ML20203E629

Trojan Nuclear Plant Decommissioning Plan

ML20203E629
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	02/28/1998
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20203E612	List: ML20203E608 ML20203E629
References
PGE-1061, PGE-1061-R04, PGE-1061-R4, NUDOCS 9802270097
Download: ML20203E629 (122)

Text

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TROJAN NUCLEAR PLANT POE 1061, " Trojan Nuclear Plant Decommissioning Plan" Revision 4 The following information is provided as a guide for the insertion of new sheets for Revision 4 into PGE 1%1, " Trojan Nuclear Plant Decommissioning Plan."

Revised pages are to be inserted as follows:

Emanave Page Insert Page hble of Ca=**n**

? age il Page 11

Pages iv through vil Pages iv through vil Pages xill and xiv Pages xill and xiv l

Secelan 1r Pages 13 through 1-6 Pages 13 through 16 Secelan 2:

Pages 2 2 through 211 Pages 2 2 through 211 Pages 2-13 through 2 29 Pages 213 through 2-29 hp Pages 2-37 and 2 38 Pages 2 37 and 2-38 Table 2.2-1 Table 2.21 Table 2.2-2, Sheets 1 - 3 Table 2.2-2, Sheets 1 - 3 j Table 2.2-4 Tc.ble 2.2-4 Table 2.2 5, Sheet 1 Table 2.2 5, Sheets 1 and 2 Figures 210 and 211 Figures 2-10 and 2-11 Secitan 3!

Pages 3 9 through 3-16 Pages 3 9 through 316 Pages 3-39 through 3 79 Pages 3 39 through 3-78 '

Table 3.1-6, Sheets 1 and 2 Table 3.1-6, Sheets 1 - 3 Secttan 5:

Pages 5-1 through 5 9 Pages 5-1 through 5 9 Table 5.1-1 Table 5.1 1 Table 5.1-2, Sheets 1 and 2 Table 5.12, Sheets 1 and 2 Table 5.31 Table 5.31 Table 5.3 2, Sheets 1 and 2 Table 5.3 2, Sheets 1 and 2 Table 5.3-3, Sheets 1 and 2 Table 5.3-3, Sheets 1 and 2 Table 5.3-4, Sheets 1 and 2 Table 5.3-4, Sheets 1 and 2 O

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e w m ass. ' PDR j

TROMN DECOAISilSS10NING PLAN 2.2.4.3 d General Decontamination and Dismantlement Considerations . . . . . . . 2-6 2.2.4.4 Qgcontamination hiethods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 2.2.4.5 Dism antlement hiethods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 2.2.4.6 Removal Seauence and hinterial Handling . . . . . . . . . . . . . . . . . . . . . . 2 10 2.2.4.7 System Deactivation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 12 2.2.4.8 Temnorary Systems to Support Decommissioning . . . . . . . . . . . . . . . . 2-12 2.2.5 DECONTAMINATION AND DISMANTLEMENT: SYSTEMS, STRUCTURES, AND COMPONENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 13 2.2.5.1 D verv i e w . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 3 2.2.5.2 Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 13 2.2.5.3 Reactor Vessel Internals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 14 2.2.5.4 R eactor Ve s s el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 16 2.2.5.5 S t e am G en e rators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 18 2.2.5.6 R eactor Coolant Pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 18 2.2.5.7 Pressurizer and Pressurizer Relief Tank . . . . . . . . . . . . . . . . . . . . . . . . 2 18 2.2.5.8 Chemical and Volume Control System . . . . . . . . . . . . . . . . . . . . . . . . . 2 19 2.2.5.9 Safets Iniection System ....................................220 2.2.5.10 Residual Heat Removal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 20 2.2.5.11 Centainment Sorav Svstem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 21 2.2.5.12 Comnonent Cooling Water System ...........................,221 2.2.5 13 Service Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 22 A

2.2.5.14 Soent Fuel Pool and Fuel Handline Eouloment . . . . . . . . . . . . . . . . . . 2 22 d 2.2.5.15 Soent Fuel Pool Cooling and Deminerali7er System . . . . . . . . . . . . . . 2 23 2.2.5.16 Condensate Deminerali7ers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 23 2.2.5.17 Steam Generator Blowdown Syrtem . . . . . . . . . . . . . . . . . . . . . . . . . . 2 24 2.2.5.18 Primary hinkeup Water System and Refueline Water Storage Tank . . 2 24 2.2.5.19 Elant E ffluent System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 24 2.2.5.20 Containment Ventilation Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 25 2.2.5.21 Hydrogen Recombiners ...................................225 2.2.5.22 Fuel Building and Auxiliary Building Ventilation Sys* cms . . . . . . . . . 2 26 .

2.2.5.23 Condensate Demineralizer Building Ventilation System . . . . . . . . . . . 2-26 2.2.5.24 Instrument andlervice Air System . . . . . . . . ..................227 2.2.5.25 Gaseous Radioactive Waste System . . . . . . . . . . . . . . . . . . . . , , . . . . . 2 27 2.2.5.26 Solid Radioactive Waste System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 27 2.2.5.27 Clean Radioactive Waste System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 28 2.2.5.28 Dirty Radioactive Waste System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 29 2.2.5.29 Radiation Af onitorine System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 29 2.2.5.30 Etocess Samoling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 30 2.2.5.31 Fire Protection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 31 2.2.5.3 2 Electrical Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-31 2.2.5.33 Containment B uildin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 32

/3 il Revision 4 O

TRO.1AN DECOMMISS10NINri PLAN f]

v 3.1.2.4.6 Summary of Environmental Results . . . . . . . . . . . . . . . . . . 3 15 3.1.3 C O N C L U S I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 15 Appendix 3.1 A, Summary of Notable Radiological Contamination Events . . . . . . . . . . . . 3 17 Appendix 3.1 B, Summary of Structural Surwy Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 22 3.2 RADI ATION PROTECTION PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 37 3.

2.1 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 7 3.2.2 RADI ATION PROTECTION OBJECTIVE . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 37 3.2.3 RADIAT10N PROTECTION AND ALARA PROGRAM POLICIES . . . . . . . 3 37 3.2.4 RADI ATION PROTECTION ORGANIZATION . . . . . . . . . . . . . . . . . . . . . . 3 38 3.2.5 MANAGEMENT RESPONSIBILITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 38 3.2.5.1 General Manager. Troian Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 8 3.2.5.2 Manager. Personnel / Radiation Protectio 5 ......................338 3.2.5.3 Engineering M anagement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 39 3.2.5.4 Managers and Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 39 3.2.6 RADI ATION PROTECTION PROGRAM IMPLEMENTATION . . . . . . . . . . 3 39 3.2.6.1 Radiatiotdrotection Equipment and Instrumentation . . . . . . . . . . . . . 3 39 3.2.6.1.1 Laboratory Radiation Protection Instrumentation . . . . . . . 3-40 3.2.6.1.2 Portable Radiation Detection Instrumentation . . . . . . . . . . 3-40 3.2.6.1.3 Portable Air Samp..ag instrumentation . . . . . . . . . . . . . . . 3-40 3.2.6.1.4 Personnel Radiation Monitoring Instrumentation . . . . . . . . 3-41 3.2.6.1.5 Area Radiation Monitoring Instrumentation . . . . . . . . . . . . 3-41

(

O) 3.2.6.2 Control of Radiation Exposure to the Publie . . . . . . . . . . . . . . . . . . . . 3-42 3.2.6.2.1 Radiological Effluent Monitoring . . . . . . . . . . . . . . . . . . . . 3-42 3.2.6.2.2 Radiological Environmental Monitoring . . . . . . . . . . . . . . 3-42 3.2,6.3 Control of Personnel Radiation Exposure . . . . . . . . . . . . . . . . . . . . . . . 3-43 3.2.6.3.1 S hi eld i ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-4 3 3.2.6.3.2 Access Control and Area Designations . . . . . . . . . . . . . . . . 3-43 3.2.6.3.3 Facility Contamination Control . . . . . . . . . . . . . . . . . . . . . 3-43 3.2.6.3.4 Personnel Contamination Control . . . . . . . . . . . . . . . . . . . . 3 44 3.2.6.3.5 Area S urveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -44 3.2.6.3.6 Personnel Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-45 3.2.6.3.7 Radiation Work Permits . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-45 3.2.6.3.8 Trai ni n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -4 f 3.2.6.3.9 Controls. Practices, and Special Techniques . . . . . . . . . . . 3-46 3.2.6.3.10 Radioactive Materials Safety . . . . . . . . . . . . . . . . . . . . . . . 3-46 3.3 RADIOACTIVE WASTE MANAGEMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-47 3.3.1 SPENT FUEL MANAGEMENT PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . 3-47 3,3.1.1 Sp.cnt Fuel Management Program Descriotion . . . . . . . . . . . . . . . . . . . 3-47 O

i iv Revision 4

TROJAN DECOMMISSIONING PLAN 3.3.1.2 Effects of Permanent Renositon Schedule on Soent Fuel hianagement Pl an . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-4 8 3.3.1.3 Licensing Activities to Support the Scent Fuel hianagement Plan . . . 3 49 3.3.2 RADIOACTIVE WASTE PROCESSING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 49 3.3.2.1 Daseous Radioactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-49 3.3.2.2 Liould Radioactive Wasts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 49 3.3.2.3 Solid Radioactive Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 50 3.3.2.4 hii x ed Wa st e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 0 3.3.3 RADIOACTIVE WASTE DISPOSAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 51 3.4 E V ENT AN A L Y S I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 3 3.4.1 O V E RV I E W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 3 3.

4.2 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 53 3.4.3 LIhflTS AND ASSUh1PTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 S$

3.4.3.1 Radionuclide Release Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 55 3.4.3.2 As s u m nt ion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 $ 6 3.4.4 RADIOLOGICAL EVENT IDENTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . 3 57 3.4.4.1 Decontamination Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 8 3.4.4.1.1 in Situ Decontamination of Systems . . . . . . . . . . . . . . . . . . 3 59 3.4.4.1.2 Surface Cler 4 Techniques . . . . . . . . . . . . . . . . . . . . . . . 3 61 3.4.4.2 Dismantlem ent Event s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 63 3.4.4.2.1 Segmentation of Components or Structures . . . . . . . . . . . . 3 63 3.4.4.2.2 Removal of Concrete . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 64 3.4.4.3 hinterial liandling Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 65 3.4.4.3.1 Dropping of Contaminated Components . . . . . . . . . . . . . . 3 66 3.4.4.3.2 Dropping of Concrete Rubble . . . . . . . . . . . . . . . . . . . . . . . 3 66 3.4.4.3.3 Dropping of Filters or Packages of Particulate hiaterial . . . 3 67 3.4.4.4 Less of Suonort System Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 67 3.4.4.4.1 Loss of Offsite Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 68 3.4.4.4.2 Loss of Cooling Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 3 68 34.4.4.3 Loss of Compressed Air . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 69 3.4.4.5 F ire E ve n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 6 9 3.4.4.6 Ex olosion Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 70 3.4.4.7 E xt emal E ve n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 71 3.4.4.7.1 E arthquake . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 71 3.4.4.7.2 F l ood in g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 72 3.4.4.7.3 Tornadoes and Extreme Windr . . . . . . . . . . . . . . . . . . . . . 3 73 3.4.4.7.4 Volcanic Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-73 3.4.4.7.5 Li ghtnin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 74 3.4.4.7.6 Toxic Chemical Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 74 3.4.5 RADIOLOGICAL OCCUPATIONAL SAFETY . . . . . . . . . . . . . . . . . . . . . . . . 3 74 nJ i y Rev..ision 4

TROL 4NDECOMMISSIONING PL4N 3.4.6 OFFSITE RADIOLOGICAL EVENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 75 .

3.4.7 NONRADIOLOGICAL EVENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 75 l i '

3.5 OCC UPATIONAL S AFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 76 l i

l 3.6 NONRADIOACTIVE WASTE MANAGEMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 76 3.C .1 A S B ES TO S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 7 3.6.2 POLYCllLORINATED BIPilENYLS (PCB) . . . . . . . . . . . . . . . . . . . . . . . . . . 3 77

- 3.6.3 M E RCU RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 7 i

3.6.4 LEA D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 7 3.6.5 OTilER PLANT WASTE MATERI ALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 78

4. PROPOSED FINAL RADI ATION SURVEY PLAN . . . . . . . . . . . . . . . . . . . . . . . . . 4 1 4.1 INTROD U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1

,1 J

4.2 FIN AL R FI E AS E CRITERI A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 [

4.2.1 LIMITS FOR LOOSE AND FIXED SURFACE CONTAMINATION . . . . . . . . 4 2 4.2.2 LIMITS FOR DIRECT EXPOSURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 4.2.3 LIMITS FOR TOTAL CONCENTRATIONS IN SOIL AND WATER . . . . . . . 4 2 I .

4.2.4 LIMITS FOR UNRESTRICTED RELEASE OF MATERIAL . . . . . . . . . . . . . . 4 3 4.3 PLANNING AND DESIGNING Tile FINAL SURVEY . . . . . . . . . . . . . . . . . . . . . . . . 4-4

+

4.3.1 QUA LITY ASSURANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 4.3.2 Tral N I N O . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 5 4.3.3 INSTRUM ENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . #

• i 4.3.4 DOC UM ENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

• 6

5. DECOMMISSIONING COST ESTIMATE AND FUNDING PLAN . . . . . . . . . . . . 51 i

5.1 DECOMMISSIONING COST ESTIMATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 -

5.1.1 COST ESTIM ATE RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 5.1.2 COST ESTIMATE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2 i 5.1.2.1 NRC (Radiologicah Decomm!ssinning Costs . . . . . . . . . . . . . . . . . . . . 5 2 5.1.2.2 Nom adiological Decommintioning Costs . . . . . . . . . . . . . . . . . . . . s . . . 5-4 5.1.2.3 Spent Fuel Management Costs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5 5.1.2,4 Financial Activity Costs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4 y vi Revision 4 l

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TROMN DECOh!AflSSIONING PLAN na (JA $.2 SPENT FUEL hiANAGEhiENT FUNDING PLAN . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5 53 DECOMMISSIONING FUNDING PL AN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6 5.3.1 CURRENT DECOMhilSSIONING FUNDING CAPABILITIES . . . . . . . . . . . . 5 6 5.3.2 T(P CG 0WNEFS' DECOMMISSIONING FUNDING PLANS . . . . . . . . . . . . 5 7 5.3.7.1 P nE Fund i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 7 5.3.2.2 kWE B /B PA Fundirig . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 8 5.3.2.3 ITAL F und i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 9

6. TF.CliNf CAh @ECIFICATIONS ANIl ENVIRONMENTAL PROTECTION PLAN..............................................................61 6.1 TECliNICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1 6.2 ENVIRONhiENTAL PROTECTION PL AN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3 QUALITY ASSURANCE PROVISIONS . . . . . . . . . . . . . ...................71 7.1 OUA LITY A M ORANC E PROG RA M . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1

8. PHYSICAL SECURITY PLAN PROVISIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 b

V 8.1 j'IlYSIC AL SECURITY PL AN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.2 FITNESS FOR DUTY PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2

9. FIRE PROTECTION PROG RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 1 9.1 FI RE P ROTE CTION PL AN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 1 APPENDIX A STATE OF OREGON ORDER APPROVING TROJAN DECOMMISSIONING PLAN PLAN REVIEW REPORT fd vil Revision 4

TROMNDECOMMISSION!NG PLAN LIST OF EFFECTIVE PAGES Section/Pase Rtrhed. Rats Title Page June 1996 Page i December 1996 Page li February 1998 Page 111 June 1997 Pages iv through vil February 1998 P;ges vill and ix June 1996 i _ Page x December 1996 Pages xi through xil June 1996 Page xiil and xiv February 1998 Section 1 i Page1 1 June 1997  ;

, Page 12 December 1996  ;

Pages 1-3 through 16 February 1998 Page1 7 December 1996 Section 2 Page 21 December 1996 Pages 2 2 through 211 February 1998 Page 2-12 _

June 1997 Pages 2-13 through 2 29 February 1998 Pages 2 30 through 2 36 June 1997 Pages 2-37 and 2 38 February 1998 Pages 2 39 through 2 43 June 1997 Table 2.21 and 2.2 2 February 1998 '

Table 2.2-3 June 1996 Table 2.2-4 February 1998

+

Table 2.2 5 February 1998 Figure 21 June 1997 Figures 2 2 through 2-9 June 1996 '

Figures 210 and 211 February 1998 Section 3 Pages 31 through 3 8 June 1996 Pages 3 9 through 3-16 February 1998 Pages 317 through 3 37 . June 1996 Page 3-38 December 1996 Pages 3 39 through 3 78 February 1998 xiii Revision 4

TROJAN DECOMMISS10NINO PL4N LIST OF EFFECTIVE PAGES Section/Page Resised Date Tables 3.1 1 through 3.1-4 June 1996 -

Table 3.15 Pages 1 of 8 through 7 of 8 June 1996 Page 8 of 8 December 1996 Table 116 February 1998 Tables 3.17 through 3.14 June 1996 :

Figures 31 through 3 39 June 1996 Section 4 June 1996 Pages 41 through 4 6 . June 1996 Table 4.21 June 1996 Section 5 Pages 51 through 5 9 February 1998 Tables 5.1 1 and 5.12 February 1998 Tables 5.31 through 53-4 February 1998 Section 6 Pages 6-1 through 6 3 June 1996 Section 7 Page 71 June 1996 O' Section 8 Pages 81 and 8 2 June 1996 Section 9 Page 91 December 1996 Appendix A June 1996 Appendix B June 1996 xiv Revision 4 h

a' TRO).4NDECOMMISSIONING PL4N i 6, An updated site specific estimate of remaining decommissioning costs; and

7. A supplement to the environmental report, pursuant to f 51.53, describing any new  !'

information or signliicant environmental change associated with the licensee's proposed temiination activities.

l

- 1.2 MAJOR TASKS. SC11RDULES AND ACTIVITIFR

}

TNP decommissioning is divided into two broad periods; transition, and decontamination and

. dismantlement. Decommissioning will be followed by site restoration. This section provides a ,

- brief description of these activities. Details are provided in Section 2.2. I 1.

2.1 DESCRIPTION

OF MAJOR ACTIVITIES The Transition Period began with permanent plant shutdown in January 1993 and will continue until spent fuel is transferred to an ISFSI. The Decontamination and Dismantlement Period will begin once the spent fuel is transferred to the ISFSi. Site restoration will begin following termination of I the Part 50 license and involves the final disposition of structures, systems, and components.

Throughout TNP decommissioning, POE may use contractors to provide specialized services or to supplement the facility staff when necessary. POE will administer the tasks and oversee contractors as delineated in Section 2.5.

Storing fuel at TNP during and after plant decommissioning significantly impacts both the process and costs associated with decommissioning. The TNP contract with DOE, " Standard Contract for Disposal of Spent Nuclear Fuel and/or liigh Level Radioactive Waste " stipulates services provided by DOE shall begin not later than January al 1998. This contract clause provides the basis for the  ;

- schedule forecast in DOE's annual acceptance priority ranking for receipt of spent fuel and/or high level waste. The published schedule specifies the first TNP shipment to be in 2002 (the fifth year of- l repository operation, assuming initial operation in 1998), and the final shipment is projected for 2018. Recognizing the uncertainty, but with no better formal estimate, the contract dates for fuel L shipment are currently being used for planning purposes. The spent fuel management plan is discussed in Section 3.3.

1.2.1.1 TramitionPeriod During this period, POE will maintain systems and components required to support decommissioning and spent fuel storage in accordance with the Facility Operating (PNsession Only) License NPF 1 and administrative procedures. Activities anticipated include assessing the functional requirements for systems, structures, and components; deactivating systems, structures, ,

and components; some dismantling, including removal of steam generators, pressurizer, and other 1-3 Revision 4 ,

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TROJAN DECOAfAflSS10NING PLAN Q'O systems / components not necessary for assuring safe spent fuel storage in the spent fuel pool; and maintaining safe spent fuel storage. PGE will also conduct detailed deccmmissioning project planning, prepare engineering specifications and procedures, procure special equipment needed to support decommissioning, and 70tiate senice contracts required for decommissioning activities.

Major activities scheduled to occur following Decommissioning Plan approval during the Transition Period include system deactivations and system / component removal not necessary for safe storage of fuel; limited decontamination of structures; reactor vessel intemals removal or I reactor vessel with intemals intact removal; decontamination and dismantlement planning; I licensing and constructing an ISFSl; and transferring the fuel from the spent fuel pool to the ISFSt.

Spent fuel will be transferred from the spent fuel pool to facilitate decontamination and dismantlement. Once the spent fuel is transferred to the ISFSI, the Transition Period ends and the Decontamination and Dismantlement Period begins.

1.2,1,2 Decontamination and Dismantlement 1

Major activities planned during the Decontamination and Dismantlement Period include j removing remaining contaminated systems and components, decontaminating structures, and a final radiation survey to verify radioactivity has been reduced to sufficiently low levels to allow for unrestricted release of the site.

Contaminated systems, components, and structural material will be decontaminated or removed and packaged. The packaged material will either be shipped to an oft site processing facility, shipped directly to a low level radioactive waste disposal facility, or otherwise handled in accordance with applicable regulations.

Decontaminating plant structures may be completed concurrent with removing equipment and systems and may include the use of a variety of techniques ranging from water washing to surface material removal. Demolishing certain buildings may be necessary based on degraded structural integrity as a result of decontamination efforts and/or removal of s3 stems and components, surrounding walls, or other barriers.

A final radiation sun'ey will be performed to determine the final condition of the site afler decontamination activities are complete. The purpose of the final radiation survey is to demonstrate radiological conditions at TNP are within the final site release criteria to support license temiination. Upon completing the final survey, a final survey report will be submitted to the NRC.

O Q 1-4 Revision 4

. - . - - -- __- -_ . ~____..

TROJ.4N DECOAfAflSSIONING PL4N 1.2.1.3 Site Restoration Nonradiological site remediation activities are scheduled to be completed following termination 3 of the Facility Operating (Possession Only) License NPF 1. Nonradiological site remediation activities are scheduled to begin around 2018 and conclude in 2019. Some site restoration activities will be completed during the Transition and Decontamination and Dismantlement l Periods of decommissioning. I 1.2.2 FINAL RELEASE CRITERIA TNP decommissioning will safely reduce radioactivity at the site to acceptable levels thereby allowing release of the site for unrestri ted use. Release criteria are discussed in Section 4.2, 1.2.3 SCllEDULE FOR DECOMMISSIONING / SITE RESTORATION ACTIVITIES A detailed schedule for decommissioning / site restoration activities is presented in Section 2.2.

The following is an overview of the current TNP decommissioning / site restoration project schedule.

January 1993 Late 1999 Transition Period i Late 1994 - Late 1995 Large Component Removal Project O Late 1996 - Late 1998 Complete planning / building an ISFSI I V Early 1997 - Late 1999 Reactor Vessel and Internals Removal l Early 1999 Late 1999 Transfer spent nuclear fuel to the ISFSI I Late 1999 - Late 2002 Decontamination and Dismantlement Period i Mid 2000 Submit application for tennination of the license I Late 2002 Complete final radiation survey I Late 2002 - Mid 2018 Caretaking l Mid 2016 - Late 2019 Demolish buildings 1-5 Revision 4

)

TROJAN DECOAIAflSSIONING PL4N l.3 DECOMhilSSIONING COST ESTihiATE AND FUNDING PLAN SUhiMARY nis section provides a summary of the final estimated costs and funding methodology for TNP l decommissioning, spent fuel management, and nonradiological decommissioning activities. The decommissioning cost estimate and the co-owners' plan for assuring availability of funds for completing these activities is provided in Section 5.

1.3.1 COST ESTihiATE SUMhiARY TNP decommissioning, spent fuel management, and nonradiological decommissioning costs are estimated at approximately $458.9 million in 1997 dollars. Costs are derived from a cost I estimate prepared by POE and TLO Services, Inc. Table 5.1 1 prnvides an itemized summary of costs.

1.3.2 FUNDING PLAN SUhth1ARY 1.3.2.1 Cunent Decommissioning Funding Canabilities Co owners separately collect and 'naintain funds for decommissioning. Funds are collected through rates and deposited to extemal trust funds. Because TNP was shutdown prematurely, the

()

V external trust funds contain only a portion of the total amount needed for decommissioning. As of December 31,1996, approximately $99 million were in the funds. Table 5.31 summarizes the status of the decommissioning trust fi 's.

l 1.3.2.2 Decommissioning Funding Plans The decommissioning trust fund cash flow and funding plan for each of the co owners are presented in Tables 5.3 2,5.3 3, and 5.3-4. The co owner funding information incorporates tmst fund contribution schedules and necessary financial assurance and bridging funds. The trust fund contribution schedu!cs are based on funding requirements for both radiological and nonradiological decommissioning costs, as well as financing costs and specific spent fuel management costs including planning, design, construction, operations and maintenance (O&M),

and decommissioning of an ISFSI.

As indicated above, decommissioning trusts will not initially contain the funds necessary to complete radiological decommissioning prior to the start of the Decontamination and Dismantlement Period. Therefore, prior to commencing this period, each co owner will secure I and maintain a financial assurance mechanism in accordance with 10 CFR 50.75, " Reporting and recordkeeping for decommissioning planning " Additional funding plan details are provided in i Section 5.

> l-6 Revision 4

! f TROJANDECOkthilSS10NINGl'L4N

' 2.2 DECOMMISSIONING ACTIVITIER.TASKE. AND SCIEDUI FR

[ 2.

2.1 INTRODUCTION

i This section presents a description of activities and tasks associated with 'INP decommissioning. 1 Included is a schedule for implementation of decommissioning activities, estimates of associated  ;

[ occupational radiation dose, and pmjected volumes of radioactive waste. The information

presented reflects initial planning for decommissioning activities. Detailed planning will precede initiation of these activities, and will include engineering design, as low as is reasonably ,,

1 achievable (ALARA) planning, and refinement of the cost, schedule, and required resources.' ,

2.2.2 ScilEDULE OF DECOMMISSIONING / SITE RESTORATION ACTIVITIES i, A decommissioning / site restoration schedule is presented in Figure 211. This schedule was 4

. used in the preparation of the decommissioning cost estimate discussed in Section 5. PGE will perform task specific schedulhg as part of the detailed planning for decontamination and ,

l dismantlement, t l TNP decommissioning can be divided into two broad periods:

n

1. Transition period; and i t

E

2. Decontamination and dismantlement period.

! Nonradiological site restoration activities, involving the final disposition of structures, systems, l ' and components, are scheduled to be completed following the termination of Facility Operating  :

. (Possession Only) License NPF 1. Some site restoration activities will be completed during the 1 Transition and Decontamination and Dismantlement Periods. I ,

j Major activities planned during the Transition Period include: 1 4

1. :- Lay up ofplant systems;

2. Removal of the steam generators and pressurizer;

, t 3.. Removal of the reactor vessel internals or reactor vessel with internals intact; - I

• 4._ Development of specific work plans and procedures; I

5. - Licensing and construction of an ISFSl; and -

2 22 Revision 4 {

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TROJ.4N DECOMkflSS10NING PL4N

6. Maintaining safe storage of spent fuel.

7. Dismantlement and removal of some systems / components not necessary for assuring safe spent fuel storage in the spent fuel pool.

The Transition Period began with permanent plant closure in January 1993 and will continue until the spent fuel is transferred to the ISfSt. Decontamination and dismantlement of the remaining facility radioactive systems, components, and structures are scheduled to be conducted upon completion of the transfer of spent fuel to the ISFSI. For planning purposes,it was assumed that spent fuel and high level waste will be stored in the spent fuel pool until 1999, at I which time it will be transferred to the ISFSt. Some dismantlement activities may occur during the Transition Period.

Major activities planned during the Decontamination and Dismantlement Period include:

1. Removal of remaining contaminated systems and components;

2. Decontamination of structures; and

3. Final radiation sun'ey.

]v Major decontamination and dismantlement activities are expected to last from late 1999 to the end of 2002. The fmal radiation survey will be conducted following decontamination and I

I dismantlement.

Some activities, such as d: tailed planning, may continue through more than one period.

2.2.3 TRANSITION PERIOD Plant closure activities were initiated following the decision to permanently cease TNP power operations in January 1993. These activities culminated with the plant in a safe transition state awaiting decontamination and dismantlement. Detailed project planning and engineering activities for the Decontamination and Dismantlement Period, as discussed in Section 2.2.4.2, will begin during the Transition Period. Plant activities will continue to be implemented in compliance with the existing possession only license and other regulatory requirements.

Removal of the four steam generators and pressurizer has been completed. These components were disposed of at the US Ecology radioactive waste disposal facility near Richland, Washington. Removal of the steam generators and pressurizer were accomplished through a new opening in the south face of the Containment Building. That opening is equipped with a door ao that the Containment Building can be maintained in a closed condition except during active

/

U 23 Revision 4

TROJAN DECOAfAf'SSIONING PLAN component removal. Prior to removal of the components from the Containmen: Building, low-density cellular concrete was placed inside each steam generator and the pressurizer. The concrete fixed intemal contaminatica and provided additional shielding.

Each component was moved via nn internal rail system out of the Containment b.ailding, loaded by a gantry crane onto a multi wheeled transporter, moved to a preparation area within the TNP Industrial Area, and prepared for river barge shipment to the Port of Benton, Washington The component, transport cradle assembly, and transporter was then moved as an integral unit by -

river barge from 'he TNP barge siip to Benton, Washington, on the Columbia River. The multi-wheeled transpotter was used to off load the barge and move the component to the US Ecology facility on the Hanford reservation.

l Piping systems that were opened during component removal were closed and/or isolated as appropriate. The Containment Building door will be controlled in accordance with the TNP security plan.

The reactor vessel may be removed with the intemals intact. If performed, removal of the reactor I vessel and internals will commence after required approvals are received and is scheduled to be I completed in 1999. I Alternatively, the reactor vessel internals may be removed separately. If performed, removal of I the internals is scheduled to be completed in 2000. Segmentation of the components will be I performed underwater in the reactor cavity. Segmented components that are classified as greater.

than Class C waste in accordance with 10 CFR 61, " Licensing requirements for land disposal of radioactive. waste," will be transferred to the spent fuel pool for storage pending disposel. .

Additional activities that were completed or are in process during the Transition Period include, but are not limited to the following: 4

1. Assessment of the functional requirements for plant systena structures, and components.

Plant systems, structures, and components needed to support safe storage of the spent fuel, support snent fuel pool cooling, and facilitate ongoing plant-activities have been identified.

2. . - Deactivation / removal of plant systems, structures, and components.

Systems, structures, and components not required to support safe storage of the spent fuel, support spent fuel pool cooling, or support of ongoing plant activities may be deactivated, drained, and removed as O 2-4 Revision 4 l

TROJAN DECOAfAllSSIONING Pl.AN

{n} appropriate. A comprehensive plant lay up program was developed and implemented at described in Section 2.2.4.7. Systems, structure,, n.d components may be decontaminated, decommissioned andar removed during the transition period provided that the cost schedules of Section 5 are adhered to and systems, structures, and components required to support decommissioning and spent fuel storage in accordance with the possession only license and other administrative and implementing procedures are maintained. The plant maintenance program in effect during the transition period consists of correctivt maintenance, preventive maintenance, and surveillance activities.

3. Redefinition of regulatory bas is for the defueled plant.

Thelermination of operations and the conversion of the operating license to a possession only license has rendered many of the existing provisions of the TNP Technical Specifications inappropriate. On July 31,1993, PGE submitted a request to revise the TNP Technical Specifications to reflect the permanently defueled status of the plant. That request was supplemented by PGE on March 8,1994. On March 31,1995, the NRC issued Amendment

'#194 to Facility Operating (Possession Only) License NPF-1. This amendment provided the TNP Technical Specifications to reflect the i f) permanently defueled condition of the plant, and regulatory requirements and L/

operating restrictions to ensure the safe storage of spent fuel in the spent fuel pool.

The Final Safety Analysis Report was revised to reflect the permanently defueled plant condition and was retitled "Defueled Safety Analysis Report."

l The DSAR was transmitted to the NRC on October 7,1993. Additional l licensing basis documents were also revised to reflect the plant's defueled condition.

4. Assessment of the plant's radiological status.

Section 3.1 presents an assessment of the radiological status of TNP. This assessment was used in developing the Decommissioning Plan.

2.2.4 DECONTAMINATION AND DISMANTLEMENT: GENERAL INFORMATION 2.2.4.1 Overview This section presents a general description of the decontamination and dismantlement period activities for TNP decommissioning. These activities involve the reduction of radioactivity to 2-5 Revision 4 l

l

TROJAN DECOAfAflSSIONING PLAN acceptable levels, allowing for release of the site for unrestricted use. This infomiation provides v

l the basis for development of the programs and procedures for ensuring safe decommissioning i

and a basis for detailed planning and preparation to be completed prior to initiating decontamination and dismantlement activities.

During this period, the remaining contaminated systems and components will be decontaminated or removed, packaged, and either shipped to an offsite processing facility, shipped directly to a low-level radioactive waste disposal facility, or handled by other altematives in accordance with applicable regulations (e.g., greater than Class C waste).

Decontamination of plant structures may be completed concurrently with equipment removal.

Decontamination of structures may include a variety of techniques ranging from water washing to surface material renmval. Contaminated structural material may be packagcd and either shipped to a processing facility, or shipped directly to a low level radioactive waste disposal facility. Altemative disposal methods, in accordance with applicable regulations, may also be used.

A comprehensive final radiation survey will be conducted following the removal or decontamination of contaminated systems, components, and structures. The survey will verify that radioactivity has been reduced to sufficiently low levels to allow the release of the site for unrestricted use. Upon completion of the final survey, PGE will submit a report to support p license termination per 10 CFR 50.82 and State of Oregon regulations. A discussion of the final V radiation survey is provided in Section 4.

2.2.4.2 Detailed Planning and Engineering Activities Detailed project plans will be developed in accordance with design control procedures to support the decontamination and dismantlement activities before they are initiated. The plans will be used to develop work packages, support ALARA reviews, aid in estimating labor and resource requirements, and track decommissioning costs and schedule.

Work packages will be used to implement the detailed plans and provide instructions for actual field implementation. The work packages will address discrete units of work and will include appropriate hold and inspection points. Administrative procedures will control work package format and content, as well as the review and approval process.

2.2.4.3 General Decontamination and Dismantlement Considerations The following general decontamination and dismantlement considerations, as applicable, will be incorporated into the decommissioning work packages. Specific considerations are presented in Section 2.2.5.

/~m 2-6 Revision 4

TROJ.4N DECOMMISSIONING PL4N Dismantlement activities will be carefully reviewed to ensure they do not impact the safe storage of fuel. This review will not be limited solely to dismantlement activities but will include the impact of extemal events. When applicable, work packages will include specific steps to physically protect the systems, strtictures, and components supporting spent fuel storage, or establish safe load paths and protective zones around these systems, structures and components.

Work packages will be implemented in accordance with administrative controls that require evaluations in accordance with the requirements of 10 CFR 50.59.

Temporary shielding will be used where practical for ALARA purposes during decommissioning activities. Some dismantlement activities may be performed under water for shielding purposes as well as contamination control.

The capability to isolate or to raitigate the consequences of a radioactive release will be maintained during decontamination and dismantlement activities. Isolation is the closure of penetrations and openings to restrict transport of radioactivity to the environment. This consideration should not preclude the removal of penetrations and attachments to containment, provided that openings are closed in a timely manner.

Airbome radioactive particulate emissions will be filtered. Effluent discharges will be monitored and quantified. Consideration will be given to the following items:

L i 1. Operation of the appropriate portions of the containment ventilation and purge system, or an approved attemate system, during decontamination and dismantlement activities in the Containment Building;

2. Operation of the appropriate portions of the Auxiliary Building and Fuel Building ventilation system, or an approved alternate system, during decontamination and dismantlement activities in the Auxiliary and Fuel Buildings;

3. Operation of the Condensate Demineralizer Building ventilation exhaust system during decontamination and dismantlement activities in the Condensate Demineralizer Building; and

4. Use oflocal high efficiency particulate air (HEPA) filtration systems for activities expected to result in the generation of airbome radioactive particulates (e.g.

grinding, chemical decontamination, or thermal cutting of contaminated components).

Work activities will be planned to minimize the spread of contamination. Contaminated liquids will be contained within existing or supplemental barriers and processed by a liquid waste processing system prior to release. To minimize the potential for spread of contamination the O 2-7 Revision 4

TROJAN DECOAIAflSS10NING PL4N following considerations will be incorporated into the planning of decommissioning work activities:

1: Isolation of electrical and pneumatic services from components prior to their dismantlement;

' 2. . Covering of openings in intemally contaminated components to confine internal contamination;-

3. - Decontamination and dismantlement of contaminated systems, structures, and components by decontamination in place, removal and decontamination, or removal and disposal;-

~4. Removal of contaminated supports in conjunction with equipment removal or decontamination of supports in conjunction with the building;

5.  : Removal of contaminated systems and components from areas vid buildings prior to structural decontamination. (Block shield walls, or portior of other walls, ceilings, or floors may be removed to permit removal of systems and components.);

6. Removal or decontamination of embedded contaminated piping, conduit, ducts, _

plates, channels,' anchors, sumps, and sleeves during area and building structural lD _ decontamination activities; d

7. Consideration oflocal or centralized processing and cutting stations to facilitate packaging of components removed in large pieces; and

8. Removal of small or compact plant components and parts intact, where feasible.

(This includes most valves, smaller pumps, some small tanks, and heat exchangers.

These components could then be decontaminated in whole or part, and reduced to smaller dimensions in preparation for disposal or release.)

2.2.4.41 Decontamination Methods Systems and components that are contaminated will typically be removed and sent to an offsite processing facility, sent to a low-level radioactive waste disposal facility, or decontaminated

' onsite and released.

Although large scale chemical decontamination is not anticipated as part of the TNP decommissioning, limited application may be used on systems or tanks to reduce radiation dose -

rates prior to dismantlement or general area decontamination.

2-8 Revision 4

TROJAN DECOMMISSIONING PLAN Other decontamination methods typically include wiping, washing, vacuuming, scabbling, p

V' spalling, and abrasive blasting. Selection of the preferred method will be based on the specific situation. Other decontamination technologies may be considered and used if appropriate.

Application of coatings and hand wiping may be used to stabilize or remove loose surface contamination. Airborne contamination control and waste processing systems will be used as necessary to control and monitor releases. If structural surfaces are washed to remove contamination, controls will be established to ensure that waste water is collected in liquid waste processing systems.

Tanks and vessels will be evaluated and, if required, flushed or cleaned prior to sectioning and/or removal to reduce contamination levels and remove sludge. The following considerations will be incorporated into tank and vessel sludge removal activities:

1. Precautions will be made to ensure that liquid inadvertently discharged from the tank is captured in a liquid waste processing system; 2, Sludge removed from the tank will be stabilized prior to shipment; and

3. Waste water will be processed and analyzed before being discharged.

Concrete that is contaminated or activated may be removed and sent to a low-level radioactive q waste disposal facility, allowed to decay below site release criteria, or handled by other methods V in accordance with applicable regulations. Removal ofconcrete should be performed using methods that control the removal depth to minimize the waste volume produced. Vacuum removal of the dust and debris with HEPA filtration of the effluent should be used to minimize the spread of contamination and reliance on respiratory protection measures.

2.2.4.5 Dismantlement Methods Diamantlement methods can be divided into two basic types: disassembly, and cutting or other dutructive methods. Disassembly generally means removing fasteners and components in an orderly non-destructive manner (the reverse of the original assembly). Cutting methods include flame cutting, abrasive cutting, and cold cutting.

Flame cutting includes the use of oxyacetylene and other gas torches, carbon are torches, air or oxy are torches, plasma are torches, cutting electrodes, or combinations of these. Most of the torches can either be handheld or operated remotely with the appropriate devices. Abrasive cutting includes the use of grinders, abrasive saw blades, most wire saws, water lasers, grit blast, and other techniques that wear away metal. Cold cutting includes the use of bandsaws, bladesaws, drilling, machining, shears, and bolt / pipe / tubing cutters.

A U 2-9 Revision 4

TROJAN DECOMMISSIONING PLAN Selection of the preferred method will depend on the specific situation. Other dismantlement q technologies may be considered and used if appropriate. Dismantling of systems will include the V removal of valves and piping for disposal, Most valves can be removed with the piping. La ger valves and valves with actuators may be removed separately for handling purposes. Valve actuators that can be decontaminated should be removed from the valves prior to pipe removal where practical.

2.2.4.6 Removal Seouence and Material Handling Removal sequence may be dictated by access and material handling restrictions or by personnel exposure considerations. In some cases a top-down approach may be used; materials and structures at the highest elevations would be removed first to allow access to components in lower levels. In other cases, different approaches may prove more efficient.

In most cases, the first items removed will be those that are not contaminated, or are only slightly contaminated, to preclude contamination by other equipment. However, personnel exposure considerations may not always allow this option. Where non-contaminated equipment cannot be removed first, covers, or other protection, will be used. Similarly, non contaminated piping should be removed from pipe chases and horizontal pipeways before cutting contaminated pipes, if this is not possible, other precautions, such as covers, will be used to minimize the spread of contamination.

q Where rapid cutting techniques are available, pipes and equipment can be reduced in place to V pieces that are manageable using light rigging or by manual lifling. Where slow cutting techniques are used, the largest pieces manageable will typically be freed and further reduced at a more convenient location.

, Most material removed from the Containment Building will be passed through the equipment hatch into the Fuel Building. The Fuel Building operating floor provides a convenient location for handling and processing of materials because of the availability of a crane and because it is a relatively open area. The areas above the holdup tanks are surrounded by shielding walls and can be used for decontamination, sorting, or packaging. The cask wash pit, spent fuel pool, and cask load pit may be available for wash down of components.

The plant is equipped with multiple cranes, hoists, and lifting and transport systems. These systems can be used to litt and transport components and equipment to support plant decommissioning activities. Forklifts, mobi!c cranes, front-end loaders, and ether lifting and transport devices can also used for plant decommissioning activities. The major installed plant cranes, hoists, and lifting and transport devices that are available to support decommissioning include:

1. Containment Building polar crane; O

(,/ 2-10 Revision 4 l

TROJANDECOAIMISSIONING PLAN

2. Fuel Building ovedead crane;-

3. ' Auxiliary Building electric hoist; ,

4. Auxiliary Building elevator;

5. Equipment room monorails;

6. Spent fuel pool bridge crane; and

7. Condensate Demineralizer Building bridge crane.

Inspection requirements for the Containment Building Polar Crane, Fuel Building Overhead Cranc, Auxiliary Building Electric Hoist, Spent Fuel Pool Bridge Crane, and the Condensate Demineralizer Building Bridge Crane are specified in Trojan Plant Maintenance Procedure MP 1 20, " Cranes, Hoists and Winches." Chainfalls and other temporary hoists are inspected and verified to be in good working condition at the time ofissue from the Tool Room. The Auxiliary Building Elevator is inspected by a State Inspector in accordance with State of Oregon requirements.

The Containment Building polar crane is capable of reaching most locations inside the Containment Building and can handle large, heavy loads. The Fuel Building overhead crane has access to a hoistway open to plant grade at the 45 fl elevation. The Auxiliary Building elevator

(~)N

(_ has access to upper floors in the building and can carry small loads.

Installed cranes and hoists may be used in conjunction with temporary or mobile lifting and transport devices to support decommissioning. In addition to the use of cranes for general material handling and movement, their use is noted in the applicable component removal descriptions.

The installed plant cranes, hoists, and other lifting devices can be decontaminated and dismantled when they no longer are required to support decommissioning activities.

In areas where considerable material movement is expected, such as the pipe penetration areas, pipeways, and pipe chases, hoisting equipment, such as winches and hoists, may be practical with blocks attached to existing building structural steel. Beam clamps and welded lugs on the steel will allow repositioning of hoisting lines throughout an area.

In some areas of the plant it may be convenient to use material handling equipment, such as forklifts or front-end loaders, for moving materials from one location to another. Small mobile cranes can be used inside plant structures for smaller equipment and materials. Wheeled carts can also be used for moving pipe, steel, and other items. Skid rails, skid ways, and air pallets may also be used for the movement oflarger equipment.

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TROJANDECOMMISSIONING PLAN Localized temporary ventilation equipment and HEPA filtration may be used to supplement building ventilation and minimize the spread ofradioactive particulate contamination.

2.2.5 DECONTAMINATION AND DISMANTLEMENT: SYSTEMS, STRUCTURES, AND COMPONENTS 2.2.5.1 Overview This section of the Decommissioning Plan presents a brief, general description of TNP systems, components, and structures that are known or considered to be internally contaminated or that may be used to support decommissioning activities.' Discussion of activities and tasks associated -

with decommissioning intemally contaminated systems, structures, and components is presented.

Also discussed are systems or components that may be used to support decommissioning.

Because external contamination is generally considered to exist on systems, structures and components located in the radiologically controlled areas (RCAs) of the plant, it is not specifically noted in the following system discussions. However, systems, components, and structures that are extemally contaminated will be decontaminated for release or disposed of as radioactive waste. Plant layout and general arrangement drawings are provided in Figures 2-1 through 2-9.

i The considerations identified in this section are based on preliminary planning and will be used l during detailed planning in the development of specific work packages. With the exception of -

'q the removal of the steam generators, pressurizer, reactor vessel with intemals intact, and removal l A) of systems / components not required for the safe storage ofirradiated fuel, full scale 1-L dismantlement of the facility radioactive systems, structures, and components is scheduled to -

begin after completion of the transfer of spent fuel to the ISFSI. If the reactor vessel with the .I

- intemals intact removal option is cancelled, the option of removing the internals by segmentation I- I will be performed during the mid 1998 to mid 2000 time frame. I This section of the Decommissioning Plan describes the major components of contaminated

- plant systems and, in some cases, a description of equipment removal considerations for system -

components. The section is intended to provide general information and guidance for work package planning'and is not required to be updated to reflect equipment removal. Table 2.2-5 provides a list of major components described in the subsections of 2.2.5 that are removed each year (beginning in 1996).

' 2.2.5.2 Reactor Coolant System The reactor coolant system (RCS) has four parallel stainless steel piping loops connected to the .

reactor vessel. The major components of the RCS are the reactor vessel, four steam generators, the pressurizer, four reactor coolant pumps, and associated valves, piping, fittings, and instrumentation. The removal of the reactor vessel is addressed separately in Section 2.2.5.4.

2-13 Revision 4 i

4 TROJAN DECOMMISSIONING PLAN The RCS is located inside the Containment Building. The system is not required to support

(, decommissioning or safe spent fuel storage. The system is internally contaminated. The following specific considerations apply.

Piping sections can be removed through open steam generator cubicles, reactor coolant pump

access hatches, or other accessways using carts or skids._ The resistance temperature detector (RTD) bypass loops were removed during the Transition Period.

2.2.5.3 Raartor Va==al Inf amml=

The reactor vessel intemah consist of an upper internals assembly and a lower intemals assembly.

l The upper intemals assembly provided structural support to the fuel assemblics, as well as -

orientation and guidance for control rod assemblies and incore instrumentation. The upper

'intemals assembly consists of an upper instrumentation conduit and support assembly, upper support plate, control rod guide tubes, and the upper core plate.

, The lower hitemals assembly (lower core support structure) consists of a core barrel, core baffles, 1.

- lower core plate and support columns, neutron shield pads and specimen holders, and lower core support plate.

O Neutron irradiation from reactor operation generated activation products in the reactor vessel intemals. The reactor vessel intemals are also contaminated. I I

The reactor vessel intemals will either be removed intact with the reactor vessel or removed l'

' separately and segmented. Removal of the reactor vessel intemals intact with the reactor vessel _I

is discussed in Section 2.2.5.4. The following discussion addresses the separate removal and _l segmentation of the reactor intemals. -1 I

Based on the neutron activation analysis, portions of the reactor vessel internals are expected to I

be Greater Than Class C (GTCC) radioactive waste per 10 CFR 61. If removed separately, _

l

. portions of the reactor vessel intemals that are classified as GTCC waste will be segmented and I stored in containers fabricated to standard fuel assembly size for interim storage in the existing 1 -i spent fuel racks. The lower level wastes will be packaged and shipped to the radioactive waste 1 disposal facility near Richland, Washington, for burial. I F

To support separate reactor vessel internals removal and segmentation, the reactor vessel head 1

-will be removed and a refueling cavity seal will be seal welded in place. -The transfer tube blank

- flange will be removed to allow for later transfer of segmented componants to the spent fuel pool for storage. Unsegmented components or large segments in casks may an - Se transferred through the equipment hatch to the spent fuel pool. After completing preparations, the reactor

. _ _ cavity and transfer canal will be flooded. The upper internals will be removed from the reactor 2-14 Revision 4 i

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TROJAN DECOMMISSIONING PLAN ,

vessel and placed in the storage stand within the cavity. Removal of the lower intemals to its O 1 storage location within the reactor cavity may be delayed pending segmentation and removal of

the upper internals.

PGE will perform a safety evaluation prior to corrmencing physical activities associated with movement of the RV intemals into the spent fuel pool. This safety evaluation will address the effects of the packaging, movement and storage of the RV internals on spent fuel pool

- performance, he activities associated with the movement of the RV intemals over the spent fuel will be conducted in accordance with plant procedures to ensure that the load restrictions and surveillance requirements of Technical Specification 3.1.4," Spent Fuel Pool Load Restrictions,"

are met. Water level in the spent fuel pool wi'.1 be maintained in accordance with Technical Specification 3.1.1," Spent Fuel Pool Water Level."

Administrative controls for segmentation will include as a minimum:

1. Segmentation of the reactor vessel internals underwater in the reactor cavity;

2. Use of a continuous air monitor to detect an increase in normal airborne activity -

levels.- The alarm setpoim will be based on maintaining worker exposure ALARA;--

A 3. Monitoring of the reactor cavity water for activity. Filtration will be provided V as required to maintain worker exposure ALARA'and ensure the potential for airbome release is maintained below the limits provided in Section 3.4.3.1;

.4. Operation of the containment ventilation exhaust via HEPA filtration during segmentation of reactor vessel intemals; and L - - .

- 5. Development of work activities for segmentation implementing the requirements of ALARA. Continuous monitoring of radiological conditions will be performed ~

- during performance of the work.

Segmentation will be suspended if administrative controls or limits cannot be maintained.

Provisions for maintaining the integrity of the reactor cavity liner and for ensuring spent fuel pool boron concentration and water level will be maintained during segmentation and transfer of

=

segmented components into the spent fuel pool will be specified in the detailed segmentation -

plan and/or' work' package.

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TROJAN DECOMMISSIONING PLAN 2.2.5.4 Reactor Vessel

(~h O The reactor vessel supported and contained the reactor core, directed coolant flow through the core, and facilitated operation, control, and handling of reactor components. The reactor vessel is a fabricated cylinder with a hemispherical bottom head and a removable hemispherical upper head. It contains the core support and other internal structures. The reactor vessel has four inlet and four outlet nozzles located in a horizontal plane below the upper head flange. The reactor vessel is carbon steel with weld deposited austenitic stainless steel on surfaces that were exposed to the reactor coolant.

The vessel is contaminated and activated. The reactor vessel may be removed with the reactor i vessel internals intact or the internals may be separately removed. Separate reactor vessel I internals removal is discussed in Section 2.2.5.3. The following specific considerations apply, I The reactor vessel upper head can be disposed ofintact, in segments, or with the reactor vessel. I Ifit is disposed ofintact, a cover plate will be installed over the bottom flange. Control rod drive mechanism (CRDM) housings and other attachments to the head will be removed and the

' penetrations sealed, if the reactor vessel upper head is disposed ofin segments, the upper head can be processed using cutting methods suitable for cutting thick, highly activated components.

The sections can then be packaged and shipped. If disposed of with the reactor vessel, the upper I head will be attached to the vessel by means of tensioned studs. The number of studs will be I n sufficient to maintain the integrity of the vessel package under transportation conditions. I The reactor vessel may also be removed intact, with or without the internals or upper head, or i sectioned. The method selected will be based on an evaluation of the ease of execution, personnel exposure, schedule impact, transportation availability, and cost.

High dose rates from the activated surfaces of the reactor vessel may require the use of special

' shielding and handling methods to ensure that personnel exposure is maintained ALARA.

Neutron irradiation from the reactor operation generated activation products in the reactor vessel.

Based on the neutron activation analysis shown in Tables 3.1-8 and 3.1-9, the reactor vessel can be disposed of as 10 CFR Part 61, Class A waste (vessel wall) and Class C waste (vessel clad).

The reactor vessel and reactor internals, packaged together, can be dirposed of as Class C waste. I The radionuclide content estimates will be verified with a radiation survey of the reactor vessel. 1 Detailed classification evaluations will be completed as a part of detailed planning of the reactor vessel removal activity.

Intact vessel removal would involve shipping the vessel, with or without the internals, to a low- I level radioactive waste disposal facility in one piece. This may require certification of an exclusive use shipping container for transporting the vessel, or the vessel may serve as its own shipping container. If removed together, the reactor vessel and internals will be shipped as a l NRC-approved Type B package. A safety analysis report wil be prepared to support NRC I

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TROJ.4N DECOMMISSIONING PLAN approval of the package. The following items will be considered if the vessel, with or without I

(~'T , the internals,is removed intact: I U

l. Removal of support attachments;

2. Removal of piping up to the nozzles and capping the nozzles;

3. Installation of a cover over the vessel fiange to reduce dose rate and control airbome radioactivity;

4. Removal of water from the vessel;

5. Application of a fixative coating or grouting to stabilize remaining intemal surface contamination:

6. Reconfiguration of the polar crane to its 400 ton capacity;

7. Selection of an appropriate lifting point on the vessel; j 8. Attachment of a skid box or upending device for handling the vessel; and p) 9. Routing the vessel through the construction opening and loading it onto a V transporter.

Segmented vessel removal would involve shipping vessel sections to a low-level radioactive waste disposal facility inside appropriate shipping containers.

Sectioning of the vessel can be performed by appropriate cutting or machining processes. The ves::el is lined with stainless steel which limits methods for sectioning. I:owever, removal of the stainless steel lining with machining, grinding, or other techniques would expose the carbon steel vessel wall. The carbon steel could then be sectioned using appropriate thermal or mechanical cutting techniques. Altematively, the vessel could be filled with grout and sliced into sections using a diamond wire saw. The wilowing items will be considered if the vessel is removed in sections:

1. Shipping the lower hemispherical head in one piece after removal and plugging of incore detector sleeves;

2. Use of a water purification system and/or an underwater vacuum cleaner in the proximity of the cutting to minimize contamination levels in the water and to i

maintain clarity when water is in the vessel during cutting operations;

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TROJAN DECOAfAflSSIONING PL4N p

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3. Use of a temporary liEPA filtration unit in the p:uximity of the cutting during cutting operations;

4. Scaling the RCS piping penetrations to minimize the transport of contamination; and

5. Use of a cover plate and shielding on the reactor flange to reduce dose rate and spread of contamination.

2.2.5.5 Steam Generators The four steam generators are venical shell and U-tube heat exchangers with integral moisture separating equipment. The reactor coolant channel head is divided into inlet and outlet plenums by an ineonel vertical panition plate extending from the head to the tube sheet. The steam generators are constructed primarily of carbon steel. The heat transfer tubes are inconel, the primary side of the tube sheet is clad with inconel, and interior surfaces of the reactor coolant channel head and nozzles are clad with austenitic stainless steel. These components are internally contaminated. The steam generators were removed during the Transition Period.

2.2.5.6 Reactor Coolant Pumns The reactor coolant pumps are single speed, centrifugal pumps driven by air-cooled,

'_(O .

three-phase induction motors. The pumps have bottom inlet, and side discharge openings and are equipped with controlled leakage seals on the shaft. The motors are mounted above *.he pumps and can be removed as separate components. A flywheel mounted on the shaft, above the motor, provided additional inertia to extend pump coastdown.

The reactor coolant pumps are internally contaminated. The following specific considerations apply.

The reactor coolant pumps, their motors, and flywheels may be removed from the Containment 13uilding and shipped off-site as integral units or as separate components. The oil reservoirs have been drained. Reactor coolant pump piping will be cut and capped as close to the pump as practical. Hatches above the pumps allow rigging and polar crane access to reactor coolant pump motors and pumps. Pump nozzle covers will be installed, and grout or fixative coatings will be applied, as necessary, to contain contamination prior to shipment.

2.2.5.7 Pressurizer and Pressurizer Relief Tank The pressurizer is a vertical, cylindrical vessel with hemispherical top and bottom heads, constructed of carbon steel that is clad with austenitic stainless steel on surfaces exposed to

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TROJAN DECOMMISSIONING PL4N 3

(V reactor coolant. Electric heaters are installed through the bottom head of the vessel. The spray nozzle, relief, and safety valve connections are located in the top head of the vessel. .

The pressurizer was removed during the Transition Period.

The pressurizer relief tank is located inside the Containment Building. It is internally contaminated. The pressurizer relief tank can be moved from its location and lifted out of the Containment Building in one piece, or it may be sectioned in place.

2.2.5.8 Chemical and Volume Control System The chemical and volume control system (CVCS) consists of several subsystems: charging, letdown and seal water, chemical control, purification and makeup, and boron recovery. Main components of the system are two centrifugal pumps, one positive displacement pump, a volume control tank, three ion exchangers, a regenerative heat exchanger, a letdown heat exchanger, an excess letdown heat exchanger, a seal water return heat exchanger, and associated valves, piping, fittings, filters, and instrumentation. Additional major components include two boric acid evaporators, three holdup tanks, two boric acid tanks, concentrates holding tank, two monitor tanks, boric acid batching tank, chemical mixing tank, and a resin fill tank, g The CVCS is located in the Auxiliary and Fuel Buildings and inside the Containment Building,

• Portions of the system are intemally contaminateu. The following specific considerations apply.

The two centrifugal charging pumps were removed and transferred to another nuclear utility.

The positive displacement charging pump is located on the 25 ft elevation of the Auxiliary Building. This pump is skid mounted and relatively compact and accessible. Monorails are installed over the pump skid. The pump skid can be removed as a unit, or the pump can be separated from its motorc speed increaser, and other equipment.

The volume control tank is surrounded by thick shielding walls and sectioning may be the best method of removal. Because of the size of the monitor tanks and holdup tanks, they will likely require sectioning for removal. The relatively thin west wall of the monitor tank room may be removed to facilitate the removal of these tanks. The access openings to the holdup tanks could be enlarged to facilitate tank removal. The boric acid tanks may also require sectioning for removal. Access plugs above the tanks may facilitate removal of sections. The chemical mixing tank can be removed in one piece. The boric acid batching tank and resin fill tank are not internally contaminated. Each tank can be removed in one piece.

The regenerative and excess letdown heat exchangers are located on the 61 ft elevation inside the Containment Building. The letdown and seal water heat exchangers are located on the 61 ft elevation of the Auxiliary and Fuel Buildings, respectively. The heat exchangers are relatively compact and can be removed from their location without sectioning. The Auxiliary and Fuel

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TROJAN DECOMMISSIONING PL4N

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V

' Building heat exchanger rooms have knockout panels in the walls. Certain heat exchanger areas e are equipped with monorails and have knockout panels in the walls.

The boric acid evaporator components are skid mounted. Components can be removed individually.

2.2.5.9 Safety Injection System The safety injection system consists of two centrifugal safety injection pumps, four safety injection accumulators, and associated valves, piping, fittings, and instrumentation.

Safety injection system pumps are located on the 5 ft elevation of the Auxiliary Building. The accumulators are located on the 45 ft elevation of the Containment Building. The following specific considerations apply.

Safety injection pumps are skid mounted and relatively compact. The skids can be removed as a unit or the pumps can be separated from their motors and other skid mounted equipment.

Monorails are provided in the pump rooms for equipment removal. The accumulators can be removed intact from the Containment Building ifinterferences are removed. The accumulator vessels can also be sectioned in place and moved to rigging pathways inside the Containment Building.

i(3 V Internal contamination should be relatively low, allowing for rapid removal of the pipe. Piping inside the bioshield can be removed in conjunction with other pipes connecting to the reactor coolant loops. Piping can be removed either through the normal access into the bioshield or through the steam generator or reactor coolant pump access openings.

, 2.2.5.10 Residual Heat Removal System The residual heat removal system contains two pumps, two heat exchangers, a common containment recirculation sump, and associated valves, piping, fittings, and instrumentation.

The residual heat removal pumps and heat exchangers are located on the 5 ft elevation of the Auxiliary Building. The system is internally contaminated. The following specific considerations apply.

Access hatches for removal of heat exchangers or tube bundles were included in the building construction. Those access hatches can be opened, interfering piping and cables removed, and the entire heat exchangers (or large portions of them) can be removed via that path.

Altematively, the heat exchangers can be sectioned in the heat exchanger room and removed in smaller pieces.

U 2-20 Revision 4

+ --

TROJAN DECOAfAflSSIONING PLAN C4 The residual heat removal pump rooms have knockout panels and are equipped with monorails to V aid in pump removal.

The recirculation sump screens, racks, and supporting structure on the 45 ft elevation of the Containment Building can be disassembled in place for removal and decontamination, or shipment and burial.

2.2.5.11 Containment Sprav System The containment spray system consists of two pumps, a sodium hydroxide tank, two eductors, two spray headers, and associated valves, piping, fittings, and instrumentation.

The containment spray pumps are located on the 5 ft elevation, and the sodium hydroxide tank is located on the 25 ft elevation of the Auxiliary Building. The spray headers are located inside the Containment Building. The system is internally contaminated. The following specific considerations apply.

The containment spray pumps can be removed intact or separated from their motors. The spray headers, spray nozzles, and other piping at the 205 ft level inside Containment can be removed by using the polar crane trolley as a platform. Pipe sections may be light enough to be rigged, using relatively lightweight rigging, and lowered to the 93 ft elevation of the Containment

q Building for removal. Scaffolding can be erected to access sections of pipe supported and routed V up the Containment wall.

Removal of the sodium hydroxide tank intact would require removal ofits cubicle walls and an obstructing motor control center. It may be more practical to remove the tank by sectioning in place and removing the pieces.

2.2.5.12 Comnonent Cooling Water System The component cooling water (CCW) system consists of three pumps, two surge tanks, two main heat exchangers, a chemical addition tank, equipment heat exchangers, and associated valves, piping, fittings, and instrumentation.

The CCW heat exchangers and pumps are located on the 45 ft elevation of the Fuel Building.

The surge tanks are located on the 77 ft elevation in the pipe penetration area. Portions of the system support spent fuel cooling and are required until the fuel is moved to the ISFSI or '

alternate cooling is established. Portions of the system have detectable levels ofinternal contamination. The following specific considerations apply.

Existing tube pulling openings in the wall north of the CCW heat exchangers can be enlarged, or the wall entirely removed to allow passage of the heat exchanger into the Fuel Building crane O

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TROJAN DECOAfAflSSIONING PLAN (3

V bay. The CCW pumps can be removed after the heat exchangers are cleared. The CCW surge tanks can be removed by enlarging the stairwell openings in the pipe penetration area for rigging, or by sectioning in place.

2.2.5.13 Service Water System The service water system supplies raw water from the Columbia River via the intake structure to the component cooling water heat exchangers, essential room coolers, emergency make-up to the spent fuel pool and component cooling water system, and other equipment. The system has three service water pumps, two service water booster pumps, equipment heat exchangers, and associated valves, piping, fittings, and instrumentation.

The service water system pumps are located in the intake structure. Portions of the service water system support spent fuel cooling and will be required until the fuel is moved to the ISFSI or attemate cooling is established. This syst:m may also be required to provide dilution for liquid radwaste discharges. The system is not internally contaminated, but

. room and equipment coolers in contaminated areas of the plant may have external contamination.

Removal of these coolers is considered to be associated with the decontamination and dismantlement of structures.

12.2.5.14 Spent Fuel Pool and Fuel Handling Eouloment q

b The spent fuel pool and fuel storage structures consist of the spent fuel pool, the spent fuel storage racks, the fuel transfer canal, the cask loading pit, and the new fuel storage area. The spent fuel pool provides for irradiated fuel storage. Additionally, the spent fuel pool provides a transparent radiation shield for personnel. Fuel assemblies are stored in stainless steel spent fuel storage racks located at the bottom of the spent fuel pool. The fuel transfer canal and cask

. loading pit. facilitated handling ofirradiated fuel by providing isolable underwater operating areas for fuel transfer evolutions. The new fuel storage area provided a protected area for dry, suberitical storage of new fuel assemblies. The fuel transfer canal and cask loading pit are connected to the spent fuel pool by transfer slots which can be closed and sealed by leak-tight gates. The spent fuel pool, fuel transfer canal, and the cask loading pit are reinforced concrete structures with seam-welded stainless steel linings. 4 The spent fuel pool is located in the Fuel Building. Fuel handling equipment is located in the Fuel Building and the Containment Building. The spent fuel pool supports spent fuel storage and will be required until the fuel is moved to the ISFSI. The fuel handling equipment in the spent

. fuel pool may be required to transfer the fuel to the ISFSI. Additionally, the spent fuel pool bridge crane may be required to support the decommissioning of the spent fuel pool. The system is contaminated. The following specific considerations apply.

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TROJ.4N DECOMMISSIONING PL4N e

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The potential for high levels of contamination exists for components removed from the spent fuel pool. The spent fuel storage racks are accessible with the Fuel Building crane and could be I

I removed from the spent fuel poo! intact for sectioning and packaging at another location. The liner of the spent fuel pool, fuel transfer canal, and cask load pit will be sectioned for removal.

The fuel transfer tube and sleeve will also be sectioned for removal. The fuel handling cranes, the fuel transfer cart, its tracks, and the upender frames will be sectioned into manageable pieces for removal. It may be possible to decontaminate and release sections of the liners, and the spent fuel storage racks.

Dismantlement of the majority of the spent fuel pool and fuel handling radioactive systems, components, and structures is scheduled to be conducted upon completion of the transfer of spent fuel to the ISFSI, Spent fuel will be in the ISFSI prior to decontamination and dismantlement of the spent fuel pool liner Other items stored in the spent fuel pool will be moved to alternate storage locations or disposed of before decontamination and dismantlement of the spent fuel pool liner, 2.2,5,15 Srient Fuel Pool Cooling and Demineralizer System The spent fuel cooling and demineralizer system removes decay heat from the spent fuel elements stored in the spent fuel pool and purifies the system water inventory. The system consists of two cooling pumps, two heat exchangers. one p;4fication pump, a skimmer pump, a O demineralizer, and associated valves, piping, and instrumentation.

b The spent fuel pool cooling and demineralizer system is located in the Auxiliary and Fuel Buildings. The system supports spent fuel cooling and water purification and is required until the fuel is moved to an ISFSI facility or alternate cooling and water purification is established.

The system is internally contaminated. The following specific considerations apply.

The pumps are small and can be removed intact. The heat exchanger room has a knockout panel in the wall, and the heat exchangers can be removed intact. The skimmer filter housing is easily accessible and can be removed intact.

2.2.5.16 Condensate Demineralizers The full flow condensate demineralizers produced high purity condensate for use in the generation of steam. The system has eight demineralizer vessels, a backwash sy stem. and associated valves, piping, fittings, and instrumentation.

The condensate demineralizers are located in the Condensate Demineralizer Building. The backwash receiver tank is potentially internally contaminated. Other portions of the system have detectable levels of contamination.

O d 2-23 Revision 4

- -- TROJA A DECOAIAllSSIONING PL4N q) 2.2.5.17 Steam Generator Blowdown System The steam generator b!owduwn system consists of a pump, a tank, a heat exchanger, two bn exchangers, and associated valves, piping, fittings, and instrumentation.

The steam generator blowdown system is located in the Auxiliary and Turbine Buildings, the Main Steam Support Structure, and the Stean Generator Blowaown Building. The .;ystem is internally contaminated. The following specific considerations apply.

The steam generator blowdown tank, located in the Steam Generator Blowdown Building, should be sectioned to facilitate removal and disposal.

2.2.5.18 Primary Makeup Water System and Refueling Water Storage Tank The primary water storage tank (PWST) is a vertical tank with immersion heaters and a floating roof. The system has two primary makeup water pumps, the primary water storage tank, and associated valves, piping, fittings, and instrumentation.

The refueling water storage tank (RWST) is a vertical tank constructed of austenitic stainless steel with immersion heaters.

(n') The primary makeup water system is located in the Auxiliary Building. The PWST and the RWST are located in the tank farm (south of the Containment Building). These systems are l intemally contaminated. The following specific considerations apply.

Due to their size, the PWST and the RWST should be sectioned to facilitate packaging and shipping, It may be possible to decontaminate and release sections of the tanks. The pumps are small and can be removed intact.

2.2.5.19 Plant Effluent System The Turbine Building sump, oily water separator, solids settling basin, and discharge and dilution structure togethe. comprise the plant effluent system. The plant effluent system provides a means of discharging plant liquid wastes while ensuring compliance with the National Pollutant Discharge Elimination System Waste Discharge Permit.

The plant effluent system components are located throughout the plant site. The system may be required to collect and dispose of waste generated by decommissioning activities. The Turbine Building sump is contaminated. Other portions of the system, such as the oily water separator, may also be contaminated; they are still being used to support ongoing plant activities. These components can be sampled for contamination when they are no longer in use. The following specific considerations apply.

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TROJAN DECOMMISSIONING PL.4N

(~

Q} Turbine Building sump contamination will be removed using concrete d: contamination techniques described in Section 2.2.4.4. Sump input and discharge piping will be checked for contamination to determine the proper method of disposal. Buried and embedded piping may be left in place ifit meets site release criteria. The ef0uent diffusion pipe can be removed by divers.

2.2.5.20 Containment Ventilation Systems The containment ventilation system consists of the following subsystems: purge supply system, purge exhaust and refueling cavity supply and exhaust system, CRDM cooling system, pressurizer compartment and incore instrumentation switching room cooling system, reactor coolant pump cooling system, reactor cavity cooling system, containment air cooler system, hydrogen control system, hydrogen mixing system, hydrogen vent system, unit heater system, and cleanup rec.irculating units. Containment purge exhaust is directed to the primary vent stack which is attached to the outside of the Containment Building. Exhaust air is monitored for radiation and is exhausted through a vent at the top of the Containment Building, i

l Containment ventilation systems are located inside the Containment Building, the Auxiliary Building, and the Main Steam Support Structure. The primary vent stack is attached to the l

outside of the Containment Building. Portions of the system will be used to maintain a habitable environment and control contamination during decommissioning. The systems are internally contaminated. The following specific considerations apply, b Using rigging clips that are embedded in the Containment Building dome, the containment air coolers can be disassembled and removed from the Containment Building. Piping can be cut into manageable lengths and lowered in the same manner. Vertical pipe runs on the Containment Building wall can be cut and lowered to the deck for additional sectioning. Access to this piping can be gained using scaffolding or baskets rigged from the polar crane. The ducting will be dismantled and removed in manageable sections.

The systems will remain in service until:

1. The individual system or component has been evaluated as not required to support further decommissioning activities; or

2. An attemate system has been established; or

3. Contaminated components have been removed and the building decontaminated.

The CRDM fans and motors, cleanup recirculating units, pressurizer compartment cooling system, and the refueling cavity supply system have been or will be partially or completely removed to support the removal of the steam generators and pressurizer.

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TROJAN DECOMMISSIONING h AN

(] ,9 5.21 Hydrogen Recombiners The hydrogen control system consists of two electric hydrogen recombiners located at the 93 ft elevation inside the Containment Building. Each recombiner consists of a thermally insulated vertical metal housing with metal sheathed electric resistance heaters, Major structural components of the recombiners are stainless steel, except the base which is carbon steel.

The hydrogen recombiners have been removed to support the removal of the steam generators and pressurizer.

2.2.5.22 Fuel Building and Auxiliary Building Ventilation Systems The Fuel Building and Auxiliary Building ventilation systems provide fer the supply, heating, cooling, and exhaust of air for the Fuel and Auxiliary Buildings. The systems include several subsystems: Fuel and Auxiliary Building supply system, Fuel and Auxiliary Building exhaust system, spent fuel pool exhaust system, maintenance area supply cooling system, space heating system, radioactive waste annex supply and retum, pump cooling units. Air is exhausted through the primary vent stack which is attached to the outside of the Containment Building. Exhaust air is monitored for radiation. The other systems provide heating and cooling for specific areas in the buildings.

( The Fuel and Auxiliary Building ventilation systems are located inside the Fuel and Auxiliary Buildings. Portions of the systems will be used to maintain a habitable environment and control ,

contamination during decommissioning. The systems are internally contaminated. The following specific considerations apply.

The systems will remain in service until:

1. The individual system or component has been evaluated as not required to support further decommissioning activities; or

2. An attemate system has been established; or

3. Contaminated components have been removed and the building is decontaminated.

! 2.2.5.23 Condensate Deminera!ieer Building Ventilation System The Condensate Demineralizer Building ventilation system provides for supply and exhaust air in the building. Supply air is provided through infiltration and, as appropriate, roof supply fans.

Exhaust air is monitored using a sample pump, sample probe, and a radioactive airborne particulate monitoring filter.

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2-26 Revision 4

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TROJ4N DECOAfAflSSIONING l'L4N The system will remain available for senice until the Condensate Demineralizer Building is no longer used to prccess radioactive waste, and the building is decontaminated.

2.2.5.24 Instrument and Service Air System The instrument and service air system supplies compressed air required for pneumatic instruments, valves, and service air outlets throughout the plant. The system has four air compressors, aftercoolers, air receivers, Olters, dryers, and associated valves, piping, fittings, and instrumentation.

The instrument and service air system is located in buildings throughout the plant. The system may be used for operation of control valves, dampers, tools, anj breathing air. As a portion of the instrument and senice air system is determined not to be required to support further decommissioning, it may be deactivated and removed. The system is not considered to be internally contaminated.

2.2.5.25 Gaseous Radioactive Waste System The gaseous radioactive waste system, which consists of the vent collection system and the gas collection system, collected and processed gaseous efnuents fron, various tanks, sumps, and plant equipment which contained Guids with entrained or dissolved radioactive gases. The system contains four waste gas decay tanks, a waste gas surge tank, two waste gas compressors, a vent collection header exhaust fan, and associated valves, piping, fittings, filters, and instrumentation.

The gaseous radioactive waste system is located in the Auxiliary and Fuel Buildings and the Containment Building. The following specinc considerations apply.

Tanks should be sectioned to facilitate removal. The waste gas compressors can be dismantled into separate components for removal and disposal. The compressor motor can be removed intact or dismantled as necessary to facilitate removal. The gas collection header and vent collection header exhaust filter housings can be removed intact.

Temporary filtration and ventilation equipment may be used to provide for venting of required tanks to eliminate the need for 'he vent collection header portion of the system. Temporary systems to support decommissioning are discussed in Section 2.2.4.8.

2.2.5.26 Solid Radioactive Waste System The solid radioactive waste system provides for storage and processing for disposal of spent demineralizer resin. expended filter cartridges. and other miscellaneous contaminated solid refuse. The system contains a spent resin storage tank, tiger lock storage tank, spent resin O 2-27 Revision 4

TROJAN DECOAIAflSS10NING PLAN (O

g transfer pump, solid radwaste process module, solid waste compactor, spent resin compactor, a Siter handling vehicle, and associated valves, piping, and 6ttings.

The solid radioactive waste system components are located in the Auxiliary and Fuel Buildings.

The solid waste compactor can be used during decommissioning to support packaging of dry active waste. The system is contaminated. The following specine considerations apply, The spent resin storage tank and tiger lock storage tank can be sec'.loned for ease of removal.

System pumps are small and can be removed intact.

2.2.5.27 Clean Radioactive Waste System The clean radioactive w;aste system was installed to collect, store, process, and dispose of contaminated liquids with low particulate and corrosive chemical content. The system processes water and monitors it during discharge. The clean radioactive waste system consists of the reactor coolant drain tank, two clean waste receiver tanks, chemical waste drain tank, auxiliary building drain tank, two treated waste monitor tanks, several pumps, the liquid radioactive waste

! lon exchangers, various Glters, and associated valves, piping, Ottings, and instrumentation. .

The clean radioactive waste system is located in the Auxiliary and Fuel Buildings and the Containment Building. The treated waste monitor tanks may be used for processing radioactive waste water generated during decommissioning activities. The system is internally (a

contaminated. The following specine considerations apply'.

The reactor coolant drain tank and system pumps are small and can be removed intact. Other system tanks can be sectioned to facilitate removal, htajor components of the clean radioactive waste evaporator skid (the vent condensa, distillate cooler, evaporator condenser, evaporator, and absorption tower) should be sectioned to facilitate removal. The skid mounted pumps and eductor are small and can be removed intact.

hiajor components of the decontamination system (ultrasonic cleaning tanks, decontamination catch tank, and spray booth) should be sectioned to facilitate removal. The pressure washer, workbench sinks, and catch tank pump can be removed intact.

Temporary water cleanup systems may be used to reduce the amount ofinstalled equipment required to remain operational. The treated waste monitor tanks can operate in conjunction with the dirty waste drain tank and various sumps to process water used to decontaminate plant buildings or components. Plumbing modi 6 cations may be required to use temporary systems.

Temporary systems to support decommissioning are discussed in Section 2.2.4.8.

Radioactive liquid efnuents will be monitored and released in accordance with the requirements of topical report PGE 1021,"Offsite Dose Calculation h!anual"(ODCht).

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2-28 Revision 4

TROJ.4N DECOAIAllSSIONING PLAN O

() 2.2.5.28 Dirty Radioactive Waste System The dirty radioactive waste system was installed to collect leakage, drains, reliefs, and condensation within the Containment, Auxiliary and Fuel Buildings for storage and processing. Fluids entering the system are discharged from the plant via the discharge and dilution structure. This system includes Containment Building and Auxiliary Building sumps, sump pumps, dirty waste drain tank, dirty waste monitor tank, filters, and associated piping, valves, and fittings.

The dirty radioactive waste system components and piping are located primarily in the Auxiliary and Fuel Buildings and the Containment Building, with several floor and equipment drains located within the Main Steam Support Structure, The dirty waste drain tank and the system sumps may be used for processing of radioactive waste water generated during decommissicaing activities. The systemis internally contaminated. The following specific considerations apply.

The system tanks should be sectioned to facilitate removal. System pumps are small and can be removed intact. The dirty waste bag filter skid will be dismantled and the bag filter housings removed intact. System sumps are concrete pits that will be decontaminated when other decontamination efforts in the various buildings are complete. Techniques for concrete i- decontamination and demolition are noted in Section 2.2.4.

Temporary water cleanup systems may be used to reduca the amount of!nstalled equipment

. O l '

required to remain operational. The dirty waste drain tank can operate in conjunction with the treated waste monitor tank and various sumps to process water used to decontaminate plant buildings or components. Plumbing modifications may be made to simplify dirty waste drain tank discharge path to the treated waste monitor tank. Temporary systems to support decommissioning are discussed in Section 2.2A.8.

2.2.5.29 Radiation Monitoring System The radiation monitoring system consists of the process and effluent radiological monitoring systems (PERMS) and the area radiation monitoring system (ARMS). The PERMS were designed to provide monitoring of gaseous and liquid effluent release paths and selected gaseous and liquid plant process systems. A process and effluent radiation monitoring (PRM) system typically consists of an in-line or off-line sample chamber, detector, associated process filters, a check source, and other applicable equipment.

The ARMS was designed to provide monitoring of general areas in and around the plant. An area radiation monitor (ARM) channel typically consists of a detector, remote / local alarms, remotellocale indicators, power supply, and a check source.

Portions of the radiation monitoring system are considered to be contaminated. The following specific considerations apply.

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'j 2-29 Revision 4

TROJ.4N DECOMMISSIONING PL4N (qj S. Estimated personnel exposure due to the transfer of fuel to the ISFSI is approximately 1.6 person rem for each of the estimated 36 casks.

2.2.7 DECOMMISSIONING RADIOACTIVE WASTE PROJECTIONS The radioactive waste management program (Section 3.3) will be used to control the generation, processing, handling, shipping, and disposal of radioactive waste during decommissioning.

Activated and contaminated systems, structures, and components represent the largest volume of low-level radioactive waste expected to be generated during decommissioning. Other forms of waste generated during decommissioning include:

1. Contaminated water;

2. Used disposable protective clothing;

3. Expended abrasive and absorbent materials;

4. Expended resins and filters; I

5. Contamination control materials (e.g., strippable coatings, plastic enclosures); and I

O V

6. Contaminated equipment used in the decommissioning process. I Tables 2.2-2,2.2 3, and 2.2-4 provide projections of waste volumes for decommissioning. The waste volume projections are conservative estimates obtained from the decommissioning cost estimate, removal of the steam generators and pressurizer, and removal of the reactor vessel with I internals intact or removal of the reactor vessel intemals. Included in this estimate is 340 3ft of I greater 11 an Class C radioactive waste from the reactor vessel internals if they were removed I separately and segmented. This waste would be disposed of with the spent fuel. I The decommissioning cost estimate (Section 5) assumes that cost effective waste volume reduction methods are limited, it also assumes significantly contaminated or activated materials are sent directly to a disposal facility. However, alternative processing methods will be evaluated during decommissioning.

2-37 Revision 4

TROJAN DECOAIAflSSIONING PL4N 2.3 - DECOMMISSIONING ORGANIZATION AND RESPONSIBILillES 2.3.1 DECOMMISSIONING ORGANIZATION The TNP organization (General Manager and above) is shown in Figure 2-10.

The Trojan Site Executive and Plant General Manager has corporate responsibility for overall

- nuclear safety and decommissioning activities et TNP. The General Manager, Trojan Plant is i ponsible for operations, maintenance, personnel / radiation protection, including the ALARA and onsite safety and hazardous materials programs, and emergency preparedness. Reporting to the General Manager, Trojan Plant are the General Manager, Nuclear Oversight; General Manager, Engineering / Decommissioning; General Manager, Plant Support and Technical Functions: Manager, Operations: Manager, Personnel / Radiation Protection; and Manager, Maintenance. The Independent Review and Audit Committee (IRAC) reports to and advises the General Manager, Trojan Plant.

- The General Manager, Nuclear Oversight, is responsible for quality assurance and quality control. The Nuclear Oversight Department is independent of other departments perfonning quality-related activities. The General Manager, Engineering / Decommissioning, is responsible for decommissioning planning and engineering, The General Manager, Plant Support and Technical Functions is responsible for purchasing, cost control, nuclear security, licensing, and I training.

Experienced and knowledgeable personnel will be utilized to perform the technical and administrative tasks required during TNP decommissioning. To the extent practicable, the decommissioning organization will include staff previously employed at TNP to capitalize on their knowledge and familiarity with the facility Contractors may be used to provide specialied services, or to supplement the facility staff, when warranted.

Each member of the facility staff will meet or exceed the minimum qualifications of ANSI

.N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel," for comparable

- positions, except for the Manager, Radiation ProtecFon, who shall meet or exceed the

. qualifications of Regulatory Guide 1.8, " Qualification and Training of Personnel for Nuclear Power Plants " Revision 2, April 1987. l p

j-2-38 Revision 4

TROJ.4N DECOAIAllSSIONING PL4N O Table 2.2 1 Radiation Exposure Projections Exposure-Activity (person rem)

Steam generators and pressurizer removal 138'

- Reactor vessel internals removal' - 50 i Dismantlement Reactor vessel' 35 i Nuclear steam supply system 51.-

Spent fuel recks 19 Balance of plant systems 165 Structures 46 Miscellaneous 20 Subtotal 336 Normal plant operations 9 Fuel transfer to ISFSI 58 Total 591 _

• Then exposure projections assume the reactor vessel internals are removed separate from the reactor 1

. vessel and segmented. The radiation exposure projection for the intact removal of the reactor vessel and the I internals together is approximately 67 person-rem. l

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_ TROJ4N DECOAltflSS10NING PLA V O Table 2,2 2 Page 1 of 3 Decommissioning Waste Classification and Volume Projections Class A Class B Class C Burial Burial Burial Volume Volume Volume item (ft)) (ft') (ft')

Reactor coolant piping 5,894 0 0 Pressurizer relief tank 625 0 0 Reactor coolant ptmps and motors 3,044 0 0 CRDMs/incore instrumentation / service 1,726 0 0 structure removal Reactor vesseP 3,799 0 3,013 i Spent fuel racks 16,551 0 0 120 V ac preferred instrument ac 1,400 0 0 125 V de power 175 0 0 4.16 kV ac power 726 0 0 480 V ac auxiliary load center 5,080 0 0 480 V ac motor control center 8,426 0 0 Chemical and volume control -10,968 0 0 Clean radwaste 5,423 0 0 Containment Building penetrations 188 0 0 Control rod drive 85 0 0 Dirty radwaste 1,613 0 0 Electric heat tracing 164 0 0 Electrical (Cables / Tray / Conduit) 60,139 0 0

• This waste projection assumes the reactor vessel internals are removed separate from the reactor vessel l and segmented.- The burial volume projection for the intact removal of the reactor vessel and the internals I together is approximately 8341 cubic feet of Class C waste. I ,

Revision 4 r

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TROJAN DECOAfiflSS10VIVG PL4N Table 2.2 2 Page 2 of 3 Decommluloning Waste Classification and Volume Projections Class A Class B Class C Burial Burial Burial Volume Volume . Volume Item (ft') (A2) (n')

l Fuel handling system 339 0 0-Fuel pool cooling and demineralizer 4,632 0 0 Fuel and Auxiliary Buildti.s h.ating, 3,661 0 0 ventilation, and att conditioning (IIVAC)

Gaseous radwaste 2,529 0 0

}{VAC 6,635 0 0 liydrogen recombiners $76 0 0 Iraegrated leak rate test instrument line 106 0 0

-Instrument and service air - 1,327 0 .0 Lighting panel supply 997 0 0 Miscellaneous components -1,936 0 0 Miscellaneous reactor coolant 3,418 0 0 Nuclear instrumentation 193 -0 0 Olly waste and stonn drains 1,882 0 0 Containment HVAC 1S,869 0 0 Primary makeup water 3,615 0 0=

Primary sampling 114: 0 0 Radiation monitoring 134 0 0 g Reactor nonnuclear instruments 245 0 0 Reactor vessel system i16 0 0 Residual heat removal 7,649 0 0 Safety injection system 7.149 0 0 Solid radwaste 370 0 0 Revision 4

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TROJ4N DECO \f tflSSIONING PL4N s

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Table 2.2 2 Page 3 of 3 Decommissioning Waste Classilleation and Volume Projections j Class A Class B Class C Durial Burial Durial Volume Volume Volume item (ft)) (f1') (fi))

Spent fuel pool 754 0 0 1 Steam generator system 3,562 0 0 Turbine Building sump pumps and 639 0 0

~

miscellaneous Component cooling water 6,115 0 0 Condensate demineralizer 2,262 0 0 Discharge and dilution 3,834 0 0 Containment spray 1,563 0 0

,3 Containment Building 13,458 0 0 Auxiliary Building 2,650 0 0 Fuel Building 4,711 0 0 hiain steam supply system and electrical 629 0 0 per:etration area Turbine Building 1,054 0 0 Process liquid radwaste 0 3,686 0 Disposal of dry active waste generated 5,942 0 0 Total 239,691 3,686 3,013 Revision 4

TROJA V MCOAf tflSS10VIVG Pl.kV O

V Table 2.2 4 Reactor Vessel Internals Removal Volume Projections Class A ClassB Class C Burial Burial Burial > Class C Volume Volume Volume Volume item (ft') (ft') (ft') (ft')

Reactor vessel internals' 2,889 231 1,500 340 l Filters, etc. 2,272 0 0 0 Dry activated waste 4,276 0 0 0 Total 9.437 231 1.500 340

• This waste projection assumes the reactor vessel internals are removed separate from the reactor vessel I and segmented. The burial volume projection for the intact removal of the reactor vessel and the internals I together is approximately 83J1 cubic feet of Class C waste, I G Revision 4

Trill 4 V l)lTH Af \flSSIO\lW1l'I 4 Y TAlllI2.2 5 Page 1 of 2 MAJOR COMPONENTS REhlOVED (IlY YEAR) 1996:

Reactor Coolant Pumps and Motors -(4)

Reactor Coolant System (RCS) Piping (up to bioshield wall)

RCS and Steam Generator support structural steel Containment Main Steam and Feedwater Piping and Supports (all B/C loops; partial A/D loops)

A Residuallleat Removallleat Exchanger B Residual lleat Removal llcat Exchanger Positive Displacement Charging Pump / Motor A Centrifugal Charging Pump Motor 11 Centrifugal Charging Pump Motor A Safety injection Pump / Motor 11 Safety injection Pump / Motor A Containment Spray Pump! Motor 11 Containment Spray Pump / Motor 11 Component Cooling Water Pump' Motor (Note: replaced with smaller pump)

Condensate Demineralizer vessels Decontamination Shop Equipment Portions of Outside lluildings (WSil warehouse, Maintenance Shop)

Low Level Radwaste Storage Building (Replaced by Condensate Demin Bldg)

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!D Revision 4

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

t TROJ4N DIT0Af tfl,WONING (*L4N TA11LE 2.2 5 Page 2 of 2 MAJOR COMPONENTS REMOVED (BY YEAR) ,

.1991t I I

A/B Residual 1leat Removal (RilR) Pumps / Motors l A/C Service Water Booster Pumps / Motors I Letdown licat Exchanger i A/B Boric Acid Evaporator skids l- :

Clean Radioactive Waste Evaporator skid I  ;

Seal Water lleat Exchanger I Dirty Waste Monitor Tank l Auxiliary Building Drain Tank and Pumps I

Waste Concentrate lloid Tanks and Pump 1 CVCS Concentrates lloid Tank and Pumps I Tiger Lock Storage Tank - I Steam Generator Blowdown Tank and Pump I Boric Acid Storage Tanks and Pumps l Boric Acid Blender and Chemical Addition Tank l Containment Post Accident Sampling System (PASS) l Containment flydrogen Analysis System (CilAS) I lloid Up Tank Recirculation Pump I Gas Stripper Feed Pumps i Speat Fuel Storage Rack (1) I New Fuel Storage Racks I '

Decontamination Area Ventilation Supply Cooling System (AB 5) l

. B/C Safety injection Accumulators 1

Control Rod Drive Mechanism Patch Panels 1 Reactor Vessel Neutroa Water Bag Racks i Flux Thimble Drive Units and Detectors I Manipulator Crane _

l Portions of Containment Ventilation Systems CS.1, CS 2, CS-3, CS-4, CS 5, CS4 i Portions of Activated Concrete around Reactor Vessel i

, Portions of Containment Wall for new 10'x10' roll up access door i Bioshield Structural Steel and Equipment Suppo ts I RCP Oil Collection System l Remaining Containment Main Steam and Feedwater Piping and Supports I Revision 4

l TMOJANDECOMMLSSIONING PL4N Figure 2-10 TROJAN ORGANIZATION INDEPENDENT TROJAN SITE EXECUTIVE AND PLANT GENERAL MANAGER -

REVIEW AND AUDIT COMMITTEE (IRAC)

Key Isecomsnessionseg Rc;___1NMies Are:

Maentensoce W

Personnetheleetwee Protection Ensergency rreperedness O b ^

GENERAL MANAGER GENERAL MANAGER ENGINEERING /

NUCLEAR OVERSIGHT L DECOMMISSIONING FUNCTIONS ll Key A -

Key Ib- .

Key A _ -

- ^*

Rm, E "*^* - Are: R,- . Are: R ,_ ~ - _

• Art:

Quality Ammance Engineering Nisclear Smurity Quakty Centrol A- **

• Utendug r J * #ew Comores i

Trainima Revision 4

- J i l

^ ~ ^

TROJAN DEComt! MON!? iN w - -

See2ItosalteselteWI18eeltee7 19881188e'2980'29e1 2es2 2ess asee 2ees aeos m ases asse'ap s arti ma'ma'ae14'arte acte att7 .ae14'ap9 m Transition Period gp Large Component g Removal Project 15F5i~~ <m p implementation ReactorVessel & <gg Intemals Removal Decontamination & (g g Dismantlement Period Final Radiation ,;gg ,

Survey l t Non-Contaminated- 3 Building Demolition i.

we eesmee ( 6 es,se '

[

Z

m. I Figure 2-11 Decommissioning / Site Restoration M G

.-_ 8=

Revision 4

TROJ4N DECOMMI.WONING PL4N major structural material composi ons including Type 304 stainless steel, pressure vessel carbon steel, concrete, and plate /rebar carbon steel.

A listing of 10 CFR 61 classification by component, one and five years after shutdown, is included in Tables 3.18 and 3.19. One year following shutdown, the radioactivity content of activated components was estimated at 4.2 x 10' Cl. Five years following shutdown, the calculated activity of the activated cornponents was approximately 2 x 10' Ci. Predominar,t radionuclides include "Fe, "Co, and "Ni.

If disposed of separately of the reactor intemals,it is anticipated that most parts of the reactor vessel I will meet Class A burial waste criteria. If the reactor inernals are removed separate from the reactor I vessel and segmented, the core baffle, core formers, and lower core plate are expected to be OTCC, I Portions of the reactor vessel internals that are GTCC waste will be stored in the ISFSt. Intact I disposal of the reactor internals with the reactor vessel will meet Class C burial waste criteria. I 3.1.2.4 Environment The environmental survey, which included representative outdoor areas, focused on the impact of l TNP operation on the environment due to the release of radioactive material. Operational and preoperational environmental monitoring data were used to measure and evaluate the impact.

Additional sampling was conducted to augment, or better define, areas requiring biased surveys.

Survey results were compared to background data to detemiine the overall consequences of TNP operation.

During Phase 1 of the site characterization plan, soil, sediment, and surface water were sampled.

Exposure rates were measured wherever soil was sampled, except where exposure rates were influenced by onsite stmetures. Paved areas onsite were scanned for beta contamination or sampled and analyzed for gamma emitters. A general site map, enclosed at the end of Section 3 as Figure 3-3, shows the site divided into zones. Survey maps depicting sample points by grid location are provided in Figures 3-4 through 3 14. Survey maps were not included for zones where no samples were collected (i.e., Zones 4,13,15, and 16).

Biased sample locations were determined from reviewing plant records that documented radiological events at TNP from 1975 to 1993 (see Section 3.1.1.2 and Appendix 3.1 A). Corrective action programs were reviewed and interviews conducted with PGE personnel to help detennine potential sample locations.

Sampling results were compared to release concentration values developed from the site release criteria of 15 mremlyr TEDE (Total Effective Dose Equivalent). The concentration of radioactive 3-9 Revision 4

,. TROJAN DECOAfAflSSIONING PLAN

\

material associated with this dose was calculated using guidance contained in draft NUREG/CR 5512. " Residual Radioactive Contamination From Decommissioning."

3.1.2.4.1 Surface Soil Survey Surface soil samples were obtained at locations on PGE property contiguous with TNP. The locations of the soll samples are outlined in the Site Characterization Report, Section 5, and can be determined from the tables and maps included in the Site Characterization Report. Each sample contained approximately I liter of material collected from a 1 fF area. Soll samples were analyzed in house for gamma emitters by gamma spectrometry and control samples were analyzed by TMA/Eberline in Albuquerque, New Mexico. Selected samples were also analyzed for "Sr.

Background soil samples were collected from four locations around TNP. Two samples were used as quality control checks and were not included with the data. The four background locations were:

l 1. PGE owned property in Prescott, Oregon (approximately 0.75 miles north northwest of TNP containment);

q 2. _ Water treatment facility in Rainier, Oregon, near radiological environmental sample V location 2 (approximately 3.8 miles nonhwest of TNP containment);

3. PGE owned property west of Highway 30 (approximately 1 mile west of TNP containment); and

4. Northwest of Kalama, Washington, near radiological environmental sample location 1IB (approximately 1.4 miles east northeast of TNP containment).

For soil background measurements, the mean background "'Cs concentration was 0.49 pCi/g with a standard deviation of 0.4 pCi/g and a range of 0.01 to 1.3 pCi/g. Substantial variation in background "'Cs concentrations was observed between varying soil types. Sandy soils found near the river contained low "'Cs concentrations, while clay soils contained higher concentrations. Background "'Cs levels will require further evaluation prior to completing the fmal site survey.

For the survey of unaffected soil areas, the mean "'Cs concentration was 0.77 pCi/g with a standard deviation of 0.86 pCi/g and a range of 0.01 to 2.94 pCi/g. Primarily, the nonnaturally occurring isotopes found in soil samples were O'Cs and "Sr. Fallout from atmospheric weapons tests and the Chernobyl accident are the major sources of"'Cs and "Sr in the environment.

O V 3 10 Revision 4

TROJAN DECOAfAflSSIONING l'LAN

"'Cs results were compared to the release concentration value for "'Cs of 6.1 pCi/g (based on 15 mrem /yr TEDE). No "'Cs results were above the release concentration value. *Sr results averaged 0.2 pCi/g with a standard deviation of 0.16 pCi/g and a range of 0.02 to 0.32 pCi/g. The "Sr levels measured during the preoperational period ranged from 0.01 to 1.28 pCi/g with a mean of 0.30 pCi/g. This average was also less than the release concentration of 1.2 pCi/g.

Blased survey soil samples were taken onsite where potential soil contamination may have occurred. Subsurface soil samples taken in 1991 from the tank farm area were also reviewed as part of the analysis. Samples were taken at 1,2, and 311 depths at 5 locations.

The predominant nonnaturally occurring isotope detected was "'Cs. One surface sample, taken from the tank fami area- also contained low levels of "'Cs (0.010 pCi/g) and "Co (0.044 pCi/g). The mean value for "'Cs in affected soll samples was 0.10 pCilg with a standard deviation of 0.098 pCi/g. De data ranged from 0.01 to 0.33 pCi/g. Radionuclides in the samples were at concentrations below the release concentration values calculated for the mixture of radionuclides. The "'Cs content of the 1991 samples was below the release concentration value.

The following table summarizes the release concentration vahies used to compare actual q measurements to ensure predicted doses would be under the 15 mrem /yr TEDE limit that would Q allow releasing the site for unrestricted use. TNp criteria was based on NUREG/CR-5512 plus background.

Guideline Values Mean NUREG/CR 5512 Background Factor TNp Criteria Nuclide (pCi/q) (pCi/g) (pCi/g)

"'Cs 0.49 5,7 6.1 "Sr 0.30 0.91 1.2 "Co 0 0.37 0.37 3.1.2.4.2 Water Survey Surface water was sampled from indicator sites on pGE property surrounding TNp. A 1 gallon sample was obtained from each site for gamma and "Sr analysis and a 60 mi sample for tritium analysis. The water samples were analyzed for gamma emitters using a gamma spectroscopy 3 11 Revision 4

TROJ.4N DECOAfAflSSIONING PL4N r~

system located onsite. Water samples were analyzed for tritium in the onsite counting laboratory. Selected samples were analyzed for"Sr.

To determine background, water samples were collected from four locations around TNP The locations included:

1. Fishhawk Lake (approximately 18 miles west of TNP containment);

2. Ponds at the intersection of Goble and Bishop Roads (approximately 3 miles southwest of TNP containment);

3. Kress Lake (approximately 1 mile cast northeast of TNP containment); and

4. Deer Island ponds (approximately 7 miles south of TNP containment).

Analyses for gamma emitters and tritium were completed on the samples. No gamma emitters other than naturally occurring radionuclides were identified in the samples. Tritium values were less than detectable. The four samples analyzed for "Sr were less than detectable.

Minimum detectable activity (MDA) for "'Cs, tritium, and "Sr was approximately 4,450, and 0.3 pCi/1, respectively.

For the survey of unaffected water areas, samples were collected from random locations in Whistling Swan and Reflection Lakes. No nonnaturally occurring radionuclides were detected in the samples by gamma spectrometry. Neither tritium nor "Sr was detected in the samples.

For the biased survey, samples were taken from Recreation Lake. No nonnaturally occurring radionuclides were detected in the samples. MDAs for the biased and unbiased survey analyses were the same.

3.1.2.4.3 Bottom Sediment Survey Bottom sediment samples were taken from PGE property around TNP. Approximately I liter of sediment was obtained at each sampling site. The sediment samples were dried and analyzed for gamma emitters using a gamma spectroscopy system located onsite. Selected sediment samples were analyzed for "Sr by TM A/Eberline.

Specific isotopic background sediment samples were not collected. Instead, soil background results were used as sediment background. Background soil samples were analyzed as part of the site characterization effort, and the mean "'Cs concentration was 0.49 pCilg. Background sediment samples will be collected as part of the final site survey, if necessary A comparison of the "'Cs concentration in preoperational sediment samples to the background soil samples O

V 3 12 Revision 4

TROJAN DECOAfAflSSIONING PLAN showed a high correlation with the sediment mean equal to 0.51 pci/g and the soil mean equal to 0.49 pCi/g.

In conducting the survey of unaffected sediment areas, samples were taken from Whistling Swan and Reflection Lakes. The mean value for "'Cs was 0.36 pC1/g with a standard deviation of 0.22 pCi/g and a range of 0.02 to 0.86 pCi/g. The unaffected area sediment samples contain

"'Cs at levels below the guideline value of 6.1 pCi/g. "Sr content of the two sediment samples sent to TMA/Eberline were 0.05 and 0.03 pCi/g. The lower level of detectability for the "Sr analysis was 0.02 pCilg. These results are within the preoperational range of"Sr which was from 0.01 to 0.44 pCi/g with a mean of 0.08 pCi/g. The "Sr content of the sediment samples was also below the release concentration value of 1.2 pCi/g.

For the biased sediment survey sample population, samples were taken from the berm and main areas of Recrea lon Lake. Results of the analyses indicate a mean of 0.28 pCi/g with a standard deviation of 0.37 pCi/g and a range of 0.04 to 1.12 pCi/g. The affected area samp!cs contain

"'Cs at levels below the telease concentration value of 6.1 pCilg. No other gamma emitters were detected.

3.1.2.4.4 Pavement Survey

(~} Pavement scans and sampling were performed. Pavement was scanned for beta contamination.

V in areas where there was interference from the RWST, a 1 n2 sample was collected and analyzed using a gamma spectroscopy system located onsite.

No specific background pavement locations were monitored for this survey. Sample locations located in the TNP park and n.creational areas were used to estimate background levels. Since these areas were unaffected by TNP operation, the survey data for these locations was determined to be an acceptable estimate of background levels of radioactive material in pavement. The mean gross beta reading was 610 dpm/100 cm2 with a standard deviation of 94 dpm/100 cm2and a range of 456 to 764 dpm/100 cm2 .

For the survey of unaffected pavement areas, randomly selected 100 ft2 sections of pavement in other areas of the TNP site which were unaffected by operations were scanned with an ESP 2 2

and 13P 100 detector. The mean value was 657 dpm/100 cm with a standard deviation of 74 dpm/100 cm . The range of measurements was from 542 to 788 dpm/100 cm2 . Measured 2

values were below the cleanup criteria of 5000 dpm/100 cm2 above background for total beta-gamma surface contamination For the biased pavement survey, the affe.ted areas consisted of pavement around the tank farm and its drainage to the west, pavement around the oily water separator, and the paved equipment laydown area around the cooling tower. Pavement samples were taken from 3-13 Revision 4

TROJ.4N DECOMMISSIONING PL4N G

V affected areas with at least two samples from each affected area. The only detectable nonnaturally occurring radionuclide found i 4 the pavement samples was "'Cs in low concentrations. The results of the biased samples exhibited a mean of 0.16 pCl/g with a standard deviation of 0.40 pC1/g and a range of 0.019 to 1.5 pCilg. "'Cs content of the biased pavement samples was similar to that found in background and indicator soll samples obtained for site characterization. One sample, taken from the curb at the southeast corner of the circulating water pump pit area, had the highest "'Cs concentration of 1.5 pCi/g. For comparison, conservatively assuming the "'Cs was from the top I crn of the concrete and covered a 100 cm2 area, then the calculated contamination level would be 799 dpm/100 cm2 which is below the cleanup criteria of 5000 dpm/100 cm2.

3.1.2.4.5 Exposure Rate Survey Exposure rues were measured at locations where affected and unaffected site characterization indicator soil samples had been collected. The measurements were made with a Reuter Stokes pressurized ion chamber instrument posi'ioned I meter above the sample site.

Data for exposure rate background was collected during preoperational surveys at TNP using a high pressure ion chamber, the same type ofinstrument us:d during the site characterization survey. The preoperational mean reading was 7.1 pR/hr with a standard deviation of 1.0 pR/hr o and a range of 5.6 to 9.4 pR/lu. The survey locations coincide with the Radiological Q Environmental Monitoring Program locations.

For the exposure rate survey of unaffected areas, surveys were taken at the maffected soil sampling locations. The exposure rate at one location was nM measured %ause of an instrument failure. Exposure rates ranged from 5.2 to 9.0 pR/hr at the une.iected area locations. The mean exposure rate was 6.4 pR/hr. Data compared favorably with preoperational data, indicating no effect from TNP operation. Readings were below the cleanup criteria of 5 pR/hr above background, or 12.1 pR/hr including background.

For the biased survey, exposure rates were measured at affected area soil sample sites where it was determined that radioactive content of surrounding structures would not influence the measurements. Measurements made at two locations were influer.ced by the RWST and were not included. Exposure rates at four locations were not measured because of radiation levels from the RWST. Exposure rates at two locations were not measured because of radiation levels from the Low Level Radioactive Waste Storage and Fuel Buildings. The values at the remaining locations ranged from 6.0 to 8.3 pR/hr with a mean of 6.8 pR/hr. This is consistent with background data and is below the cleanup criteria of 12.1 pR/hr.

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r- TROJAN DECOAIAllSS10NING PL4N k 3.1.2.4.6 Summary of Environmental Results The site characterization survey data indicated envirorunental radioactivity and radiation exposure levels were below the remediation critena and no environmental remediation is expected. This conclusion was based,in part, on the following results.

1. Surface soil samples analyzed for the scoping survey phase were below the release value of 15 mrem /yr TEDE with respect to "'Cs activity. Subsurface samples taken in 1991 from the tank farm area were also below this guideline value. The soil samples analyzed for "Sr contained less than 0.91 pCilg. The 15 mrem /yr TEDE calculation assumed 0.91 pCi/g "Sr.

2. Surface water. samples analyzed for the scoping survey did not have detectable levels of nonnaturally occurring radionuclides. Additional water samples are not necessary for site characterization, but will be performed during the final survey.

3. Bottom sediment samples analyzed for the scoping survey were below the guideline value of 15 mrem / year TEDE with respect to "'Cs content. "Sr content in sediment samples was below the 0.91 pCi/g on which the guideline value was based. No remediation is expected.

4. Pavement scans in unaffected areas measured beta contamination levels below the cleanup criteria of 5000 dpm/100 cm2 contained in the TNP Decommissioning Plan.

There are no release guidelines for activity per unit mass concentrations of radionuclides in pavement, llowever, "'Cs activity detected in pavement samples was below the release guideline level of 6.1 pCilg calculated for soil.

5. Unaffected area exposure rate results are well below the cleanup criteria of 5 pR/hr above background. The affected area samples have exposure rates below the cleanup criteria of 5 pR/hr above background. Those affected area samples not surveyed during this phase will be surveyed when radioactive material has been removed from structures that could influence the measurements.

3.

1.3 CONCLUSION

in summary, several general overall conclusions regarding the site characterization survey can be made about the four sections: structures. systems, activation, and environment.

First, plant structures contain radioactive material which will require removal prior to license termination. The contamination consists of radioactive material incorporated (fixed) into the upper layer of concrete / block and deposited on the surface (loose). Although the levels of n

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TROJAN DECOAIAfISSIONIVO PUN

radioactivity are generally low, structures, including building surfaces and piping, are considered potentially contaminated and will require, as a minimum, a wipe / wash down.

Second, some plant systems contain deposited radioactive material due to plant operation. The majority of the radioactive material is contained in reactor system piping and systems directly connected to the reactur coolant system (e.g., CVCS, safety injection system, and residusd heat removal system). Although some systems contain contamination, the systems are not expected to be greater than Class A waste.

Third, activated components contain the vast majority of the radioactive material not contained in fuel, Most activity is primarily concentrated in the vessel internals and shield wall. The reactor vessel lower internals contain the highest activity including subcomponents that, if I

removed separately from the reactor vessel and segmented, are estimated to contain I concentrations of radioactive material greater than Class C limits. If the internals are I segmented, portions of these components will be stored initially in the spent fuel pool or I transferred directly to the ISFSI. Neutron activation products have been found in samples of I containment concrete in various structures, including the reactor vessel shield wall, steam generator missile shields, and the containment wall itself. Remediation of the activated components will be required to meet the site release criteria and facilitate license tennination.

Fourth, and finally, the environmental survey results indicated that no radioactive material requiring remediation is present in the various materials sampled. The final survey will require

\ additional background data for a number of the sample media. Soil and sediment a. are indicative of at ler.st two background populations indicating that characterization of u.e sample location as to the soil type will be required. Data for direct radiation measurements will require remediation of site struc'.ures to remove interfering radiation sources. Preliminary results indicate no radioactivity at TNP has been spread to the environment outside plant systems or structures in quantities requiring remediation.

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TROJAN DECOAIAllSSIONING PLAN bi V regulatory requirements, and TNP procedures. The Manager, Personnel / Radiation Protection has overall responsibility for ensuring that radiation exposures are ALARA.

3.2.5.3 Engineering Management Engineering Management is responsible for ensuring that radiation exposures during planned engineering activities are maintained ALARA, and that engineering personnel comply with radiation protection requirements and maintain their radiation exposures ALARA.

3.2.5.4 Managers and Sunervisors Managers and supervisors have responsibilities related to the Radiation Protection Program including the following

1. Maintaining ALARA awareness and cooperating with Radiation Protection to provide individual personnel with the understanding and the means to minimize their own exposures;

2. Ensuring that personnel assigned to work with radioactive material attend required training; and

3. Ensuring personnel under their direction comply with radiation protection requirements.

3.2.6 RADIATION PROTECTION PROGRAM IMPLEMENTATION "te purpose of this section is to summarize how TNP's Radiation Protection Program will be implemented during decommissioning to maintain radiation exposure ALARA. The Radiation Protection Program is implemented and audited in accordance with approved plant procedures.

Additional details concerning TNP's Radiation Protection Program are provided in the DSAR and in radiation protection implementing procedures.

3.2.6.1 Radiation Protection Eauipment and Instrumentation The various equipment and instmmentation for conducting radiation surveys and measuring and minimizing personnel exposure are summarized in this section. Additional information on the facilities for radiation protection activities, and the procedures and equipment employed for measuring and minimizing personnel exposure, is provided in the DSAR.

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TROJAN DECOMMISSIONING PLAN 3.2.6.1.1 Laboratory Radiation Protection Instrumentation The laboratory type radiation instrumentation includes the following:

1. liigh resolution solid state detector (s) provided with lead shielding and a multichannel analyzer;

2. A beta counting system using Geiger Mueller detectors;

3. A liquid scintillation counter;

4. A low background thin window gas flow proportional counter; and

5. An alpha counting system (solid state detector).

Counting emeiencie, oflaboratory radiation detectors have been determined with certified radionuclide standards. A periodic calibration check is performed to check the emclency of"in use" laboratory radiation detectors. Additional detail regarding calibration, testing, and maintenance oflaboratory radiation protection instrumentation is provided in the DSAR and in radiation protection implementing procedures.

D i

,d 3.2.6.1.2 Portable Radiation Detection instrumentation The ponable radiation detection instrumentation available for use within the plant includes the following:

1. Alpha detectors having count rate output;

2. lonization chamber instruments equipped with a beta window and correction factor for beta measurement; and

3. Wide-range Geiger Mueller instruments having dose rate and count rate output.

Additional detail regarding the use storage, ca.libration, testing, and maintenance of portable radiation detection instrumentation is provided in the DSAR and in radiation protection implementing procedures.

3.2.61.3 Portable Air Sampling Instrumentation The portable air sampling instrumentation includes the following:

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l. Orab air samplers; and

2. Air samplers equipped with a filter and detector for the collection and counting of particulates.

The grab air sarnplers are used to collect samples of radioactive particulates for subsequent analysis in the laboratory. Periodic t.ampling oflocalized areas is conducted prior to and during personnel entry in accordance with radiation protection implementing procedures.

The air monitors collect and measure gross activity concentrations of airborne radioactivity.

These monitors are stationed as necessary in airborne radioactivity areas during personnel occupancy, and will wam ofincreasing radioactivity levels. The air monitors can also be employed for routine surveys of gross airbome radioactivity levels throughout the plant.

Additional detail regarding the use, storage, calibration, testing, and maintenance of portable air sampling instrumentation is provided in the DSAR and in radiation protection implementing procedures.

l 3.2.6.1.4 Personnel Radiation hionitoring Instrumentation q Worker radiation monitoring instrumentation includes the following:

b 1. Direct reading pocket ion chambers; l

} 2. Thermoluminescent dosimeters (TLDs); and

3. Digital alarming dosimeters (DADS).

Additional detail regarding the use, storage, calibration, testing, and maintenance of personnel radiation monitoring instrumentation is provided in the DSAR and in radiation protection implementing procedures.

3.2.6.1.5 Area Radiation hionitoring Instrumentation The ARhtS supplements the personnel and area radiation monitoring of the plant Radiation Protection Program. Radiation detectors provide local and/or remote indication and alarm of direct radiation dose rate. The ARMS measures radiation levels over the range of 1.0 x 104 to 1.0 x 10' mR/hr. Section 2.2.5.28 provides additional description of the ARh1S as well as temporary area radiation monitoring instrumentation that will be available, as necessary, to monitor the spent fuel pool and support decontamination and dismantlement activities during the decommissioning of TNP.

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The radiation monitors have been calibrated by the manufacturer. The manufacturer's calibration is traceable to certified National Bureau of Standards / National Institute of Standards and Technology or commercial radionuclide standards. Following repairs or modifications, the monitors are recalibrated at the plant with the secondary radionuclide standards. Additional details on the use, calibration, testing, and mairitenance of area radiation monitoring instrumentation are provided in radiation protection implementing procedures.

3.2.6.2 Control of Radiation Exnosure to the public 3.2.6.2.1 Radiological Efiluent Monitoring The ODCM contains the Radioactive Emuent Controls Program required by TNP Technical Specifications. Implerm:ntation of this program ensures compliance with the requiremena of 10 CFR 50.36a, " Technical specifications on emuents from nuclear power reactors;"

10 CFR 20; 10 CFR 50 Appendix 1 " Numerical guides for design objectives and limiting conditions for operation to meet the criterion "as low as is reasonably achievable" for radioactive material in light water cooled nuclear power reactor efquents;" and 40 CFR 190,

" Environmental radiation protection standards for nuclear power operations."

Installed and temporary process and emuent monitoring systems necessary to support

(~N decommissioning activities are described in Section 2.2.5.28. These systems monitor liquid

() and airborne emuent discharges from the plant per the requirements of the TNP Radiological Emuent Controls Program. The emuer.t sampling and analysis schedules comply with the NRC positions described in Regulatory Guide 1.21," Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Oaseous Emuents for Light Water Cooled Nuclear Power Plants," Revision 1. Additional details of the policy, methods, frequency, and procedures for emuent monitoring are provided in the ODCM and radiation protection implementing procedures.

3.2.6.2.2 Radiological Environmental Monitoring Monitoring, sampling, analyzing, and reporting radiation and radionuclides in the environment is performed in accordance with the Radiological Environmental Monitoring Program, which is required by TNP Technical Specifications and is incorporated into the ODCM. The Radiological Environmental Monitoring Program provides representative measurements of radioactivity in the highest potential exposure pathways and verification of the accuracy of the effluent monitoring program and modeling of envirotunental exposure pathways. The Radiological Environmental Monitoring Program is periodically reviewed to address changing plant conditions and regulatory requirements in accordance with plant procedures. Additional details of the policy, methods, and procedures associated with the Radiological Environmental O

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TROJAN DECOAIAflSSIONING PL4N Q-C Monitoring Program are provided in the ODCM and radiation protection implementing procedures.

3.2.6.3 Control of personnel Radiation Exposure Personnel radiation exposure is maintained ALARA by a combination of shielding, access control, contamination control, surveys and monitoring, work planning, training, and sound radiation protection practices implemented by TNP plant procedures. As specified in TNP Technical Specifications, the procedures for personnel radiation protection are prepared consistent with the requirements of 10 CFR 20 and are c.pproved, maintained, and adhered to for activities involving personnel radiation exposure.

3.2.6.3.1 Shielding The objective of facility radiation shielding is to teduce external doses to plant personnel,in conjunction with a program for controlling personnel access and occupancy in radiation areas, to levels which are both ALARA and within the regulations defined in 10 CFR 20. Radiation protection implementing procedures provide for evaluation of the use of temporary shielding for activities involving high dose rates.

p 3.2.6.3.2 Accet s Control and Area Designations Q

In general, access to plant buildings and the Industrial Area is controlled by locked doors or gates. Radiologically controlled access within the industrial Atea of the plant is determined by the radiation level, the degree of contamination, or the presence of radioactive materials in the various areas.

A RC A is an area where access is controlled for the purpose of protecting individuals from exposure to radiation. Within the RCA, access to areas of higher radiation or contamination levels is further controlled and defined in accordance with 10 CFR 20 and radiation protection implementing procedures. Plant procedures also describe the requirements for radiological postings advising workers of potsntial radiologival hazards at the entrance and boundaries of radiologically controlled areas.

3.2.6.3.3 Facility Contamination Control Plant and radiation protection implementing procedures direct the use of various practices and equipment to ensure general plant area contamination is controlled at the source to the greatest extent possible. Additional contamination controls are specified forjobs insolving high levels of contamination (e.g., a double step-off pad, additional surveys, etc.). Appropriate contamination centrols are used when carrying containinated tools and equipment between O

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TROJAN DECOMMISSIONING PLAN areas. Geiger Mueller count rate meters (friskers) are located within the plant so that personnel can determine if they have been contaminated prior to entering another area of the plant. The Gnal checkpoint for personnel leaving controlled areas of the plant is the access control point.

Temporary exit points may be established at remote control areaa as needed.

Airborne contamination is minimized by minimizing loose contamination levels and their sources. The use ofinstalled and temporary ventilation systems prevents the build up of air contamination concentrations. These systems are described further in Sections 2.2.4.8, 2.2.5.20, 2.2.5.22, and 2.2.5.23.

Additional details on the policy and methods for controlling general area and altborne contamination are contained in radiation protection implementing procedures.

3.2.6.3.4 Personnel Contamination Control Contamination of personnel is controlled by the use of several types of protective clothing when entering contaminated areas, in the event that levels of airborne contamination approach or exceed applicable limits, provision is made for personnel to use respiratory protective equipment. Allowances are made for the use of respiratory protective equipment, as speci6cally authorized by the NRC, in determining whether individuals in restricted areas are p - exposed to concentrations in excess of the values speci6ed in 10 CFR 20. The use of Q respiratory protection equipment is consistent with the goal of maintaining the total effective dose to personnel ALARA.

Additional details on the policy and methods for controlling personnel contamination are contained in radiation protection implementing procedures.

3.2.6.3.5 Area Surveys Radiation protection personnel perform routine radiation surveys of accessible areas of the plant. These surveys consist of contamination suneys, air samples, and external radiation measurements as appropriate for the speci6c area. Additionally, specific surveys are performed as needed for operational and maintenance functions involving potential exposure of personnel to radiation or radioactive materials. Specific activities requiring these non routine surveys include initial system opening, equipment release to uncontrolled areas, and response to radiation alarms. Additional details on the policy, methods, frequencies, and requirements for conducting both routine and non routine radiation surveys are contained in radiation protection implementing procedures. These procedures specify the types and suitability of instrumentation and methods to be employed when performing surveys and actions required when abnormal radiological conditions are discovered.

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TROJ.4N DECOAfAflSSIONING PL4N V 3.2.6.3.6 Personnel Monitoring TLDs are wom by plant personnel within radiologically controlled areas to measure radiation dose. Personnel assigned TLDs are also required to wear a direct reading pocket ion chamber on a DAD when entering RCAs. The intemal deposition of radioactive materials in personnel working in RCAs of the plant is evaluated primarily by a whole body count. Urinalyses are performed on plant personnel involved in plant activities where airborne concentrations of tritium exceed the limits of 10 CFR 20.

Additional details on the policy, methods, and frequency of personnel monitoring are contained in radiation protectica implementing procedures.

3.2.6.3.7 Radiation Work Pemiits Work in RCAs is performed under the authorization of radiation work permits issued by radiation protection personnel. These permits state protective clothing and dosimetry requirements, monitoring requirements, and special notes or cautions pertinent to the job.

These permits also specify the maximum contamination level, radiation level (including hot spot contact radiation level), and airborne radiation level for the worker to be entered under that radiation work permit, or will instruct the worker where to obtain such information. Additional details on the use of radiation work permits are contained in radiation protection implementing O procedures.

3.2.6.3.8 Training Workers requiring unescorted access to the Industrial Area receive General Employee Training which includes radiological protection fundamentals. Personnel who require access to RCAs at TNP receive radiation protection training in accordance with 10 CFR 19, Notices, instructions and reports to workers: inspection and investigations," and commensurate with the individual's responsibilities. The training process and requirements for gen:ral employees and radiation workers are summarized in Section 2.4 and are described in plant and radiation protection

• implementing procedures, in addition to radiation worker training, sepr. rate and detaited instruction in advanced radiation work practices is provided to those workers performing tasks that involve signincant exposure to radiation or quantitles of radioxtive material. The need for specialized ALARA training is evaluated during ALARA reviews or radiation work permit preparation in accordance with radiation protection implementing procedures. Specialized ALARA training includes such items as mock ups, dry runs, pre job brienngs, and other special training classes.

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TROJ.4N DECOAfAt/SS10NING PL4N

(,/ Specialized respiratory protection training is required for radiation workers who use respiratory protection devices.

3.2.6.3.9 Controls, Practices, and Special Techniques Radiation protection implementing procedures specify that during the planning phase for activities in high dose rate areas or high airborne areas. various engineering controls to rninimize exposures should be evaluated and/or implemented. These engineering controls and practices include, but are not limited to: temporary shielding; remote surveillance equipment; multi discipline input regarding ALARA goals; pre job,in progress, and post job briefings; end adequate lighting, ventilation, work space, and work area accessibility. Additional details on the implementation of controls, practices, and other techniques that are used to meet the radiation protection standards of 10 CFR 20, including ALARA, are contained in radiation protection implementing procedures.

3.2.6.3.10 Radioactive Materials Safety Equipment and fluids in ceitain plant systems became contaminated during plant operation.

These contaminated materials, together with radioactive materials contained in spent fuel, seakd sources, and instrument calibration devices, can result in radiation exposure to plant r pers nnel. Pr cedures, facilitics, and equipment for handling, processing, and disposing of i radioactive gaseous, liquid, and solid wastes are described in Sections 3.3.2 and 3.3.3.

Recognized methods for the safe handling of radioactive materials will be implemented to mair.tain potential external and internal doses ALARA.

External doses are minimized by a combination of time, distance, and shielding considerations.

Internal doses are minimized by the measurement and control ofloose contamination. The materials safety program is defined by written radiation protection procedures.

Additional details on the materials safety program are contained in radiation protection implementing procedures.

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TRo.MN occomflSSloNING l'UN V 3.3 RADIOACTIVE WASTE MANAGEMENT Radioactive waste management activities during the TNP decommissioning include activities related to spent fuel management; and gaseous, liquid, and solid radioactive waste processing and disposal.

Spent fuel will be stored onsite in accordance with the requirements of the TNP Techr,1 cal Specifications. The processing and disposal of gaseous, liquid, and solid radioactive waste will be managed in accordance with the Radiation Protection Program, Process Control Program, ODCM, Radioactive Ef0uent Controls Program, Radiological Environmental hionitoring Program, and Storage Tank Radioactivity Monitoring Program.

TNP policy for control'bf radioactive wastes is to minimize the amount of waste material generated, and to maintain the discharge of radioactive material below the design objectives provided in the ODCM. To ensure waste minimization goals are achieved during decommissioning, radiation workers will receive training in waste minimization procedures and practices. The TNP radioactive waste control program defines responsibilities and provides guidance for the minimization of radioactive wastes.

3.3.1 SPENT FUEL MANAGEMENT PROGRAM This section describes the program for rnanagement of the spent fuel at TNP during and following decommissioning until title and possession of the spent fuel is transferred to the DOE. The estimated costs and associated funding plan for implementation of the TNP fuel management program are described in Section 5.

The descriptions of the spent fuel management program and associated funding plan provided in this section and in Section 5, respectively, fulfill the requirements of 10 CFR 50.54,

" Conditions oflicenses," Paragraph (bb), which stipulates that " nuclear power reactors licensed by the NRC...shall, within 2 years following permanent cessation of operation of the reactor....

submit written notification to the Commission...of the program by which the licensee intends to manage and provide funding for the management of all irradiated fuel at the reactor until title to the i Tadiated fuel and possession of the fuel is transferred to the Secretary of Energy for its ultimate disposal in a repository."

3.3.1.1 Spent Fuel Management Program Descrintion The spent fuel (irradiated fuel) currently stored in the spent fuel pool will remain there until licensing and construction of an ISFSI is completed on the TNP site. Use of an ISFSI was determined to be the most economical method for the temporary storage of the TNP spent fuel until a DOE or other offsite facility is available. Relocation of the 3 47 Rtvision 4

T10]AN DECOAfAflSSIONING l'LAN spent fuel, and other high level radioactive waste stored in the pool, to the ISFSI would allow decontamination and dismantlement of structurcs, systems, and components throughout TNP to proceed without impacting the safe storage of the spent fuel. This action will allow earlier termination of TNP's Part 50 license.

The spent fuel, and other high level radioactive waste stored in the pool, will be relocated to the 1SFS1 prior to:

1. lleginning decommissioning activities that have the potential to adversely alTect the spent fuel pool and its contents; and

2. Ileginning decommissioning of systems and components needed for moving and loading spent fuel into casks for storage !n the ISFSI.

The ISFSI will include the capability to traasfer spent fuel from a storage cask to a shipping cask for shipment directly to an offsite repository. Since the spent fuel pool will be decontaminated and dismantled prior to the shipment of spent fuel to a permanent offsite facility, the capability to transfer the fuel from a storage cask to a shipping cask using the spent fuel pool will no longer exist.

3.3.1.2 Effects of Permancat Repository Schedule on Snent Fuel Management Plan (3

V Under the terms of the " Standard Contrect for Disposal of Spent Nuclear Fuel and/or liigh.

Level Radioactive Waste" executed between PGE and the DOE, the DOE has the responsibility, following commencement of operation of a repository, to take title to and possession of the TNP spent fuel and high level radioactive waste as expeditiously as practicable upon the request of PGE. The scope of the contract states, in part, that the semices to oc provided by DOE shcIl begin, after commencement of facility operations, not later than January 31,1998. This contract clause provides the basis for the schedule forecast in the DOE's annual Acceptance Priority Ranking and Annual Capacity Report for receipt of spent fuel and/or high level waste.

if a DOE lacility is in operation as of January 31,1998, then the first shipment from TNP would occur in 2002. Shipments are forecast to continue through 2018. This projection is based on DOE's 1991 Acceptance Priority Ranking (DOE /RW 0331P, December 1991),1991 Annual Capacity Report (DOE /RW 0328P, December 1991), and an extrapolation beyond the 10 year DOE outlook. This schedule was used to develop the decommissioning cost estimate described in Section S.

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TROJ4N DECOAIAflSSIONING PUN V 3.3.l.3 Licensing Activities to Supnort the Sprnt Fuel Management Plan POE will file an cpplication for a license for an ISFSI in accordance with 10 CFR 72, Subpart B, which specifies the NRC licensing requirements for the independent storage of spent fuel and high level radioactive waste. POE will also petition the State of Oregon to adopt rules allowing storage of spent fuel in an ISFSI. Transfer of spent fuel and high level radioactive waste from the spent fuc.) pool to the ISFSI will commence after NRC issuance of the 10 CFR 72 license, and issuance of required State of Oregon approvals.

3.3.2 RADIOACTIVE WASTE PROCESSING 3.3.2.1 Gaseous Radioadlyily Gaseous radioactivity is expected to be limited primarily to airborne radioactive particulates generated during decontamination and dismantlement activities.

Airborne radioactive particulates will be filtered through IIEPA filters in the containment ventilation system, the Auxiliary Building and Fuel Building ventilation systems, and the Condensate Demineralizer Building ventilation system, portions of which will be maintained in operation during decontamination and dienantlement activities in those buildings (see Sections p, 2.2.4.3,2.2.5.20,2.2.5.22, and 2.2.5.23). Local temporary ventilation systems with IIEPA V filtration, or other approved alternate systems, may be used in lieu of or to supplement building ventilation for activities expected to result in the generatiois of airbome radioactive particulates.

Radioactive gaseous efnuents will be monitored and release limits adhered to in accordance with the methodology and parameters in the ODCM.

3.3.2.2 Liquid Radioactive Waste IJquid radioactive waste will be generated as a result of draining, decontamination, and cutting processes during plant decommissioning.

Portions of the existing liquid radioactive waste treatment systems (plant efnuent system, clean radioactive waste system, and dirty radioactive waste system) will be maintained in operation during decommissioning to process liquid radioactive wastes by filtering, demineralizing, and providing for holdup or decay of the radioactive wastes for the purpose of reducing the total radioactivity prior to release to the environment (see Sections 2.2.5.19,2.2.5.26, and 2.2.5.27). Temporary liquid waste processing systems may also be used to process liquid radioactive waste.

Radioactive liquid ef0uents will be processed in accordance with the ODCM.

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q . TRO!.4N DECOhlAtlSSIONsNG PL4N 33.23 SolidRadioactive Watti Solid radioactive waste generated during decommissioning will include neutron activated materials, contaminated materials, and radioactive wastes. Neutron activated materials include the reactor pressure vessel, reactor vessel internals components, and the concrete biological shield. Contaminated material and radioactive wastes include pipe sections, valves, tanks, other plant equipment, concrete surfaces., contaminated air filters, wet solid wastes from the processing of contaminated water volumes (lon exchange resins, cartridge filters), and dry solid wastes (rags and wipes, plastic sheeting, contaminated tools, disposable protective clothing).

The solid radioactive waste system spent resin transfer system, filter handling vehicle, solid waste compactor, and spent resin compactor will be maintained in operation as necessary l during decommissionifig to process solid waste (see Sec' ion 2.2.5.25). Temporary solid waste processing systems may also be used.

l Solid radioactive waste will be piocessed in accordance with the TNP Radiation Protection l

Program, Process Control Program, and plant procedures. The Process Control Program provides requirements for processing radioactive wastes requiring solidification, radioactive wastes requiring high integrity containers, and low activity dewatered resins and other wet wastes to ensure that shipping and burial ground requirements ue met with respect to

( ,) solidification and dewatering. To the maximum extent practicable, solid radioactive waste will V be decontaminated and compacted to reduce tM volume to be packaged for shipment to s n offsite disposal facility.

Waste container selection will be detennined by the type, size, weight, classification, and activity level of the material to be pt.n:nged. Examples of containers used at TNP include drums, metal boxes C. vans (container vans), and high integrity containers. Other special containers may be used as required.

3.3.2.4 Mixed Wastes Mixed wastes are wastes that contain both a hazardous waste component regulated under Subtitle C of the Resource Conservation and Recovery Act and a radioactive component consisting of source, special nuclear, or byproduct material regulated under the Atomic Energy Act. Plant procedures provide guidance for the minimization, control, and storage of mixed waste in accordance with the Environmental Protection Agency (EPA)

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TROJAN DECOMMISSIONING PL4N p

Q and NRC regulations. The use of potentially hazardous materials in radiologically controlled areas will be reviewed to minimize the generation of mixed waste.

TNP currently has approximately 60 fP of mixed waste stored onsite. Mixed waste will continue to be stored onsite until a pennanent storage or disposal facility becomes available.

3.3.3 RADIOACTIVE WASTE DISPOSAL TNP is located within the Northwest Compact. Radioactive solid waste will be shipped to the US Ecology site near Richland, Washington, if removed separately from the reactor vessel and I segmented, several of the reactor vessel internals components have radionuclide concentrations I in excess of the 10 CFR 61 Class C limits. These components may initially be stored in the I spent fuel pool or transferred directly to the ISFSI. Intact disposal of the reactor internals with I the reactor vessel will meet Class C burial waste criteria. I Packaging, storage, and shipment of radioactive waste generated during decommissioning will be controlled by the TNP Radiation Protection and Process Control Programs, and plant procedures. Plant procedures include requirements for:

1. Sorting and segregation of radioactive waste, and processing to an acceptable form; O 2. Classification , f redioactive waste in accordance with Department of Transportation U (DOT) and NRL :equirements; '

1 Packaging, labeling, and marking of radioactive waste in accordance with DOT and disposal site criteria;

4. S ,nr-e a i dioactive waste;

5. Receipt survey of vehicles used to transport radioa:tive waste;

6. Contamination surveys to ensure packages shipped meet DOT requirements for smearable contamination levels;

7. Radir .on surveys, e.g., package contact, vehicle contact, specified distances from the package and the vehicle, and normally occupied po sitions in the vehicle cab for the material and package and for the transport vehicle cepending on the type of shipment (e.g, low-specific activity, exclusive-use low-specific activity, etc.);

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TROJ.4N DECOMMISSIONING PL4N

8. Shipment of radioactive waste in accord:.ns- wi'b DOT and NRC requirements; and

9. Disposal and offsite volume reduction arrar.gements.

Radioactive waste storage facilities onsite include the Condensate Demineralizer Building, and, depending on the type and radiation levels, may also include areas of the Auxiliary and Fuel Buildings. Other temporary radioactive waste storage areas may be established as necessary.

A projection of radioactive waste generation (projected waste volumes, radionuclide concentrations, waste forms, and classification) is contained in Sections 2.2.7 and/or 3.1.2.

i Shd O 3-52 Revision 4

TROJ.4N DECOAIAllSSIONING PL4N t i V 3.4 EVENT ANALYSIS 3.4.1 OVERVIEW This section presents the results of evaluations and analyses of postulated decommissioning events and evaluates the potential for adverse effects on public health and safety. This evaluation includes postulated events 16 could be significantly different from accidents that have previously been evaluated for plant operations or maintenance. The analyses consider events related to decommissioning activities, loss of suppon systems, internal events, and external phenomena. The results of the analyses indicate that decommissioning activities can be conducted in a manner that does not significantly affect public health and safety.

Section 3.4.5 discusses radiological occupational safety during decommissioning.

3.

4.2 INTRODUCTION

The evaluation methodology was developed by generally wilowing the methodology presented in NUREG/CR-0130, " Technology, Safety and Costs of Decommissioning a Referenced Pressurized Water Reactor." Decommissioning activities and other occurrences which could cause radiological events with the potential for release of radioactive material beyond the Exclusion Area Boundary were identified, n

b Potential accidents involving the storage and handling of spent fuel are not within the scope of these analyses. TNP spent fuel is currently stored in the spent fuel pool. Potential accidents involving the storage and handling of spent fuel are addressed in the DSAR. Spent fuel will later be removed from the speut fuel pool and transferred to an ISFSt. Postulated accidents involving the transfer of spent fuel to the ISFSI and potential interactions with decommissioning activities will be addressed as part of the license submittal for construction and operation of an ISFSI in accordance with 10 CFR 72.

The decommissioning activities evaluated included events with the potential for liquid and/or airborne radioactive releases. During decommissioning activities, contaminated liquids will primarily be generated inside buildings. If there is leakage of these contaminated liquids, the liquid would flow to the floor and equipment drain system, and would be disposed of through the normal plant discharge system, which is a monitored release pathway. Engineering and/or administrative controls will be established to ensure the quantity of radioactive liquids stored within plant buildings does not exceed the capacity of the available liquid waste processing equipment. Temporary tanks will be used when installed equipment is removed or unavailable, in general, the volume of water generated by decontamination efforts is expected to be relatively small. A disposable liner could be used in an existing sump cavity to collect any p

3-53 Revision 4

TROJAN DECOAfAllSSIONING Pld V i

G' water. A portable pump could be used to transfer the water to a temporary holding tank for sampling and processing prior to fmal discharge. As is the case with the installed system, a tank level alarm and indication system could be used to notify Operations personnel when the tank is filled to some pre-determined level.

Section 2.2,4.8 describes that tempmary systems for decommissioning support may be utilized, including possible u.:e of temporary liquid processing systems. As described in this section, plant design change procedures would control temporary modifications to plant structures, systems, and components.

Administrative controls contain restrictions on the amounts of radioactive material that may be stored in temporary tanks (tanks used for temporary storage of contaminated liquids located exterior to buildings). These restrictions ensure that potential liquid releases from such temporary tanks are within the limits of 10 CFR 20, Appendix B, at the nearest potable water and surface water supply in unrestricted areas. Storage ofliquid wastes during decommissioning activities will be subject to these same restrictions, or alternatively, the wastes will be stored such that releases would be contained by appropriate engineered features such as dikes, dams, and overflows routed to plant drains. Design of the engineered features as described will generally follow the guidance of Regulatory Guide 1.143," Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-(p/ ' Vater-Cooled Nuclear Power Plants," and Generic Letter 81-38, " Storage of Low-Level Radioactive Wastes at Power Reactor Sites" where applicable or appropriate in all cases, good engineering practices will be followed.

It was determined that potential consequences of such events are less than the calculated doses at the Exclusion Area Boundary from an airborne release of radioactive material.

Consequently, only events with the potential for airborne releases are discussed in detail in this section.

The events that could involve airborne releases were then grouped into one of four categories:

1. Decommissioning activity events, including decontamination, dismantlement and materials handling events (Sections 3.4.4.1 through 3.4.4.3);

2. Loss of support system events, including loss of offsite power, cooling water, and compressed air (Section 3.4.4.4);

3. Fires and explosions (Sectiod.4.4.5 and 3.4.4.6); and,

4. Extemal events (Section 3.4.4.7).

O V 3-54 Revision 4

l l

l TROJ.4N DECOAIAllSSIONING PL4N Each of these events was evaluated for its potential offsite radiological consequences. Each

was evaluated to ensure that the resulting doses at the Exclusion Area Boundary would be less than 0.5 rem TEDE. Previous calculations have indicated that a 2.07 Curie release outside a  !

- building has the potential to result in a 0.5 rem dose at the Exclusion Area Boundary. The  !

analyses in this section use this value as the limiting release quantity. The major assumptions i that were used in the analyses are described in Section 3.4.3.2.

4

, The sections describing each type of event contain a brief explanation of the reason a particular

scenario was determined to be the bounding case, This discussion is followed by a summary of the analysis of the particular scensrio.

3.4.3 LIMITS AND A3SUMPTIONS 3.4.3.1 Radionuelide Release Limits The EPA has established protective action guidelines, EPA 400 R-92-001, " Manual of

. Protective Action Guides and Protective Actions for Nuclear Incidents," October 1991, that specify the potential offsite dose levels at which actions should be taken to protect the health and safety of the public. The EPA protective action guidelines (PAGs) are limiting values e based on the sum of the effective dose equivalent resulting from exposure to ex:ernal sources

' and from the committed effective dose equivalent incurred from the significant inhalation pathways during the early phase of an event. The EPA PAG limits are:

EPA PAGs. rem TEDE 1

Thyroid Committed 5 Dose Equivalent (CDE) l Skin CDE 50 Following permanent shutdown of Trojan, PGE analyzed the potential accidents that could occur in a permanently defueled state. PGE concluded there were no potential accident l scenarios that could lead to Exclusion Area Boundary doses in excess of EPA PAGs. Based on the results of these analyses, PGE requested exemption from the offsite emergency preparedness requirements of 10 CFR 50.54(q). The NRC subsequently granted this exemption.

The Food and Drug Administration has established preventive PAGs for low impact protective actions at projected radiation doses of 0.5 rem TEDE, bone marrow CDE, or other organ CDE and 1.5 rem thyroid CDE.

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TROJAN DECOMMISSIONING PLAN (s

PGE has conducted additional evaluations to ensure that decommissioning activities described in this plan will not create the potential for new or different events that could cause doses at the Exclusion Area Boundary to exceed preventive PAG levels.

To ensure that the maximum dose at the Exclusion Area Boundary would be maintained less than or equal to these limits, calculations were performed to determine the amount of radioactive material that would have to be released to result in a limiting dose of 0.5 rem at the Exclusion Area Boundary. This calculation was performed for three possible release locations.

Release Airborne Location Activity Limit Inside a building with 2,840 Ci filtered ventilation Inside a building without 34.5 Ci filtered ventilation Outside a building 2.07 Ci U Airborne releases of the magnitude shown above potentially result in radiation exposures of 0.5 rem at the Exclusion Area Boundary using the conservative assumptions discussed in Section 3,4.3.2.

3.4.3.2 Assumntions The following assumptions have been incorporated into the TNP decommissioning event analysis:

1. Dose calculations for releases of airborne activity from an event involving contaminated components are based on dislodging ten percent of the contamination, of which one percent becomes airborne (except "'I and tritium, of which 100 percent of that dislodged becomes airborne.);

2. Airbome releases of radioactive materials are assumed to pass through a HEPA filtration system when the release originates inside a building with filtered ventilation.

HEPA filter efficiency is assumed to be 99.97%;

3. For releases that originate inside a building without filtered ventilation, the guidance of Regulatory Guide 1.78," Assumptions for Evaluating the Habitability of a Nuclear

/~'T

\') 3-56 Revision 4

I TROJAN DECOMAllSSIONING PLAN Power Plant Control Room During a Postulated Hazardous Chemical Release,"

Revision 0, was used to obtain an activity decay rate of 6.00 x 104/hr;

4. Conservative meteorological conditions (Pasquall class F with wind speed 1.13 m/sec) and release elevations are assumed;

5. Dose conversion factors were taken from EPA 520, Table 2.1, Revision 2.

6. Doses are early phase projections during the first two hours or less.

7. The north sector site boundary at 662 meters is the location for which doses are calculated.

8. Breathing rate is 3.33E-4 m 3/sec.

9. A plume depletion factor of 0.92 is used.

10. Credit was taken for diffusion due to building wake effect.

I1. Isotopic concentrations found in the primary side of the Trojan steam generators projected to June 1994 were used for dose predictions. This isotopic mixture is

(] conservative for contaminated components and activation of concrete components.

(/

The amount of activity for each isotope was determined as a percentage of the total activity present in the isotopic mixture. This percentage was then normalized to I curie, and the projected dose from an airborne release was calculated. The isotopic mixture of activated metallic components was not utilized since credible events did not result in airbome release of these materials.

3.4.4 RADIOLOGICAL EVENT IDENTIFICATION Decommissioning activities and other occurrences which could cause radiological events that were considered include the following:

1. Decommissioning activity events, including decontamination, dismantlement, and materials handling events;

2. Loss of support system events, including loss of offsite power, cooling water, and compressed air;

3. Fires and explosions; and n) b 3-57 Revision 4

TROJAN DECOAIAllSSIONING PLAN L 4. External events (e.g., earthquakes, tornadoes, etc.).

Each of these types of events was evaluated for its potential offsite radiological consequences.

The results of those evaluations are presented below.

Potential accidents involving the storage and handling of spent fuel are not within the scope of these analyses. Spent fuel from Trojan is currently stored in the onsite spent fuel pool.

Potential accidents involving the storage and handling of spent fuel are addressed in the DSAR.

Spent fuel will later be removed from the spent fuel pool and transferred to an ISFSI.

Postulated accidents involving the transfer of spent fuel to the ISFSI and potential interactions with decommissioning activities will be addressed as part of the license submittal for construction and operation of an ISFSI in accordance with 10 CFR 72.

To ensure dismantlement activities will not impact the safe storage of spent fuel, dismantlement will be imp!cmented in accordance with administrative controls that require, in part, an evaluation of activities in accordance with the requirements of 10 CFR 50.59. Work packages will include specific steps to physically protect the systems, structures, and components supporting spent fuel storage, or establish safe load paths and protective zones around these systems and structures.

N Sources of radioactive material associated with decommissioning activities can be separated (d into two categories; contamination and activation products.' Activation products are contained within structures and components and are therefore not available for airborne release during most credible events. Contamination ir considered to be more readily available for airbome ,

release and constitutes the major concern for offsite doses. Section 3.1 provides a more detailed discussion of the radiological characterization of the site.

3.4.4.1 Decontamination Events Plant systems, structures, and components may be decontaminated in order to reduce worker radiation exposure rates or to allow release of materials from the plant. Decommissioning decontamination events relate to activities associated with hand-held water lance surface cleaning operations and chemical Jecontamination practices. The events could involve items such as:

1. Gross leakage ofin situ decontamination equipment; and

2. Accidental spraying ofliquids containing concentrated contamination.

Decontamination methods typically use liquids to remove radioactivity from the surface (e.g.,

chemical decontamination, high pressure water washing). Contaminated liquid wastes C 3-58 Revision 4

TROJAN DECOMMISSIONING PLAN C\

U generated during decommissioning operations will be sent to the plant liquid waste storage system or to other tanks that are designated for temporary storage. Administrative controls l contain restrictions on the amounts of radioactive liquids that may be stored in temporary tanks.

These restrictions ensure that potential liquid releases from such temporary tanks are within the limits of 10 CFR 20, Appendix B at the nearest potable water and surface water supply in unrestricted areas. Storage ofliquid radioactive waste generated during decommissioning activ; ties will be subject to these same restrictions, or altematively, the wastes will be stored such that releases would be contained by appropriate engineered features such as dikes, dams, and overflows routed to plant drains.

Liquid radioactive wastes generated during decommissioning tvill be filtered and/or demineralized in a liquid radioactive waste system so that liquid releases remain within the limits established by tre Facility Operating (Possession Only) License NPF-1, the ODCM, and 10 CFR 20.

3.4.4.1.1 In Situ Decontamination of Systems Although large scale chemical decontamination of systems is not anticipated as part of TNP decommissioning, limited application may be used on systems or tanks to reduce radiation dose rates prior to dismantlement or general area decontamination.

O V Chemical decontamination is typically performed by recirculation of a decontamination solution throughout a system or tank until am. lysis of samples indicates that the desired decontamination level has been achieved. One system volurae is used to minimize the quantity of contaminated liquids, thus an extended recirculation time of several days may be required.

The system is typically heated to 80*C and may be pressurized.

This in situ decontamination method is performed on piping systems which form a closed loop.

There is normally no interface between the circulating solution and the air outside the system.

This provides a low potential for airbome release outside the system. However, any fluid system has the potential for leaks.

To provide a bounding analysis for chemical decontamination the methodology and assumptions of NUREG/CR-0130 are applied. The assumptions used in the calculation are:

1. Chemical decontamination of any system or component will not exceed one week of continuous operation;

2. A leak rate of I gpm (3.81/ min) may go undetected during decontamination operation; f3 O 3-59 Revision 4 m

TROJAN DECOAfAt/SSIONING PLAN

^O 4 3. ' The decontamination solution density is approximately that of water (62.4 lb/ft' or 1.

g/cm));

4. Airbome droplet concentration generated due to stream or drop type leaks is assumed i

to be 10 mg/m3(this results in an airbome release fraction of 10 5); and, l

! 5. ' For spray type leaks, a maximum of 0.3% of the solution will be in the size range that I could be transported as airbome activity.

4 l .- A leak during chemical decontamination would be in the form of a stream, drops, or spray.

p At 80T, it is anticipated that a solution is in the vapor phase. A stream or drops ofliquid -

do not present a significant airbome release potential. Using the above assumptions, a

[ maximum concentratioTi (Ci/l) for decontamination solutions can be determined to ensure the

!- releases do not result in conditions which would exceed applicable limits. - No credit is taken L for building ventilation or filtration.

For conservatism, it was assumed that 2.07 Ci (the limit for an outside release location) is the maximum airborne activity limit. The maximum solution activity associated with a

decontamination leak is calculated by the equation:

O - Maximum Concentration = Airborne Activity Limit i V (Leak Rate)(Airborne Release Fraction)(Leak Duration)

. Stream or Drops Tvoe Leak l For leaks in the form of streams or drops, using a release fraction of 10-5 and a leak duration of L-- one week, the maximum concentration of the decontamination solution is

l Maximum Concentration = 2.07 Ci (3.81/ min)(10-5)(10,080 min)

j. TMaximum Concentration = 5.4 Ci/l(20.5 Ci/ gal)

! Assuming the entire system contamination inventory was contained within the volume of fkid 5

released (10,080 gal), systems containing s 2.1 x 10 Ci would not result in airborne activities in excess of applicable limits. Excluding the reactor vessel internals, there are no single -

.' - . systems or structures that contain activity of this magnitude.

p k-E.

4 v 3-60 Revision 4

?

p TROJAN DECOMMISSIONING PL4N Snray Tyne Leak A spray type leak would provide the maximum airbome release potential since the contamination would already be airbome and can be immediately entrained in local airflow.

Assuming an airborne release fraction of 0.3% (0.003) and a leak duration of one week, the maximum concentration of the decontamination solution is:

Maximum Concentration = 2.07 Ci (3.81/ min)(0.003)(10,080 min)

Maximum Concentration = 1.8 x 10' pCi/l(6.8 x 104 pCi/ gal).

Assuming the entire contamination was contained within the volume of fluid released (10,080 gal), system activities s 690 Ci would not result in airborne activities in excess of applicable limits. Chemical decontamination on systems which contain s 690 Ci will not result in airbome release activities which exceed applicable limits. There are no single systems at TNP with contamination levels as great as 690 Ci.

l The above calculations are conservative. They assume that the entire decontamination solution,

[]

V up to a maximum of 10,080 gal (1 gpm for I week) are released to the environment, as spray.

No credit is taken for operator action. Spray type leaks could be terminated by securing recirculation and depressurizing the system. For stream or hop type leaks, system depressurization should result in decreased leak rate. Additional measures such as leak isolation or collection could be used to further minimize consequences.

3.4.4.1.2 Surface Cleaning Techniques Several different techniques can be employed in decontamination of surfr.ces. These typically include wiping, washing, vacuuming, and waterjets (e.g., the use of hand-held high-pressure water jets). The high-pressure water cleaning is considered to provide the greatest potential for airbome activity generation and is discussed below.

The principle mechanism for airborne activity entrainment is the suspension ofliquid droplets c ntaining contamination. When the lance is sprayed initially, it produces droplets up to 300 pm in size, the size of fogs or mists. The spray particles then break into smaller particles when they impact on a surface. Thus there are significant amounts of small airborne droplets with considerable variation in the amount airborne.

Direct data is not available to define the quantity of droplets formed. A conservative estimate is made by assuming that a sufficient quantity of droplets are generated to maintain an airborne A

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TROJAN DECOAIAf/SSIONING PL4N O

V liquid concentration of 10 mg/m' with vigorous mixing in air. This is the maximum mass concentration that is found in air velocities less than 0.046 m/s. There is fairly constant weight in a spray of 10 to 20 pm particles of about 10 mg/m' even with gross entrainment oflarger particles. Since 10 pm particles and smaller are in the respirable size range, they are potentially hazardous. The quantity of radioactivity in these airbome droplets is influenced by many factors such as the quantity of radioactivity on the surface, the ability of the liquid to entrain radioactivity, and the contact between the liquid and the surface.

The operating parameters for the high-pressure spray vary with the requirements of the ,

situation. For the surfaces considered in this analysis,23 to k Vmin and 46.5 to 65 m 2/hr are reasonable estimates of the rage of solution flow rate and surface cleaning rate respectively.

To determine the maximum release potential, the largest cleaning rate (65 m2 /hr) with the lowest solution flow rate (231/ min) is used in the calculation.

Tables 3.1-4,3.1-5, and 3.1-6 provide a summary of contamination levels of various systems, buildings, and components and estimated contaminated surface areas. The spent fuel pool contains the highest activity lesel of systems, components, or structures which could be expected to be decontaminated by high pressure water lance. Using the methodology provided in NUREG/CR-0130 and the TNP specific values, the following values associated with water lance cleaning of the spent fuel pool can be calculated:

O O Total Surface Area (m )

pCi/m2 2 24,071 4150 Contamination Removed (pCi/ min) 4.5 x 102 Solution Concentration (pCi/l) 195 Solution Concentration (pCi/cm') 0.195 i

Airborne Concentration (pCi/m3 ) 1.95 x 10-3 Aerosol Generation Rate (pCi/ min) 52.8 Total Operation Time (min) 22,220 Total Generated Airbome Activity 1.2 Ci The above values are greater than could be expected based on several conservative assumptions. NUREG/CR-0130 assumes a decontamination efficiency of 90%; for conservatism 100% of the activity was assumed to be removed. The total airbome activity assumes that airborne activity generated remains suspended and is not removed by ventilation filtration, nor does it precipitate out for removal by water collection system. The above analysis also assumes water lance operations are continuous for approximately 370 hours0.00428 days <br />0.103 hours <br />6.117725e-4 weeks <br />1.40785e-4 months <br /> and are not terminated upon detection ofincreased airbome activity. Even under these conservative assumptions, the 0.5 rem TEDE limits would not be exceeded, b 3-62 Revision 4

I I

TROJAN DECOAfAflSSIONING PL4N O

G 3.4.4.2 Dismantlement Events Dismantlement events relate to activities associated with segmenting of components or structures and the removal of concrete. The potential for airbome release for each of these activities must be reviewed separately.

3.4.4.2.1 Segmentation of Components or Structures Segmentation of components or structures can be accomplished by disassembly, or cutting or other destructive methods.

Disassembly of components or structures does not result in destruction of material. The potential for radioactive material release is limited to dislodging contamination. Disassembly events are therefore considered bounded by the material handling event discussed in Section 3.4.4.3.

Segmentation of components or structures by cutting or other destructive methods (e.g., sawing, grinding, or plasma cutting) can result in releasing airbome activation products in addition to dislodging contamination. Detailed planning of dismantlement activities will ensure that systems, structures, and components that are contaminated or activated will be dismantled using methods that minimize the release and spread of contamination.

k>' Although activated metallic components contain the greatest activity levels, the potential for airborne release is limited to small fractions of material directly affected by cutting operations.

The reactor vessel intemal components contain the largest quantity of activated materials. The I reactor vessel intemals will either be disposed ofintact with the reactor vessel or removed I separately and segmented. Due to the highly radioactive nature of these components, I dismantling the intemals will be performed underwater if the segmentation option is chosen, i The water will not only provide shielding from direct radiation exposure but will also provide a retention media for the small particulate material r: leased during segmentation. Although the potential for release of airborne activity is considered remote, administrative controls will be established to ensure that airborne activity generated during segmentation will not exceed a small fraction of the activity limits corresponding to an Exclusion Area Boundary dose of 0.5 rem as discussed in Section 3.4.3.1. These controls are discussed in Section 2.2.5.3.

The dismantlement of the RCS piping is considered to provide the bounding analysis for generation of airborne activity. Dismantlement activities cannot be performed underwater, as may be done with the reactor vessel intemals, and the RCS piping contains the greatest I contamination of the systems to be dismantled,221 Ci. The guidance provided by NUREG/CR-0130 was used to determine the amount of activity that could be generated during Il

%J 3-63 Revision 4 J

TROJANDECOMMISSIO? VG PLAN segmentation. To determine the total activity generated from segmentation the following equation was used:

Total Activity Generated = (Surface Contamination Level)(Kerf width) x x(Length ofID per cut)(Number of Cuts)

To determine the maximum generated activity the following values were used:

Surface contamination activity for the RCS was 55 pCi/cm2 . This is the highest activity level calculated for the RCS and was used for determining contamination levels of CRDMs and reactor coolant pumps.

Kerf width was 0.95 crti. This is consen ative since it is the largest kerf of possible cutting methods that may be employed.

The length ofID per cut 78.7 cm (31.7 in). This is conservative since it is the largest sized section of piping.

The number of cuts is assumed to be 55. This is an estimated number of cuts needed to segment the RCS piping and is considered conservative for this analysis since the airborne release during a segmentation event would only involve the cuts being performed just prior to pd the release.

Using the above equation a maximum release to the Containment Building atmosphere is calculated to be 0.71 Ci. This activity is conservatively assumed to become. airborne. This limit is significantly below the 2,840 Ci limit for a release from inside a building with filtered l ventilation such as the containment building and is well below the 2.07 Ci assumed for releases outside a building. Based on this analysis it is concluded that dismand . ment activities cannot result in events which would exceed the 0.5 rem TEDE limits.

3.4.4.2.2 Removal of Concrete Three techniques are available for the removal of concrete. These are rock cplitting, explosives, and electric or pneumatic hammers. Each of these techniques has a different potential for the release of airborne radioactivity and may be used within different areas of the plant.

The bounding analysis for concrete removal is considered to be the primary shield wall. The primary shield wall, located inside the Contairunent Building, contained an estimated 351 Ci of total activity one year after plant shutdown. For the purposes of this bounding analysis explosives are assumed to be used in dismantlement of the primary shield wall.

3-64 Revision 4

m TROMN DECOAfAllSSIONING PLAN Airborne activity would be generated during the various activities of primary shield wall dismantlement. The airborne release potential for these activities is discussed below.

Long, deep holes would be drilled for the insertion of explosives. They would be drilled frc,m the top of the primary shield wall, down, parallel to the inside of the concrete surface.

The primary shield wall contains an estimated 351 Ci within the 885,000 kg (1,950,000 lb) of concrete, resulting in an average activity of 0.397 pCi/g. If the airborne concentration of dry material is assumed to be 10 mg/m 2, and it is conservatively assumed this can be dispersed throughout the 5.66 x 10' m' containment volume, then a maximum of 566 g of dust could be suspended. The maximum activity that could be airborne would be 225 Ci. This maximum suspension of activity is well below the 2.07 Ci limit discussed in section 3.4.3.1 for maintaining Exclusion Area Boundary exposures below 0.5 rem.

The above calculations are conservative since they do not account for local contamination control methods. These methods could include providing local tenting and HEPA filtration of l

the concrete drilling area, as well as dust minimization by spraying water on the area. Based on the above, it is not credible that drilling of concrete could result in a release which would i

exceed the applicable limits.

Q For the blasting of the primary shield wall, a postulated aerosol concentration of 100 mg/m2 is (V used. This concentration is ten times that assumed in the discussion of drilling above. The entire volume of the Containment Building (5.66 x 104m2 )is considered available for suspension of material. The concrete is assumed to have 0.397 pCi'g of activity. Based on the above assumptions the total Containment Building atmosphere could contain 0.0023 Ci. This activity is significantly less than 2.07 Ci limit discussed in section 3.4.3.1 for maintaining Exclusion Area Boundary exposures below 0.5 rem.

3.4.4.3 Material Handling Events Material handling events involve activities associated with lifting and transporting parts of systems, structures and components once removed from the facility. Material handling events encompass those events that could potentially occur during movement of radioactive materials from their installed location to a location outside of the structure. The events could involve such items as:

1. Dropping of con'.aminated components;

2. Dropping of concrete rubble; and

3. Dropping of filters or packages of particulate material.

(3 U 3-65 Revision 4 1

U

TROJAN DECOAfAllSSIONING PL4N g

) 3.4.4.3.1 Dropping of Contaminated Components fable 3.1-6 contains a listing of contaminated systems and provides activity levels of con /.4mination. Radioactive material contained as activation products within the metal lattice tatrix of plant components or structures is not considered to be releasable as airbome particles via a materials handling event. Therefore, no further consideration of materials handling events involving activated metal components is necessary. The highest contamination activity within a system (not including activation)is contained in the RCS piping, approximately 221 Ci.

The assumptions contained in Section 3.4.3.2 conservatively assume that 10% of the contained activity of a contaminated component becomes dislodged, of which one percent becomes airbome. To provide an extremely conservative analysis, it is assumed that the entire activity of tla RCS piping is contained within a single component being handled.

Using this 221 Ci as the bounding activity,0.221 Ci could become airborne. This amount of airborne activity is below the limit of 2.07 Ci which was determined to result in offsite radiation exposures of 0.5 rem for a release location outside a building. Therefore, dropping a contaminated component can not result in Exclusion Area Boundary doses exceeding applicable limits.

3.4.4.3.2 Dropping of Concrete Rubble p)

The primary shield wall contains an estimated 351 Ci of activity. Due to physical and process constraints the primary shield wall will be dismantled and packaged in portions, it is not credible that the entire volume of concrete rubble associated with the primary shield wall would be involved in a drop event, however, it does provide a bounding case for the analysis. The assumptions contained in Section 3.4.3.2 conservatively assume that 10% of the contamination of a contaminated component becomes dislodged, of which one percent becomes airbome. The majority of activity in the primary shield wall consists of activation products which are not easily dislodged or as likely to become airbome as contamination on a surface. Therefore, it is conservative to assume 10% of the contained activity of the primary shield wall becomes dislodged, of which 1% becomes airborne. These ascumptions would result in 0.351 Ci becoming airbome. This amount of airbome activity is below the limit of 2.07 Ci which was determined to result in Exclusion Area Boundary radiation exposures of 0.5 rem for a release location outside a building. Therefore, dropping of concrete rubble cannot result in Exclusion Area Boundary doses exceeding applicable limits.

) 3-66 Revision 4

TROJANDECOMMISSIONING Pl.AN

- 3.4.4.3.3 Dropping of Filters or Packages of Particulate Material .

Dropping of filters or packages containing particulate material has the potential to create an

airborne release by dislodging material. To minimize this potential the Radiation Protection

Program provides administrative controls on the packaging and movement of radioactive -

material which include:

1. Radioactive materials removed from contaminated areas will be contained, surveyed -

and labeled to allow appropriate control of the material;

2. Radioactive liquid samples or sources will be properly contained and should be --

transported by, or under the cognizance of, radiation protection personnel; and 3; Movement of adioactive mrterial should be made by the most practical direct route.

l The worst case scenario would involve dropping material in a radioactive waste storage area.

This location is considered bounding since the petmtial release could include additional -

containers affected by the impact of the drop.

The consequences associated with the dropping of filters or packages of particulate material is considered bounded by the worst case fire scenario discussed in Section 3.4.4.5 for the following reasons:

.1. The fire scenario is assumed to affect all containers within the storage area, whereas a .

drop event would be limited to the impact area; and

2. 'A fire results in greater release of airbome activity than drop events.

- Therefore, dropping of filters or packages of particulate material will not result in a release of radioactive material that would exceed 0.5 rem limit as discussed in Section 3.4.3.1.

3.4.4.4 - Loss of Suonort System Events -

Electric power, cooling water, and compressed air systems provide support to decommissioning activities. Loss of these systems could potentially affect many other systems and plant areas simultaneously. Each of thrse events is evaluated below.

3-67 Revision 4

TROJAN DECOMMISSIONINGl'lA 4 3.4.4.4.1 Loss of Offsite Power Offsite power is used to energize tools, cranes, lighting, and air filtering equipment used during decommissioning activities. The following events can result from a loss of offsite power:

1. Tools, lighting, and air filtering equipment are de-energized; and

2. Cranes are de-energized.

A loss of power to tools and lighting being used for decommissioning will result in the interruption of work activities, but does not result in the release of radioactivity.

l A loss of power to plant ventilation and filtering systems could result in the disruption of air flow paths and effective utilization of HEPA filters.- In the event ofloss of offsite power, work activities with the potential for airborne contamination will be suspended.

A loss of offsite power could result in loss of power to material handling equipment.

Occupational Safety and Health Administration (OSHA) regulations require that crane hoisting units be equipped with a holding brake. A holding brake is a brake that automatically prevents motion when power is off. The Containment Building Po'ar Crane, Fuel Building Overhead Crane, Auxiliary Building Electric Holst, Spent Fuel Pool Bridge Crane, and the Condcasate Demineralizer Building Bridge Crane are equipped with h61 ding brakes. Although loss of power is not expected to result in crane or hoists failure, this event would be bounded by the material handling events analyses provided in Section 3.4.4.3.

3.4.4.4.2 Loss of Cooling Water Cooling water may be supplied to air compressors and the decommissioning cutting equipment and tools. The following events result from a loss of cooling water:

1. Compressed air is lost if an alternate cooling water supply is not established to the station air compressors within a short time period. The consequences of a loss of compressed air are presented in Section 3.4.4.4.3: and

2. Cutting operations that use cooling water will stop. This does not adversely affect contamination control.

A loss of cooling water does not result in events leading to a significant release of radioactive material to the environment during decommissioning activities. Therefore, public health and safety are not adversely affected by a loss of cooling water event.

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TROJ.4N DECOAfAflSSIONING PL4N

. (.

U 3.4.4.4.3 Loss of Compressed Air Compressed air is supplied by the station air compressors to operate pneumatic valves and dampers and to power pneumatic tools. The following events occur upon a loss of compressed air:

1. The liquid discharge control valve for plant effluents fails in a closed position terminating liquid releases;

2. Decommissioning pneumatic tools shutdown. This terminates potential releases from activities using these tools; and

3. Pneumatic ventilation exhaust fan dampers fail in a closed position terminating airborne and gaseous release via those paths. Since this event is not postulated to occer coincident with an event involving abnormal releases of radioactive material, there would be no significant impact on offsite releases.

A loss of compressed air does not result in events leading to significant releases of radioactive material to the environment during decommissioning activities. Therefore, public health and safety are not adversely affected by a loss of compressed air event.

, g 1

() 3.4.4.5 Fire Events I

A fire event could affect several plant systems, structures, and components simultaneously.

Combustible materials can be ignited by either external ignition sources (e.g., oxyacetylene torches) or internal ignition sources (e.g., spontaneous combustion). Adequate fire protection features will be maintained through implementation of the fire protection program discussed in Section 9, thereby minimizing the potential of occurrence of a fire. These features include:

1. Fire detection and suppression systems and equipment;

2. Fire barrier maintenance and control;

3. Personnel training and qualification;

4. Fire Protection Program procedures; and

5. Control of transient combustible materials and ignition sources.

Following permanent shutdown of TNP, a deactivation program was undertaken which included, in part, the removal of carbon from ventilation filters, the removal of oil from non-()

)

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1ROJAN DECOAIAtlSSIONING PLAN V' essential pumps and motors, and the deenergization of electrical power to non essential equipment. This has reduced the amount of combustibles in the facility and the potential for fires.

A calculation was performed to determine the maximum expected release that would occur in the event of a fire. The worst case fire is assumed to occur in the Condensate Demineralizer Building, which stores low level radioactive waste. A source term based on historical waste generation was assumed. This was considered conservative as discussed below.

Waste generated during decommissioning includes contaminated clothing, decontamination materials resulting from physical or chemical removal of radioactive contamination, work containment materials (plastic tents, etc.), contaminated residue from cutting and grinding, and contaminated components. These activities are basically similar to the types of work and materials generated during refueling and maintenance outages. The size of the work force that will be generating radioactive waste during decommissioning activities is expected to be less than the number typically employed during a refueling and maintenance outage. It is unlikely that a greater volume of combustible radioactive material will be generated during decommissioning than was present during plant operations. Wastes of this type will be produced at a slower rate than during normal outages and will be shipped offsite on a regular basis rather than accumulated over long periods of time, furher reducing the probability of a n large source term being available for conflagration.

V The results of this analysis determined that the offsite dose from a worst case fire involving radioactive material was approximately 5 mrem whole body. This is well below the 0.5 rem limit as discussed in Section 3.4.3.1.

3.4.4.6 Explosion Events During decontamination and dismantlement activities portable gas bottles may be used in support of welding or cutting activities or as fuel for vehicles such as fork lifts. The use, movement, and storage of portable bottles of combustible gases (e. g., acetylene, hydrogen, and propane) in the plant will be controlled by administrative procedures. These procedures will establish requirements ensuring that the planned use, movement, or storage meets appropriate fire protection codes and safety standards. In addition, administrative procedures will control the use ofignition sources within the plant. These measures will serve to limit both the residence time of portable bottles within the plant and the potential for ignition sources in close proximity to the bottles.

NUREG/CR-5759, " Risk Analysis of Highly Combustible Gas Storage, Supply, and Distribution in PWR Plants" provides a discussion of the risks associated with the use of bottled gases in plants. The report states that a review of historical information for safety-A V 3-70 Revision 4

TROJAN DECOMMISSIONING PL4N related plant areas did not identify any incidents of explosions of bottles. The report further explains that based on discussions with explosion experts, the explosion of an individual bottle of hydrogen containing 200-250 standard cubic feet of gas in a confined space, could result in the breach of fire doors and concrete block walls.- However, this would serve to dissipate the energy and no widespread damage would result.- 1JUREG/CR-5759 concludes the risk to plant safety from the explosion of portable gas bottles is not significant.

The aforementioned administrative controls, in conjunction with the information provided in NUREG/CR-5759, provide the basis for a determination that appropriate measures are/will be provided to minimize the potential for explosion events. Given tl'e limited potential for widespread damage as described in NUREG/CR 5759, it can be concluded that the consequences of the explosion event are bounded by the postulated fire event discussed in Section 3.4.4.5. ~

l3.4.4.7 External Events A review of extemal events was done to evaluate the effects of natural and manmade events on the radiological consequences of decommissioning activities. The hazards associated with these events are assumed to be consistent with those that could have occurred during TNP operation. Several external events were identified as having potential applicability to TNP decommissioning:

l. Earthquake;

2. External flooding;

3. Tornadoes and extreme winds;

4. - Volcanic activity;  ;

5. Lightning; and

- 6. Toxic chemical event.

Such events are of extremely low probability. A discussion for each of the above listed events follows.

3.4.4.7.1 Earthquake A seismic event during decommissioning could initiate a materials handling event similar to

_ those described in Section 3.4.4.3; The analysis in Section 3.4.4.3 concludes that the bounding 3-71 Revision 4

TROJAN DECOAfAllSSIONING PLAN O

b material handling event results in an Exclusion Area Boundary dose that is significantly less than the 0.5 rem limit.

Structures whose failure during a seismic event could significantly affect the spent fuel pool or spent fuel integrity are seismically qualified. Decommissioning activities which could impact the seismic qualification of these structures / components will be evaluated. One of the purposes of these evaluations is to ensure that the dismantling activities do not result in a configuration that could fait during a seismic event collapsing onto or into the spent fuel pool or affect the

- seismic qualification of the spent fuel pool or spent fuel integrity. The consequences of a seismic event on the safe storage of spent fuel has already been analyzed with the results provided in Section 6.3 of the DSAR. This analysis concluded that a seismic event would not result in Exclusion Area Boundary doses which would exceed the applicable limits.

Following transfer of the spent nuclear fuel and high level radioactive waste to the ISFSI, seismic qualification of the spent fuel pool will no longer be required.

During the Large Component Removal Project, the Containment Building was detensioned and i the opening in the south face was covered by a roll up door, An evaluation was performed on the Containment Building structural integrity in this configuration. It was concluded that under bounding environmental loading conditions, the Containment Building will remain stable and

( n the reinforced concrete and liner plate integrity will be maintained.

IU 3.4.4.7.2 Flooding As discussed in the DS AR, the water surface level of the maximum flood level is calculated to be 41 fl mean sea level (MSL). The TNP yard elevation is 45 ft MSL. This level is sufficient to be considered safe from projected floods. Access to the Auxiliary Building, Fuel Building, Containment Building, and Condensate Demineralizer Building are above the maximum expected flood level, if storage of radioactive material outside of the structures becomes necessary, it will be limited to areas with an elevation or protection equivalent to an elevation of 45 ft MSL. Alternatively, the dedicated capability to relocate stored radioactive material to a protected elevation within 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> (minimum warning time for flood peak following failure of Grand Coulee dam) will be maintained. In the unlikely event that a maximum flood level was experienced during decommissioning, loss of offsite power is considered to be the only potentially significant resultant decommissioning event. The consequences associated with loss of offsite power was discussed in Section 3.4.4.4.1.

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TROJAN DECOAfAflSSIONING PLAN b 3.4.4.7.3 Tomadoes and Extreme Winds Dismantling activities take place within structures that were designed to withstand credible meteorological conditions for the area. The Contaimaent Building was modified by the Large l Component Removal Project as described below.

During the Large Component Removal Project, the Containment Building was detensioned and the opening in the south face was covered by a roll up door. An evaluation was performed on l the Containment Building structural integrity in this con 0guration. It was concluded that this con 6guration does not adversely affect the ability of the structure to withstand tomado forces.

Postulated radiological releases as a result of breaching the opening door during a tomado or extremely high winds are bounded by other analyzed events.

Storage of radioactive material is normally limited to locations within the Fuel Building, Radwaste Annex, and Condensate Demineralizer Building. The storage of radioactive material will be administratively controlled to ensure adequate protectior. to prevent airbome releases.

These administrative controls are contained in the Radiation Protection Program.

Based on the administrative controls, tomadoes and extreme winds are not expected to initiate conditions that would result in releases that would exceed 0.5 rem TEDE limit.

O

() Tomadoes or extreme winds could initiate a loss of offsite power event. The analysis in Section 3.4.4.4.1 concludes that the 0.5 rem TEDE limit is not exceeded in the event of a loss of offsite power.

3.4.4.7.4 Volcanic Activity Section 2.5.6 of the DSAR provides a discussion of the probability and credible effects of volcanic activity. The credible effects associated with volcanic activity are identified as:

1. Ash fall:

2. Air blast, debris avalanche and pyroclastic flows; and

3. Mud flow - flooding.

An impending ash fall at TNP (irrespective of volume) would activate preparations for cleanup and maintenance of necessary systems. If necessary, dismantling activities could be suspended to minimize activities that could result in airbome contamination. Ash fall is not considered to be an initiating event for any event resulting in offsite radiological release.

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TROJAN DECOMMISSIONING PLAN

'N The distance of TNP from areas of known volcanic activity makes damage from air blast, debris avalanche, and pyroclastic flows remote and is not considered a credible initiating event for decommissioning events, i Mud flow and flooding resulting from volcanic activity have been analyzed and peak flood l levels in the TNP area are not expected to exceed 30 ft MSL. This level is less than that i expected for the Probable Maximum Flood level discussed in Section 3.4.4.7.2.

l 3.4.4.7.5 Lightning A lightning strike could result in the loss of offsite power or fire event. The loss of offsite power event is discussed in Section 3.4.4.4.1 and the effect of an onsite fire is provided in section 3.4.4.5. -

3.4.4.7.6 Toxic Chemical Event Section 2.2.3.2 of the DSAR provides a description of toxic chemical hazards. Toxic chemicals are a personnel safety concern. In the event of a toxic chemical event affecting plant personnel, decommissioning activities would be suspended and personnel evacuated as necessary. A toxic chemical event has the potential to initiate a radiological event. The most severe radiological

^N event that could be initiated would be if a personnel injury resulted in an event involving a (V loaded crane or hoist. A toxic chemical event is therefore considered an initiating event for a material handling event which is discussed in St.ction 3.4.4.3.

3.4.5 RADIOLOGICAL OCCUPATIONAL SAFETY Radiological events could occur which result in increased exposure of decommissioning workers to radiation. However, the occurrences of these events are minimized or the consequences are mitigated through the implementation of the Radiation Protection Program and the Permanently Defueled Emergency Plan.

The Radiation Protection Program is applied to activities performed onsite involving radioactive materials. A primary objective of the Radiation Protection Program is to protect workers and visitors to the site from radiological hazards during decommissioning. The program requires PGE and its contractors to provide sufficient qualified staff, facilities, and equipment to perform decommissioning activities in a radiologically safe manner.

Activities conducted during decommissioning that have the potential for exposure of personnel to either radiation or radioactive materials will be managed by qualified individuals who will implement program requirements in accordance with established procedures. Radiological hazards will be monitored and evaluated to maintain radiation exposures ALARA.

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TROJAN DECOAIAllSSIONING PUN h

(d The Radiation Protection Program at TNP implements administrative dose guidelines for TEDE to ensure personnel do not exceed federal 10 CFR 20 dose limits for occupational exposure to ionizing radiation. Radiation protection training will be provided to occupationally exposed individuals to ensure that they understand their responsibility to follow procedures and to maintain their individual radiation dose ALARA.

TNP work control procedures will ensure that work specifications, designs, work packages, and radiation work permits involving potential radiation exposurc or handling of radioactive materials incorperate effective radiological contrc!s.

The Permanently Defueled Emergency Plan retains onsite emergency reconse capability. This capability includes relocation of personnel from a radiologically affected t ca, if necessary.

Implementation of the Radiation Protection Program and the Permanently Defueled Emergency Plan ensures that potential radiological events affecting occupational health and safety will be sufficiently minimized and mitigated.

3.4.6 OFFSITE RADIOLOGICAL EVENTS Offsite radiological events related to decommissioning activities are limited to those associated l with the shipment of radioactive materials. Radioactive shipments will be made in accordance n)

( with applicable regulatory requirements. The radioactive waste management program and the Nuclear Quality Assurance Program assure compliance with these requirements. Compliance with these requirements ensures that both the probability of occurrence ar d the consequences of an offsite event do not significantly affect the public laalth and safety.

3.4.7 NONRADIOLOGICAL EVENTS Decommissioning TNP may require different work activities than were typically conducted during normal plant operations. However, effective application of the TNP safety program to decommissioning activities will ensure worker safety. No decommissioning events were identified that would be initiated from nonradiological sources that could significantly impact public health and safety.

Hazardous materials handling will be controlled by the Hazardous Material Control Program using approved plant procedures. There are no chemicals stored onsite in quantities which, if released, could significantly threaten public health and safety.

Flammable gases stor:d onsite include combustible gases used for cutting and welding. Safe storage and use of th::se gases and other flammable materials is controlled through the Fire Protection Program and plant safety procedures.

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TROJAN DECOMMISSIONING PLAN

/~N Q Plant safety procedures and off-normal instructions have been established which would be implemented if a nonradiological event occurred at TNP. Implementation of these programs and procedures ensures that the probability of occurrence and consequence c,f onsite nonradiological events do not significantly affect occupational or public health and safety.

Plant safety procedures provide personnel safety rules and responsibilities and control both chemical and hazardous waste identification, inventory, handling, storage, use, and disposal.

3.5_ OCCUPATIONAL SAFETY TNP has, and will continue to maintain, an industrial occupational safety program. The intent of the safety program is to ensure a safe and healthy working environment, maintain employee safety awareness, promote safety as an integral part of facility activities, and comply with occupational safety and health regulatory standards. Safety at TNP is viewed as part of the overall plant performance and is an integral part of the activities at the plant. Individuals requiring unescorted access to the Industrial Area receive training in plant safety and chemical safety as part of General Employee Training.

' Die TNP safety program is addressed in plant safety procedures, which reflect local, state, and federal safety codes and standards. The safety procedures outline acceptable safe work and housekeeping practices, ensure that employees are adequately trained and prepared to perform h)

V theirjobs safely, and ensure that employees are adequately informed of chemical hazards they may encounter on the job. Personnel are respons" ale fer maintaining a safe work environment.

Individuals assigned the responsibility for the industrial safety programs at the plant will meet with the plant staff on a periodic basis to discuss safety program activities, procedure changes, accident causes, safe procedures, and other subjects pertinent to promoting safety.

3.6 NONRADIOACTIVE WASTE MANAGEMENT Nonradioactive regulated waste materials expected to be handled during decommissioning include asbestos, polychlorinated biphenyls (PCBs), mercury, and lead. Other nonradioactive waste materials include steel components (e.g., piping, valves), electrical components (e.g.,

wiring, motors), and structural materials (e.g., concrete, beams). Handling and disposal of nonradioactive regulated waste materials will be controlled by the TNP chemical safety program. This program provides for evaluation of regulated substances and approval of methods for their handling and disposal. Work vcill be done in accordance with the TNP work control process. This process ensures that decommissioning activities receive appropriate safety and technical reviews. The proposed work will be reviewed for ALARA concerns if the systems, sNetures, or components are in the RCA, and for coordination with other projects.

3-76 Revision 4 l

TrOJANDECOAfAllSS10NING PLAN Nonradioactive paste materials will be transported by approved or licensed transporters as required, and shipped to permitted solid waste landfills or licensed hazardous waste facilities.

3.6.1 ASBESTOS Ast stos containing materials include Marinite board, used in the plant as a fire barrier; electrical cable with a wrap containing asbestos; piping systems with a wrap containing asbestos; the cooling tower mist climinators and distribution piping fabricated from an asbestos cement material; and roof flashing sea' ant containing asbestos fibers. Other materials that are suspected of containing asbestos will be sampled and analyzed before work is done on the inaterial.

Asbestos material will be removed and disposed ofin accordance with plant safety procedures, federal and state OSilA regulations, and federal and state hazardous air pollutant and solid waste regulations.

3.6.2 POLYCHLORINATED BlPHENYLS (PCB)

The rod control cabinet capacitors in the Control. Building may contain a small amount (approximately two liters) of PCBs.

OV PCBs and PCB items will be handled and disposed ofin accordance with federal and state PCB regulations.

3.6.3 MERCURY Mercury is contained in some plant components. Vendor technical manuals and plant walkdowns will be used to identify components that contain mercury metal, Mercury metal will be collected and sent for recycling as plant equipment is removed. If not sent for recycling or reclamation as scrap metal, the mercury will be disposed of as a hazardous waste in accordance with federal and state hazardous waste regulations.

3.6.4 LEAD Lead is contained in lead based paint:, which may have been used as a primer for some steel surfaces at TNP, and lead sheets used as a radiological shield material.

Lead containing materials will be removed and disposed ofin accordance with plant safety procedures, OSHA regt.lations, and federal and state hazardous waste regulations.

( 3-77 Revision 4

TROJAN DECOAfAflSSIONING PL4_N Nonradioactive metals with lead based paints or coatings will be sent as scrap metal to a dealer who will accept lead painted metal; otherwise they will be disposed of as hazardous waste, Sandblast materials used to remove lead based paints will be handled and disposed of as hazardous waste unless they pass a toxic characteristic leaching procedure test.

3.6,5 OTIIER PLANT WASTE MATERIALS Other plant waste materials, including batteries (e.g., lead-acid, nicad) and refrigerants from chillers and air conditioners, will be sent to a recycling facility, or disposed ofin accordance with normal waste d:sposal practices for nonradioactive nonregulated solid waste, Some industrial solid wastes (e.g., treated wood poles) may need special permits before disposal in .

solid waste landfills, t

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TROJAN DECOAfAflSSIONING PLAN

( Table 3.16 Page1of3 System Hurial Volume and Surface Activity Projections System Volume (ft') Activity'(Cl)

Reactor coolant piping 5,894 221 Pressurizer relicf tank 625 <1 Reactor coolant pumps and motors 3,044 134 Control rod drive mechanisms /incore 1,726 83 instrumentation /sem{ce structure Reactor vessel 6,812' i12 i Reactor vessel internals 4,960' 245,9 l Spent fuel pool and racks 17,305 .30+

120-V ac preferred instrument ac 1,400 <1 125.V de power 175 <1 4.16 kV ac power 726 <1 4FO V ac auxiliary load center 5,080 <l 480 V ac motor control center 8,426 <1 Chemical and volume control 10,968 25 Clean radwaste 5,423 14 Containment building penetrations 188 <1 Control rod drive 85 <1 Dirty radwaste 1,613 <1 Electric heat tracing 164 <1 Electrical (Cable / Tray /Condult) 60,139 <1 6

Fuel handling system 339 Fuel pool cooling and demineralizer 4,632 S.6 Fuel and auxiliary building heating, ventilation, and air 3,661 (1 conditioning (llVAC)

Gaseous radwaste 2,529 <1 O j -

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, 1MOMN DECOMMISSIONING PL4N -

Table 3,16 Page 2 of 3 -

System Burial Volume and Surface Activity Projections i

System , Volume (ft') Activity'(Cl)  ;

liVAC 6.635 <1 Hydrogen recombiners 576 <1 Integrated leak rate test instrument line 106 <1 Instrument and service air 1,327 <1 1,lghting panel supply 997 <1

, Mis cIlaneous components 1,936 <l Miscellaneous reactor coolant 3,418 <1 Nuclear instrumentation 193 <1 Oily waste and storm drains 1,882 <1 Containment liVAC 18,869 <1 Primary makeup water 3,615- <l -

Process sampling .I14 4 Radiation monitoring 134 <1 Reactor nonnuclear instrumee.s .245 <1 Reactor vessel system- 116 Residual heat removal 7,649 _36 Safety injection system 7,149 7 Solid radwaste - 370 -<1 Steam generator system 3,562 <1 Turbine building sump pumps and miscellaneous 639- <l Component cooling waterd 6,115 <l Condensate demineralizersd 2,262 <1 Discharge and dilution d 3,834 <1 O

Revision 4

h 1

n0MN DECOMMI55l0NING PUN Table 3.14 P se 3 of 3 -

l l

System Burial Volume and Surface Activity Projections System Volume (fF) Activity * (Cl) -

Containment sprayd 1,563 <1 Total 219,220 1070.5

• Does not include activation, b

To be determined.

Activity included with reactor coolant piping.

d Site characterization surrey results identified these systems as contaminated.

e These waste projections assume the reacter vessel intemals are removed separate from the reactor vessel and segmented. The burial volume projection for the intact removal of the reactor vessel and the intemals together is approximately 8341 cubic feet of Class C waste.

O O

Revision 4 -

l TRonNt>rcoMMissioNINGl'uN

5. DECOMMISSIONING COST ESTIMATE AND FUNDING PLAN in accordance with 10 CFR 50.82, this section provides an updated cost estimate for the decommissioning alternative selected by the TNP co owners, a comparison of the estimate with prer,ent funds set aside for decommissioning, and a plan for assuring the availability of adeounte funds for completion of decommissioning.

l 5.1 DECOMMISSIONING COST ESTIMNIE This section provides the results of and basis for a cost estimate prepared by POE with I assistance from TLO bervices, Inc. (TLO) for the decommissioning of TNP. Incorporated into I this cost estimate are costs of activities involved in radiological decommissioning necessary for termination of TNP's Part 50 license, as well as expenditures necessary to complete nonradiological site restoration activities. The costs of removal and disposal of nonradioactive structures and materials beyond that necessary for license termination have been identified separately from radiological decommissioning costs.

Also separately identified are cost projections and funding requirements for the onsite management ofirradiated fuel until possession and title of the irradiated fuel is transferred to DOE for ultimate disposal. The description of the spent fuel management costs and associated funding plan provided in this section, together with thc description of the spent fuel mana ement program in Section 3.3.1, fulfil the requirements of 10 CFR 50.54(bb), which stipulate that " nuclear power reactors licensed by the NRC ... shall, within 2 years following permanent cessation of operation of the reactor .... submit written rotification to the Commission ... of the program by which the licensee intends to manage and provide funding for the management of all irradiated fuel at the reactor until title to the irradiated fuel and possession of the fuel is transferred to the Secretary of Energy for its ultimate disposal in a repository."

5.1.1 COST ESTIMATE RESULTS As indicated in Table 5.1 1, the costs (in 1997 dollars) for the selected decommissioning I attemative are estimated to be approximately $236,057,000 for radiological decommissioning I activities, approximately $53,737,000 for nonradiological decommissioning activities (site I restoration), and approximately $157,395,000 for spent fuel management, Costs associated I with securing and maintaining decommissioning financial assurance and bridging funds are projected to total approximately $11,665,000. A detailed schedule of TNP's decommissioning i and spent fuel management costs, totaling approximately $458,854,000 is provided in I Table 5.12 and described in Section 5.1.2.

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. . - _ . _ _ _ _ _ _. . --_ -_-.-- _ - _____. . . . - - - - ._ = _ _ -

TROJAN NTOMMISSIONING PL4N 5.1.2 COST ESTIM ATE DESCRIPTION The initial Decommissioning Plan decommissioning cost estimate was based largely on the l TNP specific cost estimate performed for PGE by TLG Services,Inc. in May 1994. The methodology used to develop the cost estimate followed the approach presented in AIF/NESP- 1 036, " Guidelines to Producing Decommissioning Cost Estimates," and the DOE

" Decommissioning llandbook." These guidance documents utilize a unit cost factor method for estimating decommissioning activity costs. Unit cost factors incorporate site specific considerations whenever practicable. Using plant drawings and inventory documents, i quantities and volumes of the equipment and material to be removed during decommissioning were estimated. Unit cost factors were applied to the volumes and quantities to estimate the i

" activity dependent" costs. " Period dependent" costs were determined from a critical path schedule based on the removal activity duration.

I During 1996, the decommissioning cost estimate was updated to reDect the way I decommissioning is actually being perfonned in addition, the ecst estimate has been updated I to account for work performed through 1996, as well as adjusted projected costs, and is now I stated in 1997 dollars. l The results of PGE's decommissioning cost estimate have been incorporated into Table 5.1 2, which provides a comprehensive expenditure schedule for the decommissioning of TNP. This

(]

V table incorporates an annual breakdown of projected costs associated with radiological and nonradiological decommissioning, spent fuel management, and decommissioning expenditure financing activities. The decommissioning cost estimate expenditure schedule contained in Table 5.12 is described in the remainder of this section.

5.1.2.1 Nite (Radiologican necommissioning costs The cost schedule for NRC decommissioning activities is incorporated into Table 5.1 2, which renects the results of the decommissioning cost estimate for TNP. Consistent with current NRC policy, the TNP decommissioning cost estimate considers NRC decommissioning costs to be only those costs associated with normal decommissioning activities necessary for termination of the Part 50 license and release of the site for unrestricted use. The decommissioning cost estimate does not include in NRC decommissioning costs those costs associated with spent fuel management or the disposal of nonradioactive structures and materials beyond that necessary to terminate TNP's Part 50 license.

52 Revision 4

TROMN DECOMAflSS10NING PIAN NRC decommissioning activity costs are separately identified in Table 5.12 as large component removal activities and other radiological decommissioning costs, the latter of which are incorporated into the column entitled *DECON Planning /DECON/ License Termination."

Hurial costs were derived from POE modelinD and analysis oflow level radioactive waste I disposal costs in July 1994, which more conservatively reflect projected burial rates. POE also used r!te specific data to independently analyre and project costs associated with the separate I removal and segmentation of the reactor vessel internals, the intact removal of the reactor I vessel and internals, and development of the decommissioning plan. The decommissioning I costs incorporated into Tables 5.1 1 and 5.12 reflect the estimated costs associated with the i option of removal of the reactor vessel with internals intact. Controls exist to ensure funds are I expended consistent with t3 k visions of 10 CFR 50.82(a)(8).

The rule will be met by execWnd .<4ndard ongoing financial controls. Throughout the budgetary process and budget year, costs associated with new projects or activities are evaluated to detennine their correct cost classification, ie, fuel management, radiological, nonradiological decommissioning, capital, etc. As a result, only costs which meet the intent of the Decommissioning Plan are submitted for reimbursement from the decommissioning trust.

The activities described in the Decommissioning Plan satisfy the definitions of" decommission" and " major decommissioning actlvity" from 10 CFR 50.2.

Periodic reports are also prepared and submitted to the ODOE that compares costs by major J classification to the decommissioning cost estimate. Plant personnel review variances and impacts,if any, are examined.

Corporate finance personnel review PGE's trust fund activity and balance periodically and for Trojan Co-owners. Any significant activity which is inconsistent with the Decommissioning Plan would be brought to the attention of Trojan management.

During 1996, the decommissioning cost estimate was recast in such a manner that actual costs can be related more casily to the cost estimate. Information will be available to access the validity of the cost estimate and/or Trojan's ability to complete decommissioning tasks within i cost estimate totals, Periodically, variances between the estimate and actual costs will be reviewed as they relate to the total cost estimate to provide assurance that the cost estimate continues to be reasonable. This complies with 10 CFR 50.82(a)(8)(i)(A).

The decommissioning cost estimate has been updated to reflect costs in 1997 dollars and I adjustments made during the updating process. The cost estimate has been updated to account I for work performed through 1996 where TNP expended funds for decommissioning planning, I the Large Component Removal Project, and other decommissioning activities. The estimate I also reflects the one-piece reactor vessel (with intern ts intact) removal concept, final radiation 1 53 Revision 4 Q w .

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1 TROJAN DECOhihilSSIONING PLAN  ;

survey funding adjustment, revised estimate for removal of cooling tower asbestos and I structure, and delayed ISFSI fuel loading. I 1

Costs required to maintain spent fuel in a safe storage condition are not funded by the trust fund while the spent fuel remains in wet storage. Once the spent fuel is transferred to dry storage, there are suflicient trust fund annual contributions to more than cover annual costs. This is described in the Decommissioning Plan, Sections 5.2 and 5.3.2 and Table 5.12. This complies with 10 CFR 50.82(a)(8)(i)(II).

Finally, PGE calculates the finarmlal assurance amount required to complete radiological decommissioning. The trust fund balance, and a letter of credit, must equal the amount needed to complete radiological decommissioning. This calculation must be completed periodically and the letter of credit adjusted. The " bridge" funds are described in the Decommissioning Plan, Section 5.3. This complies with 10 CFR 50.82(a)(8)(i)(C).

5.1.2.2 Nonradiological Demmmissioning Costs Although not required by NRC regulations, the decommissioning cost estimate for TNP incorporates nonradiological decommissioning costs, as indicated in Table 5.12. The TNP decommissioning cost estimate considers nonradiological decommissioning costs to be those costs associated with site remediation and demolition and removal of uncontaminated structures. The decomrnissioning cost estimate does not include in nonradiological L decommissioning costs those costs associated with spent fuel management or NRC decommissioning activities.

Nonradiological site remediation costs were identified and incorporated into the cost estimate I based on a study conducted for PGE in April 1994 by Cil2M 11i11, an engineering firm specializing in environmental remediation and water treatment. The methodology that Cil2M liill used to estimate the nonradiological site remediation costs was consistent with the methods used by EPA and State of Oregon under their site clean up programs.

The relatively larger projected expenditures in 2018 and 2019 for nonradiological decommissioning activities (Table 5.1-2) reflect the intent to perform the majority of the site restoration and uncontaminated building demolition activities after the spent fuel and other high level radioactive waste have been transferred to an offsite repository in 2018. Significant activities planned prior to this time include removal and disposal of asbestos contained in the cooling tower, as well as annual activities related to nonradiological site remediation during the i Transition and Decontamination and Dismantlement Periods. I

( 5-4 Revision 4

TRO.lAN DECOMMISSIONING PL4N 5.1.2.1 Soent Fuel Management Costs implementation costs associated with the spent fuel management plan described in Section 3.3.1 are reflected in the projected cost schedule for the onsite management of ,

irradiated fuel detailed in Table 5.1 2. Spent fuel management costs begin with ongoing spent fuel pool operation, surveillance, and maintenance activity costs, and continue through ISFSI l planning, construction, and operation until possession and title of the irradiated fuel is transferred to the DOE for ultimate disposal (assumed in this estimate to be completed in 2018).

As indicated in Table 5.1 2, spent fuel pool operation expenditures are projected to end in 1999 I

- as a result of the transfer of the spent fuel pool contents to the ISFSI. Costs associated with ,

onsite management of the spent fuel will then involve ISFSI operation, maintenance, and I surveillance expenditures. Finally, upon transfer of the ISFSI contents to an offsite repository, spent fuel management costs end in 2018 with final experditures necessary for ISFSI I decommissioning activities.

POE has analyzed spent fuel operations and maintenance costs related to storage in both the spent fuel pool and the ISFSI. The methodology used in this analysis considered plant specific values, as appilcable, for labor, material, and outside professional services requirements as well as for other distributed items such as overheads, property and liability insurance, regulatory fees, fire protection activities, and power usage. The results of this analysis were then incorporated into the decommissioning cost study. I 5.1.2.4 Finanelal Activity Costs Additional costs will be incurred by each TNP co-owner as nccessary during decommissioning to secure and maintain assurance that adequate funds will be available to complete radiological decommissioning of the TNP site, and to secure loans or c,ther " bridging" mechanisms to augment existing funds to cover near term decommissioning costs. The financial assurance costs (e.g., letter of credit and standby trust fees) indicated in Table 5,1 2 are based on the basis points and projected amount of required financia: assurance appropriate for each co-owner as described in Section 5.3, " Decommissioning Funding Plan." The loan costs in Table 5.1-2 are based on the interest rate and loan amount appropriate for each TNP co-owner requiring financial bridging as described in Section 5.3. The method which each co-owner will use to provide the required financial assurance mechanism and bridging iunds is described in detail in Section 5.3.

5.2 SPENT FUEL MANAGEMENT FUNDING PLAN Spent fuel management costs are segregated in Table 5.1 2 into spent fuel pool operation costs and dry storage (ISFSI) costs. Ongoing costs associated with the storage of spent fuel and other high level radioactive waste in the spent fuel pool are currently

( 5-5 Revision 4

TROMN DECOMMISSIONING PLAN i incorporated into the TNP O&M budget, and are expected to continue to be funded in this manner un't the contents of the spent fuel pool are transferred to the ISFSI. Costs associated with dry storage activities, including ISFSI planning, construction, O&M, and decommissioning, as reflected under the column heading " Dry Storage" in Table 5.1 2, will be funded with decommissioning trust funds collected for that purpose. Additional details on the decom.missioning trust fund collections for each TNP co owner are provided in Section 5.3.

5.3 DFCOMMISSIONING FUNDING PLAN

, 5.3.1 CURRENT DECOMMISSIONING FUNDING CAPABILITIES Each of the TNP co-owners separately collect and maintain funds for the decommissioning of TNP. These funds are collected through rates and deposited to external trust funds in accordance with 10 CFR 50.75. Ilowever, the external trust funds were established assuming that the total collected funds at the expected time of decommissioning would be sufficient to pay both radiological and nonradiological decommissioning costs. Because the TNP was shutdown prematurely, the external trust funds established by the TNP co-owners currently contain only a portion of the total amount needed for site radiological decommissioning.

Table 5.3 1 summarizes the status of the TNP co-owners' decommissioning trust funds as of t December 31,1996.

I iO The NRC's general policy requires, prior to the start of the Decontamination and l Dismantlement Period, either funds needed for decommissioning (as the term " decommission" I is defined in 10 CFR 50.2, " Definitions") to be available or an appropriate financial vehicle to be secured and maintained that will assure the availability of adequate funds for completion of NRC (radiological) decommissioning. As indicated above, the tmsts established by the TNP co-owners for decommissioning will not contain the funds necessary for completion of radiological decommissioning prior to the start of the Decontamination and Dismantlement l Period. Thus prior to commencing this period, each TNP co owner is required to secure a i financial assurance mechanism allowed by 10 CFR 50.75. This financial assurance must be maintained until temiination of TNP's Part 50 license. Furthermore, during the I Decontamination and Dismantlement Period each co-owner's decommissioning trust fund I balance is projected to be reduced to a point where it will be necessary in certain instances to borrow or otherwise provide "bndging" funds to complete decontamination activities and allow scheduled collections to restore the decommissioning trust fund balance.

S-6 Revision 4

TROMN DECOMAtISSIONING PLAN l 5.3.2 TNP CO 0WNERS' DECOMhilSS10NING FUNDING PLANS Each of the TNP co-owners has established a program in conjunction with specified goals for the collection of funds for the decommissioning of TNP. Each TNP co-owner maintains a I ,

decommissioning fund collection schedule which ensures that each co-owner's podion of the decommissioning activity expenditures will be fully funded. These trust fund contribution I schedules are based on funding requirements for both radiological and nonradiological decommissioning costs, as well as financing costs and specific spent fuel management costs including planning, design, constmetion, O&M, and decommirsioning of an ISFSI. These I collection schedules do not include funding for spent fuel pool O&M costs since these costs are being paid with O&M budget funds rather than decommissioning trust funds. The decommissioning trust frnd cash flow for each of the TNP co-owners, based on the expenditure schedule in Table 5.1 2 .nd the co owner contribution schedules, is described below. I 5.3.2.1 PGE Funding Table 5.3 2 provides PGE's decommissioning trust fund cash flow in nominal dollars (2.84% escalation) during decommissioning. The trust fund expenditures described in this table I are PGE's share (67.5%) of the expenditures described in Table 5.1 2, with the exception of spent fuel pool O&M costs since these costs are being paid with O&M budget fund; rather than decommissioning trust funds. The trust ftmd contributions listed in Table 5.3 2 are based upon PGE's decommissioning trust fund contribution schedule which ensures that PGE's portion of I the decommissioning activity expenditures will be fully funded.

Projected requirements for bridging funds have been incorporated into PGE's decommissioning trust fund cash flow. As previously discussed, PGEi. external trust fund currently contains only a portion of the total amount needed for PGE's share of site radiological decommissioning costs. Based on the decommissioning trust fund cash flow analysis presented in Table 5.3-2,

bridging funds will be required in the year 2000 to complete decontamination activities and l

allow scheduled collections to restore the decommissioning trust fund balance Projected interest on bridging funds has also been incuporated into PGE's trust fund cash flow as indicated in Table 5.3-2.

In addition, because the trusts established by the TNP co-owners for decommissioning will not contain the funds necessary for completion of radiological decommissioning prior to the start of the D: contamination and Dismantlement Period, each TNP co-owner must secure I a financial assurance mechanism allowed by 10 CFR 50.75, and maintain this assurance until I

termination of TNP's Pr.rt 50 license. PGE's financial assurance mechanism will consist of the decommissioning trust fund balance together with a letter of credit. Because financial assurance will be maintained only for NRC decommissioning activhies, the methodology used to determine the size of the letter of credit ensures that if a given amount of the 5-7 Revision 4 l

TROJAN DECOMMISSIONING l'L4N decommissioning trust fund is used for non NRC activities during a current year, the portion of the Gnancial assurance provided by the letter of credit must be increased by the same amount.

This methodology can be summarized as follows:

L,, = Ti . T3 + T3 where L,, = Letter of Credit Portion of Financial Assurance Needed for Current Year T i = Total costs of remaining NRC activities T 3= Current decommissioning trust fund balance T 3= Portion of trust balance planned for non NRC costs during current year Financial assurance for remaining NRC decommissioning activities will be calculated at the beginning of each year and will be periodically reviewed during each year to ensure that an adequate level of financial assurance is maintained.

5.3.2.2 EWER /BPA Funding IlPA is obligated through Net Billing Agreements to pay costs associated with EWEB's share of TNP, including decommissioning and spent fuel management costs. BPA will fulnll the decommissioning funding obligations of EWEB, including providing Snancial assurance for EWEB's portion of decommissioning costs in a manner stipulated in 10 CFR 50.75(e)(3)(iv) for O' Federal government licensees. Table 5.3 3 provides BPA/EWEB's decommissioning trust fund cash now in nominal dollars (2.84% escalation) during decommissioning. The trust fund I expenditures described in this table are BPA/EWEB's share (30%) of the expenditures described in Table 5.1 2, with the exception of spent fuel pool O&M costs since these costs are being paid with O&M budget funds rather than decommissioning trust funds. The trust fund contributions listed in Table 5.3 3 are based upon BPA/EWEB's decommissioning trust fund I contribution schedule which ensures that BPA/EWEB's portion of the decommissioning activity expenditures will be fully funded.

Projected requirements for bridging funds have been incorporated into BPA/EWEB's decommissioning trust fund cash Dow. As previously discussed, BPA/EWEB's external trust fund currently contains only a portion of the total amount needed for BPA/EWEB's share of site radiological decommissioning costs. Based on the decommissioning trust fund cash Dow analysis presented in Table 5.3-3, bridging funds will be required to complete decontamination activitiet and allow scheduled collections to restore the decommissioning trust fund balance.

These bridging funds are not expected to incur interest costs since BPA, as a government entity, will provide the additional decommissioning funding when necessary according to the schedule listed in Table 5.3 3.

58 Revision 4

TROMN DECOAfAflSSIONING PL tN As allowed by 10 CFR 50.75(e)(3)(iv), BPA, as a Federal government entity fulfilling the decommissioning fundinD obligations of EWED, a licensee, will provide financial assurance in the form of a statement ofintent. The statement ofintent will contain a reference to the TNP decommissioning cost estimate described in Section 5.1, indicating that funds for radiological decommissioning will be obtained when necessary.

5.3.2.3 PP&L Funding Table 5.3-4 provides PP&L's decommissioning trust fund cash flow in nominal dollars (2.84% escalation) during decommissioning. The trust fund expenditure., described in this table I are PP&L's share (2.5%) of the expenditures described in Table 5.1 2, with the exception of spent fuel pool O&M costs since these costs are being paid with O&M budget fundt rather than decommissioning trust funds. The trust fund contributions listed in Table 5.3-4 are based upon PP&L's decommissioning trust fund contribution schedule which ensures that PPAL's portion I of the decommissioning activity expenditures will be fully funded.

Based on the decommissioning trust fund cash flow analysis presented in Table 5.3 4, PP&L's decommissioning trust balance will remain adequately funded during decommissioning such that bridging funds will not be required.110 wever, because the trusts established by the TNP co-owners for decommissioning will not contain the funds necessary for completion of n radiological decommissioning prior to the start of the Decontamination and Dismantlement I fy Period, PP&L must secure a financial assurance mechanism allowed by 10 CFR 50.75, and I maintain this assurance until termination of TNP's Part 50 license. PP&L's financial assurance I mechanism will consist of the decommissioning trust fund balance together with a letter of credit. The methodology for determining the size of the letter of credit is as described in Section 5.3.2.1, "PGE Funding."

5-9 Revision 4

TROJANDECOAfAflSS10NING Pl.4N Table 5.1 1 Total Decommissioning Costs Radiological, Nonradiological (Site Restoration),

Spent Fuel Management, and Financing (1997 dollars) l Radiological (NRC) Decommissioning Costs $

Large Component Removal 17,919,000 l DECON Flanning/DECON/ License Termination 218,138,000 i Total 236,057,000 I Nonradiological Decommissioning Costs $

Site Restoration 53,737,000 , I Total 53,737,000 l Spent Fuel hianagement Costs $

Spent Fuel Fool Operation /hiaintenance (1998 99 non-tmst fund 23,598,000 i expenditures)

ISFSI Construction and Decommissioning 62,925,000 l ISFSI Operation /hiaintenance 70,872,000 l Total 157,395,000 i Financing Costs $

Financial Assurance 496,000 l Decommissioning Loans 11,169,000 1 Total 11,665,000 i Total Decommissioning and Fuel hianagement Costs $458,854,000 I Total Trust Fund Expenditures $435,256,000 i O Redsion 4

l l

l Decommisslorl itemized j 1

Total Expenditures -

Year Total NRC Tcual Total Soent Total Combined Decommtssa ing Nonradologral Fuel Finanung Expendnurci Lipencnures Decommtmoning Management A tsvny Lapendnures Lapendnures Lapendnures 1993 . . . .

1994 8 034 0 8 034 1995 15 971 0 i 1.107 11 028 1996 8 574 404 3.161 12.229 1997 19 482 472 7,479 27.433 1998 34 427 463 29 961 64 til 1999 70 300 0 38 601 72 108 977 20m 3% Ol$ 7 677 $ 103 789 , 48534 2001 27 123 8228 3 739 1 738 40 828 7002 16 119 875 3 736 2 214 22.944 7003 1 062 762 3 729 2 312 7J65_

20ru 137 3 718 I 865 59?O 2n 3 337 3.703 1 172 5 412 xxw 337 3 681 883 4 901 2007 337 3 655 394 4 386

?<v)g 331 36?I ?5 3 983 e.a. m .

A .

i TROJAN DECOhihilSSIONING PLAN l

'ANSTEC i Table 5.12 (APERTUR$ 1 ing Cost Estimate for Trojan Nuclear Plant

)ccommissioning Expenditure Schedule CARD I  :

l l (1997 $ x 1000) Also A M 084 l l l

NRC (Radiologica0 Nonradiological Spent Fuel Management Finatu;ing Activities Decc.mmissioning Occommissioning I

i

' Dry storage $ pent Fuel Poul OAM l

" C cd'A..i ni ^$4Te'.M7at' gn3..fo.'.n b5121 J.PJ,B, Mnf aemo (9l{P4, j n.n ana, 1

1ermmation o

iggi, I

3m Aw l l is m: m u o, I

J ,4so, .m4 su 3 isi l 1

l 19 482 472 7,479 l l um m i. 128 im n y, I to win o 2s 104 i tio 11.19: 72 l l vm 3s ois tus s uo m 3,o I i nm em tm m iun l l i. n0 ns 3m 33 2 m ._ l l im m tm u 3m I

! m 3m im I m rm im I m 3w m l

! m 33s 3u l m 3 e, 3s I Revision 4 Page 1 of 2

~

k

Decommissio Itemized r I Total Expenditures Year Total NRC Total Total Swnt Decommissoning Total Combin Nortadiologkal Fuel Financing Expendicirie Lapenditures Decommissoning Management Activity Espenditures Lapenditure Expenduures 2009 337 3,381 1 1. 01(-

l

?olo 337 3.533 3 87C 2011 337 3.476 3 81 2012 l 337 3 476 3813 l

_2011 337 3 476 3 813 2014 l 137 3 476 3 813 2015 337 3 476 3 8131 2016 337 3 476 3 813' 2017 337 3 476 3.813 2018 10 9R1 10 951 21 936 2019 14.132 14 132 2020 337 13T 2021 813 813 7022 A06 806 2023 3 023 3 025 t

Total 236 057 $ 3 'r37 137 395 I t 661 458 834

\..

s

ANSTEC Tuble 5.12 APERTURE 3 Cost Estimate for Trojan Nuclear Plant CARD I I

tommissioning Expenditure Schedule Also AvaHeble on l (1997 $ x 1000) Apertitre Card l

l NRC (Radiological) Nonradiological Spent Fuel Management Decommissioning Financing Actisities Decommissioning l

Dry $turage $ pent Fuel Pool O&M Comp!knt / . iw es a n Removal I i and Ma nSinYng na misuoning Termmannn g i 33, 3 sei t I m 3m I m 3 c6 I m s o. I m 3 es I m 3 es I m s o. I m 3 es I m 3 c6 I io .is v is3 - -n -

I i4 m I m I m I u I s o:s l l

i, .i . m i3. 33 m n o,s 20 m 23 so. - n i6. I f -

Revision 4 Page 2 of 2

TROJANDECOAIAllSSIONING l'l.AN Table 5.31 Status of Decommissioning Trust Funds as of December 31,1996 l Trojan Co Owner Fund Dalance as of12/31/96 I Portland General Electric (PGE) $78,879,000 1 Eugene Water & Elec7ric (EWEB)/ $15,945,000 i Bonneville Power Administration (DPA)

Pacific Power & Light (PP&L) $4,569,000 1 Total $99,393,000 i O

O ,

4 Revision 4

()

8 O

IROJAliDECOMMISS70MNG PIAV Table 5,3-2 Portland General Electrie Decomanssioning Trimat Fund Cash now (Neanimal 5 s 1999)

PGE Trust PGE Trust PGE Trust PGE Trust Bridge Bridge Letter of Ixtter of Fund Fund Fund Net Fund EOY Funds Funds Credit Credit Expenditures Contributions Earnings Balance Intere4 Fee A B C D E F G 11 I 1996 78,879 l 1997 (18,517) 14.041 3.388 77,791 1 1998 (36,821) 14.041 2,512 57,522 1 1999 (69,328) 14.041 101 2,262 63,484 I (74) 2000 (35,053) 14.041 0 0 18,983 620 66,633 I (233) 2001 (29,513) 14,041 0 0 15,622 1,790 42,877 I (150)  ;

2002 (16.096) 14,041 0 0 2,114 2,486 16,984 I (59) 2003 (4,434) 14,041 0 0 (9,59:) 2,718 4,435 I (16) 2004 (3,330) 14.041 0 0 (10,711) 2,269 I 2005 (3,411) 14.041 0 0 (10,630) 1,717 1 20L% (3,490) 14.081 0 0 (10.551) 1,136 1 2007 (3,565) 14.041 114 2.617 (7.974) 521 I 2008 (3,636) 14,os1 568 13.069 (521) 34 1 2009 (3,701) 14.041 1.062 24,437 (34) 2 1 2010 (3,759) 14.04i 1,577 36,294 (2) l NOTEI: Ibsitive numbers irx!icate cash flow imo trust fuml; negative numbers indicate cash I'.ow out cf trust furxi. hge I of 2 :

NOTE 2: Current EOY balance = previous year EOY balance + current year A + B + C + E + H. Revision 4 I

O O O 1

?ROJANMCORDG55F0fttMMAN 1

Table 5.3-2 Partimed General Doctrie DeesammaissismingTrust Fund Cash Row (Namshnsi$ m leasp FGE Trust PGE Trust FCE Trust - FCETrust Brite Artge amener er .s anseer Faed Fund ' Femmi Met Fund EOY Fames Funds Creet Csuet Expendeures Cenertneless Earnings N Inessest Fee-A B C D E F G N I 2011 (3,809) 14.041 2,115 48,640 0 2012 (3,918) 11,521 2,558 58,802 1 2013 (4,029) - 2.494 57,268 I 2014 (4,143) 2,424 55,548 1 2015 (4,261) 2.344 53,632 1 2016 (4,382) 2,255 31,505 I 2017 (4,506) 2,154 49,153 1 2018 (26,661) 1,034 23,526 1 2019 (17,664) 275 6.137 I 2020 (433) 269 5,712 1 2021 (1,074) 232 5,131 1 2022' (1,095) 194 4.229 I 2023 (4,229) 0 0 I I

, Tseal (314,858) 222,136 27,670 (13.295) 13.293 1 ($32) I NOTEI: Pbsitive nusebers indicaec cash fkw i:seo trust fund; negative W i=d=r* cash flow out of trust fund. Page 2 of 2 :

1. 40TE 2: Current EOY balsace = previous year EOY baissce + current year A + B + C + E + H. Revisiosi4 I m

O O O 1ROJANDECOMMIS90NING PLAN Taide 5.3-3 EWER /BFA A- ~~

• _ Trwt Fund Cash How (Nennimal $ x leeM EWEB/BFA EWEB/BFA EWEB/BFA EWEB/BFA Bridge Bridge 14tter of IAtter of Trust Fund Trut Fund Trust Fund True Fund Foods Funds Credit

• Crede Fee

• Expenditures Centributions Earnings EOY B a==re Interest

• A B C D E F G H I 1996 15.W5 l 1997 (8.230) 2.374 386 10.475 I 1998 (16.258) 2.493 0 0 3.290 1 1999 (30.70I) 2.618 0 0 28.083 1 2000 (15.475) 2.749 0 0 12.726 l 2001 (13.050) 650 31 681 13.050 - l 2002 (7.128) 650 67 1.398 7.128 1 3 2003 (1.964) 650 102 2.150 1.961 I 2004 (1.480) 650 140 2.940 1.480 I 650 179 3.769 1.516 1 2005 (1.516) 650 221 4.640 1.551 1

'2006 (1.551) 650 265 5.555 1.584 1 2007 (1.584) 2008 (1.616) 650 310 6.515 1.616 l 650 358 7.523 1.645 1 2009 (1.645) 2010 650 409 8.582 1.671 I (1.671) 2011 (1.689) 650 462 9.694 1.689  !

650 517 10.861 1.741 1 2012 (1.741)

NOTE 1: Positive numbers indicate cash flow irmo trust fund; negaine numbers mdicate cash flow out of trust fund.

PageI of2 :

NOTE 2: Current EOY balance = premus year EOY balance + current year A + B + C + E. Revision 4 1

p a kJ O 1ROJANDECUMMIS57057NG PLAN Table 5.3-3 EWEB/BPA Decosninissioning Trust Fund Cash How (Nosseinal $ x 1990)

FWEBIBPA EWEB/BPA EWEB/BPA EMTIL'BPA Bridge Bridge Ixtter of Letter of

. Trmt Fund Tret Fund Trust Fund Tret Fund Funds Fonds Credit Credit Fee

• Expenditures Contributions Earnings E0Y Balance Interest

• A B C D E F G H I 2013 (1.791) 650 575 12.086 1.791 1 2014 (1.84I) 650 637 13.373 1.841 1 2015 (1.894) 650 701 14.724 1.8M i 2016 (1,W8) 650 769 16.143 1.W8 I 2017 (2.003) 650 840 17.633 2.003 1 2018 (11.R49) 650 881 7.315 0 1 2019 (7.851) 650 366 4RO O I 2020 (192) 650 24 %2 0 1 2021 (477) 650 48 1.183 0 1 2022 (487) 650 59 1.405 0 I 2023 (1.880) 0 0 0 475 I I

Total (139.512) 14.534 8.347 90.686 I

• BPA will provide bridging funds as im3wy from their operating budget, and thus will incur no loan costs.

Financial assurance wdl be provih .ya statement of intent as allowed by 10 CFR 50.75.

NOTE 1: Positive numbers indicate cash flow into trust furxt negative numbers indicate cash flow out of trust fund. Page 2 of 2 NOTE 2:. Current EOY balance = prevous year EOY balance + current year A + B + C + E. Revision 4 I i

p m d bm U TROJANDECOMMISSIOh7NG PLAN Table 5.3-4 Pacific Power & Lidd DecennaissionungTrust Fund Cash flow (Noenimal $ x 1999)

PP&L Trust PP&LTnnt k'P&L Trust PP&LTrust Bridge Pridge Letter of Letter of Fund Fund Fund Fund EOY Funds Funds Credit Credit Fee Expenditures Contributions Earnengs Balance Interest A B C D E F G H I 1996 4.569 1 1997 (1.232) 5% 208 4.141 1 1998 (1.355) 535 156 3.477 1 1999 (2.560) 535 99 1.551 1.286 (2) I 2000 (l.295) 535 41 832 992 (5) 1 2001 (I 092) 535 I3 288 751 (4) I 2002 (5%) 338 0 30 337 (2) I 2003 (165) 338 0 203 134 I (1) 2004 (123) 338 6 424 I 2005 (126) 338 16 652 1 2006 (129) 338 26 887 l 2007 (132) 338 37 1.130 1 2008 (135) 338 48 1.381 I '

2009 (l37) 338 59 I.64I I 2010 (139) 338 71 1,911 1 NOTE 1: Positive numbers indicate cash flow into trust fund; negative numbers indicate cash flow out of trust fund. J NOTE 2: Current EOY balance = previous year EOY balance + current year A + B + C. I NOTE 3: The amount shown as the trust fund expenditure for 1997 is the amount Pacific Power and Ligts withdrew Page 1 of 2 l  ;

from its decommissioning trust fund in 1997 and prior years.

Revisiort 4 I t

O O O TROJANDECOMMIS$10h7NG PIAV Table 5.3-4 PacNic Fwwer & light Decesumessdoming Triad Fund Cash Row (Nousimal $ x leem M&L Trust N&LTrust FP&LTrust N&L Trud Bridge Bridge 14eeer d Ineser er y Fund Famd Fund F M EOY Feeds Funds Creet Credit Fee F_;;

"^

__ o Cenergpations Earnings Balmace Imeerist A B C D E F G H I 2011 (141) 337 83 2,190 1 2012 (145) 95 2,140 1 2013 (149) 93 2,084 I 2014 (153) 90 2.021 I 2015 (159) 87 1.949 1 20I6 (162) 81 1.871 1 2017 (1671 80 1,7&t i 2018 (987) 58 855 1 2019 (654) 24 225 1 2020 (16) 10 219 I 2021 (40) 9 188 l 2022 (41) 8 155 1 2023 (157) 3 1 I I

Total (12,187) 6.115 - 1.504 (14) i NOTE 1: Positive numbers indicate cash flow imo trust fund; regattve numbers mihcase cash flow out of trust fund. I NOTE 2: Current EOY balance = previous year EOY balance + current year A + B + C. I NOTE 3: The amount shown as the trust fund w

• i-s for 1997 is the amount Pacifs Pbwer and Light withdrew  % 2 of 2 from its awom-miomng trust fund in 1997 and prior years. Revisiott 4 i L