protective Can'(Cross-section) 18 i

5 Typical Core Loading Diagram (Plan View) 19 i

6 Core ll Loading Diagram - Nine Arrays with Zero Pin 20 f

Pitch Separation

?

7 Core lll Loading Diagram - Nine Arrays Separated by One 21 Pin Pitch l

'8~ Cores XIll, XIV, XV, XVil, and XIX - Nine Unit Assemblies 22 .

Separated by One Pin Pitch and Boral Plates l

9 MERIT Model 23 t

I i

i

- li -

Rev 2 ;

August 1998 ;

l i

~

1

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

LIST OF TABLES Table No. Title Eage 1 Effect of Outer Water Gap on K-infinity 5 2 Effect of Fuel Rod Pitch on K-infinity 5 3 Effect of Poison Inner Dimension and Pitch on K-infinity 6 4 Effect of Boral Thickness on K-infinity 7 5 Effect of Fuel Density on K-infinity 7 6 Optimal Pitch Search for Poison Tube Design 8 ,

7 Characteristics of Critical and Exponential Lattice 9 Experiments and the CASMO-2E Calculated K4ffective 8 Comparison of K-effective for the Eight Cores From Reference 3 11 9 Summary of General Electric High Density Fuel Storage Rack 13 Experience

/

l k

-iii-Rev 2 August 1998

A. INTRODUCTION s

. item 75 of NRC letter dated January 23,1985 requested the following information:

" Discuss the likelihood of a reactivity accident in the spent fuel storage pool due to heavy 4

load drop or seismic event. If sufficient likelihood (_,10 per year) of such events exists, then, assuming step and/or ramp reactivity insertions in the stored spent array due to reduction in undermotion of stored fuel in the pool, in tum due to fuel reconfiguration initiated by a heavy load drop or strong seismic event, calculate offsite radiological consequences assuming:

a) upward spray of all pool water without the presence of the building roof, and b) pool boiling without spray and without the presence of the building roof."

PGandE's response dated February 28,1985 stated the following: "PGandE is actively evaluating design alternatives that would prevent possible criticality due to seismic and heavy load events." This report provides a complete response to item 75.

This report describes the design, fabrication, and safety analysis performed for the addition of neutron-absorbing material in the Humboldt Bay Power Plant (HBPP) Unit 3 spent fuel storage pool. The purpose of the modification is to ensure subcriticality following any event which results in a rearrangement of fuel assemblies from the existing criticality safe storage rack configurations.

l .

t l This modification consists of enclosing each fuel assembly in a can fabricated from a neutron-absorbing material, so that a k-effective greater than 0.95 can not be achieved for any possible fuel configuration.

The criticality analysis associated with this project was prepared by Pacific Gas and Electric Company (PGandE). The goal of this analysis was to find the appropriate boron loading to ensure suberiticality. The General Electric Company performed an independent analysis using their own approved calculational methods and has verified the PGandE results.

l j

l 1

i l

l l

l Rev 2 August 1998 i

4 5 .

~

B. OVERALL DESCRIPTION i 1. Existino Rack Confiouration The HBPP spent fuel storage racks have a total capacity of 486 fuel assemblies. This i

includes 351 central pool locations in 88 groups of 4*, and 135 peripheral pool locations

in 45 groups of 3. The central racks are designed to individually support each fuel l assembly. The peripheral racks support fuel assemblies in groups of three.

The central storage racks ( Figure 1 ) are constructed of aluminum and consist of pairs of storage units approximately 5 feet high and 12 inches square. Each storage unit is able

to hold four fuel assemblies. The peripheral racks are similarly constructed except that they can hold either three fuel assemblies or one full fuel ctorage can.

i The fuel storage racks are welded and/or bolted to cross members of aluminum channels.

l The fuel storage racks are spaced to be " criticality safe."

~

There are currently 390 irradiated fuel assemblies in the HBPP spent fuel storage pool, with exposures ranging from 1,307 to 22,876 MWD /MTU.

l

2. Proposed Modifications '

l 4 '

! In order to preclude criticality in the spent fuel storage pe6l following an event which results in movement or damage to the fuel assembly storage rpcks, each fuel assembly l

! will be enclosed in a can fabricated from a neutron-absorbing material. The can will l j contain an areal density (0.005 gm/cm') of boron (B-10) such that a k-effective greater

. than 0.95 cannot be achieved for any possible configuration.

1 A drawing of the can is shown in Figures 2,3 and 4. The walls of the can will be

fabricated from Boral*" Three bands will be attached at the top, middle, and bottom of the l can to provide structural strength. Additional support may be provided by corner angles, j as necessary, as shown in Figures 2 and 3. A band will be attached to the bottom of the .

can to prevent the fuel assembly from coming out of the bottom. The top band will be l l fabricated with locking tabs which will be bent over to prevent inadvertent removal of the l j fuel assembly from the can. This design will ensure that the poisoned material is an l

integral part of the fuel assembly.

i F

• (However, pool location 64-07 cannot os used due to a bolt protruding into the bottom of this location and inadvertent use of this iocation is prevented by a triangular plate welded over the top.)

Rev 2 August 1998

_ _ . _ __ - - _ _ ~ . . _ _ _ _ _ _ _ _ _ _ _ _ _

1 C. MATERIAL CONSIDERATIONS Most of the material used in fabrication of the fuel bundle enclosure can is Boral, which is ,

a thermal neutron poison material composed of boron carbide and 1100-alloy aluminum.  !

Boron carbide is a compound having a high boron content in a physically stable and l chemically inert form. The 1100 alloy aluminum is a lightweight metal with high tensile strength which is protected from corrosion by a highly resistant oxide film. The boron l

carbide and aluminum are chemically compatible and suited for long-term use in the radiation, thermal, and chemical environment of the HBPP spent fuel storage pool.

i The Boral is provided in flat sheets and is formed to enclose the full length of each of the four sides of each individual fuel assembly. Physical integrity of the poisoned can is maintained by use of type 304 stainless steel bands which are attached to the Boral with aluminum rivets and encircle the can at the bottom, the approximate center, and the top.

The materials contained in the Boral, as well as the stainless steel, are compatible with all parts of the spent fuel sforage system, including the fuel assemblies, the cooling system, the cleanup system, the pool liner, and the storage racks. The useful life of the Boral will  ;

exceed 40 years when in contact with the storage pool water. The corrosion resistance of Boral is provided by the protective film on the aluminum cladding that is an integral part of the Boral panels. Testing performed by the Boral supplier confirms that the effects are ,

negligible from general corrosion, galvanic corrosion of the Boral/ stainless steel interface,

• pitting corrosion, stress corrosion, and intergranular corrosion.

Boral is manufactured under the control and surveillance of a computer-aided quality assurance / quality control program that conforms to the requirements of 10 CFR 50, Appendix B, entitled " Quality Assurance Criteria for Nuclear Power Plants." l Boral has been licensed by the USNRC for use in BWR and PWR spent fuel storage i

racks, and is also used around the world for spent fuel shipping and storage containers.

l l-Rev 2 August 1998 l

! l D. NUCLEAR CONSIDERATIONS j l

i 1- Overview i l

The criticality analysis for these proposed modifications was performed by PGandE using  ;

the CASMO-2E (Ref.1) computer code. These calculations were performed using a i conservative set of assumptions and resulted in a maximum k-infinity of 0.894. i

- An independent analysis was then performed by General Electric using their MERIT code. [

The results of that evaluation indicated a maximum k-infinity of 0.884. j The detai!s, assumptions, and code inputs for each of the analyses are described in the ,

following sections. j i

2 PGandE Criticality Analysis With CASMO-2E l CASMO-2E is a multi-group, two-dimensional (2-D) transport theory, fuel assembly analysis code. It was used to design a B-10 loading for poison cans to be attached to the .

fuel assemblies in the HBPP Unit 3 spent fuel storage pool. A 25-energy group library, j supplied with CASMO and based on ENDF/B-Ill cross-sections, was used. A worst case i

, analysis was performed to bound all possible fuel assembly rearrangements by analyzing an infinite array of the most reactive fuel assembly in its most reactive configuration. The effects of moderation between assemblies and within assemblies were analyzed to obtain -

the most reactive geometry. Additionally, effects of uncertainties in fuel density and l poison can design were analyzed and inciuded in a conservative manner. The following I conservative assumptions were made: l

1. All fuel was assumed to have the highest as-built enrichment (2.52% U-235) and contain the greatest U-235 mass (GE Type lll-4).

2. All fuel assemblies were at beginning of life (BOL), cold and clean, and l contained no gadolinia (no credit for exposure, fission products, or burnable poisons ).

3. No credit was taken for neutron absorption in the materials of the fuel i storage racks, the fuel channel, or the aluminum outside of the B4C

! containing core of the Boral.

4. The 2-D transport calculation assumed an infinite array of infinitely tall fuel L assemblies, thus bounding all geometries (no credit for radial or axial  !

leakage).

l

5. Optimal moderation was imposed by varying the gap between assemblies,

the inner dimension of the poison can, and the fuel rod pitch within the poison can.

i Rev 2 August 1998

a l

' Achieving Optimal Moderation a.

The effect of fuel assembly separation was investigated by analyzing several gap thicknesses between assemblies with as-built lattice 1

- dimensions at several B-10 loadings. These results, shown in Table 1, indicate the most reactive situation to be the zero separation case. This is due to the fact that water outside the poison cans serves as a flux trap. j Table 1 EFFECT OF OUTER WATER GAP ON K-INFINITY L

l (Model - As-Built Lattice, Poison Can 60 Mils Thick, inner Dimension = 4.54 inches) i 4 l Outer Water Gap

- B-10 Loading (gm/cm ) i (cm) 0.003 0.005 0.010 0.0 0.92992 0.86875 0.79089 0.5 0.86889 0.80850 0.73443 l

'1.0 0.81790 0.75703 0.68492 2.0 0.72857 0.66708 0.59912 i Using the results from Table 1, a B-10 loading of 0.005 gm/cm was .

chosen for further investigation. The fuel rod pitch was perturbed to test  ;

the effects of moderation within the poison cans. Table 2 results show the i assembly to be undermoderated within the poison cans as increasing pitch increases k-infinity.

Table 2 1 I

EFFECT OF FUEL ROD PITCH ON K-INFINITY 2

(Model - 0.005 9 gm B-10/cm , Zero Outer Water Gap, Poison Can 60 Mils Thick, inner Dimension = 4.54 inches)

Pitch K-infinity )

102% as-built pitch 0.87237 100% as-built pitch 0.86875 98% as-built pitch 0.79619 l

Rev 2 I August 1998

The maximum pitch possible is determined by the inner dimension of the poison can. The poison can must fit within a fuel rack storage cell so the 5.125-inch inner dimension of the i largest cell serves as an absolute upper bound on the poison can design. The pitch was l

. varied within the poison can area to find the highest k-infinity. The results'of this optimal pitch search are shown in Table 3 as well as results for two other poison can dimension cases. These results indicate a maximum k-infinity due to moderation occurs with a maximum poison inner dimension of 5.125 inches and a pitch of 2.1168 cm (98% of the maximum pitch possible for this case).

Table 3 EFFECT OF POISON INNER DIMENSION AND PITCH ON K-INFINITY  ;

(Model - 0.005 gm B-107cm', Zero Outer Water Gap, Poison Can 60 Mils Thick) 1 Poison Inner Dimension 4.8935 in. 5.039 in. 5.125 in.

Maximum Pitch (2.06 cm) (2.125 cm) (2.16 cm)

% of Maximum Pitch 100 0.88908 0.89097- 0.89238 99 0.89149 0.89575 0.89771 98 0.89056 0.89695 0.89943 97 0.88557 0.89292 0.89580 95 0.87558 0.88379 0.88702 90 0.84465 0.85513 0.85939 83 0.80635 79 0.75818 74 0.70369 69 0.64559 L

The Boral poison was modeled as being 60 mils thick with the mass density l

. needed to obtain an areal density of 0.005 gm B-10/cm'. Using a mass density i typical of Boral manufacturing, a thickness of 11 mils was necessary to reach the i same areal density. This case was explicitly modeled at the previously determined j optimal moderation conditions to account for a lack of conservatism in the model due to Boral thickness. The results are shown in Table 4.

i Rev 2 August 1998 l

Table 4 EFFECT OF BORAL THICKNESS ON K-INFINITY 2 <

[Model - 0.005 gm B-10/cm , Zero Outer Water Gap, Poison Can Inner Dimension l

= 5.125 inches, Pitch = 2.1168 cm (optimal)]

l Boral Thickness K-Infinity l

l 60 mils 0.89943 i

11 mils 0.90217 l

1 The HBPP Unit 3 fuel has a nominal density of 10.3 gm/cc with an upper bound of i

10.5 gm/cc. The final consideration of the worst case analysis was to model the extreme fuel density in the optimal moderation. The maximum k-infinity was found to be 0.90624. Table 5 illustrates the magnitude of this effect.

l l Table 5 l

I EFFECT OF FUEL DENSITY ON K-INFINITY

[Model - 0.005 gm B-10/cm', Zero Outer Water Gap, Poison Can 11 Mils Thick, ,

! Inner Dimension = 5.125 inches, Pitch = 2.1168 cm (optimal)]

Fuel Density (om/cc) K-Infinity

( 10.3 0.90217 f 10.5 0.90624

b. Analysis of Final Design Additional analyses were performed to model the actual dimensions of the poison can as designed. The design for the poison can specifies an outer dimensica of 5.0 inches and a total Boral thickness of 100 mils. The tube material will consist of roughly 16 mils of a mixture of 35 weight percent B4C and 65 weight percent aluminum, sandwiched between two aluminum sheets 42 mils thick. The CASMO-2E model neglects the sandwiching aluminum and conserves the inner dimension i of the poison tube design. Results of the optimal pitch search are shown in Table

6. Using optimal moderation and the extreme fuel density results in a maximum k-infinity of 0.894 for the design.

Rev 2 -

August 1998

Table 6 OPTIMAL PITCH SEARCH FOR POISON TUBE DESIGN (Model - 0.005 gm B-10/cm', Zero Outer Water Gap, Poison Can 16 Mils Thick, Inner Dimension = 4.8 inches)

?

Fuel Density (gm/cm*)

Pitch 10.3 10.5 2.022 0.88823 -

2.00178 0.89006 0.89400 l 1.995 -

0.89412 1.99167 0.89018 0.89407 1.98156 0.88850 -

U

c. PGandE Benchmark of CASMO-2E The CASMO-2E prediction of k-effective was tested against 61 experiments using the 25-group production cross-section library. These experiments are uniform cold critical or exponential water-moderated UO2 lattices reported by Strawbridge and  :

Barry (Ref. 2) and byj Price (Ref. 3). Table 7 lists these experiments by case i number as presented in Reference 2 and by page number as presented in Reference 3. All these cases are UO2 fuel pins with enrichment ranging from 1.3 to 4.0 w/o U-235, an H20

U ratio of from 2.10 to 9.3, and natural boron concentration from zero to 3396 ppm. The mean k-effective value for 59 independent experiments (cases 18 and 19 and pages 169 and 170 of References 2 and 3, ,

respectively, are repeated measurements on identical lattices) is 0.9981. The i standard deviation is 0.0100.

An analysis of eight critical cores of close proximity water-moderated fuel storage experiments (Ref. 4) was conducted with the 25-group production cross-section library. These cores are composed of nine assemblies of 14 by 14 fuel pins each with boron / aluminum separation sheets between them, and borated water as 4 moderator. (See Figures 5-8). The CASMO-2E/PDQ evaluation of k-effective for these eight cores is presented in Table 8. The B-10 loading and two sets of KENO results (for comparison) are also listed. The calculated k-effective values have a mean of 1.0014 and a standard deviation of 0.0030.  !

d. Previous Use of CASMO to Support Licensing Activities Yankee Atomic Electric Company and Northern States Power are currently j performing reload licensing using NRC-approved (Refs. 6, 7) CASMO-based  !

physics methods (Refs. 8,9).

l Duke Power Company has submitted a partially CASMO-based physics method I topical (Ref.10) for NRC review.

Rev 2 August 1998

TABLE 7 Shnet 1 of 2 l

Characteristics of Critical and Exponential Lattice Experiments and the CASMO-2E Calculated K-effective  ;

Boron CASMO-

• Case H2 0:U Fuel Pellet Clad concen- Lattice Critical 2E >

Number Buckling k-Clad Clad Thickness tration Pitch orPaga Endchment Volume Density Diameter cm ppm cm m2 effective

' Number weight % Ratio 0/cm em Material OD cm  ;

1 1.311 3.02 7.53 1.5265 Al 1.6916 0.0711 0.00 2.205' 28.37 0.99438 2 1.311 3.95 7.53 1.5265 Al 1.6916 0.0711 0.00 2.359' 30.17 0.99765 3 1.311 4.95 7.53 1.5265 Al 1.6916 0.0711 0.00 2.512' 29.06 0.99748 7.52 0.9855 Al 1.1506 0.0711 0.00 1.558' 25.28 0.99379 i 4 1.311 3.93 5 1.311 4.89 7.52 0.9855 Al 1.1506 0.0711 0.00 1.652' 25.21 0.99340 6 1.311 2.88 10.53 0.9728 Al 1.1506 0.0711 0.00 1.558' 32.59 0.99776 3.58 10.53 0.9728 Al 1.1506 0.0711 0.00 1.652' 35.47 0.99757 l 7 1.311 8 1.311 4.83 10.53 0.9728 Al 1.1506 0.0711 0.00 1.806* 34.22 0.99741 SS-304 0.8594 0.04085 0.00 1.0287 40.75 1.00394 i 9 2.700 2.18 10.18 - 0.7620 10 2.700 2.93 10.18 0.7620 SS-304 0.8594 0.04085 0.00 1.1049 53.23 1.00421 3.86 10.18 0.7620 SS-304 0.8594 0.04085 0.00 1.1938 63.26 1.00199 11 2.700 ,

12' 2.700 7.02 10.18 0.7620 SS-304 0.8594 0.04085 0.00 1.4554 65.64 1.00909 0.00 1.5621 60.07 1.01249 I 13 2.700 8.49 10.18 0.7620 SS-304 0.8594 0.04085 l

10.38 10.18 0.7620 SS-304 0.8594 0.04085 0.00 1.6891 52.92 1.01015 i 14 2.700

• 2.50 10.18 0.7620 SS-304 0.8594 0.04085 0.00 1.0617 47.5 1.00173

15 2.700 l 2.700 4.51 10.18 0.7620 SS-304 0.8594 0.04085 0.00 1.2522 68.8 0.99663 16 17 3.699 2.50 10.37 0.7544 SS-304 0.8600 0.0406 0.00 1.0617 68.3 1.00657 i i

4.51 10.37 0.7544 SS-304 0.8600 0.0406 0.00 1.2522 95.1 1.00473 l 18 3.699 3.699 4.51 10.37 0.7544 SS-304 0.8600 0.0406 0.00 1.2522 95.68* 1.00318 i 19 0.7544 SS-304 0.8600 0.0406 456.1 1.2522 74.64* 0.99873 i

'20 3.699 4.51 10.37 '

3.699 4.51 10.37 0.7544 SS-304 0.8600 0.0406 709.1 1.2522 63.66" 0.99698 21 22 3.699 4.51 10.37 0.7544 SS-304 0.8600 0.0406 1261.4 1.2522 40.99* 0.99544 3.699 4.51 10.37 0.7544 SS-304 0.8600 0.0406 1332.7 1.2522 38.39" 0.99485 i l 23 24 3.699 4.51 10.37 0.7544 SS-304 0.8600 0.0406 1475.2 1.2522 33.38" 0.99349 25 4.020 2.55 9.46 1.1278 SS-304 1.2090 0.0406 0.00 1.5113 88.0 0.99674 26 4.020 2.55 ~ 9.46 1.1278 SS-304 1.2090 0.0406 3396.3 1.5113 17.2 1.00019 ,

, .34 4.020 2.14 9.46 1.1278 SS-304 1.2090 0.0406 0.00 1.450 79.0 0.99208

! 37 2.460 2.84 10.24 1.0297 Al 1.2060 0.0813 0.00 1.5113 70.10 1.01783 42 3.000

• 2.64 9.28 1.1268 SS-304 1.2701 0.07163 0.00 1.555 50.75 0.99233

'43 3.000 8.16 9.28 1.1268 SS-304 1.2701 0.07163 0.00 2.198 68.81 0.98588 ,

44. 4.020 2.59 9.45 1.1268 SS-304 1.2701 0.07163 0.00 1.555 69.25 1.00209 l 45- 4.020 3.53 9.45 1.1268 SS-304 1.2701 0.07163 0.00 1.684 85.52 0.99700 46 4.020 8.02 9.45 1.1268 SS-304 1 2701 0 07163 0.00 2.198 92.84 1.01207 [

l 47 4.020 9.90 9.45 1.1268 SS-304 1 2701 0 07163 0.00 2.381 91.79 1.00051 50 2.460 2.84 10.24 1.0297 Al 1.2060 0.0813 1677.2 1.5113 20.2 1.00323 51 ' 2.070 2.06 10.38 1.524 Al 1.6916 0.07112 . 0.00 2.1737 58.0 1.05321 52 2.070 3.09 10.38 1.524 Al 1.6916 0.07112 0.00 2.4052 80.6 1.00749 53 2.070 4.12 10.38 1.524 Al 1.6916 0.07112 0.00 2.6162 85.7 0.99453 54 2.070- 6.14 10.38 1.524 Al 1.6916 0.07112 0.00 2.9891 77.0 0.98924 l 55 2.070 8.20 10.38 1.524 Al 1.6916 0.07112 0.00 3.3255 61.6 0.98467 3

" Hexagonal lattices; all others are square.  :

"These bucklin9s were not measured directly but were inferred from critical loadings.

1.

i Rev 2 >

(.

l Sheet 2 of 2 TABLE 7 (cont'd) {

Characteristics of Critical and Exponential Lattice Experiments I and the CASMO-2E Calculated K-effective  ;

! i

Case Boron CASMO-l Number H2 0

U Fuel Pellet Clad concen- Lattice Critical 2E i orPage Enrichment Volume Density Diameter Clad Clad Thickness tration Pitch cm Buckling k- ,

Numoer weight % Ratio g/cm cm Material OD em cm opm m2 ,9y,cggy, !

165 3.006 2.990' 9.299 1.128014 SS-304 1.26746 0.0696722 0.0 1.718818' 56.6 0.99154 j 166 3.006 2.990' 9.299 1.128014 S S-304. 1.26746 0.0696722 670.3 1.718818' 36.71 0.98999  ;

167 3 006 2.990' 9.299 1.128014 SS-304 1.26746 0.0696722 1336.5 1.718818' 18.26 0.98908 168 3 006 3.700* 9.299 1.128014 SS-304 1.26746 0.0696722 0.0 1.819402' 65.81 0.98637 169 3.006 3.700* 9.299 1.128014 SS-304 1.26746 0.0696722 471 2 1.819402" 46.41 0.98667 ,

170 3 006 3.700* 9 299 1.128014 SS-304 1.26746 0.0696722 471.2 1.819402' 45.00 0.99109 171 3 006 3.700* 9 299 1.128014 SS-304 1.26746 0.0696722 995.2 1.819402' 26.20 0.98991 172 3.006 3.700* 9.299 1.128014 SS-304 1.26746 0.0696722 1349.0 1.819402' 14.62 0.98925 173 3 006 4,740* 9.299 1.128014 SS-304 1.26746 0.0696722 0.0 1.957324' 70.49 0.99029 l 174 3 006 4.740' 9.299 1.128014 SS-304 1.26746 0.0696722 431.0 1.957324' - 46.34 0.99107 175 3.006 4.740' 9.299 1.128014 SS-304 -1.26746 0.0696722 806.0 1.957324* 27.70 0.99142 L 176 3 006 4.740' 9.299 1.128014 SS-304 1.26746 0.0696722 1144.0 1.957324* 12.94 0.99019 177 3 006 4.740' 9.299 1.128014 SS-304 1.26746 0.0696722 0.0 2.169668' 70.22 0.99598 l 178 3.006 6.490' 9.299 1.128014 SS-304 1.26746 0.0696722 289.1 2.169668' 47.61 0.99489 l 179 3.006 6.490* 9.299 1.128014 SS-304 1.26746 0.0696722 604.3 2.169668' 25.22 0.99502  ;

180 3.006 6.490' 9.299 1.128014 SS-304 1.26746 0.0696722 772.7 2.169668' 15.05 0.99277 181 3.006 9.229' 9.299 1.128014 SS-304 1.26746 0.0696722 0.0 2.465578' 61.73 0.99834 l 182 3 006 9.229' 9.299 1.128014 SS-304 1.26746 .0.0696722 173.0 2.46557P 41.18 1.00086 l 183 3 006 9.229' 9.299 1.128014 SS-304 1.26746 0.0696722 260.5 2.465578*' 32.41 0.99961 i 184. 3.006 9.229' 9.299 1.128014 SS-304 1.26746 0.0696722 390.9 2.465578* 20.51 0.99633

! 185 3.006 9.229' 9.299 1.128014 SS-304 1.26746 0.0696722 540.5 2.465578' 6.04 0.99796 "Hexagonallattices; all others are square.

~

These bucklings were not measured directly but were inferred from critical loadings.

" Recalculated by PGandE to agree with definition given in Reference [2].

1 Rev 2 August 1998

TABLE 8 Comparison of k-effective for the 8 Cores From Reference [4]

Boron Loading.

B-10 Density in Boral(* Sheets.

Core grams /cm2 Number B&W KENO (# B&W " Measured

• N.S&E KENO *) PGandE 1.0001 i.0005 .995 *.004 1.0039 0 ll . 1.007 i.004 1.0000 i,0006 1.009 i.004- 1.0054 0 lli .999 i.004 4

1.008 *.005 1.0000 i.0001 1.008

• 006(d 1.0034 5.582 x 10 Xill 1.011 i.006

, 1.003 i.005 4

1.0012 5.803 x 10 XIlla 4

XIV 1.003 i.004 1.0001 *.0001 .999 i.004 1.0001 4.348 x 10

.997 *.004 4

XV .995 i.005 .9998 i.0016 .996 i.005 0.9956 1.387 x 10 4

XVil .993*.005 1.0000 i.0010 .997 i.004 .9997 0.837 x 10 4

l XIX .991 i.004 1.0002 *.0010 .995 i.003 1.0021 0.346 x 10 (dReference 4, Tables IX and XI.

• ) Reference 5, Table Ill.

(" Cases Xill and Xilla are " combined" by Reference 5. The soluble boron concentration in these cases are different; 15 and 18 ppm.

(4B-10 is 19.8 a/o of Boron.

Rev 2 August 1998 l

3. General Ehetric Criticality Analysis With MERIT MERIT is a Monte Carlo program which solves the neutron transport equation as an eigenvalue or a fixed source problem. This program was written for the analysis of fuel lattices in thermal nuclear r: actors. A geometry of up to three space dimensions and neutron energies between 0 and 10 MeV can be handled. MERIT uses cross-sections processed from the ENDF/B-IV library tapes.

A check was made of the results of the PGandE optimum moderation configuration.

The fo!!owing assumptions and input values were used in this analysis:

1. 2.52% enriched fuel (fuel density 10.5 gm/ce, reduced to 10.0422 gm/cc to include gap) ,

2. Fuel pellet radius - 0.63373 cm

3. Zirconium 2 clad - outer radius 0.71501 cm I

4. Rod pitch - 1.995 cm l 1

5. Square poison can (outside dimension 12.7 cm) on each bundle

6. Channel thickness- 0.253 cm (0.10668 cm Al,0.03965 cm Boral core,0.10688 cm Al)

7. Boral core 35 w/o boron carbide,65 w/o aluminum

8. B-10 areal density - 0.005 gm/cm' ,

1

9. Infinite array of fuel bundles of infinite length )

1

10. Water density - 1.0 grn/cm' l

The MERIT case was run for 35,000 neutron histories and predicted a k-infinity of 0.878767 i 0.00313 (1 a). The MERIT code has been benchmarked with numerous criticality experiments and his been shown to underpredict k by 0.005 i 0.002 (1 c). Thus, the MERIT-predicted lattice k-infinity for the 5-inch poison can with all uncertainties added would be 0.883767 t 0.00371 (1 o ).

This is a very conservative upper limit for this case since it assumes the maximum fuel density, the minimum thickness Boral core in the can wall, and that the fuel pins in all cans can expand to the optimum pitch even though they are held in the fuel bundle design pitch by the upper and lower tie pictes and the fuel spacers.

A sketch of the MERIT model is shown in Figure 9. A copy of the input file for MERIT is given in the Appendix.

j a. MERIT Benchmarking i

The qualification of the MERIT program rests upon extensive qualification studies including Cross Section Evaluation Work Group (CSEWG) thermal reactor benchmarks (TRX-1, -2, -3, and -4) and Babcock and Wilcox (B&W) UO2 and PuO2 criticals, Jersey Central experiments, CSEWG fast reactor benchmarks (GODIVA, JEZEBEL), the KRITZ experiments, and comparison with attemate calculational methods. Boron was used as solute in the moderator I

Rev 2 August 1998 l

in the B&W UO2 criticals, and cs a solid control curt:in in the J:rsey Central experiments.

The MERIT qurlification progrcm h:s estrblished a bits of 0.005 i O.002 (1 a) A k with respect to the above critical experiments. Therefore, MERIT underpredicts k-effective by approximately 0.5 percent A k.

b. Previous Use of MERIT to Support Licensing Activities MERIT has been used to license Boral-poisoned high density fuel storage racks at several reactor sites and/has been reviewed and checked by the NRC and found to be acceptable.

These sites are listed in Table 9. ,

Table 9

SUMMARY

OF GENERAL ELECTRIC HIGH DENSITY FUEL STORAGE RACK EXPERIENCE i

Plant Scope of Work Status Monticello 13 rack _s, storage capacity Licensed and in use since 2,237 spaces April 1978 Browns Ferry 57 racks, storage capacity Licensed and in use since 1,2, and 3 10,413 spaces Sept.1978 Hatch 30 racks, storage capacity Licensed and in use since 1 and 2 6,026 spaces April 1980 Brunswick 10 racks, storage capacity Licensed and in use i 1 and 2 3,642 December 1983 ,

Hartsville 60 racks, storage capacity Approved for installation A1,A2,B1,B2 11,804 spaces (Plant canceled) through GESAR ll FDA July 1983 Phipps Bend 30 racks, storage capacity Approved for installation 1 and 2 5,902 spaces (Plant cancelled) through GESAR 11 FDA July 1983 Kuosheng 6 racks, storage capacity Scheduled for 1985 1 and 2 1,326 spaces installation t

Rev 2 August 1998

E. CONCLUSIONS As demonstrated in the proceeding analyses, the proposed Boral cans will provide a neutron-ebsorbing material as an integral part of the HBPP fuel assemblies, and will ensure that k-effective will be less than 0.95 for the worst possible rearrangement of fuel. This analysis was done using conservative assumptions and was independently checked by General Electric.

l l

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August 1998

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1 I Rev 2 August 1998

REFERENCES I

1. M. Edenius, A. Ahlin, H. Haggblom, "CASMO-2 A Fuel Assembly Burnup Program Users Manual," Studsvik/NR-81/3 with Revision 1984-09-01.

2. L. E. Strawbridge and R. F. Barry, " Criticality Calculations for Uniform Water - l Moderated Lattices", Nuclear Science and Engineering,21, pp. 58-73 (1965).

3. Glenn A. Price, " Uranium - Water Lattice Compilation Part 1, BNL Exponential Assemblies," BNL-50035, December 30,1966.

4. Hoovier et al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," Nuclear Technology,51, pp. 217-237 December 1980. ,

l

5. S. E. Turner and M. K. Gurley, " Evaluation of AMPX-KENO Benchmark Calculations for l High-Density Spent Fuel Storage Racks," Nuclear Science and Engineering, SQ, pp. j 230-237 (1982).

l

6. Letter from D. B. Vassallo (NRC) to J. B. Sinclair (YAEC), File NVY-82-157, Docket

1. 50-271, September 15,1982.

7. Letter from R. A. Clark (NRC) to B. M. Musoif (NSP), February 17,1983.

8. E. E. Pilat, " Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980.

9. " Qualification of Reactor Physics Methods for Application to Prairie Island Units,"

NSPNAN-8101, December 1982.

10. Duke Power Company, " Nuclear Physics Methodology for Reload Design," DPL-NF-2010, April 1984.

l Rev 2 August 1998

APPENDIX MERIT INPUT LISTING 0436S/0011K A-1 Rev2 ,

August 1998

r MERIT 01C -

MARCH 1,1979 PG&E Safe Store MERIT (Input File MERIT 03) 5.0 inch Poison Can 0.005 gm B-10/cm2 Optimum Fuel Rod Pitch 1.9950 cm Fuel Density 10.5 gm/cc i

l i

A-2 Rev 2 August 1998 .

19 JULY 1985 PAGE1 PG&E SAFSTOR MERIT l

r CARD 2 NPRBTP 1234 IDENT. NO. OF PROBLEM TAPE LSTRT- 0 0 = INITIAL START, IMCT=B l 1 = RESTART, NO CHANGES, IMCT=A l (IF LCOPY = 1,IMCT=B) 2 = RESTART BUT DO INPUT CALCULATIONS, IMCT=B i

3 = RESTART BUT NO INPUT CALCULATIONS EXCEPT FOR MATERIAL INFORMATION, IMCT=B  !

LCONT 0 0 OR 1 = GO ON TO MONTE CARLO AFTER BMCIN. '

2 = GO ON TO BMCOUT AFTER BMCIN.

LSTOP O O = DO COMPLETE PROBLEM. ,

1 = DO INPUT ONLY. [

2 = DO INPUT AND MONTE CARLO ONLY. ,

LCOPY 0 1 = COPY TAPE A TO TAPE B AND USE B.

(ACTIVE ONLY IF LSTRT = 1.)

O = FOR NORMAL POSITIONING TO THE LAST RESTART CASE. [

LSKNS' O 1 = START THE TALLIES ANEW. [

N = USE THE NTH BATCH FOR RESTART. i LSKSV O MUST = 77 IF LSKNS DOES NOT = 0 OR LSKNS WILL NOT BE PERFORMED.

i ICXTSV O O = RESTART TAPE WILL BE SAVED. -

66 = RESTART TAPE WILL NOT BE SAVED.

i IMPARD 0 55 = READ NEW SET OF WEIGHTING PARAMETERS WITH NORMAL RESTART.

i IDUMP 1 1 = DUMP MONTE CARLO BLANK COMMON.

I NERCV 10 NO. OF RECOVERABLE ERRORS BEFORE TERMINATING.

(IF EQUALTO O NERCV WILL BE SET TO 10) [

NTSPL 1 NUMBER OF GEOMETRY TEST PLOTS.

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A-3 Rev 2 August 1998 i

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

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PG&E SAFSTOR MERIT 19 JULY 1985 PAGE 2 CARD 3 NBTCH 35 NUMBER OF BATCHES. '

NPTPB 1000 NUMBER OF PARTICLES IN EACH BATCH.

RTHRT 0. THE RATIO OF WElGHT LEAVING THE THERMAL TALLY RANGE TO THAT OF  !

ENTERING. (IF 0.0, ALL PARTICLES FOLLOWED. IF 1.0, NONE FOLLOWED.)

ETHRT 0. THE ENERGY TO WHICH NEUTRONS MUST BE SLOWED DOWN BEFORE THEY ARE ["

USED FOR THE THERMALTALLY RANGE.

ETHRTX 0. THE MAXIMUM ENERGY NEUTRONS MAY ACHIEVE WHILE CONTRIBUTION TO THE THERMAL FLUX TALLY. [

CARD 4 LPRB 0 0 = FISSION DESCENDENT CALCUI.ATION. i 1 = DIRECT SOURCE CALCULATION.

MOOTH 293 MODERATOR TEMPERATURE IN DEGREES KELVIN.

(USED TO INTERPOLATE THERMAL HYDROGEN SCATTERING KERNEL)

NBGPX 0 NUMBER OF BROAD ENERGY GROUPS. ,

(IF ONLY ONE, USE NPGPX = 0.) l JMAXX 4 NUMBER OF MACRO ENERGY GROUPS.  !

NHMAX 4 NUMBER OF REGIONS.

MMAX 5 NUMBER OF MATERIALS NSPRG 1 NUMBER OF SPECIAL REGIONS. [

NFMAX 5 NUMBER OF TALLY REGIONS.  ;

NLKTLY 0 NUMBER OF LEAKAGE TALLY SETS. i LTALY 0 NUMBER OF SETS TO BE TALLIED.  ;

NBNDX 6 NUMBER OF BOUNDARIES.

NOMGX 0 NUMBER OF ALBEDO SETS.

CARD 4A NRX,NRY 3, 3 MAX NUMBER OF RODS IN X AND Y IN LATTICES I NWTZX 0 NUMBER OF WElGHTING ZONES i NERGX 0 MAX NUMBER OF ENERGY WlEGHTING RANGES NBR 15 SUM OF BOUNDARIES FOR REGION SPECIFICATION j

~

MINIMOE5ARG OF E$CH MACRIfl5IFI6 ~dioliPI~~~~~~ I GROUP ENERGY GROUP ENERGY 1 1.000000E+06 3 6.250000E-01  !

2 5.530800E+03 4 0. l A-4  ;

Rev 2 l August 1998  ;

PG&E SAFSTOR MERIT 19 JULY 1985 PAGE 3 THE FOLLOWING ISOTOPES HAVE BEEN LOADED FROM CCT TAPE NO.1234 ISOTOPE ENDF/B SIGMA ID. NO. NAME ID. NO. DATED POTENTIAL 1 H1 1269 21 JUNE 1976 2.0447E+01 10 B10 1273 05 MAY 1976 2.1060E+00 11 B11 1160 05 MAY 1976 5.0350E+00 12 C12 1274 05 MAY 1976 4.7300E+00 16 016 1276 04 MAY 1976 3.7040E+0b 131 AL 1193 23 APR 1976 1.3480E+00 401 ZlRC-2 1284 23 APR 1976 6.1580E+00' 2351 U235 1261 21 APR 1976 1.1500E+01 2381 U238 1262 12 APR 1976 1.05995+01

_ _ _:2__ __ _::: _

MATERIAL DESCRIPTION MATTERIAL NO. 1 FUEL ONE NO. ISOTOPES 1, TEMP. 293, FIS SPEC. 1, RF 6.3373E-01, DG 1.0000E+00, HEAW AWT.236.0 ISOTOPE CONCENTRATION SIG M EFF. ETHRM IH W M LTHRMM LANMM LINMM 2351 5.719000E-04 6.9430E+02 0. 0 0 0 0 2381 2.183800E-02 7.8847E+00 0. 0 0 0 0 16 4.471100E-02 0. 1.275E+00 0 1 0 0 MATTERIAL NO. 2 ZlRC NO. ISOTOPES 3, TEMP. 293, FIS SPEC. 1, RF 0. DG 0. HEAW AWT. 236.0 l ISOTOPE CONCENTRATION SIG M EFF. ETHRM IH W M LTHRMM LANMM LINMM 401 4.333300E-02 0. O. 0 0 0 0 A-5 Rev2 August 1998 i

i PG&E SAFSTOR MERIT. 19 JULY 1985 PAGE 4 I MATTERIAL NO. 3 MODERATOR NO. ISOTOPES 2, TEMP. 293, FIS SPEC. 1,RF 0. DG 0. HEAW AWT. 236.0 ISOTOPE CONCENTRATION SIG M EFF. ETHRM IH W M LTHRMM LANMM LINMM 16 3.344400E-02 0. 1.275E+00 0 1 0 0 1 6.688800E-02 0. 2.102E+00 0 2 0 0 MATTERIAL NO. 4 BORAL NO. ISOTOPES 4, TEMP. 293, FIS SPEC. 1, RF 0. DG 0. HEAW AWT. 236.0 -

e ,

ISOTOPE CONCENTRATION SIG M EFF. ETHRM IH W M LTHRMM LANMM LINMM 12 9.580320E-03 0. 2.000E+00 0 1 0 0 11 3.071440E-02 0. 2000E+00 0 1 0 0 '

10 7.587600E-03 0. 2.000E+00 0 1 0 0 131 3.637750E-02 0. O. 0 0 0 0 MATTERIAL NO. 5 ALUMINUM NO. ISOTOPES 1, TEMP. 293, FIS SPEC. 1, RF 0. DG 0. HEAW AWT. 236.0 i

ISOTOPE CONCENTRATION SIG M EFF. ETHRM IHWM LTHRMM LANMM LINMM 131 6.026610E-02 0. O. 0 0 0 0 l

i A-6 Rev2 i August 1998

19 JULY 1985 PAGE5 PG&E SAFSTOR MERIT GEOMETRY INPUT DIMENSIONS ARE IN CM.

BOUNDARY DATA BOUNDARY TYPE ALBEDO DESCRIPTION 3 PLANE, Y = 0. + 1.000000E+00 . X i 4 2 1 3 PLANE, X = 6.350000E+00 3 2 3 Pt.ANE, Y = 0. ,

4 2 0 PLANE, Y = 1.06600E-01 5 2 0 PLANE, Y = 1.463300E-01 -- . ..

6 2 0 PLANE, Y = 2.530100E-01 REGION DATA REGION BOUNDARY SPECIAL BOUNDARY MATERIAL TALLY SET G.D BH C1 Q2 3 1 3 3 1 1 1 0 1

1 2 1 0

-1 6 4 0 4 0 5 5 1 1 2 0 2

1 4 3 0 1 2 2 0

-1 3 2 0 4 0 4 4 1 1 3 0 3

1 5 4 0 1 2 3 0

-1 4 2 0 4 4 0 5 5 1 1 4 0 1 6 1 0 1 2 4 0

-1 5 3 0 A-7 Rev2 August 1998

PG&E SAFSTOR MERIT -------- -

19 JULY 1985 . PAGE 6 SPECIAL INPUT FOR A SQUARE LATTICE FOR REGION 1 LTSPS 1 TYPE OF SPECIAL REGION (1 = SQUARE LATTICE OF CIAD FUEL RODS)

NRX 3 NUMBER OF ROWS OF RODS IN THE X DIRECTION NRY 3 NUMBER OF ROWS OF RODS IN THE Y DIRECTION MATCL 2 MATERIAL NUMBER OF CLADDING IFTSL 2 TALLY REGION OF CLADDING MATFL 1 MATERIAL NUMBER OF FUEL IFTFL 1 TALLY REGION OF FUEL XC 1.3615 X COORDINATE OF LOWER LEFT CORNER ,

YC 1.3615 Y COORDINATE OF LOWER LEFT CORNER * ,

DXC 1.9950 LATTICE SPACING IN THE X DIRECTION DYC 1.9950 LATTICE SPACING IN THE Y DIRECTION RDF .6337 RADIUS OF THE FUEL ROD 1 RDC .7150 RADIUS OF THE OUTER EDGE OF CLADDING MATXYS MATERIAL NUMBERS OF ALL THE RODS IN REGION 1 Y/X 1 2 3 3 1 1 1 2 1 1 1 1 1 1 1 6 IFTXYS TALLY REGIONS OF ALL THE RODS IN REGION Y/X 1 2 3 3 1 1 1 2 1 1 1 1 1 1 1 IMPORTANCE WElGHTING, SPLITTING AND RUSSIAN ROULETTE PARAMETERS.

NO IMPORTANCE WElGHTING OR SPLITTING.

WEIGHTING ZONE PARAMETERS FOR WEIGHTING ENERGY RANGE 1.

ZONE MINIMUM WEIGHT SURVIVAL WEIGHT SPLITTING WElGHT MINIMUM BEAM SURVIVAL BEAM IMPORTANCE STRENGTH STRENGTH WEIGHT 1 2.5000E-01 7.5000E-01 0. 2.0000E-01 5.0000E-01 0.

INTERPOLATED VALUES FOT H-1 SCATTER AT 293.0 REQUIRED 0 SECONDS A-8 Rev 2 August 1998

PG&E SAFSTOR MERIT 19 JULY 1985 PAGE 59 CELL REACTION RATE REACTION TYPE TOTAL GROUP 1 GROUP 2 GROUP 3 -GROUP 4 TOTAL 3.04364E+04 .006 1.96827E+03 .014 7.95721E+03 .020 9.64494E+03 .021 1.08660E+04 .006 SCATTER 2.91760E+04 .006 1.75583E+03 .013 7.84468E+03 .020 9.37172E+03 ~ .021 1.02038E+04 .006 CAPTURE 6.37941E+02 .023 6.95692E+00 - .081 2.22827E+01 .009 2.36576E+02 - .012 3.72126E+02 - .019 ABSORPTION 9.94269E+02 .034 3.30537E+01 .029 2.57872E+01 .010 2.73224E+02 .013 6.62204E+02 .035 FISSION 3.56327E+02 .020 2.60967E+01 .022 3.50444E+00 .014 3.66489E+01 .025 2.90077E+02 .021 N. FISSION 8.72429E+02 .049 7.35453E+01 - .065 8.59410E+00' .033 8.86489E+01 .059 7.01641E+02 .051 INELASTIC 2.64653E+02 .014 1.77911E+02 .013 8.67418E+01 .046 0. 0.000 0. 0.000 N2N 1.48065E+00 .094 1.48065E+00 .094 0. 0.000 0. ' O.000 0. 0.000 K-INFINITE = 8.78767E-01 .0313 K-EFFECTIVE =8.78767E-01 .0313 I

K EFF. FOR EACH BATCH USED.

8.7947E-01 8.6257E-01 8.9046E-01 8.6271E-01 9.1854E-01 8.8148E-01 8.7113E-01 9.1 J79E-01 8.7015E-01 8.7890E-01 8.9316E-01 8.7674E-01 8.8646E-01 8.5324E-01 8.8181E-01 8.7346E-01 8.4323E-01 8.8469E-01 8.8689E-01 8.8340E-01 8.8470E-01 8.7112E-01 8.5400E-01 8.4042E-01 8.8121E-01 9.0602E-01 8.7108E-01 8.8927E-01 8.8568E-01 8.8793E-01 l

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• Rev 2 >

August 1998  ;

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l- l 1

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i APPENDIX IC Humboldt Bay Power Plant Unit 3 i

SAFSTOR Baseline Radiation Study l-l l

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IC-i - Rev 2 August 1998

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SAFSTOR BASELINE RADIATION STUDY TABLE OF CONTENTS  :

1.0 Scope .... . ... . .. .. . ... . .. . . . . . . . . . . .. .. .. . . Page 1 2.0 Discussion. .. . ... . . . . . . . . . . . . . . . . . . . . .. . . Page 1 3.0 Summary. . . . . . . . . . . . . . . . . . . . . . . .. ....... .. . . . . . . . . . . . . . . . . . .. . Page 1 l

4.0 Status of Specific Areas 4.1 Access Shaft -66' elevation, REDT area . .. .... ... .. .. . Page 2 4.2 Access Shaft -66' elevation (exclusive of REDT area). . . . . . . . . . . . . . Page 2 4.3 Access Shaft -54' elevation. . . . . . . . . . . . . . . . . . . . . . . .. . . .. Page 3 4.4 Access Shaft -44' elevation . .... . . .. ... .. .. . .. . . . .. .. Page 3 4.5 Access Shaft -34' elevation... . .. ... . . . . . . . . ... . . . . . . . . Page 4 4.6 Access Shaft -24' elevation .. . . . . . .. . . . . . . . . .. ... .. .. Page 4 4.7 Access Shaft -14' elevation (exclusive of Shutdown Heat Exchanger Room) . .... . . . . . . . . .. . . . . . . . . . . . .. . . . . . . . . . . . . . . . . ... . . . Page 5 4.8 Access Shaft -14' elevation, Shutdown Heat Exchanger Room . .. .. . . Page 5 4.9 Access Shaft -2' elevation (exclusive of Cleanup Heat Exchanger Room) . . .. ... . ... . . . . . . . . .... .... .. . .. . . . . . . . . Page 5 4.10 Access Shaft -2' elevation, Cleanup Heat Exchanger Room.. . .. .. Page 6 4.11 Access Shaft -2' elevation, Cleanup Demineralizer Cell . .. . . . . Page 6 4.12 Escape Hatch -66' to +12' elevation. . . . . . . .. .. .. Page 7 4.13 Refueling Building +12' elevation. .. . . .. .. . . .page 7 4.14 New Fuel Storage Vault.. . .. . . . . .. . Page 8 4.15 Turbine Building Drain Tank Vault . . . .. . . . Page 8 4.16 Valve Gallery -14' elevation.. . .. .. . .. . Page 9 4.17 Valve Gallery -8' elevation.. . . .. . .. . Page 9 4.18 Valve Gallery -2' elevation.. . . . . . .. Page 10 4.19 Pipe Tunnel North +6' elevation. . . . .. . . Page 10 4.20 Pipe Tunnel South +6' elevation.. ... . . . . .. Page 10 4.21 Pipe Tunnel Condenser Area +6' elevation. . ... . . . . Page 11 4.22 Pipe Tunnel Condenser Area +12' elevation. . . . .. . . .. .. Page 11 4.23 Air-Ejector Room . . . . . . . . . . . . . . .. .. Page 11 4.24 Condensate Pump Room . . . . . . . . . . . Page 12 4.25 Instrument Vault in Demineralizer Pipe Gallery.. . ... .. .. . Page 12 4.26 Condensate Demineralizer Pipe Gallery...... . . . . .... .... . . Page 12 4 27 Condensate Demineralizer Cubicle.. . . . . . .. . . . .. . . Page 13 4 28 Condensate Demineralizer Regeneration Room. . . . . . . .. . Page 13 4.29 Radwaste Treatment Building -Tankage Area... .. . .. . . Page 14 4.30 Radwaste Treatment Building - Waste Concentrator Area .. .. . . .. . Page 14 4.31 Radwaste Treatment Building - Sump Area . . . . . . . . . .. .. . Page 14 4.32 Radwaste Treatment Building - Operating Area and Hallway.. .. . .. . Page 15 4.33 Radwaste Treatment Building - Filter Room.. . . . . . . . .. .. . . Page 15 4.34 Radwaste Treatment Building Roof... . .. . . . . . .. . . . . . . Page 16 4.35 Radwaste Treatment Building - CWT Tank Room . . ... .. . . Page 16 4.36 High Level Storage Vaults. . . . . . Page 16 4.37 Low Level Storage Building. . . .. . . . . . . . . . . . . Page 17 4.38 Radwaste Handling Building . . . . . ..... . Page 17 4.39 Yard ' Upper Area'. ... ... . .. . . . . . . Page 18 4.40 Yard - From Liquid Radwaste Building to Refueling Building. .. . . Page 18 IC-li Rev 2 August 1998

SAFSTOR BASELINE RADIATION STUDY TABLE OF CONTENTS (Continued) 4.41 Yard - Southeast Area... .. . ... . . . .. . . . . . . Page 19 4.42 Calibration Facility. . . . . . . . .. . . . . . . Page 19 4.43 Hot Shop.. . . . . . . . . . . . .. . .. . . . . . . Page 19 4.44 Seal Oil Room . . . . . . . . . . . . . . . . . .Page 20 4.45 Hydrogen Yard .. . . . . . . . . ... .. . . . .. ... . . . . . ... . Page 20 4.46 Reactor Feedwater Pump Room . . . . . .. . . . . . . .. .. .Page 21 4.47 Propane Engine-Generator Room... . . . . . .. . . .. .Page 21 4.48 Recombiner Vault. .... ... . . . . . . . . . . . . . . . . . . . .. .. ..... .. Page 21 4.49 Stack...... . .. .. .. . . . .. . . . . . . .. . . . .. ... . Page 22 4.50 Generator and Exciter Housing. ... ... . .... .. . . . . . . . . . . . . . . . Pag e 22 4 51 Turbine Enclosure. . .. ... .. .. . . . . . . . . . . . .. . .... . Page 23 4 52 Turbine Laydown Area . .... .. .. . . . . . . . . . . . . . . . . . ...Page 23 4.53 Hot Lab. ..... . . . . . .: . . . . . . . . . . . . . .. .. . . . . . . . . . .Page 23 4.54 L a undry . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . page 24 4.55 Security Area (Eastern portion of +27' elevation). ... .. . .. . .. .. .Page 24 4.56 Access Control (Western portion of RCA at +27' elevation) .. . . .. .Page 24 4.57 Control Room and Instrumentation Shop .. . ... .... .. . .. . . . ... .. Page 25 4.58 Count Room . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... . . page 25 4.59 Batteries, Multi-Zone Fan Area and Hot Lab ' Attic' . . ... . .Page 25 4.60 Suppression Chamber. .. . . . .. . . . .. Page 25 5.0 Onsite Radioactive Waste Inventory.. . . . .. . . .Page 26 6.0 Environmental Radioactivity Levels . . .Page 26 l l

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IC-iii Rev 2 August 1998

i 5/26/89 SAFSTOR BASELINE RADIATION STUDY 1.0 SCOPE This document describes the radiological status of Humboldt Bay Power Plant Unit 3 and its environs at the beginning of SAFSTOR.

2.0 DISCUSSION Section 5.0 of the SAFSTOR Decommissioning Plan for the Humboldt Bay Power Plant, Unit No. 3, dated July 1984, describes a radiation survey to establish activity levels and nuclide concentrations in the plant and its environs at the beginning of SAFSTOR.

This SAFSTOR Baseline Radiation Study is a review of surveys of plant areas and system components, to provide a baseline for comparison with surveys during the SAFSTOR period.

The review of each area summarizes radiation and radioactivity measurements. The usual ,

summary items are removable (beta-gamma) contamination, gamma radiation levels, beta radiation levels (if found), date(s) of survey, a general description of the potential for personnel contamination, and any additional remarks. ,

The study includes a section which discusses the inventory of on-site radioactive waste, and a section which describes previous measurements of environmental radioactivity. ,

3.0

SUMMARY

The radiological status of various plant areas ranges from those areas which require no significant control to those where personnel internal / external exposure might be significant.

Areas (that can be entered without tools or mobile lifting equipment) considered to have a high potential for personnel contamination are:  ;

+ Access Shaft -66' elevation, REDT Area

+ Access Shaft -14' elevation, Shutdown Heat Exchanger Room  !

+ Access Shaft -2' elevation, Cleanup Heat Exchanger Room l

+ Access Shaft -2' elevation, Cleanup Demineralizer Cell  :

+ Valve Gallery -14' elevation

+ Valve Gallery - 8' elevation

+ Air-Ejector Room

+ Condensate Demineralizer Regeneration Room

+ Specific Portions of the Refueling Building +12' elevation l

+ New Fuel Storage Vault + Turbine Building Drain Tank Vault

+ Radwaste Treatment Building - Waste Concentrator Area .

+ Radwaste Treatment Building - Sump Area

+ Radwaste Treatment Building - Filter Area (during filter changes)

+ Radwaste Treatment Building - Concentrated Waste Tank Room

+ Radwaste Treatment Building - Resin Disposal Tank Room

+ Offgas Filter / Holdup Pipe Tunnel IC-1 Rev 2 August 1998

5/26/89 Areas (that can be entered without tools or mobile lifting equipment) considered to have a potential for significant personnel exposures are:

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+ Access Shaft -66' elevation, REDT Area

+ Access Shaft -14' elevation, Shutdown Heat Exchanger Room

+ Access Shaft -2' elevation, Cleanup Heat Exchanger Room

+ Access Shaft -2' elevation, Cleanup Demineralizer Cell

+ Condensate Demineralizer Cubicle

+ Condensate Demineralizer Regeneration Room

+ Radwaste Treatment Building - Waste Concentrator Area

+ Radwaste Treatment Building - Sump Area I

+ Radwaste Treatment Building - Resin Disposal Tank Room l

+ Offgas Filter / Holdup Pipe Tunnel l

+ Radwaste Treatment Building - Resin Disposal Tank Room '

4.0 STATUS OF SPECIFIC A8EAS l 4.1 Access Shaft -66' elevation, REDT area  ;

4.1.1 Most surfaces in this area show levels of removable contamination that range from 2,000 to 20,000 dpm/dm'. Higher activity levels are located around the REDT pumps (70,000 dpm/dm') and the drainage trench to the sump (60,000 dpm/dm 2),

The sump interior should be considered highly contaminated, although numerical i activity levels are unknown.

1 4.1.2 Gamma radiation levels are generally 20 mR/hr to about 40 mR/hr, increasing toward the piping on the SW wall (about 120 mR/hr at 18"). There are numerous I hot spots on the piping, with highest contact dose rates ranging from 250 to 600 mR/hr. '

4.1.3 Beta radiation fields have been found in this area, associated with the drainage trench and the REDT (80 mrad /hr).

4.1.4 This summary is based on surveys taken during the period of 11/18/86 through 6/16/88.

4.1.5 This area has a high potential for personnel contamination, as well as the potential for undesirable radiation exposures.

4.1.6 Note that some of the particulate radioactive contamination found in the sump came directly from the reactor (during control rod drive changes and incore ' flushing').

Also, the loop seal in the vent line from the REDT may show increasing radiation levels if there is any vertical migration of material from higher elevations.

4.2 Access Shaft -66' elevation (exclusive of REDT area) 4.2.1 Most of this area is free of removable contamination (less than 500 dpm/dm 2),

Areas of contamination have been found on the walls and on the overhead surfaces. The highest accessible removable contamination (up to 16,000 dpm/dm2) was found in the drainage area around No. 2 Core Spray Pump.

4.2.2 4.2.2 Gamma radiation levels in the generally accessible area range from about 5 mR/hr near the manlift to about 15 mR/hr under the Drywell. There are numerous IC-2 Rev 2 August 1998

5/26/89 hotspots on the piping, with higher contact dose rates ranging from 70 to 300 mR/hr.

The contact dose rate on the control rod drive lifting machine cable spool is 60 mR/hr.

4.2.3 Beta radiation fields have been found around the Core Spray pump drainage area (up to 20 mrad /hr). Beta radiation fields have been found on the control rod drive lifting machine (160 mrad /hr on the cable spool).

4.2.4 This summary is based on surveys taken during the period of 11/18/86 through 6/16/88.

4.2.5 This area has moderate potential for personnel contamination from the sources discussed above, as well as from the contamination in the REDT area.

4.2.6 Note that in addition to the sources discussed above, the painted surfaces contain fixed contamination. During almost every refueling outage, this area was contaminated when control rod drives were exchanged. When decontamination was not successful, the floor was repainted.

4.3 Access Shaft -54' elevation 4.3.1 Floor surfaces in this area show low levels of removable contarnination (at or below 500 dpm/dm' as of 6/7/88). Surfaces and piping around the control rod drive 2

hydraulic accumulators show levels from 2,000 to 6,000 dpm/dm w th occasional 2

spots to 16,000 dpm/dm .

4.3.2 Gamma radiation levels currently range from about 6 mR/hr in the center of the area to about 10 mR/hr in front of the accumulators on the east and west walls. Except for a hot spot on an overhead valve west of the manlift (350 mR/hr contact in 1986),

the hot spots on the accessible piping are less than 20 mR/hr. The accumulators have contact dose rates of 250 to 400 mR/hr (1986).

4.3.3 No beta radiation fields were reported in this area.

4.3.4 This summary is based on surveys taken during the period of 1/16/86 through 6/7/88. Over this time period, dose rates have appeared to decay by approximately 25%.

4.3.5 This area has moderate potential for personnel contamination from the various piping surfaces. The potential for personnel radiation exposure is low for routine inspection, but moderate for any work in contact with the accumulators / piping.

4.3.6 Note that the surfaces below the grating on the west side of the area were not surveyed. Because this area collected contamination whenever the accumulators were exchanged, there may be sources of removable contamination there.

4.4 Access Shaft -44' elevation 4.4.1 Floor surfaces in this area generally do not have detectable levels of removable 2

contamination (less than 500 dpm/dm as of 6/7/88). Typical surface contamination on piping and equipment is about 2,000 dpm/dm , with occasional hot spots to 2

12,000 dpm/dm IC 3 Rev 2 August 1998

5/26/89 4.4.2 Gamma radiation levels currently range from about 3 mR/hr in the center of the area to about 5 mR/hr (waist high) at the east and west walls. Higher levels of 10 mR/hr can be found at floor level above the accumulators, and contact dose rates up to 15 mR/hr were reported on piping on the east side of the area.

4.4.3 No beta radiation fields were reported in this area.

4.4.4 This summary is based on surveys taken during the period of 1/16/86 through 6/7/88. Over this time period, dose rates appeared to decay by approximately 25%.

4.4.5 This area has moderate potential for personnel contamination from the various piping surfaces.

4.5 Access Shaft -34' elevation 4.5.1 Floor surfaces.in this area generally do not have detectable levels of removable contamination (less than 500 dpm/dm2 ). Most other surfaces have low 2

contamination levels (at or below 1,000 dpm/dm ), except for a few spots of contamination (e.g. 8,000 dpm/dm' on the edge of the ladder hatch way).

4.5.2 Gamma radiation levels are currently about 1 mR/hr in the east side of the area, decreasing to 0.2 mR/hr in the west side of the area. Highest contact dose rates on piping in this area are 5 mR/hr (SE corner) and 3 mR/hr (NE wall).

4.5.3 No beta radiation fields were reported in this area.

4.5.4 This summary is based on surveys taken during the period of 1/6/86 through 6/3/88.

4.5.5 This area has low to moderate potential for personnel contamination.

4.6 Access Shaft -24' elevation 4.6.1 Surfaces in this area generally do not have detectable levels of removable 2

contamination (less than 500 dpm/dm ), and except for a few measurements (2,000 to 3,000 dpm/dm'), contamination levels are at or below 1,000 dpm/dm2, 4.6.2 Gamma radiation levels are currently about 0.6 rr.R/hr in the east side of the area, decreasing to 0.2 mR/hr in the west side of the area. Highest contact dose rates on piping in this area are 5 mR/hr (SE corner, east side) and 3 mR/hr (SE wall, west side).

4.6.3 No beta radiation fields were reported in this area.

4.6.4 This summary is based on surveys taken during the period of 1/16/86 through 6/3/88.

4.6.5 This area has low to moderate potential for personnel contamination.

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l 4.7 Access Shaft -14' elevation (exclusive of Shutdown Heat Exchanger Room)

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4.7.1 Surfaces in this area are either " clean" or have barely detectable levels of  ;

removable contamination. Except for a few measurements of approximately 2,000 J 2

dpm/dm*, contamination levels are at or below 1,000 dpm/dm , j 4.7.2 Gamma radiation levels in the main room vary from a low of 0.2 mR/hr (near stair) to 1 1 mR/hr at the east and west ends. There is a hot spot of 10 mR/hr (contact with overhead pipe) at the east end of the area. The highest dose rate at contact with the Low Pressure Core Flooding line (shielded elbow on south wall) is 20 mR/hr.

The general area dose rate in the access room to the piping chase is about 3 mR/hr, with hot spots on the piping (20 mR/hr in pipe chase, and 90 mR/hr in SE corner).

l 4.7.3 No beta radiation fields were reported in this area.

i 4.7.4 This summary is based on surveys taken during the period of 1/16/86 through 5/20/88.

4.7.5 This area has moderate potential for personnel contamination. I 4.8 Access Shaft -14' elevation, Shutdown Heat Exchanger Room 4.8.1 Based on a limited number of samples, the removable contamination level of the floor 2

in this room is approximately 3,000 to 6,000 dpm/dm ,

t 4.8.2 Gamma radiation levels in this area vary considerably due to the complexity of the piping. The general area is expected to be about 30 to 50 mR/hr, with hot spots on ,

piping of a few R/hr. l 4.8.3 Beta radiation has been reported for this area, but except for open systems (e.g. l drains) it does not appear to be significant as compared to the gamma fields. ,

l 4.8.4 This summary is from data collected 9/26/88, as well as from data collected May, 1984.

y 4.8.5 Because of the scarcity of information, this area is considered to have a moderate to {

l high potential for personnel contamination. i e

4.9 Access Shaft -2' elevation (exclusive of Cleanup Heat Exchanger Room)  :

4.9.1 Surfaces in this area are either " clean" or have bsci/ detectable levels of  ;

removable contamination. Except for one measurement of approximately 2,000 1 dpm/dm', contamination levels are at or below 1,000 dpm/dm .

4.9.2 Gamma radiation levels in the main room vary from a low of 0.3 mR/hr (foot of stair) to about 2 mR/hr elsewhere. Dose rates of up to 10 mR/hr can be found on a few pipes. There is some " shine"(5 mR/hr) at the door to the Cleanup Heat Exchanger Room.

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5/26/89 4.9.3 4.9.3 No bnta was observed in this area, except for a floor drain toward the i East end of the room (8 mrad /hr beta).

4.9.4 This summary is based on surveys taken during the period of 1/16/86 through 5/20/88.

4.9.5 This area has low to moderaie potential for personnel contamination.

4.10 Access Shaft -2' elevation, Cleanup Heat Exchanger Room e

4.10.1 Based on a limited number of samples, the removable contamination level of 2

surfaces in this room may be as high as 20,000 dpm/dm (as of May 1984), although .

subsequent data suggests a decreasing trend to about 3,000 dpm/dm'(September )

1988).

4.10.2 Gamma radiation levels in this area vary considerably due to the complexity of piping. The general area is expected to be about 40 to 60 mR/hr, with hot spots on '

piping of a few R/hr. ,

4.10.3 Beta radiation was not reported for this area, but might be expected at any open systems (e.g. drains).

4.10.4 This summary is from data collected 9/26/88, as well as from data collected May, 1984.

4.10.5 Because of the scarcity of information, this area is considered to have a high potential for personnel contamination.

4.10.6 To aid interpretation of surveys, it should be noted that the new (PG&E West) heat exchangers are relatively free of contamination and internal radioactivity, as compared to the old ones that were abandoned in place (about 1976).

4.11 Access Shaft -2' elevation, Cleanup Demineralizer Cell 4.11.1 The removable contamination level of surfaces in this room is highly variable, 2

ranging from 1,000 to 100,000 dpm/dm (beta-gamma)2 Removable alpha contamination was also reported at about 200 dpm/dm ,

4.11.2 Gamma radiation levels in the cell are generally highest at floor level (due to radiation from components placed here for SAFSTOR). Highest contact dose rates are the Demineralizer Inlet piping (250 mR/hr) and the drum containing Scram Dump Tank Level instrumentation (420 mR/hr at floor level).

4.11.3 Beta radiation has been reported for this area, but except for open systems (e.g.

drains) it does not appear to De significant as compared to the gamma fields.

4.11.4 This summary is based on a survey taken 12/30/86, with other data from May 1984.

4.11.5 This area is considered to have high potential for personnel contamination and undesirable personnel radiation exposure.

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5/26/89 4.11.6 Note that there are 6 drums stored here, containing contaminated components (such as the hydraulic filter housings, the cleanup pump, and scram dump tank level instrumentation).

4.12 Escape Hatch -66' to +12' elevation 2

4.12.1 The removable contamination levels range from less than 500 to 1,000 dpm/dm from the +12' to the -14' elevations. Contamination levels range from 1,000 to 3,500 dpm/dm2from the -34' to -66' elevations.

4.12.2 No gamma or beta readings are available, but because there are no radioactive systems in this area, no significant radiation is expected.

4.12.3 This summary is based on a survey taken on 9/9/86.

4.12.4 This area has low to moderate potential for personnel contamination.

4.13 Refueling Building +12' elevation 4.13.1 This area has varying levels of removable contamination. Contamination levels are j l generally less than 500 dpm/dm', with spots of removable contamination at levels of 1,000 to 3,000 dpm/dm'. Higher levels are found at the following locations:

2 a) Removable contamination at levels of 25,000 to 60,000 dpm/dm was found on the crane main hook cable spool, traveling block and main hook. Note that the more highly contaminated portions are enclosed in a vinyl bag. i b) The surfaces of the Spent Fuel Pool' overflow scuppers' (in the Southwest and Southeast corners) had 200,000 to 400,000 dpm/dm of removable contamination. Note that these areas are now below the sheet metal flashing around the pool. ,

c) The Spent Fuel Pool Recire. Pumps have indicated vaging smear results, l

ranging from a typical level of 10,000 to 20,000 dpm/dm .to a high of 260,000 l dpm/dm'.

l d) Some of the piping below the West erid of the Emergency Condenser has i been found to have removable contamination levels in the range of 15,000 to 25,000 dpm/dm2, l

e) Smears of the two hydraulic pumps (prior to decontamination) indicated 2

removable contamination levels of 10,000 to 30,000 dpm/dm . Although 2

more recent smears show reduced levels (e.g.1,000 dpm/dm ), it is not clear that all of the contamination has been removed.

f) The 'Washdown Area' has removable contamination in the range of 2,000 to 2

20,000 dpm/dm indicated 15,000 (including) a smear of the top of the concrete blo dpm/dm .

4.13.2 Gamma radiation levels are generally less than 1 mR/hr, increasing to about 3 to 5 mR/hr around the pool, the recirculation pumps, and the hydraulic system pumps.

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5/26/89 Various hotspots are located on piping, particularly piping near the hydraulic pumps (about 50 mR/hr), on piping near the Recirc Pumps (20 to 50 mR/hr) and on the recirc. pump discharge line, about 8' above floor level (100 mR/hr).

4.13.3 Beta rad!: ton fields have been measured at the floor drains (12 to 32 mrad /hr) and at the SFP Recirc Pumps (survey data illegible).

4.13.4 This summary is based primarily on surveys performed from 8/18/86 through 6/17/88.

4.13.5 Most of this area has a moderate potential for personnel contamination, but specific locations have high potential for personnel contamination.

4.13.611 should be noted that this area was contaminated throughout the life of the plant, and that the floor was routinely repainted to fix any contamination remaining after decontamination. A similar treatment was applied to the exterior of the Fuel Transfer Cask There is also a variety of previously contaminated components stored in this area, including the (interior of) the Fuel Transfer Cask, the fuel handling tools (in the washdown area), the Extension Tank (inside the sheet metal can on top of the reactor shield plug), and some of the slings (Northeast corner of area). For a final note, the 4" steel floor plate (between the pool and the reactor) was installed under contaminated conditions, so that the floor underneath it should be assumed contaminated.

4.14 New Fuel Storage Vault 4.14.1 Floor surfaces have typical removable contamination levels of from 1,000 to 10,000 2

dpm/dm . Higher values have been observed on a valve on the Spent Fuel Pool 2

Cooler (18,000 dpm/dm ),

4.14.2 General area gamma dose rates are 10 to 25 mR/hr. Contact dose rates on the

' cans' containing the spare control rod drives show hotspots (up to 160 mR/hr).

1 4.14.3 A beta radiation field of 20 mrad /hr was found at the sample scupper in the Northwest corner of the area.

i 4.14.4 This summary is based on surveys performed from 4/10/86 through 1/22/88.

4.14.5 This area has a high potential for personnel contamination.

4.14.6 Note that this area was used to store contaminated (but not irradiated) fuel bundles.

The spare control rod drives are stored (in ' cans') in this area.

4.15 Turbine Building Drain Tank Vault 4.15.1 Floor surfaces have removable contamination levels ranging from 3,000 to 9,000 2

dpm/dm . Walls have removable contamination levels ranging from 10,000 to 2

170,000 dpm/dm . The West wall has the highest readings (170,000 dpm/dm2 beta-gamma,40 cpm alpha) The West end of the tank has the highest readings (90,000 dpm/dm',20 com alpha. The pump located in the SE corner has removable 2

contamination nevels of 100,000 dpm/dm .

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I 5/26/89 4.15.2 General area gamma dose rates are 10 to 25 mR/hr. Contact dose rates on the bottom of the tank range from 30 to 150 mR/hr. There is a 60 mR/hr hot spot on a l pipe in the overhead (in the middle of the South wall).

4.15.3 No beta radiation fields were found with the exception of the floor drain (80 mrad /hr).

4.15.4 This summary is based on surveys performed from 4/10/86 through 1/22/88.

4.15.5 This area has a high potential for personnel contamination.

1 l 4.16 Valve Gallery -14' elevation 4.16.1 The removable contamination levels range from less than 500 to 2,500,000 2

dpm/dm2 w th a general average of about 2,000 to 10,000 dpm/dm The highest i contamination was found at a valve (assumed to be the 8" Shutdown System Return l to Reactor Motor Operated Valve). At this location, the smear indicated 0.5 mR/hr gamma,60 mrad /hr beta, and 1,200 dpm alpha.

l 4.16.2 Gamma radiation levels in this area are generally 10 to 20 mR/hr. The major i hotspot is at the valve (and line) indicated above, with a contact reading of 40 mR/hr.

l 4.16.3 Beta radiation was primarily found at the hotspot at the valve described above (200 mrad /hr). In addition, the two floor drains were observed to be beta sources (12 to 64 mrad /hr).

4.16.4 This summary is primarily based on a survey taken on 5/22/86 and 5/23/86.

4.16.5 This area has moderate to high potential for personnel contamination.

4.17 Valve Gallery -8' elevation 2

4.17.1 The removable contamination levels range from less than 500 to 30,000 dpm/dm with a general average of about 5,000 dpm/dm'. One smear was found to have alpha with 200 dpm/dm2.

4.17.2 Gamma dose rates in this area are generally 10 to 30 mR/hr. A variety of hotspots were observed on piping and valves, up to 220 mR/hr.

4.17.3 Beta radiation was primarily found at a hotspot at a valve near the center of the area (possibly the feedwater isolation valve) with a reading of 320 mrad /hr.

4.17.4 This summary is primarily based on a survey taken on 5/21/86 and 5/23/86.

4.17.5 This area has moderate to high potential for personnel contamination.

4.18 Valve Gallery -2' elevation IC-9 Rev 2 August 1998

l 5/26/89 4.18.1 Tha removablo contamination lovels range from less than 500 to 18,000 dpm/dm 2 2

with a general average of about 2,000 dgm/dm . No removable alpha activity was I observed.

4.18.2 Gamma dose rates in this area are generally 10 to 30 mR/hr, except around the t l Emergency Condenser Condensate Return Valve (East of Reactor). A variety of ]

hotspots were observed on piping and valves, up to 280 mR/hr. l l 4.18.3 No beta radiation was observed in this area. l

4.18.4 This summary is primarily based on a survey taken on 5/21/86 and 5/23/86.

4.18.5 This area has moderate potential for personnel contamination.

4.19 Pipe Tunnel North +6' elevation 4.19.1 This area 2

is considered to be free of removable contamination (less than 500 i dpm/dm ),  ;

4.19.2 Gamma radiation dose rates are approximately 1 mR/hr. Higher dose rate (2 to 6 mR/hr) are found near piping.

l 4.19.3 No beta radiation fields have been detected.

i 4.19.4 This summary is based on surveys performed from 1/14/86 through 6/2/88.

4.19.5 This area is considered to have a low potential for personnel contaminations.

i 4.19.6 Note that although most of the area is clean, there is contamination in the drainage trench which underlies the ' seismic patch plate' in the middle of the room.

4.20 Pipe Tunnel South +6' elevation 4.20.1 This area has low to moderate levels of removable contamination (generally at or below 1,000 dpm/dm2, w th a few locations up to 4,000 dpm/dm').

i 4.20.2 General gamma radiation dose rates in the area under the turbine steam exhaust trunk range from 5 mR/hr up toward 35 mR/hr closer to the condenser. Dose re'.c3 near the North end of the feedwater heaters and associated piping are approximately 25 mR/hr (at 18") with hotspots up to 150 mR/hr.

4.20.3 No beta radiation was noted in this area on the most recent survey, but some tieta was observed during earlier surveys (at a floor drain and at the packing of valves located along the West wall).

4.20.4 This summary is based on surveys performed from 7/4/86 through 7/14/86.

4.20.5 This area has low to moderate potential for personnel contamination.

l 4.21 Pipe Tunnel Condenser Area +6' elevation IC-10 Rev 2 August 1998

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5/26/89 l

4.21.1 The surfaces in this g::neral area have levels of contamination of about 1,000 to L 4,000 dpm/dm2 . The contamination levels under the condenser have not2 been well i defined, but are expected to be in the range of 1,000 to 10,000 dpm/dm ,

4.21.2 Away from the condenser, the area dose rates vary from about 10 to 50 mR/hr,

depending on the radiation levels of the nearby piping. The bottom of the condenser has contact readings varying from 100 to 250 mR/hr, typically dropping to less than 100 mR/hr at 18". The space under the condenser probably has

! doserates in the range of 100 to 200 mR/hr.

l l 4.21.3 A reading of 15 mrad /hr for Beta radiation was observed at the floor drain near the

! center of the East wall.

I i l

4.21.4 This summary is primarily based on surveys performed from 7/16/86 through l l 7/17/86. I 4.21.5 This area is considered to have a moderate potential for personnel contamination.

( 4.22 Pipe Tunnel Condenser Area +12' elevation

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4.22.1 This area hac generally low levels of removable contamination (less than 500 2

dpm/dm2 ), except that 1,000 dpm/dm of removable beta contamination was 2 observed on the Nonh condenser waterbox doors (associated with 2 dpm/dm of removable alpha activity).  ;

4.22.2 Gamma radiation dose rates range from 2 to 15 mR/hr, with the higher dose rates directly in front of the access door on the East side of the area.

i 4.22.3 No beta radiation fields were detected.

4.22.4 This information is primarily based on surveys performed on 4/7/86 through 7/16/86, and on 5/25/88.

4.22.5 This area is considered to have a low potential for personnel contamination.

4.23 Air Ejector Room 4.23.1 This area is considered to be a contaminated area. Removable contamination 2

levels on most surfaces are generally 1,000 to 10,000 dpm/dm . The highest removable contamination levels are associated with the turbine gland seal exhaust fans and condenser. One smear of the equipment pedestal read 24 mrad Beta. A 2

different survey at the same point indicated 165,000 dpm/dm beta-gamma and 150 dpm/dm2 alpha.

~4.23.2 Gamma radiation levels in the general area are about 1 to 5 mR/hr, with equipment / piping contact readings up to 30 mR/hr. There is one spot on (in) the concrete floor under the offgas flow meter loop seal (North side of the air ejector) that produced a contact reading of 100 mR/hr.

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4.23.3 Beta radiation has been observed on the equipment pedestal noted above (72 mRadihr), on a floor drain (24 mrad /hr), and at the hot spot on the floor (1200 mrad /hr).

4.23.4 This information is based on surveys performed between 5/15/86 and 4/25/87.

4.23.5 This area has a high potential for personnel contamination.

I 4.24 Condensate Pump Room 4.24.1500This area, except dpm/dm is g)enerally considered for some portions to bedrive of the electric freevacuum of removable pump (6,000 contamin to 2

10,000 dpm/dm ) and the pits for the two condensate pumps (30,000 dpm/dm2),

4.24.2 Gamma radiation fields are in the range of 1 to 5 mR/hr. Hot spots on piping are typically 20 mR/hr, with the exception of one hot spot on piping of 60 mR/ht.

4.24.3 Recent survey information on beta fields is not available. However, a survey performed on 5/20/86 showed beta fields of 20 mrad /hr in one of the pits for the ,

condensate pumps.  !

4.24.4 This information is based on surveys performed from 1/17/86 through 3/4/88.

4.24.5 This area is considered to have a low potential for personnel contamination. j 4.25 Instrument Vault in Demineralizer Pipe Gallery 4.25.1 This area is considered to be essentially free of removable contamination. l Contamination levels are at or below 1,000 dpm/dm' 4.25.2 Gamma radiation readings in this area are less than 0.5 mR/hr.

4.25.3 No beta radiation was found in this area. l 4.25.4 This summary is based on surveys performed 12/29/86 and 6/13/88.

4.25.5 This area has a low potential for personnel contamination.

4.26 Condensate Demineralizer Pipe Gallery 4.26.1 This area is not considered to have significant removable contamination (less than i 500 dpm/dm2),

4.26.2 Gamma radiation levels about 1 to 2 mR/h in the Southern side of the area, j increasing toward the door to the Regeneration Room, and toward the Condensate piping. There are hot spots on the bottom of the condensate demineralizer strainers (of about 15 to 30 mR/hr).

4.26.3 No beta radiation fields exist with the exception of the West floor drain which measures about 1 mrad /hr.

4.26.4 This summary is based on surveys performed from 1/17/86 through 8/4/88.

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5/26/89 4.26.5 This area has a low potential for personnel contamination.

4.27 Condensate Demineralizer Cubicle l

4.27.1 Removable contamination levels in this area range from less than 500 to 2,000 dpm/dm' 4.27.2 Gamma radiation levels in the general area are 20 to 50 mR/hr when the resin in #1  !

Demineralizer is not contaminated (see below). Otherwise, the general levels can i be about 200 to 300 mR/hr. Typical levels for the contact dose rate profile of the demineralized tank are approximately 40 mR/hr at the top and bottom and 2 to 3  !

R/hr at the middle.

1 4.27.3 No beta radiation fields were reported in this area.

4.27.4 This information is based on surveys performed from 1/9/87 through 8/4/88.

4.27.5 This area has a low to moderate potential for personnel contamination, and has the potential for undesirable radiation exposures.

4.27.6 Note that the #1 (nearest to the entrance) demineralizer tank was converted to serve as the Spent Fuel Pool demineralizer. Because of radioactivity continuously removed from the recirculated water, the contact radiation levels will increase during l

the lifetime of each resin replacement. The contact dose rates appear to change at about 200 mR/hr for each month of operation.

l 4.28 Condensate Demineralizer Regeneration Room 4.28.1 This area is considered to be a highly contaminated area, with contamination levels ranging from 2,000 to 200,000 dpm/dm'.

4.28.2 General area gamma radiation levels range from 6 to 30 mR/hr. Contact dose rates l

on the laundry waste tank (Northwest corner of room) range from 2 to 45 mR/hr. '

l Contact dose rates on the cation tank (Northeast corner of room) range from 30 to l

800 mR/hr. Contact dose rates on the anion tank (Southeast corner of room) range l from 6 to 40 mR/hr. The higher dose rates generally are found at or under the bottom of the tanks. Contact dose rates on the overhead resin piping are generally J 30 mR/hr, with a hot spot of 100 mR/hr near the cation tank.

4.28.3 Beta radiation fields show 40 mrad /hr at the floor drain,60 mrad /hr at the sight j glass and 600 mrad /hr at a scupper (due to approximately 1/2 gallon of resin, noted ,

to be in the scupper)._ l 1

4.28.4 This summary is based on surveys performed from 9/8/86 through 12/22/86. i l

l 4.28.5 This area has a high potential for personnel contamination, as well as the potential for undesirable radiation exposures.

l 4.29 Radwaste. Treatment Building - Tankage Area l

l

l. IC-13 Rev 2 August 1998

i 5/26/89  !

4.29.1 This area has generu!Iy low levels of removable contamination (less than 500 dpm/dm') with the exception of the level instrumentation piping (for the CWTs and 3 the RDT), near No. 3 WRT (3,000 to 6,000 dpm/dm'). Removable contamination '

has also been found (20,000 to 50,000 dpm/dm2 ) on the concentrator feed line  :

valves (at the corner of the concrete wall, near the entrance to the tankage area). l 4.29.2 Gamma radiation area doses rates range from 2 to 14 mR/hr. Highest readings are in No. 3 WRT area. Lowest readings are in No.1 WRT and No.1 WHT area.

Contact dose rates on Nos.1 and 2 WHTs are less than 5 mR/hr. Contact dose  ;

rates on No.1 WRT are 3 to 12 mR/hr. Contact dose rates on No. 2 WRT are 35 to 70 mR/hr. Contact dose rates on No. 3 WRT are 35 to 150 mR/hr.

4.29.3 Beta radiation fields have been detected at the scupper near No. 3 WRT (52 mrad /hr) and on the Concentrator Drip Pump (22 mrad /hr). )

1 4.29.4 This summary is based on surveys performed from 1/8/86 through 8/26/88.  !

4.29.5 This area has a low to moderate potential for personnel contamination.

4.29.6 There are contaminated hoses (reserved for future solidification projects) stored in plastic at the West side of the area. Note that although the typical liquid radioactive waste ( in Nos.1 and 2 WRT or in Nos.1 and 2 WHT) is at about 1 x 10-5 pCi/ml, the material in No. 3 CWT has radioactivity levels of about 0.3 pCi/ml (Cs-137 and Co-60). Note also that the pipe trench (under the walkway at the South side of the area) was decontaminated, and then painted to fix remaining contamination.

l 4.30 Radwaste Treatment Building - Waste Concentrator Area i 4.30.1 This area has high levels of removable contamination, with surfaces in this area  ;

showing widely varying levels of contamination. The levels range from 1,000 to i 330,000 dpm/dm', with the highest levels near the concentrator.

4.30.2 Gamma radiation dose rates are approximately 20 mR/hr. Dose rates near the l concentrator are approximately 45 mR/hr.

4.30.3 Beta radiation fields have been measured on both pumps (30 to 50 mrad /hr) and the floor drain (400 mrad /hr).

l 4.30.4 This summary is based on surveys performed from 2/4/86 through 7/1/88.

4.30.5 This area has a high potential for personnel contamination, and it has the potential for undesirable personnel beta exposure .

4.31 Radwaste Treatment Building - Sump Area i

4.31.1 This area has high levels of removable contamination. Levels at the sump range from 26,000 to 70,000 dpm/dm' and 46 dpm/dm2 alpha. Levels at the floor drain 2

show 350 dpm/dm' beta-gamma and 250 dpm/dm alpha. Levels near the Concentrator Feedpump range from 6,000 to 26,000 dpm/dm2, I IC 14 Rev 2 i

Suaust 1998 _

5/26/89 4.31.2 Gamma radiation dose rates are variable. Dose rates are highest at floor level (20 mR/hr) and decrease with height (3 mR/hr). A hot spot at the sump measures 40 mR/hr. The floor drain measures 55 mR/hr. Piping below the CCW Heat Exchanger has contact dose rates of about 30 to 50 mR/hr.

4.31.3 Beta radiation fields have been measured at the floor drain (2,000 mrad /hr), the sump (60 mrad /hr), and the Concentrator Feedpump (280 mrad /hr).

4.31.4 This summary is based on surveys performed from 1/8/86 through 7/1/88.

4.31.5 This area has a high potential for personnel contamination and undesirable personnel beta exposures.

4.32 Radwaste Treatment Building - Operating Area and Hallway 2

4.32.1 This area shows removable contamination levels from less than 500 dpm/dm to 2

about 2,000 dpm/dm .

4.32.2 Gamma radiation dose rates are generally less than 2 mR/hr. There are several hot spots (4 to 9 mR/hr).

l 4.32.3 Beta radiation was found at the floor drain between the radwaste pumps (110 mrad /hr), at the sample scupper (220 mrad /hr) and on the pump bases (60 to 70 mrad /hr) 4.32.4 This summary is based on surveys performed from 1/8/87 through 3/2/88.

l 4.32.5 This area has a moderate potential for personnel contamination.

t

! 4.33 Radwaste Treatment Building - Filter Room 2

4.33.1 This area shows removable contamination levels from less than 500 dpm/dm to l 1,000 dpm/dm' 4.33.2 Gamma radiation dose rates are generally 1 to 5 mR/hr. At the time of these surveys, the filter ' pig' had a surface reading of about 30 to 50 mR/hr. Note that with current (1989) filters, the ' pig' contact dose rate is about 2 to 5 mR/hr.

4.33.3 Beta radiation has been detected at the floor drain (40 mrad /hr).

4,33.4 This summary is based on surveys performed from 3/15/88 through 8/8/88.

I 4.33.5 This area has low potential for personnel contamination during normal conditions, the area is considered to have moderate to high potential for personnel contamination during the process of replacing expended filter cartridges.

4.34 Radwaste Treatment Building Roof IC-15 Rev 2 August 1998

l 5/26/89

! 4.34.1 This area has low contamination levels (less than 500 dpm/dm').

4.34.2 Gamma radiation dose rates are less than 2 mR/hr. The resin transfer line from the Radwaste Demineralizer to the Resin disposal tank has general contact reading of about 5 mR/hr. The hotspot (about 100 mR/hr) that had been indicated on the Resin Disposal Tank vent line has been removed (by flushing back into the tank).

4.34.3 No beta radiation fields have been detected.

4.34.4 This summary is based on surveys performed from 2/5/86 through 9/26/88.

4.34.5 This area has low potential for personnel contamination, except possibly in the area  ;

of the trash compactor when waste handling is in progress.

4.35 Radwaste Treatment Building - CWT Tank Room 4.35.1 This area haswery high removable contamination levels. Smears of walls and piping measured by ' pancake GM' indicate 40,000 to 100,000 dpm/dm2 , but most smears of the floor were measurable only with dose rate instruments (reading 8 to 15 mrad /hr per dm2),

4.35.2 Gamma radiation dose rates decrease with height. The ceiling has the lowest dose rate. General area dose rates are on the order of 40 mR/hr. Contact dose rates on the tanks range from 30 to 150 mR/hr. Contact dose rates on piping range from 30 to 120 mR/hr.

4.35.3 Beta radiation was noted (320 mrad /hr) for a survey taken while the floor was flooded with concentrated liquid waste. After washing the area, no beta radiation fields were detected (using an instrument bagged to prevent contamination). Based on the readings from the smears described above, beta fields of 20 to 200 mrad /hr would probably be observed near the floor.

4.35.4 This summary is based on surveys performed from 6/30/86 through 7/31/86.

4.35.5 This area has a high potential for personnel contamination, as well as personnel beta and gamma exposure.

4.35.6 Note that this area has been flooded with concentrated liquid waste at least twice in the history of the plant. The tanks were left essentially empty, so that this problem is not expected to reoccur.

l 4.36 High Level Storage Vaults 4.36.1 No.1 Vault has removable contamination levels which range from less than 1,000 to l 2,000 dpm/dm . No. 2 Vault has removable contamination levels which range from less than 1,000 to 7,000 dpm/dm2 . No. 3 Vault has removable contamination levels which range from less than 1,000 to 9,000 dpm/dm2 . One smear showed 20 dpm/dm2 alpha.

IC-16 Rev 2 August 1998

5/26/89 4.36.2 No.1 Vault gamma radiation dose rates are less than 0.2 mR/hr with the exception of a hot spot on the floor of 1.2 mR/hr gamma and 60 mrad /hr beta. No. 2 Vault gamma radiation dose rates are less than 0.2 mR/hr with the exception of the floor .

drain which shows 5 mR/hr gamma and 52 mrad /hr beta. No. 3 Vault gamma radiation dose rates are less than 0.6 mR/hr. Beta fields up to 140 mrad /hr were detected on the east floor and wall.

l 4.36.3 This summary is based on surveys performed from 1/31/86 through 10/20/86.

4.36.4 This area has moderate potential for personnel contamination.

4.36.5 Note that these vaults will be used for storage of higher dose rate waste (in 55-gallon drums).

4.37 Low Level Storage Building 2

4.37.1 This area is c5nsidered free of removable contamination (less than 500 dpm/dm ),

4.37.2 Gamma radiation dose rates are generally less than 1 mR/hr. Note that dose rates in this building depend on the materials stored in the area, as there are no other significant radiation sources in the area.

4.37.3 No beta radiation fields have been detected.

4.37.4 This summary is based on surveys performed from 1/7/86 through 6/24/88.

4.37.5 This area has a low potential for personnel contamination.

4.37.6 The West end of the structure is used as a staging area for the accumulation of waste for packaging. This part of the area is also used for storage of ladders and scaffoldings that have fixed contamination. The East end of the structure is primarily used for 'SAFSTOR' storage of tools / equipment that is contaminated.

4.38 Radwaste Handling Building 4.38.1 This area is considered to be free of removable contamination (less than2 500 dpm/dm') with the exception of the reactor head shield (4,000 dpm/dm beta-gamma 2

and 43 dpm/dm alpha).

4.38.2 Gamma radiation dose rates are generally less than 0.2 mR/hr except for roped area in SE corner (5.0 mR/hr).

4.38.3 No beta fields have been detected with the exception of the reactor head shield (20 mrad /hr).

4.38.4 This summary is based on surveys performed from 1/31/86 through 8/9/88.

4.38.5 This area has low potential for personnel contamination.

4.39 Yard ' Upper Area' IC-17 Rev 2 August 1998

5/26/89 l 4.39.1 This area is considered to be generally free of removable contamination (less than 500 dpm/dm2),

I l 4.39.2 in most of the area, gamma dose rates are less than 0.2 mR/hr. Higher dose rates l

l are found between the Low Level Storage Building and the Solid Waste Handling '

Building (about 3 mR/hr) due to material stored in both structures. j 4.39.3 No beta radiation fields have been detected.

l l 4.39.4 This summary is based on surveys performed from 4/3/86 through 9/26/88. i I 4.39.5 This' area is not considered to have significant potential for personnel contamination.

4.39.6 Most of this area (with the exception of an area about 100' x 100' in the Northwest corner of the yard) had the pavement removed, soil sampled and discarded (if contaminated)- The remaining soil was graded, selectively back-filled and then ,

paved. In general, the soil below the pavement in this area is considered to have  !

less than 1 pico-Curie /gm of Cs-137 or Co-60.

4.40 Yard - From Liquid Radwaste Building to Refueling Building 4.40.1 This area is considered to be generally free of removable contamination (less than 2

500 dpm/dm ),

4.40.2 In most of the area, gamma dose rates are less than 0.2 mR/hr. Higher dose rates are found along the South face of the Liquid Radwaste Treatment Facility Enclosure (generally 0.5 to 2.5 mR/hr, with 3.4 mR/hr at ground level, in front of the middle roll-up door), around the Condensate Storage Tank (up to 10 mR/hr at contact with tank base) and around the amputated piping Southeast of the Stack (about 5 mR/hr). Hot spots were identified at the piping South of the CST (100 mR/hr) and i I

i on the amputated piping by the Stack (30 mR/hr), but some of this piping was removed since the survey (3/30/88).

4.40.3 No beta radiation fields have been detected.

4.40.4 The summary is based on surveys performed from 1/1/86 through 9/26/88.

4.40.5 This area is not considered to have significant potential for personnel contamination.

4.40.6 Most of this area (with the exception of concreted sections) had the pavement removed, soil sampled and discarded (if contaminated). The remaining soil was l graded, selectively back- filled and then paved. The dirt bank at the East end of the area was coated with concrete to stabilize the bank. In general, the soil below the l

, pavement / concrete in this area is considered to range from less than 1 to about 10 '

l pico-Curie / gram of Co-60 or Cs-137. i l

4.41 Yard - Southeast Area IC-18 Rev 2 l August 1998

5/26/89 4.41.1 This area is considered to be generally free of removable contamination (less than 500 dpm/dm').

4.41.21n most of the area, gamma dose rates are less than 0.2 mR/hr. Higher dose rates are found at the entrance to the Condensate Pump Room (0.5 mR/hr), to the Condensate Demineralizer Pipe Gallery (0.8 mR/hr) and to the Condenser Bay (4.8 mR/hr).

4.41.3 No beta radiation fields have been detected.

4.41.4 The summary is based on surveys performed from 1/9/86 through 8/31/88.

4.41.5 This area is not considered to have significant potential for personnel contamination.

4 41.6 Most of this ar,ea had the pavement removed, soil samp'.cd and discarded (if significantly contaminated). The railroad tracks and ties between the buildings and the gate were removed. The remaining soil was graded, selectively back-filled and then paved. In general, the soil below the pavement / concrete in this area is considered to range from less than 1 to about 100 pico-Curie / gram of Co-60 or Cs-137.

4.42 Calibration Facility 4.42.1 This area is considered to be essentially free of removable contamination (less than 2

500 dpm/dm ),

4.42.2 Gamma radiation dose rates are generally low (ranging from 0.2 mR/hr to 0.6 mR/hr), with hot spots on the source locker (6 mR/hr) and the cask for the 'old' Co-60 source (4 mR/hr).

4.42.3 No beta, radiation fields have been detected. Note there is a Uranium slab source in the source locker that reads about 120 mrad /hr.

4.42.4 This summary is based on surveys performed from 1/22/86 through 9/18/86.

4.42.5 This area has a low potential for personnel contamination.

4.42.6 There is a high potential for undesirable personnel exposures due to the presence of the two Co-60 sources (one in indefinite storage pending disposal, and the other installed in the source calibration well). Note that when the source in the well is unshielded and moved to its highest location there is a dose rate of about 25 R/hr at the opening of the well, and the potential of several hundred mR/hr on top of the roof.

4.43 Hot Shop 4.43.1 The area on the clean side of the step-off-pad is considered to be essentially free of 2

removable contamination (less than 500 dpm/dm ). Most of the remaining area of the shop has generally low ievels of removable contamination (less than 1,000 IC-19 Rev 2 huced WR@

l 5/26/89 2

dpm/dm ) with the exc:ption of the lathe (up to 3,000 dpm/dm2 beta-gamma and 64 2

dpm/dm alpha) and with exception of the washdown area (up to 40,000 dpm/dm2 and 20 dpm/dm alpha).

4.43.2 Gamma radiation dose rates are less than 0.2 mR/hr, with the exception of the floor drains (5 mR/hr).

4.43.3 No beta radiation fields have been detected, with the exception of the lathe (0.4 mrad /hr) and the floor drains (12 mrad /hr at the one in the washdown area).

4.43.4 This summary is based on surveys performed from 1/10/86 through 8/12/88.

4.43.5 Most of this area nas a low potential for personnel contamination. The lathe has a moderate potential for personnel contamination, and the washdown area has a high potential for personnel contamination.

l 4.44 Seal Oil Room -

4.44.1 This area is generally considered to be free of removable contamination with the 2

exception of the center floor drain (2500 dpm/dm beta-gamma and 11.8 dpm/dm2 alpha).

4.44.2 Gamma radiation levels are less than 0.2 mR/hr.

i 4.44.3 No beta radiation fields have been detected with the exception of the center floor drain (0.4 mrad /hr). ,

4.44.4 This summary is based on surveys performed from 1/6/86 through 9/10/88.

4.44.5 This area has a low potential for personnel contamination.

4.45 Hydrogen Yard 4.45.1 This area2 is considered to be free of removable contamination (less than 500 dpm/dm ), with the exception of the instrument vault, where 12,000 dpm/dm2 has been observed.

4.45.2 Gamma radiation levels are less than 0.2 mR/hr.

4.45.3 No beta radiation has been detected.

4.45.4 This summary is based on surveys performed from 1/6/86 through 9/10/88.

, 4.45.5 This area is generally considered not to have potential for personnel contamination.

l 4.45.6 Note that the Battery Room (batteries removed) was constructed about 1976, and is l believed to be not contaminated. At the time of the latest survey; the generator

! Hydrogen Cooler (Closed Cooling Water heat exchanger) is stored in a box to the West of the Battery Room. The Hydrogen Cooler is believed to be uncontaminated.

IC-20 Rev 2 August 1998

5/26/89 4.46 Reactor Feedwater Pump Room 4.46.1 500This area dpm/dm isexception

. An p)enerallyis theconsidered to be offree trou0hs and scuppers of removable the two reactor contam feedwater pumps where some removable contamination was found (approximately 2,000 dpm/dm' ) ,

4.46.2 Gamma radiation levels are generally less than 0.2 mR/hr, except in the vicinity of the feedwater pumps and piping. Radiation levels near the pumps and piping are 1 mR/hr. Contact dose rates on the piping are 2 to 5 mR/hr. An overhead elbow SE of No. 2 Reactor Feedpump has a contact dose rate of 20 mR/hr.

4.46.3 No Beta radiation fields have been detected in this area.

4.46.4 This summary is based on various surveys taken during the period of 1/6/86 through '

9/10/88.

4.46.5 This area is considered to have low potential for personnel contamination, exclusive of any material retained on the interior of the feedwater piping. No significant contamination has been found in the overhead portions of the room. ,

4.46.6 Note that the below grade conduit run from pullbox 02 towards the radwaste areas has been contaminated.

4.47 Propane Engine Generator Room 4.47.1 This area is considered to be free of removable contamination (less than 500 dpm/dm2),

4.47.2 Gamma radiation levels are less than 0.1 mR/hr.

4.47.3 No beta radiation fields have been found in this area.

4.47.4 Tnis summary is based on surveys performed from 2/4/86 through 7/6/88.

4.47.5 This area is not considered to have potential for personnel contamination.

4.47.6 During operation, this area was outside the Radiological Control Area, but access was changed during preparation for SAFSTOR. To expedite maintenance, this area will probably be released for uncontrolled access, since no radioactivity has been detected.

4.48 Recombiner Vault 4.48.1 This area is considered to be free of removable contamination (measurements show less than 500 dpm/dm 2). ,

4.48.2 Gamma radiation dose rates are less than 0.1 mR/hr except for a field of 0.1 mR/hr at the end of the tunnel toward the original offgas filter. This field is probably due to radiation levels south of the wall.

IC-21 Rev 2 '

August 1998

5/26/89 4.48.3 No beta radiation fields have been found in this area.

4.48.4 This summary is primarily based on a survey performed 6/23/88.

4.48.5 This area is not considered to have personnel contamination potantial, since the area was constructed clean and the offgas treatment equipment was not connected to contaminated plant systems. (The sump pump discharge was connected). No radioactivity has been detected in this area.

4.49 Stack 4.49.1 The normally accessible levels of the stack (+0', +12' and +26'6") are considered to be free of removable contamination (at or below 500 dpm/dm').

4.49.2 Gamma radiation levels are generally less than 0.06 mR/hr on the +12' elevation.

The +26'6" elevation radiation levels are generally less than 0.2 mR/hr, increasing toward a P-trap on the south wall of the area. Contact dose rates on the P-trap are 1 mR/hr. The +0, elevation radiation levels are generally less than 0.5 mR/hr, except toward a hot spot on the elbow of the Drywell Purge line on the south wall of the area. This line has a contact dose rate of 80 mR/hr.

4.49.3 Beta radiation has been detected on the Drywell purge fan and on the floor drains.

4.49.4 This summary is based on various surveys taken during the period of 1/11/86 through 7/8/88.

4.49.5 This area has low potential for personnel contamination from the interior surfaces of the liquid / gaseous portions of the gas scrubber system.

4.49.6 Note that there appears to be a significant amount of radioactive material collected in the Drywell Purge line. The dose rate at the P-trap on the upper elevation is probably the result of material washed down from the upper level of the stack. This trap is supposed to have water in it, but since it is probably dry, any air flow through it may be a source of airborne contamination.

4.50 Generator and Exciter Housing 4.50.1 This area (including the access on both sides, and the catwalk to the control room) is free of removable contamination (less than 500 dpm/dm').

4.50.2 Gamma radiation dose rates are less than 0.2 mR/hr.

4.50.3 No beta radiation fields have been found in this area.

4.50.4 This summary is based on surveys performed from 1/20/86 through 2/8/88.

4.50.5 This area is not considered to have potential for personnel contamination. The area is now outside the Radiological Controlled Area (released on 2/8/88).

IC-22 Rev 2 Suqus99998

5/26/89 4.51 Turbine Enclosure 4.51.1 This area is considered to be generally free of removable contamination (less than 500 dpm/dm2 ). The contamination levels within the protective cover over the turbine control valve mechanism are not available.

4.51.2 Gamma radiation levels range from <0.2 mR/hr (South end of area) to about 1.5 mR/hr (at the grating toward the North end of the area).

4.51.3 No beta fields have been found in this area.

4.51.4 This summary is based on surveys performed from 1/9/86 through 4/27/88.

4.51.5 This area is considered to have low potential for personnel contamination.

4.51.6 Although smear surveys of portions of the turbine surface do not show contamination, the turbine is roped off as a contaminated area, because of the complexity of the piping precluded a completely detailed survey.

4.52 Turbine Laydown Area 4.52.1 This area is considered to be generally free of removable contamination (less than 500 dpm/dm').

4.52.2 Gamma radiation levels are less than 0.2 mR/hr.

4.52.3 No beta fields have been found in this area.

4.52.4 This summary is based on surveys performed from 9/8/86 through 4/27/88.

4.52.5 This area is considered to have low potential for personnel contamination.

4.52.6 The material in this area (' Rail Car' and the I-beams) have low levels of fixed contamination, but do not appear to have significant removable contamination.

4.53 Hot Lab 4.53.1 This area is considered to be reasonably free of removable contamination (less than 500 dpm/dm') with occasional exceptions, such as the sinks (2,000 and 4,000 dpm/dm') and at least one drawer (2,500 dpm/dm').

4.53.2 Gamma radiation dose rates are less than 0.2 mR/hr.

4.53.3 No beta radiation fields have been detected other than the sink in the Northeast corner of the room (12 mrad /hr).

4.53.4 This summary is based on surveys performed from 1/11/06 through 6/30/88.

IC-23 Rev 2 August 1998

i I l

5/26/89 4.53.5 This area is considered to have a low potential for personnel contamination.

4.54 Laundry 4.54.1 This area is considered to be reasonably free of removable contamination (less than  ;

l 1,000 dpm/dm') with the exception of the internal parts of the washers, dryers, and I the exhaust to the ventilation filter. Some contamination has been detected, such l as the inside of No.1 Dryer (5,000 dpm/dm') and an area in front of No. 2 Dryer (1,000 dgm/dm 2),

i 4.54.2 Gamma radiation dose rates are leas than 0.2 mR/hr. l 4.54.3 No beta radiation fields are detected.

4.54.4 This summary is based on surveys performed from 1/11/86 through 6/30/88.

4.54.5 This area is considered to have low potential for personnel contamination.

4.54.6 Note that several bags of slightly contaminated (but laundered) cloth coveralls are l stored in this area. l 4.55 Security Area (Eastern portion of +27 elevation) 4.55.1 This area2 is considered to be free of removable contamination (less than 500 ,

dpm/dm ).

4.55.2 Gamma radiation dose rates are less than 0.2 mR/hr.

4.55.3 No beta radiation fields have been detected, with the exception of the floor drains (less than 0.8 mrad /hr).

4.55.4 This summary is based on surveys performed from 1/11/86 through 7/21/88.

4.55.5 This area is not considered to have significant potential for personnel contamination.

4.56 Access Control (Western portion of RCA at +27' elevation) 4.56.1 This area2 is considered to be free of removable contamination (less than 500 dpm/dm ),

4.56.2 Gamma radiation dose rates are less than 0.2 mR/hr.

l 4.56.3 No beta radiation fields have been detected. Note that there is a potential for beta exposure from the Sr-90 check source mounted on the Turbine Shield wall.

4.56.4 This summary is based on surveys performed from 1/11/86 through 8/3/88.

4.56.5 This area is not considered to have significant potential for personnel contamination.

IC-24 Rev 2 August 1998

l 5/26/89 l

4.57 Control Room and Instrumentation Shop 4.57.1 This area is considered to be free of removable contamination (less than 500 dpm/dm').

4.57.2 Gamma dose rates are less than 0.2 mR/hr.

4.57.3 No beta radiation fields have been detected.

4.57.4 This summary is based on surveys performed from 2/6/86 through 8/8/88. l l

4.57.5 This area is not considered to have potential for personnel contamination.

4.58 Count Room l

l l 4.58.1 This area is considered to be free of removable contamination (less than 500 dpm/dm ),

2 l l

4.58.2 Gamma radiation dose rates are less than 0.2 mR/hr.  ;

4.58.3 No beta radiation fields have been detected.

l l

4.58.4 This information is based on surveys performed from 2/21/86 through 8/19/88.

i 4.58.5 This area is considered to have a low potential for personnel contamination.

4.59 Batteries, Multi-Zone Fan Area and Hot Lab ' Attic' 4.59.1 The portion of the area accessible from the Counting Room is considered to be free 2

of removable contamination (less than 500 dpm/dm ). The portion of the area accessible from the ladder near the Hot Lab door is considered to have low levels of 2

removable contamination (less than 1,000 dgm/dm ).

4.59.2 Gamma radiation dose rates are less than 0.2 mR/hr.

4.59.3 No beta radiation fields have been detected.

4.59.4 This summary is based on surveys performed from 10/7/86 through 8/29/88.

4.59.5 This area is not considered to have significant potential for personnel contamination.

4.60 Suppression Chamber 4.60.1 The suppression chamber is closed with bolted hatchways.

4.60.2 This is a highly contaminated area. Contamination levels average 40,000 dpm/dm' with highest measurements of 100,000 to 2,000,000 dpm/dm' IC-25 Rev 2 August 1998

5/26/89 4.60.3 Gamma radiation levels are on the order of 3 to 11 mR/hr. An I-beam showed a hot spot of 140 mR/hr.

4.60.4 There are a few beta radiation fields of 4 mrad /hr. The I-beam hot spot measured 560 mrad /hr.

4.60.5 This summary is based on surveys performed from 1/27/86 through 3/25/86.

! 4.60.6 This area has a high potential for personnel contamination, for high airborne contamination and undesirable personnel exposures.

5.0 ONSITE RADIOACTIVE WASTE INVENTORY l

l 5.1 As of January,1989, the most significant inventory of radioactive waste is the resin in the resin disposal tank. The contents of the tank is considered to be about 70 Curies (primarily Cs-137), distributed in 196 cubic feet of resin. At the currently estimated rate at ,

which the Spent Fuel Pool Demineralizer resin requires replacement, the tank Cs-137 '

inventory will increase at a rate of about 15 Curies per year (and about 40 cubic feet per  ;

year). i 5.2 The inventory (1/89) of liquid concentrated waste in No. 3 Waste Receiver Tank is about i 2500 gallons of liquid totaling about 1.7 Curies for Cs-137 and Co-60 (approximately 60:40 l for Cs-137:Co-60).  ;

5.3 There are a variety of waste packages onsite, pending shipment. As of January,1989, this l consisted of 27 drums and 12 boxes. One of the drums contains filter cartridges from the l

l original Spent Fuel Pool Filter, and has a contact dose rate of about 1.3 R/hr. Two other drums (containing radwaste system filter cartridges) have contact dose rates of 110 mR/hr and 12 mR/hr. These three drums are stored in the No. 3 High Level Storage Vault. The rest of the packages are stored in the Solid Waste Handling Building. The estimated activity of the drums in the vault is 1.5 Curies, and the estimated activity of the other packages is less than 0.5 Curies.

5.4 These activity estimates are made on the basis of gamma emitting nuclides. A review of I the mixtures used for shipping indicates that the combined quantities of other nuclides (H- )

3, C-14, Ni-63 and Sr-90) will normally be less than the combined activity of Cs-137 and l Co-60 (except that after about 5 years, Ni-63 may be about equal to Co-60).

6.0 ENVIRONMENTAL RADIOACTIVITY LEVELS 6.1 The Annual Environmental Monitoring Report and the SAFSTOR Environmental Report (Attachment 6 to the SAFSTOR license amendment request) provide comprehensive documentation of environmental radioactivity levels.

6.2 A concrete ' pad' adjacent to the Unit 2/3 fencelir.e was found to have fixed contamination (Cs-137). Further surveys / sampling will be required to define the extent of the j contamination.

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Appendix 11 l SAFSTOR Operator Training and Certification Program 1

ll.1 INTRODUCTION This program describes the training and certification for Supervisors and Operators associated with the maintenance at Humboldt Bay Power Plant Unit 3 in the SAFSTOR mode consistent with its possession-only license. l ll.2 APPLICABILITY The Unit 3 Technical Specifications require that certain operations associated with the maintenance and handling of reactor spent fuel be performed by or under the supervision of persons certified by the Plant Manager or his delegate. The following members of the plant staff (as a minimum) shall be certified in accordance with this program:

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j o Plant Manager  ;

o ' Power Plant Engineer ,

o Supervisor of Operations

• Shift Foremen ,

• Selected operators who shall be performing duties requiring certified operators

• Training Coordinator j

! 11.3 INITIAL CERTIFICATION -

! Certification candidates shall participate in a training program covering the following topic areas:

• Reactor Theory (as applicable to the storage and handling of spent reactor fuel) ,

• Spent Fuel Handling and Storage Equipment - Design and Operating Characteristics

• Monitoring and Control Systems

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• Radiation Protection

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• Normal and Emergency Procedures

• Administrative Controls applicable during the SAFSTOR period ReectorTheory training will include characteristics of the stored spent fuel, subcritical multiplication, factors affecting reactivity and criticality, and the basis for fuel handling I

restrictions and procedures.

The design and operating characteristics will include training in the functions and use of fuel handling tools, cranes, the spent fuel storage pool, and pool service systems and equipment.

Prior to shipments of spent fuel this training will include shipping casks, cask handling equipment, and procedures, ll-1 Rev 2 l August 1998

Monitoring and Control Systems will include training on the spent fuel pool level monitoring systems, criticality monitors, and Unit 3 Area Radiation Monitors.

Radiation protection training will include theory of radioactive emissions, control of radiation exposure, use of radiation detection and monitoring equipment, protective clothing and respiratory protection, and contamination control procedures. Training will emphasize the principles and practices associated with maintaining exposures as low as reasonably achievable (ALARA).

Normal and Emergency Procedure Training will include the Emergency Plan and any operations  ;

and emergency procedures associated with the operation of Unit 3 systems and equipment during SAFSTOR. This area shall also include training in the handling and processing of 1 radioactive wastes, l Administrative Control Training will include the Unit 3 Technical Specifications, Security Plan, Quality Assurance Plan and plant administrative procedures associated with the operation, i s surveillance, and mainten'ance of Unit 3. I Plant training programs will be implemented primarily through classroom instruction. Training I aids such as videotapes or films may be used. However, an instructor will participate in at least 50 percent of a lecture series. Supervised self-study and on-the job training may be used to supplement classroom training.  ;

A quantitative measure of the portion of training that can be attributed to'each training method cannot be made. General employee training will be conducted primarily utilizing classroom '

training. Due to the small staff and anticipated low turnover, initial fuel handler certification i training programs will be tailored to the needs of the individual, initial certification training will l l be approximately 50 percent classroom instruction or supervised self-study with an instructor  !

l available to monitor progress and answer questions. The remainder of the initial certification training program will be on-the-job training. Proficiency training for certified fuel handlers will be accomplished through a preplanned lecture series which will be primarily classroom training and on-the-job self-study assignments. Adequacy of the training methods will be evaluated through the use of annual written and oral examine.tions.

l Satisfactory completion of the training shall be based on passing of a comprehensive written l l examination including each of the above areas and an oral examination. Minimum passing grade for the written examination shall be 70 percent in each area and 80 percent overall. The oral examination shall be administered by a member of the plant management staff. Results of the oral examination shall be on a pass / fail basi If, during initial certification, a weakness is j noted which did not result in the individual failing either the written or oral examination, the l

individual will be given remedial training. The period of remedial training shall not exceed 3 months, and during that period the individual will not be permitted to perform duties requiring fuel handic certification. Satisfactory completion of the remedial training will be evaluated

! through an oral examination of the weak area given by a member of the plant management staff

! other than the original examiner.

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. Examinations shall be conducted to evaluate an individual's overall depth of knowledge and l understanding of the plant and of the possession-only license requirements, ll-2 Rev 2 August 1998

11.4 PROFICIENCY TRAINING AND TESTING The frequency of proficiency training shall be such that all six of the topics discussed in the '

SAFSTOR Decommissioning Plan are covered in a 2-year period. If, in the course of the annual training,- an infrequently performed activity or procedure is planned for which training i should be conducted, it may be moved from its regularly scheduled time to accommodate the  ;

present circumstances. This training will be in addition to any previously scheduled topic. ,

Additionally, any significant areas of weakness noted on the annual exams shall be given priority in the training schedule.

Annual examinations shall be used to demonstrate the proficiency of certified personnel. l Examinations will be similar to but not as comprehensive as the initial certification examinations. i Minimum passing grade for proficiency examinations shall be 70 percent in each section and 80 percent overall. Oral examinations shall be on a pass / fail basis.  ;

ll.5 CERTIFICATION I

Upon successful completion of the initial certification training program, the Plant Manager or his '

delegate shall certify the individual as a Certified Fuel Handler. Normally an employee will complete the initial certification within one year after entering the program. After initial certification, personnel will be recertified every 2 years based on the successful completion of l

the Proficiency Training and Testing Program.

Initial certification written and oral examinations will cover each of the six areas included in the initial certification program. Emphasis on these examinations will be equally divided among all six areas. )

Recertification testing will cover all six ereas of the initial certification. The emphasis will be l l placed on topics taught since the last annual examination and on current conditions at the plant.  !

I II.6 PHYSICAL REQUIREMENTS As a prerequisite to acceptance into the training program and for recertification, a candidate

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must successfully pass a medical examination designed to ensure that the candidate is in  ;

I generally good health and is otherwise physically qualified to safely perform the assigned work. l Minor correctable health deficiencies, such as eyesight or hearing, will not oer se prevent certification. }

The medical examination will meet or exceed the requirements of ANSI Standard N546-1976, l "American National Standard - Medical Certification and Monitoring of Personnel Requiring Operator License for Nuclear Power Plants." l l I l  !

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I 11.7 DOCUMENTATION Initial Certification and Proficiency Training shall be documented and maintained for certified  !

personnel for a minimum of 5 yt,ars. The records shall include the dates and periods of j training, results of all quizzes and examinations, copies of written examinations, oral examination records, and information on results of physical examinations.

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