Text

Trojan ISFSI Safety Analysis Rept
	ML20135B534
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	11/25/1996
From:	; PORTLAND GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20135B473	List: ML20135B469; ML20135B507; ML20135B534;
References
	NUDOCS 9612050049
	Download: ML20135B534 (53)
	v • d • e

. . - - - -. - - . - . - . - - . - .

l INSTRUCTION SHEET .

The following information is provided as a guide for the insenion of new sheets for changes to - j

. the " Trojan Independent Spent Fuel Storage Installation Safety Analysis Report," dated '

'OV 4

November 25,1996 Remove InWII )

Table of Contents Table of Contents 4- ]

pages i through xi pages i through xi List of Tables 1-3 List of Tables 1-3 a

! List of Figures 1-3 List ofFigures 1-3 i

i List ofDrawings 1 List ofDrawings 1 List of Effective Pages List of Effective Pages pages 1 through 7 pages 1 through 10 i.

Chapter 1 Chapter 1 page 1-3 page 1-3 Chapter 3 Chapter 3 3

page 3-2 page 3-2 page 3-14 page 3-14 pages 3-16 through 3-19 pages 3-16 through 3-19 Table 3.2-5 Table 3.2-5 Chapter 4 Chapter 4

page 4-3 page 4-3

, pages 4-5 and 4-6 pages 4-5 and 4-6 pages 4-10 and 4-11 pages 4-10 and 4-11

page 4-32 page 4-32 ,

l

! Table 4.2-la Table 4.2-la !

Figure 4.2-6 Figures 4.2-6a and 4.2-6b l Chapter 5 Chapter 5

- pages 5-1 through 5-3 pages 5-1 through 5-3 l

t Chapter 7 Chapter 7 j page 7-8 page 7-8 Chapter 9 Chapter 9

page 9-10 page 9-10 f Table 9.2-1 pages 1 of3 and Table 9.2-1 pages 1 of 3 and 2 0f 3 2of3 9612050049 961127 PDR ADOCK 05000344 P PH

i l

l Troian Independent Spent FuelStorasve installation Saferv Analysis Report h rh l V

TABLE OF CONTENTS ,

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF INSTALLATION . . 1-1 l

1.1 INTRODUCTION

. ... .... . . . . . 1-1 l 1.2 GENERAL DESCRIPTION OF THE INSTALLATION . . . .... .. . 1-2 l 1.3 GENERAL SYSTEMS DESCRIPTION , .. . .. 1-2 i 1.3.1 STORAGE SYSTEM BASKETS .. . . . 1-3 1.3.2 STORAGE SYSTEM CONCRETE CASK .. . .. 1-3 1.3.3 TRANSFER EQUIPMENT . . .. ..... . 1-4 j 1.3.3.1 Transfer Cask . . .. .. . 1-4 1.3.3.2 Transfer Station . .. . .. . 1-4 1.3.3.3 Auxiliarv Systems . . . . .. 1-5 1.3.4 AUXILIARY EQUIPMENT . .1 - 5 1.3.4.1 Vacuum Drving System .. . . 1-5 1.3.4.2 Semi-Automatic Weldine System . . .. . 1-5 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS . 1-5 1.5 MATERI AL INCORPORATED BY REFERENCE . 1-6 2.0 SITE CHARACTERISTICS .. 2-1 l N 2.1 GEOGRAPHY AND DEMOGRAPHY , . 2-1 2.1.1 SITE LOCATION . . . . 2-1 2.1.2 SITE DESCRIPTION .. .. . . . 2-2 2.1.2.1 Other Activities Within the ISFSI Site Boundarv . 2-4 2.1.2.2 Boundaries for Establishing Effluent Release Limits . . 2-4 2.1.3 POPULATION DISTRIBUTION AND TRENDS . . . 2-5 2.1.4 USES OF ADJACENT LANDS AND WATERS 2-6 2.2 NEARBY INDUSTRIAL. TRANSPORTATION AND MILITARY FACILITIES

. .. .. . . .. . . 2-8 2.2.1 LOCATIONS AND ROUTES .. .. 2-8 2.

2.2 DESCRIPTION

OF PRODUCTS AND MATERIALS . . 2-11 2.2.3 EVALUATION OF POTENTIAL ACCIDENTS 2-12 2.2.3.1 Exolosions 2-12 2.2.3.2 Toxic Chemicals 2-14 l 2.2.3.3 Eites . . . . . 2-14 2.2.3.4 Aircraft Imnacts . .. 2-16 1 2.2.3.5 Cooling Tower Collanse . 2-18 l 2.2.3.6 Air Pollutants . 2-19 2.3 METEOROLOGY . .. . 2-19 2.3.1 REGIONAL CLIMATOLOGY 2-19 2.3.1.1 General Climate 2-19 i November 25,1996

I l

Trvjan independent Spent FuelStorage Installation Saferv Anahsis Report h r

2.3.1.2 Severe Weather ....... ...... .. .... . ... 2-20 2.3.2 LOCAL METEOROLOGY . . . . ........... . . . . 2-21 2.3.2.1 Nogaal and Extreme Values of Meteorolonical Parameters . 2-21 i

~ 2.3.2.2 Patential Influence of the ISFSI on Local Meteorology ... 2-22

{

2.3.2.3 .Tooographic Description . . . . . . . . . . . . . . .. ..... .. 2-22 2.3.3 ONSITE METEOROLOGICAL MEASUREMENTS PROGRAM . 2-23 2.3.4 DIFFUSION ESTIMATES .... . ............. .. . ... 2-23 2.4 HYDROLOGIC ENGINEERING . . . . . . . . . . ......... . ..... 2-25 2.4.1 HYDROLOGIC DESCRIPTION . . ...... . . . . . . . . . . 2-26 2.4.1.1 Site and Facilities . . . . . . . . . . . . ... . ..... ..... 2 26 2.4.1.2 Hydroschere . . . ... ......... . .. ...... . 2-27 1 2.4.2 FLOODS . . . . . . . . . . . . . . . . . . . . . .... ... .... . 2-28 2.4.2.1 Flood Historv . .. . . . .. ....... ...... . . 2-28 2.4.2.2 Flood Design Considerations . . . ...... . ... 2-28 2.4.2.3 Effects of Local Intense Precipitation . . . . . . .. . . . 2-29 2.4.3 PROBABLE MAXIMUM FLOOD (PMF) OF STREAMS AND RIVERS

... . . .. .... . . . . . . . . ... . .... . .... ... . 2-30 2.4.4 POTENTIAL DAM FAILURES . .. .. . . . .. .... . 2-31 2.4.4.1 Seismically Induced Dam Failure .. .... . . . 2-31 2.4.4.3 Soirit Lake Blockage Failure . . . . . , . . . . . . . . . . . . 2-3 2 2.4.5 PROBABLE MAXIMUM SURGE FLOODING . ......... . 2-32 2.4.6 PROBABLE MAXIMUM TSUNAMI FLOODING ,.. .. . . 2-3 3 2.4.7 ICE EFFECTS .. ... ....... .................... . . 2-33 2.4.8 FLOODING PROTECTION REQUIREMENTS ...... . . . 2-33 2.4.9 ENVIRONMENTAL ACCEPTANCE OF EFFLUENTS . . . . . 2-34 2.5.1 REGIONAL AND SITE CHARACTERISTICS . . . . . .. .. . 2-34 2.5.2 CONTAMINANT TRANSPORT ANALYSIS . . . .... .. . . 2-35 2.6 GEOLOGY. SEISMOLOGY AND GEOTECHNICAL ENGINEERING . . . 2-36 l 2.6.1 BASIC GEOLOGIC AND SEISMIC INFORMATION , . . 2-36 i 2.6.1.1 Regional Geolonv . . ... . .. . . .. .. .. 2-37 i 2.6.1.2 Site Geolony - . ... .... .......... .... . 2-39 l 2.6.2 VIBRATORY GROUND MOTION . .... ... . . 2-42 i 2.6.2.1 Seismicitv . ... .. .. . .. .. . , 2-42 2.6.2.2 Geolonic Structures and Tectonic Activity . . . . 2-42 2.6.2.3 Maximum Earthauake Potential . . . . . . . . . . 2-43 2.6.2.4 Seismic Margin Earthauake ... ........ .. ,. 2-44 2.6.3 SURFACE FAULTING . . .. .. . .. . . . . 2-46 2.6.4 STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONS

.... ... ... . ..... .. .. .... ... ....... 2-47 2.6.4.1 Geological Foundation Evaluation . . . . . . ... . . 2-48 2.6.5 STABILITY OF SLOPES . . ..... . .. . . 2-49 ii November 25,1996

O Twian Independent Spent FuelStorage Installation Safety Analysis Report @ '

2.6.6 VOLCANOLOGY . . . . . . . ... .. ........ .. . .. 2-50 2.7

SUMMARY

OF SITE CONDITIONS AFFECTING CONSTRUCTION AND ;

OPERATING REOUIREMENTS . . . . . . . ... .. .. . ... 2-51

2.8 REFERENCES

. . . . . . . . . . . . ... . ....... .... ........ .. 2-52 3.0 PRINCIPAL DESIGN CRITERIA . . . .. .. . ....... ...... ... . 3-1 3.1 PURPOSES OF INSTALLATION . . . .. .. ...., ........ ..... . 3-1 3.1.1 MATERIALS TO BE STORED . . . . .... .. ... ... ... 3-1 ,

3.1.1.1 Intact Fuel Assemblies . ..... . .. ..... ... . 3-1 3.1.1.2 Failed and Partial Fuel Assemblies ... ... . ...... 3-2 '

3.1.1.3 Fuel Debris . .. .... .... ...... .. ... .. .. . 3-2 3.1.1.4 GTCC Waste . .. .. .. .. ...... . ..... 3-2 ,

3.1.2 GENERAL OPERATING FUNCTIONS . . . . . . . . . . . . 3-3 3.2 STRUCTURAL AND MECHANICAL SAFETY CRITERIA . . ..... 3-3 ,

3.2.1 TORNADO AND WIND LOADINGS . . . . .... .. 3-4 l 3.2.1.1 Aeolicable Desien Parameters . . .. ..........

~

.. 3-4 1 3.2.1.2 Determination of Forces on Structures .. . . . 3-5 3.2.1.3 Ability of Structures to Perform Despite Failure of Structures Not Designed for Tornado Loads .. . . .. 3-5 '

3.2.1.4 Tomado Missiles . . .. .. ... ....... . 3-5

"% 3.2.2 WATER LEVEL (FLOOD) DESIGN . . . ........ ... .. 3-6

(% - 3.2.2.1 Flood Elevations . . .. . . .. . . . 3-6 3.2.2.2 Flood Protection . . . . ... . . . . . 3-6 3.2.3 SEISMIC DESIGN . . .. . . . . . . . . 3-6 3.2.3.1 Input Criteria . . . .. . , .. .... .. . 3-6 3.2.3.2 Seismic-System Analyses . . . . . ... 3-9 3.2.4 SNOW AND ICE .. .. . . ....... ... .. . .... 3-11 3.2.5 COMBINED LOAD CRITERIA . . . . . .. . . .. . 3-12 3.2.5.1 Load Combinations and Design Strencth - Concrete Cask 3-13 3.2.5.2 Load Combinations and Design Strencth - PWR Basket. GTCC Basket. and Basket Overnack . .. . 3-13 3.2.5.3 Load Combinations and Design Strencth - Transfer Cask . 3-14 3.2.5.4 Load Combinations and Design Strencth - Failed Fuel Can . 3-14 3.2.5.5 Load Combinations and Design Strencths - Fuel Debris l Process Can Caosule . .. . .. ... . 3-14 3.3 SAFETY PROTECTION SYSTEMS . .. . . . . . 3-15 3.3.1 GENERAL . . . . .. .. .. . ... . . . . . . 3-15 3.3.2 PROTECTION BY MULTIPLE CONFINEMENT BARRIERS AND SYSTEMS ... .. ...... . . .. 3-17 3.3.2.1 Confmement Barriers and Systems ... . .. . . 3-17 3.3.2.2 Ventilation-OfTgas . . , 3-18 iii November 25.1996

Trojan independent Spent FuelStorare Installation Saferv Analvsis Report h O 3.3.3 PROTECTION BY EQUIPMENT AND INSTRUMENTATION SELECTION . . . . . . . .. .

... . . . . . ... .. . 3-18 3.3.3.1 Eauioment . . . .. ...... .. . .... ...... . 3-18 3.3.3.2 Instrumentation . . . . . ....... .... . . .... . 3-19 3.3.4 NUCLEAR CRITICALITY SAFETY . .... .. .. .... 3-19 3.3.4.1 Control Methods for Prevention of Criticality . . ~

. . . 3-19 3.3.4.2 Error Contingency Criteria . . ........... .. ... 3-20 3.3.4.3 Verification Analyses . . . . ......... . . .. . 3-20 3.3.5 RADIATION PROTECTION . . . . ..... ..... . 3-20 3.3.5.1 Access Control . .... .... .. ., ..... . ... . 3-20 3.3.5.2 Shielding . . .. .... . .. ... ... ... . . 3-20 3.3.5.3 Radiolonical Alarm Systems . . . ... . . . .. 3-21 3.3.6 FIRE AND EXPLOSION PROTECTION . . . . . . . 3-21

' 3.3.7 MATERIALS HANDLING AND STORAGE .. . . . . 3-22 3.3.7.1 Spent Fuel or GTCC Waste Handlina and Storace ~

.... 3-22 3.3.7.2 Radioactive Waste Treatment .. .. .'. . .. 3-23' 3.3.7.3 Waste Storage Facilities .. ... ... . . 3-23 3.3.8 INDUSTRIAL AND CHEMICAL SAFETY .. . . . . 3 23 3.4 CLASSIFICATION OF STRUCTURES. SYSTEMS AND COMPONENTS 3-23 3.5 DECOMMISSIONING CONSIDERATIONS .... . .. . . ... 3-24 s 3.6

SUMMARY

OF DESIGN CRITERIA . . .. .. ...... .... . 3-24

3.7 REFERENCES

. . . . .... . . .. .. . . . 3-25 4.0 INSTALLATION DESIGN . ...... . ... . ........ .. .. 4-1 4.1

SUMMARY

DESCRIPTION . . . . . . .... .. ... . .. . .. 4-1 4.1.1 LOCATION AND LAYOUT OF INSTALLATION . . , . .4-1 4.1.2 PRINCIPAL FEATURES . ... . . . . ... .. 4-1 4.1.2.1 Site Boundarv .. ... .. . . . . .4-1 4.1.2.2 Controlled Area . . ... .... . .. . . . . 4-1 4.1.2.3 Site Utility Sucolies and Systems . . .. .. 4-2 4.1.2.4 S12Iage Facilities . . . . . . .. . . . . .. .. 4-2 4.1.2.5 Stacks . ... . . . . . .. ... . . . 4-2 l 4.2 STORAGE STRUCTURES . . . . . .. . ..... .. . . . . 4-2 4.2.1 STRUCTURAL SPECIFICATION . . .. 4-3 4.2.2 INSTALLATION LAYOUT ... . . . . . 4-4 4.2.2.1 Buildine Plans and Sections . ..... ... ... 4-4 4.2.3 CONFINEMENT FEATURES . .. . .. .... . 4-4 4.2.4 INDIVIDUAL UNIT DESCRIPTION . . . . . 4-5 4.2.4.1 Functional Descriotion . . . . ... .. ..... . . 4-5 4.2.4.2 Component Descriptions . ... ... . .. . 4-5 4.2.4.3 Design Bases and Safety Assurance . . . 4 - 12 iv November 25.1996

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

1 Trojan Independent Spent FuelStorage installation Safety Analysis Report i

4.2.5 STRUCTURAL EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . 4 -' 13 4.2.5.1 Weights and Centers of Gravity . ... .... . '4 - 14 4.2.5.2 Mechanical Properties of Materials . . . . . ... .. . . . . . 4 - 14

. 4.2.5.3 Basket Analysis Under Normal Loads . . . . . , ... . . 4 - 14 4.2.5.4 Concrete Cask Analysis under Normal Ooeratino Loads ~

. 4 19 4.2.6 THERMAL EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . 4 - 21 !

4.2.6.1 Summarv of Thermal Properties of Materials . . . . . . . . . 4 - 22 .

4.2.6.2 Thermal Models for Normal Storane Conditions . . . . . . . . 4 - 23 4.2.6.3 Air Flow and Tenipediare Calculation . . . . . . .. . . . . . . . 4 - 24 ;

4.2.6.4 Concrete Cask Body and PWR Basket Exterior Thermal Model- *

......... .................. ..... ........... 4 - 26 4.2.6.5 PWR Basket Thermal Hydraulics . .. ......... .. 4 -29 4.2.6.6 Maximum Temneratures . . . . .... .. . . . . ... . 4-31 l 4.2.6.7 Minimum Temocratures . . . . ..,,. ... . 4-31 4.2.6.8 Maximum Internal Pressure . . .. ... . ..... ... . 4 -32 .

4.2.6.9 Evaluation of Cask Lifetime Performance under Normal !

Conditions of Storage . ..... . .. . . . . . 4 -32 4.

2.7 CRITICALITY EVALUATION

. . . . . . . . . . ., .. 4 -32 ,

4.3 AUXILIARY SYSTEMS . . ...... ..... ..... . .......... . . 4 - 33 :

4.3.1 VENTILATION AND OFF-GAS SYSTEMS . . . . . . . . . . . . . . 4 - 33 4.3.2 ELECTRICAL SYSTEMS . ........ .... .. ... . . . . 4 - 34 !

4.3.3 AIR SUPPLY SYSTEMS , , ,. .. . .. .. ... . . 4 -34 4.3.4 STEAM SUPPLY AND DISTRIBUTION SYSTEM . . . ... . 4 - 34 ;

4.3.5 WATER SUPPLY SYSTEM , . . ... . ....... .. . . ~ 4 - 34 ,

l 4.3.6 SEWAGE TREATMENT SYSTEM . . . . . . . . . . . . . . . . . 4 - 35 4.3.7 COMMUNICATION AND ALARM SYSTEMS . .. , , . . 4 - 35 4.3.8 FIRE PROTECTION SYSTEM . . . . . . . ..... .. ......4-35 1 4.3.9 MAINTENANCE SYSTEMS . . . . . ...... ....... ... . 4 -35 ,

4.3.10 COLD CHEMICAL SYSTEMS . .... .... . . . 4 -36 !

4.3.11 AIR SAMPLING SYSTEMS . ..... . ......, ... 4 -36 4.4 DECONTAMINATION SYSTEMS . . . . . . . . . . . . . . ..... . ... 4-36 4.4.1 EQUIPMENT DECONTAMINATION . . . . . . . .. .. 4 -36 4.4.2 PERSONNEL DECONTAMINATION .. .. .. .. . 4 -36 4.5 SHIPPING CASK REPAIR AND MAINTENANCE . ... .... 4 -37 l 4.6 CATHODIC PROTECTION . ..... . . . . .. .. . . ... . . . . . 4 - 37 l 4.7 SPENT FUEL AND HIGH-LEVEL RADIOACTIVE WASTE HANDLING i OPERATION SYSTEMS . . . ... .... .... . . . . 4 - 37 l

4.7.1 STRUCTURAL SPECIFICATIONS . . . . . . . . .. . .....4-38 4.7.2 INSTALLATION LAYOUT . .. .. ..... .... .. . 4 -39 4.7.2.1 Building Plans and Sections ... ... . . 4 -39

- 4.7.2.2 Confinement features . . .... .... . . . . 4 -39 i

y November 25,1996 1

l l

l

1 1

Trojan Independent Spent Fuel Storage Installation Safety Anaksis Report .

O 4.7.3 INDIVIDUAL UNIT DESCRIPTION . . . . . . . . 4 -39 4.7.3.1 Transfer Cask Descriotion .. . . . . ... 4 -39 j 4.7.3.2 Transfer Station ..... .. ... . .. ... .... 4 -40 1 4.7.3.3 Air Pad System . . . .. . . . . . . . 4 - 41 4.7.3.4 Lifting Yoke .... . .. ...... ... ... 4 -41 4.7.3.5 Hoist Rings . . . .... . . . . ... 4-41 4.7.3.6 Mobile Cranes . . . . . .. .. . . . . . 4 - 41 4.7.4 STRUCTURAL ANALYSIS OF THE FUEL HANDLING COMPONENTS . . . . ... . . ... . . . . .. .. .. 4-42 4.7.4.1 Transfer Cask Lin . .. . . . . . 4 -42 4.7.4.2 Air Pad System . . . . . . . 4 - 52 4.7.4.3 Lifling Yoke . . . . . . . . 4 - 53 4.7.4.4 Hoist Rings .. . . . . . . 4 - 54 4.7.5 THERMAL EVALUATION DURING FUEL TRANSFER ... . 4 - 55 4.7.5.1 Transfer Cask Heat Transfer Modes . .. .. 4 - 55 4.7.5.2 Basket Thermal Hydraulic Model . .. . ... 4 - 56 5.0 OPERATIONS . ..... .. . . . . . . . . . . 5-1 5.1 GENER.AL DESCRIPTION . . .. . .... .5-1 5.1.1 OPERATION DESCRIPTION ,5-1

/ 5.1.1.1 Failed Fuel. GTCC Waste. and Fuel Debris Process Can l Capsule Loading ... . .. . . . . . . 5-2 l 5.1.1.2 Basket Loadine and Sealing Ooerations . . . . 5-3 5.1.1.3 Transfer to Storage Area Ooerations . .... .. . 5-5 5.1.1.4 Maintenance Operations . . .. ... 5-6 5.1.1.5 Off-Normal Event Recovery Onerations .. .. 5-6 5.1.1.6 Off-Site Transfer Onerations ... . ... . 5-7 5.1.2 FLOWCHARTS . .. .. . . . . . . 5-8 5.1.3 IDENTIFICATION OF SUBJECTS FOR SAFETY ANALYSIS 5-8 5.1.3.1 Criticality Prevention . . . . 5-8 5.1.3.2 Chemical Safety . . . 5-8 5.1.3.3 Ooeration Shutdown Modes . .5-8 5.1.3.4 Instrumentation . . . . .5-9 5.1.3.5 Maintenance Techniaues . .. 5-9 5.1.3.6 Heavy Loads Procedures . . . 5-9 5.2 SPENT FUEL HANDLING OPERATIONS . . . . . 5 - 10 5.2.1 SPENT FUEL HANDLING AND TRANSFER . 5 - 10 5.2.1.1 Functional Descrintion . 5 - 10 5.2.2 SPENT FUEL STORKGE . . . . . . 5 - 12 5.2.2.1 Insoection ana' Surveillance Program . . . 5 - 12 5.2.2.2 Saf'ety Features . 5 - 12 vi November 25.1996

s t

Trojan Independent Spent FuelStoraxe Installation Saferv Analvsis Report h (3

5.3 OTHER OPERATING SYSTEMS

.... . . . 5 - 13 5.3.1 SYSTEM OPERATIONS . . . . . 5 - 13 5.3.2 COMPONENT / EQUIPMENT SPARES . . . . . . 5 - 13 5.4 SUPPORT SYSTEM OPERATION . ... .. . . .. . . . 5 - 13 5.4.1 INSTRUMENTATION AND CONTROLS . ... . . 5 - 13 5.4.2 SPARES . . . ... . . . 5 - 14 5.5 CONTROL ROOM AND CONTROL AREAS . . . . . 5 - 14 5.6 ANALYTICAL SAMPLING . .. . . . . . 5 - 14 6.0 SITE-GENERATED WASTE CONFINEMENT AND MANAGEMENT . .6-1 6.1 ONSITE WASTE SOURCES . ... . . 6-1 6.2 OFFGAS TREATMENT AND VENTILATION . . . 6-1 6.3 LIOUID WASTE TREATMENT AND RETENTION . 6-1 6.4 SOLID WASTES .. . . . .. .. . 6-2 6.5 RADIOLOGICAL IMPACT OF NORMAL OPERATIONS -

SUMMARY

6-2 7.0 RADIATION PROTECTION . . . . .. 7-1 7.1 ENSURING THAT OCCUPATION RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) ... . .. .7-1 7.1.1 POLICY CONSIDERATIONS . . 7-1 7.1.2 DESIGN CONSIDERATIONS . . . 7-1 7.1.3 OPERATIONAL CONSIDERATIONS . 7-2 7.2 RADIATION SOURCES . . .. 7-5 7.2.1 CHARACTERIZATION OF SOURCES 7-5 7.2.1.1 Fuel Gamma Source . . . 7-5 7.2.1.2 Fuel Neutron Source . .. . 7-7 .

7.2.1.3 Non-Fuel Recion Gamma Sources . . . 7-8 7.2.1.4 Greater Than Class C Waste . . 7-8 7.2.1.5 Fuel Debris . ... . . 7-9 7.2.1.6 Non-Fuel Bearine Comoonents . .7-9 7.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES 7-9 7.3 RADIATION PROTECTION DESIGN FEATURES 7 - 10 7.3.1 INSTALLATION DESIGN FEATURES 7 - 10 ;

7.3.2 SHIELDING . . 7 - 12 7.3.2.1 Radial and Axial Shieldine Confieurations 7 - 12 7.3.2.2 Shieldine Evaluation . . . . 7 - 14 7.3.3 VENTILATION . . . . 7-17 7.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION . 7 - 18 7.4 ESTIMATED ON-SlTE COLLECTIVE DOSE ASSESSMENT 7 - 18 7.5 BAQTATION PROTECTION PROGRAM . 7 - 19 vii November 25,1996

. - . - . - - . - . . = - - - . - . . . . . - . . . - - - - - . . . - - - . - . - - - . -

t j Tmian Independent Spent FuelStorage installation Safety Analysis Report @

t

\

l j '

7.5.1 ORGANIZATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 19

7.5.2 RADIATION PROTECTION EQUIPMENT, INSTRUMENTATION, 1 AND FACILITIES . . . . . . . .... ....................7-20 i 7.5.2.1 Radiation Protection Instrumentation . . . . . . ..... ... 7 - 20 ,

j 7.5.2.2 Area Radiation Monitorine Instrumentr. tion .. .. . 7 - 20 -

l 7.5.2.3 Radiation Protection Facilities . . . . . . . . . .. ... .. 7 -20 I

} 7.5.3 RADIATION PROTECTION PROCEDURES . . . . . . . . . . . . 7 - 21 1 7.5.3.1 Control of Radiation Exoosure to the Public . . . .. . 7 - 21

7.5.3.2 Control of Personnel Radiation Exposure (Occupationah . 7 - 21

7.5.3.3 Records and Reoorts . . . ............. . . . . . . . . . . 7 - 24

i 7.6 ESTIMATED OFF-SITE COLLECTIVE DOSE ASSESSMENT . . . . .. 7 - 24 !

7.6.1 RADIOACTIVE EFFLUENT AND ENVIRONMENTAL MONITORING l PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 2 4 .

I 7.6.2 ANALYSIS OF MULTIPLE CONTRIBUTION .. .. . . . 7 - 25

. 7 - 25 7.6.3 ESTIMATED DOSE EQUIVALENTS . . . .

i 7.6.4 LIQUID RELEASE ..... . . . .. . .. . .. . 7 - 25

7.7 REFERENCES

. . . . . . . . . . ........ ..... . . . . . . . . ... . . 7 - 26 i l 8.0 ACCIDENT ANALYSIS .. . .. . ... ... . ... ... . . . .. .. 8-1

8.1 NORMAL AND OFF-NORMAL EVENTS . . . ..... . 8-2 8.1.1 OFF-NORMAL STRUCTURAL ANALYSIS . . ... . .8-3 8.1.1.1 PWR Basket Off-Normal Handling Load . . .. 8-3

) 8.1.2 OFF-NORMAL THERMAL ANALYSIS . . . . . . . . . . . 8-5

, 8.1.2.1 Severe Environmental Condition .. .. ... . 8-5 8.1.2.2 Blockage of One-Ha'f of the Air inlets .. ..... . .. 8-6 i 8.1.3 OFF-NORMAL CONTAMINATION RELEASE . ...... . 8-8 i 8.1.3.1. Small Release of Radioactive Particulates from Exterior of i Baskets ... . .. .... . . . ... . 8~8 j 8.1.3.2 Radiological Imoact from Off-Normal Qoerations .. . 8 - 10 1

8.2 ACCIDENTS .... .. . .. .. . .. . ..... .. .... 8 - 10 l

$ 8.2.1 FAILURE OF FUEL PINS WITH SUBSEQUENT BREACH OF PWR BASKET . .. .. . .... . .. . . . 8 - 10 !

8.2.1.1 Cause of Accident .... . . .. ... .. . 8 - 11 !

8.2.1.2 Accident Analysis .. ... .. .... ...... . . . . 8 - 11 8.2.1.3 Accident Dose Calculations .. .. . . . . . . . . . . . 8 - 12 8.2.2 MAXIMUM ANTICIPATED HEAT LOAD . . . ...... .. . . 8 - 13 1 8.2.2.1 Cause of Accident . ........ . . . . 8 - 13 l 8.2.2.2 Accident Analysis . . . .... ... . . .. . 8 - 13 l 8.2.2.3 Accident Dose Calculation . . ... . . .. .. . . 8 - 14 I 8.2.3 CONCRETE CASK OVERTURNING EVENT . . . .. . 8 - 14 8.2.3.1 Cause of Accident . . . . . . ... . . . .... . 8 - 14 viii November 25,1996 i

4

0 Troian Independent Spent Fuel Storage Installation Safety Analysis Report h O 8.2.3.2 Accident Analysis . . .. . . . . 8 - 14 8.2.3.3 Accident Dose Calculation . . .. . . 8 - 18 8.2.4 TORNADO . . . . . . . . .. ...... . . ... ... . 8 - 18 l 8.2.4.1 Cause of Accident . ....... . .. . . . . 8 - 18 I 8.2.4.2 Accident Analysis . . . .. . . . .... .. . . . . . 8 - 18 8.2.4.3 Accident Dose Calculations . . . . . . ... . 8 - 29 8.2.5 EARTHQUAKE EVENT . . . . . . . . . . . . . . . . . . . . . .. 8 - 29 8.2.5.1 Cause of Accident . . . ... ...... ... . .8-30 8.2.5.2 Accident Analysis . . .... . ... . 8 8.2.5.3 Accident Dose Calculations . . . .... . 8 - 33 8.2.6 PRESSURIZATION . . . . . . . .. ... .. . . . . 8 - 33 8.2.6.1 Cause of Accident .. . . . .... . .. 8 - 33 8.2.6.2 Accident Analysis . . . .. . . .. . 8 -33 8.2.6.3 Accident Dose Calculations . . . 8 -34 8.2.7 FULL BLOCKAGE OF AIR INLETS . . . . . . . 8 - 34 8.2.7.1 Cause of Accident . ... . . . . .... 8 - 34 8.2.7.2 Accident Analysis . . . . .. 8-35 8.2.7.3 Accident Dose Calculations . .. . 8 -35 8.2.8 EXPLOSIONS OF CHEMICALS, FLAMMABLE GASES, AND MUNITIONS . ... . ... . . .. ... . 8-36 A 8.2.8.1 Cause of Accident .. . . . 8-36 U 8.2.8.2 Accident Analysis .. . . . 8 -36 8.2.8.3 Accident Dose Calculations . . . . 8-38 .

8.2.9 FIRES .. . . .. . 8-38 l 8.2.10 COOLING TOWER COLLAPSE . . . . . 8-39 !

8.2.11 VOLCANISM .. . . . . 8-39 8.2.11.1 Cause of Accident . . .. . 8-39 I 8.2.11.2 Accident Analysis .. .. . . .. 8-39 8.2.11.3 Accident Dose Calculatip.ns . . . . 8 -40 8.2.12 LIGHTNING . . . . . 8-40 8.2.12.1 Cause of Accident .. .... . . . .8-40 8.2.12.2 Accident Analysis .. . .. . .. . . . 8-40 l 8.2.12.3 Accident Dose Calculations 8 - 41 1

I i

ix November 25.1996

t Troian Independent Spent Fuel Storage Installation Saferv Analysis Report h O 8.2.13 OVERPACK OPERATIONS AND OFF-SITE SHIPPING EVENTS. . ... ... . . .. . . . .. 8 - 41 I 8.2.13.1 Interference During Raisine the Basket from Concrete Cask into Transfer Cask ..... . .. ... . . . . . . . . 8 - 41 8.2.13.2 Interference During Basket Lowering into a Concrete

. Cask . . . . . . . . . . . . . . . . . . .. . . . . . . . . 8 -42 8.2.13.3 Basket Drop into Concrete Cask . . . . . . . . . . . . . . 8 - 43 8.2.13.4 Loaded Shinoing Cask Drop . , . . . . . . ... . .8-46 8.2.14 NATUPAL GAS TURBINE COMBINED CYCLE POWER PLANT EVENTS .... . ..... ... . . ....... . . .8-49 8.2.14.1 Cause of Accident . . .. . 8 - 50 8.2.14.2 Accident Analysis . . . . . ... . . . 8 - 50 8.2.14.3 Accident Dose Calculations . .. .. .... .... 8 - 54 8.3 SITE CHARACTERISTICS AFFECTING SAFETY ANALYSIS . . . .. 8 - 55

8.4 REFERENCES

. . .. ..... .. .... . .. . ... . . . . 8 - 56 9.0 CONDUCT OF OPERATIONS . . .. . .. ... . . .... . .. . 9-1 9.1 ORGANIZATIONAL STRUCTURE . . . .. ... ... . . .. 9-1 9.1.1 ISFSI CONSTRUCTION AND FUEL LOADING ORGANIZATION . . . . . . .. 9-2 9.1.1.1 Corocrate Oreanization . . .. ... .. 9-2 9.1.1.2 Site Oruanization 9-2

) . ...

9.1.1.3 Interrelationships with Contractors and Suppliers 9-4 9.1.1.4 Technical Staff . . . .. . .. . . .9-4 9.1.2 ISFSI OPERATION ORGANIZATION . . . .. . 9-5 9.1.2.1 Corocrate Organization .. . . ..... . 9-5 i 9.1.2.2 Operating Organization . . . ... .. .. . 9-5 9.1.2.3 Succession of Authority .. 9-5 l .. . .

l 9.1.2.4 Coroorate Support . . ... .. . 9-6 9.1.2.5 ISFSI Safety Review Committee . .. .. . . . 9-6 9.1.3 PERSONNEL QUALIFICATION REQUIREMENTS . . 9-6 9.1.4 LIAISON WITH OUTSIDE ORGANIZATIONS . . . ... .. 9-7 9.2 PRE-OPERATIONAL AND STARTUP TESTING , . . .. ... 9-7

! 9.2.1 ADMINISTRATIVE PROCEDURES FOR CONDUCTING TEST PROGRAM . . . . . .... . .. ...... .. . . 9-8

9.2.2 TEST PROGRAM DESCRIPTION .. . .. .. .. . 9-8 i 9.2.3 TEST DISCUSSION , .. .. .. .. . .. .. . 9-9 9.2.3.1 Physical Facilities . . . .. . . .9-9 9.2.3.2 Ooerations . . . ... . .. .... . . . . .. 9 - 11 9.2.3.3 Test Resoonse . . ... . . . . . . 9 - 12 9.2.3.4 Corrective Action . . . . .. . 9 - 12

' x November 25.1996

m ,h, A *_ A_# _. e+,-.h.__._e.._,. _-,__4.4u.i__J,a.,.4 ._ .a ,L , me_4,e b-.,, m m1._- = _h._. __m_a - . ..w..as a 4 - - <m,,. _

Tmfan Independent S.nent FuelStorage Installation Saferv Analysis Report f l O 9.3 TRAINING PROGR.AMS . . . . . . . . . . . . . . . 9 - 12 9.3.1 TRAINING PROGRAM DESCRIPTION 9 - 13 9.4 NORMAL OPERATIONS . . . . . . . . . . . 9 - 14 9.4.1 PROCEDURES . . . . . . . . . . . . . . . . . . 9 - 14 9.4.2 RECORDS . . . . . . . . . . . . . . 9 - 14 9.5 EMERGENCY PLANNING . . . . . . . . . . . . . . . . . 9 - 15 9.6 ISFSI DECOMMISSIONING PLAN . . . . . . . . . . . . 9 - 15 9.6.1 DECOMMISSIONING PROGRAM . . . . . . . 9 - 16 9.6.2 COST OF DECOMMISSIONING . . . . . . . . 9 - 16 9.6.3 DECOMMISSIONING FACILITATION . . . . . 9 - 16 9.6.4 RECORD KEEPING FOR DECOMMISSIONING 9 - 17 9.7 PHYSICAL SECURITY PLAN . . . . . . 9 - 17 10.0 OPERATING CONTROLS AND LIMITS . . . . . . . . 10 - 1 11.0 QUALITY ASSURANCE . . . . . . . . , . 11 _ 1 i

O g

1 Tmian Independent Spent FuelStorare Installation Saferv Analvsis Report h (3 l Q,I' LIST OF TABLES Page1of3 I Table Title 2.1-1 Public Facilities and Institutions 2.1-2 1994 Land Use Census j Nearest Location to Trojan Within a Five-Mile Radius ;

2.2-1 Nearby Industrial Facilities 2.7-1 Summary of Site Conditions Affecting Constmetion and Operating Requirements 3.1-1 Fuel Characteristics l

3.1-2 Design Maximum Radiological Characteristics of Stored Material 3.1-3 Design Maximum Thermal Characteristics of Stored Material 3.1-4 Physical Parameters of Fuel Assembly Inserts r

Q) 3.2-1 Wind and Tornado Design Specifications 3.2-2 Design Basis Tornado Generated Missiles 3.2-3 Design Load Combinations 3.2-4 Summary of Load Combinations 3.2-5 PWR Basket, GTCC Basket, and Basket Overpack Design Criteria 3.6-1 Summary of Concrete Design Criteria 4.2-1 Fabrication Specification Summary PWR Basket, GTCC Basket, and Basket Overpack 4.2-2 Concrete Cask Fabrication Summary 4.2-3 Conformity to Requirements 4.2-4 Weights and Centers of Gravity 4.2-5 Mechanical Properties of Steels Used in the Storage System 4.2-6 Properties of Concrete Used in Concrete Cask November 25,1996 (v )

O Troian independent Spent FuelStorage installation Saferv itnalysis Report -

O Page 2 of 3 LIST OF TABLES (continued)

Table Title 4.2-7 Summary of Maximum Basket Thermal Stresses (ksi) 4.2-8 Basket Maximum Stress Evaluation 4.2-9 Summary of Maximum Storage System Temperatures (Wit 1out Basket Overpack) 4.2-10 Concrete Cask Structural Load Combination Summary 4.2-11 Summary of Maximum Concrete Cask Thermal Stresses 75"F Ambient Air, Normal Operation 4.2-12 Summary of Cask Thermal Evaluation 4.2-13 Thermal Properties 4.2-14 Summary of Storage System Cooling Air Flow Analysis 4.7-1 Transfer Station Fabrication Specification Summary 7.2 Major Isotopic Contributors to Fuel Source Term 7.2-2 Fuel Region Gamma Source Strength (gammas /sec-assembly) for Various Burnup and Cooling Time Combinations (including control component source)

-7.2-3 Fuel Region Neutron Source Strengths (neutrons /sec-cask) 7.2-4 Relative Burnup Level and Source Strengths for PWR Assembly Axial Sub-Sections -

7.2-5 Non-Fuel Region Co* Gamma Source Strengths (y/sec-cask) 7.2-6 GTCC Gamma Source (y/sec-cask) 7.3-1 Elemental Densities (gm/cc) 7.3-2 Elemental Densities (Atoms / barn-cm) 7.3-3 Neutron Energy Group Flux-to-Dose Conversion Factors (mrem /hr per neutron /cm2,3ec)

November 25,1996 4

J

Troian independent Spent Fuel Storare installation Safety Analysis Report @-

O LIST OF TABLES (continued) Page 3 of 3 Table Title 7.3-4 Gamma Flux-to-Dose Conversion Factors)

(mrem /hr per y/cm2-sec) 7.4-1 Maximum Expected Dose Rates for the Storage Cask System (Fuel) 7.4-2 Maximum Expected Dose Rates for the Storage Cask System (GTCC) 7.4-3 Estimated Personnel Exposure Doses while Operating the Fuel Cask System 7.4-4 Estimated Personnel Exposure Doses while Operating the GTCC Cask System 8.0-1 Design Basis Normal and Off-Normal Events 8.0-2 Design Basis and Beyond Design Basis Infrequent (Accident) Events 8.1-1 PWR Basket Stresses (ksi) Resulting From Off-Normal Handling Events 8.1-2 Summary of Cask Thermal Evaluation 8.1-3 Summary ofImpact from Off-Normal Operations 8.2-1 PWR Basket and Basket Overpack Stresses (ksi) Resulting From Accident Pressurization 8.2-2 Beyond Design Basis Accident Dose Calculations 8.2-3 Regulatory Guide 1.76 Design Basis Comparison 8.3-1 Summary of Site Characteristics Affecting the Safety Analysis 9.1-1 ISFSI Stafling Qualifications 9.2-1 Pre-Operational, Startup, and Other Tests 9.6-1 ISFSI Decommissioning Costs

[ November 25,1996

l 1

Trojan independent Spent FuelStorare Installation Safety Analysis Report n

V LIST OF FIGURES Page1 of3 Figure Number Title 1.1-1 Location ofTrojan Nuclear Plant 1.1-2 Location ofISFSI at TNP Site 1.3-1 Storage System 2.1-1 Site Location <

l 2.1-2 PGE Property, ISFSI Site, and Controlled Area Boundary l

2.1-3 ISFSI Layout 2.1-4 1990 Population Distribution within 10 miles 2.1-5 2000 Project Population Distribution within 10 miles 2.1-6 2010 Projected Population Distribution within 10 miles i

2.2-1 NearbyIndustrial Activity !

(

v 2.2-2 LongviewIndustrial Activity 2.3-1 Topographical Cross Sections (hf aximum Elevations) SSE, S, SSW 2.3-2 Topographical Cross Sections (hiaximum Elevations) SW, NW, l WNW l l

2.3-3 Topographical Cross Sections (hf aximum Elevations) W, WSW, !

NNW j 2.3-4 Topographical Cross Sections (hf aximum Elevations) N, hWE, NE 2.3-5 Topographical Cross Sections (hf aximum Elevations) ENE, E i 2.3-6 Topographical Cross Sections (hiaximum Elevations) ESE, SE 2.4-1 ISFSI Topography 2 4.2 ISFSI Site Drainage 4.2-1 PWR Basket 4.2-2 GTCC Basket November 25,1996

Troian Independent Svent FuelStorare installation Safety Analysis Report @

O LIST OF FIGURES (continued) Page 2 of 3 i

Figure Number Title 4.2-3 Basket Overpack 4.2-4 Concrete Cask '

4.2-5 Failed Fuel Can 4.2-6a Fuel Debris Process Can l 4.2-6b Fuel Debris Process Can Capsule l 4.2-7 GTCC Can 4.2-8 Weights and Center of Gravity 4.2-9 Concrete Cask Thermal Model 4.2-10 PWR Basket Heat Transfer Model 4.2-11 PWR Basket Temperature Distribution (100*F Ambient Air) l 4.7-1 Transfer Cask 4.7-2 Transfer Cask Lining Tmnnion Design :

4.7-3 Transfer Station ,

l 4.7-4 Lining Yoke j 4.7-5 Transfer Cask ANSYS Model 4.7-6 Transfer Cask Shield Door Rail Design 4.7-7 Transfer Cask Temperature Contours (75'F Ambient Air) 5.1-1 Operations Sequence 5.1-2 Operations Sequence for Leaking Basket Recovery 5.1-3 Shipping Cask Operation Sequence 7.2-1 PWR Fuel Assembly Axial Burnup Profile 7.3-1 PWR Basket in Transfer / Concrete Cask (Top Axial Model)

O V

November 25,1996

/

Trojan independent Spent FuelStorare Installation Saferv Analvsis Report h 4

LIST OF FIGURES (continued) Page 3 of 3 i !

I Figure Number Title I i

7.3-2 PWR Basket in Concrete Cask (Radial Model) i i 7.3-3 GTCC Basket in Transfer / Concrete Cask (Top Axial Model) 7.3-4 GTCC Basket in Concrete Cask (Radial Model) i 7.3-5 PWR Basket in Transfer Cask (Bottom Axial Model) 4 l 7.3-6 PWR Basket in Transfer Cask (Radial Model) !

7.3-7 GTCC Basket in Transfer Cask (Bottom Axial Model) 7.3-8 GTCC Basket in Transfer Cask (Radial Model) l 7.3-9 Summary of Dose Rates for Concrete Cask (PWR) 7.3-10 Summary of Dose Rates for Concrete Cask (GTCC) 7.3-11 Summary of Dose Rates for Transfer Cask )PWR) 8.1-1 Concrete Cask Temperature (100*F Day) l 8.1-2 Concrete Cask Temperature (-40 F Day) 1 8.2-1 Concrete Cask Temperature (125"F Short Term) 8.2-2 Missile / Cask Impact Geometry 8.2-3 Cask Tip-Over Geometry 8.2-4 Outlet Air Temperature for Full Inlet Blockage 9.1-1 ISFSI Organization - Construction and Fuel Loading !

i 9.1-2 ISFSI Organization - Operation '

l i

j O

November 25.1996 ,

a Trojan independent Spent Fuel Starace Installation Saferv Analysis Report

O List of Drawings Page1 ofI

]

i Drawing Title PGE-001-Sheet 1/1-Rev 0 PWR Basket Assembly >

PGE-002-Sheet 1/1- Rev 0 Concrete Cask Assembly PGE-003-Sheet 1/1- Rev 0 GTCC Basket Assembly PGE-004-Sheet 1/1- Rev 0 Transfer Cask Assembly PGE-005-Sheet 1/1- Rev 0 Transfer Cask Lifting Yoke PGE-006-Sheet 1/1- Rev 0 Basket Overpack Assembly O

November 25,1996

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

Tmian independent Spent Fuel Storare Installation Safety Analysis Report h LIST OF EFFECTIVE PAGES ,

INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT l

Page Number Revision Title Page March 26,1996 Table ofContents i-xi November 25,1996 l

List of Tables 1-3 November 25.1996 List of Figures 1 - 3 November 25,1996 List of Effective Pages 1-10 November 25,1996 List of Drawings Page 1 November 25,1996 1-1 and 1-2 July 15,1996 1-3 November 25,1996 1-4 through 1-6 July 15,1996 Figure 1,1-1 Original

Figure 1.1-2 Original

j i

Figure 1.3-1 Original

l 2-1 through 2-56 July 15,1996 Table 2.1-1 July 15,1996 Table 2.1-2 July 15,1996 i Table 2.2-1 July 15,1996 i

Table 2.7-1 July 15,1996 i ,

! Figure 2.1 1 Original *

March 26,1996 - page not November 25,1996 marked with date 1 ,

d j .

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

1 1

i i

Troian Independent Spent FuelStorare Installation Safety Analysis Report i

LIST OF EFFECTIVE PAGES INDEPENDENT SPENT FUEL STORAGE INSTALLATION '

i SAFETY ANALYSIS REPORT l Page Number Resision Figure 2.1-2 Original

Figure 2.1-3 Original

Figure 2.1-4 Original

i Figure 2.1-5 Original

j Figure 2.1-6 Original

l Figure 2.2-1 Original

l Figure 2.2-2 Original

Figure 2.3-1 Original

Figure 2.3-2 Original

Figure 2.3-3 Original * ,

Figure 2.3-4 Original

l Figure 2.3-5 Original

Figure 2.3-6 Original

Figure 2.4-1 Original

i l

~

Figure 2.4-2 Original

3-1 July 15,1996 3-2 November 25,1996 3-3 through 3-13 July 15,1996

March 26,1996 - page not November 25,1996 marked with date 2 :

)

c l

I Trojan Independent Spent Fuel Stora,ee Installation Saferv Analvsis Report ,

LIST OF EFFECTIVE PAGES l INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT i

Page Number Resision 3-14 November 25,1996 l '3-15 July 15,1996 3-16 through 3-19 November 25,1996

]'

3-20 through 3-25 July 15,1996 i Table 3.1-1 July 15,1996

! Table 3.1-2 July 15,1996 f' Table 3.1-3 July 15,1996 4

Table 3.1-4 July 15,1996 Table 3.2-1 July 15,1996 i Table 3.2-2 July 15,1996 i

Table 3.2-3 July 15,1996 Table 3.2-4 July 15,1996 Table 3.2-5 November 25,1996

. Table 3.6-1 July 15,1996 f 4-1 through 4-2 July 15,1996 4-3 November 25,1996 4-4 July 15,1996 4-5 through 4-6 November 25,1996

March 26,1996 - page not November 25,1996 marked with date 3

t Troian Independent Spent FuelStorage Installation Saferv Analvsis Report h LIST OF EFFECTIVE PAGES 1

INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Page Number Revision 4-7 through 4-9 July 15,1996 4-10 through 4-11 November 25,1996 4-12 through 4-31 July 15,1996 4-32 November 25,1996 4-33 through 4-57 July 15,1996 Table 4.2-1 July 15,1996 Table 4.2-1a November 25,1996 Table 4.2-2 July 15,1996 Table 4.2-2a July 15,1996 Table 4.2-3 July 15,1996 Table 4.2-4 July 15,1996 Table 4.2-5 July 15,1996 Table 4.2-6 July 15,1996 Table 4.2-7 July 15,1996 Table 4.2-8 July 15,1996 Table 4.2-9 July 15,1996 Table 4.2-10 July 15,1996 Table 4.2-11 July 15,1996

March 26,1996 - page not November 25,1996 marked with date 4

l 1

l Troian independent Spent FuelStorage installation Saferv Analysis Report i

LIST OF EFFECTIVE PAGES INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT l i

i

! Pane Number Revision

\

Table 4.2-12 July 15,1996 l

Table 4.2-13 July 15,1996 i Table 4.2-14 July 15,1996 Table 4.7-1 July 15,1996

! Figure 4.2-1 Original

Figure 4.2-2 Original

i Figure 4.2-3 Original * ,

Figure 4.2-4 Original

Figure 4.2-5 July 15,1996 Figure 4.2-6a November 25,1996 Figure 4.2-6b November 25,1996 Figure 4.2-7 Original

Figure 4.2-8 Original

Figure 4.2-9 Original *

, Figure 4.2-10 Original

l i Figure 4.2-11 Original

Figure 4.71 Original

Figure 4.7-2 Original

1

March 26,1996 - page not November 25,1996

[

marked with date 5 e

it

- - w - ,

b Tmian independent Spent FuelStorage installation Saferv A nalvsis Report i

, LIST OF EFFECTIVE PAGES l

INDEPENDENT SPENT FUEL STORAGE INSTALLATION I

SAFETY ANALYSIS REPORT I

! l l

Page Number Revision

! Figure 4.7-3 Original

Figure 4.7-4 Original

Figure 4.7-5 Original

l Figure 4.7-6 Original

l Figure 4.7-7 Original

l 5-1 through 5-3 November 25,1996 5-4 through 5-14 July 15,1996 l

Figure 5.1-1 Original

Figure 5.1-2 Original * ;

l' Figure 5.1-3 Original

6-1 through 6-2 July 15,1996 7-1 through 7-7 July 15,1996 7-8 November 25,1996 7-9 through 7-26 hily 15,1996 Table 7.2-1 July 15,1996 Table 7.2-2 July 15,1996 Table 7.2-3 July 15,1996 4

Table 7.2-4 July 15,1996

March 26,1996 - page not November 25,1996

i

.O marked with date 6 1

~ - , ,

i l

Trojan Independent Spent Fuel Storage Installation Saferv Analysis Report h LIST OF EFFECTIVE PAGES INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT 1

i Page Number Revision !

Table 7.2-5 July 15,1996 j Table 7.2-6 July 15,1996 l 1

Table 7.3-1 July 15,1996 '

Table 7.3-2 July 15,1996 i

Table 7.3-3 July 15,1996 Table 7.3-4 July 15,1996 Table 7.4-1 July 15,1996 Table 7.4-2 July 15,1996 ;

i Table 7.4-3 July 15,1996 l Table 7.4-4 July 15,1996 l Figure 7.2-1 Original

i Figure 7.3-1 Original

Figure 7.3-2 Original

Figure 7.3-3 Original

Figure 7.3 4 Original

Figure 7.3-5 Original

4 Figure 7.3-6 Original

Figure 7.3-7 Original *

March 26,1996 - page not November 25,1996 marked with date 7

Troian Independent Spent Fuel Storne Installation Safety Analysis Report LIST OF EFFECTIVE PAGES INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Paue Number Revision Figure 7.3-8 Original

Figure 7.3-9 Original

Figure 7.3-10 Original

Figure 7.3-11 Original

Figure 7.3-12 Original

I 8-1 through 8-56 July 15,1996 Table 8.0-1 July 15,1996 Table 8.0-2 July 15,1996 Table 8.1-1 July 15,1996 Table 8.1-2 July 15,1996 Table 8.1-3 July 15,1996 Table 8.2-1 July 15,1996 Table 8.2-2 July 15,1996 Table 8.2-3 July 15,1996 Table 8.3-1 July 15,1996 I

Figure 8.1-1 Original

Figure 8.1-2 Original

Figure 8.2-1 Original

November 25,1996

March 26,1996 - page not O marked with date 8

Troian independent Spent FuelStorage Installation Saferv A nalysis Report LIST OF EFFECTIVE PAGES INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT Page Number Revision Figure 8.2-2 Original

Figure 8.2-3 Original

Figure 8.2-4 Original

9-1 through 9-9 July 15,1996 9-10 November 25,1996 9-11 through 9-17 July 15,1996 Table 9.1-11 July 15,1996 Table 9.2-1, pages 1 and 2 November 25,1996 Table 9-2.1, page 3 July 15,1996 Table 9.6-1 July 15,1996 Figure 9.1-1 Original

Figure 9.1-2 Original

10-1 July 15,1996 11-1 July 15,1996 Drawings:

PGE-001-Sheet 1/1 PWR Basket 0 Assembly PGE-002-Sheet Concrete Cask 0 Assembly

March 26,1996 - page not November 25,1996 O marked with date 9 l

l

4 i !

9 0

' Troian independent Spent FuelStorage Installation Saferv A nalysis Report ,

a i

LIST OF EFFECTIVE PAGES l INDEPENDENT SPENT FUEL STORAGE INSTALLATION SAFETY ANALYSIS REPORT s :

Pane Number Revision PGE-003-Sheet 1/1 GTCC Basket 0 Assembly PGE-004-Sheet 1/ITransfer Cask 0 j Assembly PGE-005-Sheet 1/ITransfer Cask 0

- Lifling Yoke i PGE-006-Sheet 1/1 Basket 0 l Overpack Assembly l

l i

i November 25,1996 p

March 26,1996 - page not 10 v marked with date

i i

i ,

Tmian independent Spent Fuel Storare installation Saferv Analysis Report ,

!O used during storage and transfer of spent fuel and GTCC waste. Figure 1.3-1 provides an 1

overview et he basket and Concrete Cask. {

4 I

i ,

L 1.3.1 STORAGE SYSTEM BASKETS

. The Trojan ISFSI storage system utilizes two types of baskets, the PWR Basket and the GTCC j Basket. The baskets are metal containers that are seal-welded closed. Both baskets serve as a '

I confinement boundary for the materials stored within the baskets.

i l j

i The PWR Basket is a fuel storage canister designed to provide safe storage ofintact spent fuel, failed fuel and fuel debris. The PWR Basket consists of an internal sleeve assembly, an outer shell assembly, a shield lid and a structural lid. The internal sleeve assembly is fabricated from ;

high strength steel plates formed into an array of 24 square storage sleeves, each holding one PWR spent fuel assembly. The cells aie sized to accommodate storage of control components within the fuel assembly. The PWR Basket relies only on geometry for subcriticality during storage. !

Assemblies containing damaged fuel, process can capsules containing fuel debris, fuel assembly I hardware, process cans containing fuel assembly hardware, and a fuel rod storage container are l placed in a failed fuel can in the PWR Basket. Fuel debris is placed in process cans, which are l ,

i placed in a process can capsule, prior to placement in the failed fuel can in the PWR Basket. The I four peripheral cells in each PWR Basket can accommodate failed fuel cans as well as spent fuel l l' assemblies.

l I

The GTCC Basket is designed to provide safe storage of GTCC waste. GTCC waste is placed in canisters and then placed into the GTCC Basket. The GTCC Basket does not contain an internal sleeve assembly. The GTCC Basket accommodates 28 individual canisters designed for GTCC waste.

A Basket Overpack is provided to be used in the unlikely event of a leaking basket.

1.3.2 STORAGE SYSTEM CONCRETE CASK The Concrete Cask provides structural support, shielding, and natural circulation cooling for the basket. The basket is stored in the central steellined cavity of the Concrete Cask. The Concrete 1-3 November 25,1996 1

b Trolan Independent Spent FuelStorare Installation Safety Analvsis Rennrt h

(

,7 b sheath. RCCAs consist of 24 absorber rods which can be insened into the thimble guides. A total of 61 RCCAs will be stored in the ISFSI. BPRAs are similar to RCCAs but consist of fewer absorber rods (9 to 20). Thimble plugs were used to " plug" thimble guides which did not contain absorber rods or sources during reactor operation. Sources are similar in shape to absorber rods but a portion of the length contains a secondary neutron source. The primary design concern associated with these components is weight. Table 3.1-4 summarizes the physical characteristics of the inserts.

The main physical parameters of concern are the fuel assembly dimensions and weight, and envelope (cross-sectional dimension). These parameters establish the mechanical and stmetural design aspects of the Concrete Cask and basket. The thermal and radiological characteristics establish the shielding and thermal aspects of the design.

3.1.1.2 Failed and Partial Fuel Assemblies Ten (10) partial fuel assemblies and one (1) fuel rod storage container, which contain intact, l suspect, or failed fuel rods, will require storage.

p

( 1 3.1.1.3 Fuel Debris Fuel debris consists ofloose fuel pellets, fuel pellet fragments, and fuel assembly fragments (ponions of fuel rods, portions of grid assemblies, etc.). The quantity of fissile material contained as fuel debris will not exceed 10 kg per basket. This limit is imposed to satisfy the license conditions of the TranStor Shipping Cask (Reference 1). An additional limit for fuel debris of no more than 20 curies of plutonium is imposed to meet the offsite transportation requirements of 10 CFR 71.63.

3.1.1.4 GTCC Waste GTCC waste consists of activated core components consisting mainly of segmented reactor internals. GTCC waste characteristics such as weight and curie content are addressed in Table 3.1-2. GTCC waste is not stored in the same basket with spent fuel.

O 3-2 November 25,1996 N)

Troian Independent Spent FuelStorare Installation Safety Analysis Report ,

O 3.2.5.3 Load Combinations and Design Streneth - Transfer Cask The Transfer Cask is a special lining device designed and fabricated to the requirements of ANSI 14.6 (1993) and NUREG 0612 (1980). The criteria for its load-bearing components are:

Maximum principal stress during the lia (with 10% dynamic load factor) will be less than S3 /3 or Sj5.

Load bearing members of the Transfer Cask shall be subject to drop weight test (ASTM E208) or charpy impact test (ASTM A370) per ANSI 14.6 (1993) paragraph 4.2.6. ;

3.2.5.4 Load Combinations and Design Strength - Failed Fuel Can The Failed Fuel Can is designed to be placed into one of the four corner storage locations of the

/ PWR Basket. The Failed Fuel Can is designed to allow for water draining and vacuum drying during PWR Basket closure. Once placed into its storage location, the Failed Fuel Can is not subjected to externalloadings applicable to ASME Senice Level A (normal). Specific structural design criteria and load combinations are not applicable.

1 3.2.5.5 Load Combinations and Design Strengths - Fuel Debris Process Can Caosule l The Fuel Debris Process Can Capsule material and welds are selected based on ASME Section l III, Division I, Subsection NG (1992). The Fuel Debris Process Can Capsule is stmeturally l analyzed for external pressure, internal pressure, dead weight, thermal stresses, and drops. The l stresses calculated by classical equations are less than the allowable stresses provided in ASME, l Section 111, Division I, Subsection NG (1992) for senice levels A and D. l 3 - 14 November 25,1996

1 i

Troian Independent Spent Fuel Storage Installation Saferv Analysis Report .

o, N ,)

2. To provide adequate heat transfer so that the fuel clad temperature does not exceed allowables under design conditions.

l The primary functions of the Concrete Cask are:

1. To protect the basket from weather and postulated environmental events such as earthquakes and tornado missiles,

2. To provide adequate heat transfer for the PWR Basket and GTCC Basket, and

3. To provide adequate shielding (together with a PWR Basket or GTCC Basket) to meet 10 CFR 72 requirements.

The primary functions of the Transfer Cask are:

1. To serve as a special lifting device meeting the requirements of NUREG-0612 (1980)/ ANSI 14.6 (1993) for movement of a PWR Basket or GTCC Basket, and

2. To provide radiation shielding to minimize exposure rates during transfer

() operations.

The primary function of the Transfer Station is to prevent the Transfer Cask from falling or overturning during Basket transfer operations.

The primary function of the Failed Fuel Can is to provide a containment boundary for failed fuel l such that the failed fuel will be constrained within its PWR Basket storage location. The primary l function of the Fuel Debris Process Can Capsule is to provide a containment boundary for fuel 1 debris. Constraining failed fuel and fuel debris to fixed storage locations is required to maintain l i I

the assumptions in the criticality analysis and heat transfer modeling.

As discussed in the following sub-sections, the ISFSI design incorporates features addressmg each of the above design considerations to assure safe operation during fuel loading, storage system handling, and storage.

I 3 -16 November 25,1996 (A)

?

0, Trojan Independent Spent Fuel Storace Installation Safety A nalvsis Report p

4 i G/

3.3.2 PROTECTION BY MULTIPLE CONFINEMENT BARRIERS AND SYSTEMS 3.3.2.1 Confmement Baniers and Systems Oregon Administrative Rule (OAR) 345-26-0390 prohibits the storage of spent nuclear fuel or radioactive materials other than that generated or used in the operation of the Trojan Nuclear Plant. Spent nuclear fuel and fuel related material will be confined within PWR Baskets. GTCC waste will be confined within GTCC Baskets.

The PWR Basket and GTCC Basket are designed to provide a confinement barrier for spent nuclear fuel and GTCC waste in accordance with the general design criteria requirements of 10 CFR 72, Subpart F. Each basket is a stainless steel seal welded enclosure. The basket shield lid and stn2cturallid closures are accomplished by multi-pass welding. The PWR Basket and GTCC Basket confinement barriers are designed in accordance with ASME,Section III, Subsection NC (1992).

(3 The PWR Basket internals, which are used to constrain fuel assemblies and Failed Fuel Cans l V during storage, are designed in accordance with ASME,Section III, Subsection NG (1992). The l PWR Basket internals provide 24 storage locations. The four (4) corner locations are designed slightly larger to accommodate a Failed Fuel Can. l The Fuel Debris Process Can Capsule provides a containment boundary for fuel debris within the l PWR Basket. It is designed using the guidance in ASME, Section Ill, Subsection NG (1992) l (see Section 3.2.5.5). l The Failed Fuel Cans and GTCC Cans do not provide a confinement boundary and are considered to function as part of the basket internals. The Failed Fuel Can is designed in accordance with applicable portions of ASME,Section III, Subsection NG (1992). The GTCC Can is designed in accordance with applicable ponions of ASME, Section 111, Subsection NF (1992).

In the unlikely event of a PWR Basket or GTCC Basket confinement boundary failure, the affected basket may either be repaired or sealed within a Basket Overpack. The design criteria

) 3-17 November 25,1996

0, Troian independent Spent Fuel Storage Installation Saferv A nalysis Report C/

for the Basket Overpack are the same as those specified for the PWR Basket and GTCC Basket confinement boundary.

The PWR Basket must be designed to withstand credible drop accidents without damaging the stored fuel (i.e., the storage cells do not deform such that they bind the fuel). The PWR Basket must also be designed to provide confinement in the event of a fuel clad failure.

3.3.2.2 Ventilatiori-OfTgas The ISFSI is designed to confine radioactive n.aterials within a sealed enclosure for the life of the facility. There are no radioactive releases Juring normal operations or credible accidents. In the unlikely event a leaky basket must be placed in a Basket Overpack, evacuation of the Basket Overpack and backfilling with helium would be required. The operation is discussed in Chapter 5. A suitable filtration system such as a high efficiency particulate air (HEPA) filter would be used for the vacuum system vent path during this evolution.

,q 3.3.3 PROTECTION BY EQUIPMENT AND INSTRUMENTATION SELECTION L.) ,

3.3.3.1 Eauinment 1

The equipment / components that have been identified as imponant to safety for the ISFSI are:

1. Concrete Cask,

2. PWR Basket,

3. GTCC Basket,

4. Basket Ovegack,

5. Fuel Debris Process Can Capsule, l

6. Failed Fuel Can,

7. Transfer Cask, and

8. Transfer Station

/D 3-18 November 25,1996 N.l

4 l Trojan Independent Spent Fuel Storare Installation Safety A nalysis Report h i O The design criteria for the PWR Basket, GTCC Basket, and Basket Overpack are sununarized in

Table 3.2-5. The design criteria for the Concrete Cask are summarized in Table 3.6-1.

]

! 3.3.3.2 Instrumentation l

A temperature monitoring device is provided for each of the air outlet vents per storage cask (four per cask). The temperature monitoring devices are commercial grade. Additional discussion of temperature monitoring is provided in Section 5.1.3.4 and Section 5.4.1.

! 3.3.4 NUCLEAR CRITICALITY SAFETY ]

l l

i The storage system is designed to maint tin suberitical conditions (Kg s 0.95) under normal

! handling and storage conditions, off-normal handling and component functioning, and

, hypothetical accident conditions.

4 3.3.4.1 Control Methods for Prevention of Criticality 4

O Subcritical conditions are to be maintained by PWR Basket internal geometry. The PWR Basket internals will establish fuel assembly spacing. The design will assume a fuel assembly enrichment equal to or greater than the maximum initial fuel assembly enrichment that will be stored (3.56 wt% U235). No credit will be taken for burnup or fbel assembly control inserts.

Although neutron absorbing material is incorporated into the PWR Basket internals design, it is not credited in the criticality analysis for dry storage conditions.

Table 3.1-1 lists the fuel characteristics. Loose pellets and fuel debris are placed in Fuel Debris l Process Cans which are placed in a Fuel Debris Process Can Capsule. Administrative controls l limit the amount of fuel debris which can be placed within a basket. l There are no criticality control requirements for the GTCC Baskets.

3 - 19 November 25,1996

l Trojan imferendent Spent Fuel Storaer Installation Saferv Analysis Repon h n

(] Table 3,2-5 l

PWR Basket, GTCC Basket, and Basket Overpack Design Criteria Comnonent (Applicable Code or Criteria) Criteria l PWR Basket, GTCC Basket, or Basket Overpack P, < 1.0 S, Normal Operation - Senice Level A Pt + Pb< 1.5 S, ;

ASME Section III, Subsection NC (shell) P + Q < 3.0 S, l ASME Section III, Subsection NG (internals)

PWR Basket, GTCC Basket, or Basket Overpack P, < l.2S,(shell), < 1.5S,(cells)

Off-Normal Operation - Senice Level C Pt+ Po < l.8S,(shell), < 2.25S,(cells)

ASME III, Subsection NC (shell)

ASME III, Subsection NG (internals)

PWR Basket, GTCC Basket, or Basket Overpack P, < 2.4 S, or 0.7 S o !

Accident Condition - Service Level D (whichever is less)

ASME III, Subsection NC (shell) Pt + P b< 3.6 S, or 1.0 S, ASME III, Subsection NG (internals) (whichever is less)

(\s) I 1 I

f-~3 November 25,1996 (v )

l Troian inci, penclent Spent FuelStarare Installarmn Saferv Analysis Report '@

i t l

'O' 4.2.1 STRUCTURAL SPECIFICATION I

The design criteria of the sto: age structures and components account for both normal and off-normal conditions, including a range of credible and postulated accidents. The principal design criteria for the ISFSI is in accordance with Title 10, Code of Federal Regulations, Part 72 l (10 CFR 72), and ANSI /ANS 57.9. The design criteria for the ISFSI is presented in Chapter 3. ;

The design codes for the major ISFSI storage stmetures and components are summarized in the ;

following table. I Component Governing Code / Standard i l

Storage Pad ACI 318 (1983)

PWR Basket Confinement boundary - AShiE, Section Ill, Subsection NC Internal assembly - ash 1E, Section Ill, Subsection NG GTCC Basket Confinement boundary - AShiE,Section III, Subsection NC l Basket Overpack ash 1E,Section III, Subsection NC ;

(3 1 V Fuel Debris Process ash 1E, Section 111, Subsection NG (used as guidance - see Section l Can Capsule 3.2.5.5) l Failed Fuel Can AShiE, Section 111, Subsection NG Concrete Cask ACI 349 and ANSI 57.9 Section 3.4 provides the criteria used to classify structures, systems and components, imponant to safety.

The ISFSI Storage Pad meets the requirements of ACI 318 and is capable of supporting the loads associated with the array of Concrete Casks and transfer equipment. The ISFSI Storage Pad is not classified as important to safety. Its function is to provide a slab-on-grade supporting surface for the Concrete Casks, Transfer Station and shipping cask. It also provides a smooth level surface to allow operation of the air pad system.

3 Applicable revision of governing code / standard is provided in Chapter 3.

,o 4-3 November 25,1996

Trolan independent Spent Fuel Storare Installation Saferv A nalysis Repon O

~

Basket Overpack. Section 3.3.2 discusses the design criteria applicable to these ISFSI i components.

The cladding ofintact fuel assemblies provides an additional confinement boundary. The Fuel

- Debris Process Can Capsule provides a containment boundary for fuel fragments and is stored l within a PWR Basket. The Failed Fuel Cans provide a containment boundary for failed fuel

assemblies to constrain these assemblies and associated components within its PWR Basket

{ storage location. Constraining this material to fixed storage locations is required to maintain the assumptions in the criticality analysis and heat transfer modeling.

i The design requirements for confinement barriers and systems are further discussed in Section 3.3.2.

i' 4.2.4 INDIVIDUAL UNIT DESCRIPTION The ISFSI is comprised of up to 36 individual storage systems. Each storage system consists of i

' O d

a Concrete Cask containing either a PWR Basket or GTCC Basket. In the unlikely event a PWR Basket or GTCC Basket fails to maintain a confinement boundary and can not be repaired, a i Basket Overpack is available. The Concrete Casks are arranged on the Storage Pad as discussed in Section 4.2.2.1.

4.2.4.1 Functional Description i

The primary functions of the ISFSI storage system components are discussed in Section 3.3.1.

4.2.4.2 Component Descriotions 4.2.4.2.1 Description of the PWR Basket The PWR Basket is a transportable cylindrical container consisting of an outer shell assembly, a shield lid, a structural lid, and an internal basket assembly. The basket shell provides the confinement boundary and is designed to withstand credible accidents without loss ofintegrity.

4-5 November 25,1996

Troian independent Spent Fuel Storare Installation Saferv Analysis Report .

U,o The shell exterior is coated with a gloss epoxy coating for ease of decontamination following loading operations.

The PWR Dasket internal assembly is fabricated from steel plates formed into an array of 24 square storage cells. Four (4) of the outer corner cells are slightly larger to allow accommodation of a Failed Fuel Can. Intact fuel assemblies, with or without inserts, may be l stored in any of the storage locations. The internal assembly uses structural tubes to provide support for the storage cells during a postulated drop accident. Neutron absorbing poison sheets are also used in the constmetion of the PWR Basket internal assembly, however they are not credited in the criticality analysis for dry storage conditions.

Section 5.1.1 discusses the operations associated with basket loading and installation of the shield lid and structural lid. The steel shield lid contains two penetrations to allow for vacuum drying and helium backfilling of the basket internal atmosphere. Prior to lowering the shield lid onto the basket after loading is complete, a pipe is threaded into one of the two penetrations.

When the shield lid is in place, the pipe length is such that it extends to the bottom of the basket to facilitate water removal. Upon completion of water removal, a pipe plug is threaded into the drain pipe penetration. The other penetration utilizes a quick disconnect fitting to allow q connection to a vacuum drying and helium backfilling system. The shield lid is seal welded to

) the basket shell. The first and final shield lid weld passes are inspected by dye penetrant testing.

The basket is then hydrostatically tested at approximately 7.3 psig.

Following completion of hydrostatic testing, a steel structural lid containing a penetration allowing access to the shield lid penetrations is placed on top of the shield lid and seal welded to the basket shell and to the shield lid (where exposed by the structural lid penetration). The first and final weld passes are dye penetrant checked.

Upon completion of the vacuum drying and helium inerting of the internal basket atmosphere, the shield lid and structural lid penetrations must be sealed. The quick disconnect fitting is relied upon to maintain the helium atmosphere until the penetration closure plates are installed. The shield lid penetrations are isolated by two steel plates inserted into the structural lid access penetration. The steel plates are inserted individually and seal welded to the sides of the structurallid penettation. The tirst and final weld passes for each of these closure plates are dye penetrant checked.

4-6 November 25,1996

(

i Troian Independent Spent FuelStorare Installatwn Saferv Analvsn Report g

t ,

An air flow path is formed by the openings at the bottom (air entrance), the air inlet ducts, the gap between the basket exterior and the Concrete Cask interior, and the air outlet ducts at the top. The air inlet and outlet vents are steel-lined penetrations that take non-planar paths to minimize radiation streaming. A shield ring is provided over the basket-liner annulus to reduce the dose rate at the top of the cask.

The cask lid is fabricated from a steel plate which provides additional shielding to reduce the skyshine radiation. The cask lid also provides a cover and seal to protect the basket from the environment and postulated tornado missiles. The lid is bolted in place and is provided with a locking wire with a lead seal.

l The bottom of the Concrete Cask is covered with a steel plate which minimizes loss of cask concrete during a bottom drop accident. The Concrete Cask has reinforced chamfered corners at the top and bottom to minimize damage during handhng. i 1

l The cask is constructed by pouring concrete between a re-usable form and the inner metal liner.

The reinforcing bars and air flow embedments are installed and tied prior to pouring.

O 1 A summary of fabrication requirements is presented in Table 4.2-2. Figure 4.2-4 provides a description of the Concrete Cad 4.2.4.2.5 Failed Fuel Can The Failed Fuel Can is designed to contain partial or complete fuel assemblies with failed or suspect rods. The internal square opening accommodates a fuel assembly without inserts. The l Failed Fuel Can will also be used to store a fuel rod storage container, Fuel Debris Process Can l Capsules, fuel assembly hardware (non-fuel bearing components), and Fuel Debris Process Cans l that contain fuel assembly hardware (non-fuel bearing components). The outside dimensions l allow the Failed Fuel Can to fit in one of the four oversized storage locations within a PWR Basket.

The shell of the Failed Fuel Can is fabricated from carbon steel. Near the bottom of each side of the shell assembly are two screened vent holes. These vent holes enable vacuum drying of the canister. The vent holes also expose the contents of the Failed Fuel Can to the helium atmosphere of the PWR Basket.

4-10 November 25,1996

1 Troian Independent Spent FuelStorare Installation Safety Analysis Report h The lid is bolted in place and is designed to be lified using a fuel handling tool. The lid bottom also has vent holes to facilitate draining.

l i Carbon steel components of the Failed Fuel Can are coated with radiation resistant, high

temperature, hard surface inorganic zinc coating. Figure 4.2-5 provides a description of the 4

Failed Fuel Can.

~

s 4 2.4.2.6 Description of Fuel Debris Process Can and Capsule l l

! \

4 1

The process can, shown in Figure 4.2-6a, is the container used to process the organic media and I

! fuel debris located in the Spent Fuel Pool. The process can is constructed of 300 series stainless 'l

, steel for corrosion resistance. The process can has 5 micron metallic filters in both the can l

bottom and lid. These filters allow removal of water and organic media by high temperature l l steam, while retaining the solid residue from the processed media and fuel debris inside the l

! process can. l l 1 After high temperature steam processing, up to five (5) process cans are placed inside the l l process can capsule shown in Figure 4.2-6b. The process can capsule is constructed of 304 l 4

stainless steel for corrosion resistance and is inerted with helium. The process can capsule l provides a sealed containment for the fuel debris. The process can capsule is designed to be l lifted by normal fuel handling tools. l The process cans may also be used to store fuel assembly hardware (non-fuel bearing l b components). These process cans will be not placed in a process can capsule, but will be directly l placed inside a failed fuel can. The process cans containing fuel assembly hardware (non-fuel l bearing components) will not be processed by high temperature steam because there will be no l organic media to remove. Water will be removed from the process can through the metallic l filters during the basket vacuum drying process. l 4.2.4.2.7 GTCC Can The GTCC Can is designed to contain GTCC waste for placement within a GTCC Basket. Up to 29,000 lbs. contained within 28 GTCC Cans can be placed in a GTCC Basket.

The shell of the canister is fabricated from steel. Two vent holes are located near the bottom of each side of the shell assembly and on the bottom plate allowing draining and vacuum drying of the container.

O 4-11 November 25,1996 1

Trojan Independent Spent FuelStorage Installatwn Safety Analvsis Report '

4

(

4.2.6.8 Maximum Internal Pressure The basket is backfilled with helium so that at the conditions present during normal operations the internal pressure is at approximately atmospheric pressure. The pressure calculated for different ambient conditions is presented in Table 4.2-9 and stresses are included in the structural analysis in Section 4.2.5. The worst case internal pressure occurs during a postulated accident where fuel rods inside the basket are breached and release their fission gases. This case and the resulting pressure and stresses are described in Section 8.2.6.

4.2.6.9 Evaluation of Cask Lifetime Performance under Normal Conditions of Storage As shown in the preceding sections, the storage system operates within the thermal design limits.

Therefore, no degradation due to temperature effects on materials or components is expected during the lifetime of the cask.

4.

2.7 CRITICALITY EVALUATION

p C The criticality evaluation was performed using the KENO-Va module of the SCALE-4.1 code package (Reference 9).The model analysis was based on Westinghouse 17x17 standard fuel.

The ISFSI storage system will also contain B&W 17x17 fuel which is considered to be bounded by the Westinghouse fuel. The only significant difference in these two types of fuel assemblies is the B&W assembly has a slightly smaller fuel pellet diameter. The smaller pellet size makes the B&W assembly slightly less reactive and should therefore be bounded by Westinghouse analysis.

1 The four corner cells of a PWR Basket may contain a Fuel Debris Process Can Capsule inside a l failed fuel can. A fuel debris mass limit of 10 kg per PWR Basket will be administratively l controlled. A limit of 10 kg of fuel debris is significantly less than the fuel mass of an intact fuel assembly (-460 kg uranium). The 10 kg of fuel mass will not be nearly as reactive as an intact fuel assembly no matter how the fuel debris is arranged within the Fuel Debris Process Can l Capsule (s). The 10 kg of fuel mass will not cause thermal, structural or shielding problems no I matter how it is distributed within the Fuel Debris Process Can Capsule (s). l The parameters of concern for criticality evaluations are initial enrichment, burnup, moderation, poisons, and geometry. These parameters combined produce the reactivity of the system which

] 4-32 November 25,1996

O v 1mjan independent Spent FuelStomge Installation Safety Analysis Report h

Table 4.2-la ASME Code Deviations ASME Section III Subsection NC ,

Requ;.m.a..t Exceptiort' Justification Section 3211.1 Establishes scope for vessel design Requirement of NC-5250 (referenced in this subparagraph) to radiograph all Category C joints is not met for the basket closure urkts. Radiographic examination of these field welds is not feasible; redundant feakage berners used in lieu of this ,

requirement. Lugs and comer support tubes are not attached to the shell with continuous welds (per NC-4267) due to lack of accessibihty. Lugs are not loaded wtiile the shell is serving as a pressure boundart Detailed drop analysis includes the actual wrid configuration and considers influence on the shell as applicable. Set ofcalculations, drawings, and specifications is used in lieu of the Design Rerert per Appendix C. Ilowever. the information is omvided and the intent of the Code is met 3223.2 Requires Design Report in Appendix C A set of calculations, drawings, and specificatons is used in lieu of the Design Report per Appendix C. Iloweser, the format information is omvided and the intent of the Code is met Derribes permissible types of wrlded inints As stated above. basket elmure welds (Category C) can not be radicarsched per NC-2553 3252 3254 Refers to NC-4267 for structural attachment Lugs and comer support tubes are not attached to the shell with continuous wekts (per NC-4267) due to lack of accessibihty.

welds. Lugs are not loaded while the shell is serving as a pressure boundary. Detailed drop analysis includes the actual wrld configuration and considers influence on the shell as applicaNe.

Requirements for category D weld joints in Valve covers (Cat. D welds) are fillet welds and do not meet figure 4266. Small diameter and wrld size provide large factors of 4266 vessels designed to NC-3200 safety. Lid reinforcement is provided and redundant valve covers are installed for the ITVR basket (see NC-32521 4267 Types of attachment welds allowed in vessels Lugs and corner support tubes are not attached to the shell with continuous welds due to lack of accessibihty. Lugs are not '

designed to NC-3200 loaded while the shell is sersing as a pressure boundary. Detailed drop analysis includes the actual wekt configuration and !

considers influence on the shell as applicaNe [see NC-3211.11 Category C wekled joints for vessels Structural tid welds are not radiographed due to limited accessibihty and redundant PWR basket closure. Welds are tested using 5253 I designed to NO3200 require RT liquid penetrant or magnetic particle examination and helium leak testing.

examinction. [See NC-3211.1]

6113 Requires pressure test in presence of The inspection is performed but not by a Code certified inspector. Acceptable because the vessel is not inspector. N-stamped 8100 References NCA-8000 for narneplate, Code Stamping is not provxicd. The basket is not a part of a nuclear power system as defined by stamping, and n iw1 require 6 ents.

NCA-1110.

ASME Section 111 Subsection NG Requ;.m..m.t Exceptiort rJustification Section The basket intemals and failed fuel can do not use SA materials. Materials used meet ASTM specifications tiint are identical to l 2121 Requires use of ASME(SA) materials.

w..wmding ASME specifications. All materials are supplied as important to safety, CMTRs are provided 4110 General requirements specify use of materials NG-2000 is not entirely rnet (see NG-212I).

per NG-2000 Means ofcertification ormatenals. ASTM materials are utili7ed for NG components (see NG-2120 4121 ft100 References NCA-8000 for nameplate, Code Stamping is not prmided. The basket is not a part of a nuclear power system as defined by stamping. and mut requirements. NCA-1110.

November 25,19%

e. _. . - _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ _ _ _ _ - _ - _ _ .__ . - _ _ - . . _ _ _ - _ _ _ - _ - _ _ .

07.50 e

/ \

/ i N i p l \

g V L '. - ...___ e J.' w"J .

\ I i

\ /

\ / ,

N /

N % #

. e .

WWA.A M?-A4

4) l i

l W

)

b .L

. \

l a

'! l e i 4 l e ,

l l FILTER MEDIAS INCONEL j -

i i

r y ~ 'n l -----_: _

f! \

e

.O t29.0')

i l

l

\ /

\

fi FILTER ME0!As INCC NEL i

i e

a

,1 I

l l CETAIL I f i l l

_ - --- .c

./

i! .

l k) irp

(___._..___ s l/ f O TROJAN ISF.9! I V SAFETY ANALYSIS RERRT FIGURE 4.2-6a f FUEL DEBRIS PROCESS CAN '

_ _ . _ _ _ . - _ . _ _ . _ _ _ _ . _ . .. _m.. . _ _ _ _ _ _ .

4-

.gj.

PENT FUEL TOOL *

, m INTERFACE f.

08.50' MAX O.&

08.00 I.D.

--j I I WELD JOINT SHIELD PLUG

, WM5 TO SHELL 4-PROCESS CAN jMy '

OTY 5 l

SHIELD PLUG. STAINLESS STEEL WITH SCHRACER PURGE AND BACKFILL PORT VALVE a.1-16o"

.m.

J t l

l rTa l

< n.

i j TROJAN ISFSI j SAFETY ANALYSIS REPORT l

FIGUAE 4.2-6b FUEL DEBRIS PROCESS CAN CAPSULE

Troian Independent Svent FuelStorare Installation Safety Analysis Report '

O 5.0 OPERATIONS 5.1 GENERAL DESCRIPTION The methods and sequences described below define the operational controls which personnel performing spent fuel loading and storage activities will implement to assure that operations utilize the passive safety features of the Trojan ISFSI design described in Chapter 4. Fuel loading and basket sealing operations (including non-destructive examination and pressure testing) will be performed within the Fuel Building in order to utilize the existing systems and equipment for heavy lifts, radiation monitoring and controls, decontamination and any necessary ;

auxiliary support (i.e., electrical, crane, senice air, etc.). Fuel handling and cask loading operations in the Fuel Building will be performed in accordance with Portland General Electric Company's 10 CFR 50 license for the Trojan Plant. Storage at the ISFSI will be subject to the l requirements of the ISFSI license issued in accordance with 10 CFR 72. Once the loaded I storage cask is placed on air pallets in the Fuel Building Bay and moved to the ISFSI concrete slab area, operational activities are essentially limited to monitoring proper decay heat removal.

5.1.1 OPERATION DESCRIPTION O

The following sections describe the spent fuel handling, basket sealing, and cask loading l activities relevant to the operation of the Trojan ISFSI. As previously described in Chapter 3, the Trojan ISFSI will contain intact and failed spent nuclear fuel assemblies, fuel debris, and GTCC waste. The PWR Basket is vertically loaded with fuel assemblies and/or special canisters l i designed to hold failed fuel or fuel debris. The GTCC Basket is loaded with a special canister l containing segmented GTCC waste. Section 5.1.1.1 describes the operational controls for l loading the individual canisters. Section 5.1.1.2 describes operational controls for loading the individual basket.

l i

Specific procedures will define and control classification criteria, loading sequence and individual basket / cask inventory. Fuel / debris /GTCC waste will be visually inspected as it is loaded to verify that each assembly / item conforms to the established classification criteria. As a minimum, item identification and/or serial numbers will be verified and recorded. Fuel loading operations will be videotaped to visually record fuel assembly serial numbers and to provide an independent record ofloaded inventory. Fuel will be examined to verify that pellets are stmeturally contained within the cladding or it will be placed in a Failed Fuel Can. Additional l 5-1 November 25,1996

Trolan independent svent FuelStorage in.stallation Safety Analysis Report .

O procedures will control placement and use ofimpact limiters, allowable travel path inside the l Fuel Building, and limit lifting heights to assure compliance with bounding analysis.

5,1,1,1 Failed Fuel. GTCC Waste. and Fuel Debris Process Can Caosule Loading l Special containers are used to segregate failed fuel, fuel debris, and GTCC waste within the confines of the PWR Basket or GTCC Basket. The individualized containers provide containment propenies by constraining the material to fixed storage locations which maintains the assumptions in the criticality analysis and heat transfer modeling.

Failed fuel is contained in special cans designed to fit in one of four oversized peripheral storage sleeves of the basket. Failed or suspect fuel that cannot structurally contain pellets within the cladding will be placed in a Failed Fuel Can. Because the cans are open to the internal basket atmosphere, they are vacuum dried and backfilled at the same time as other basket contents.

Fuel debris is contained within a specially designed and fabricated canister (process can capsule) l O which has been sealed as part of the fuel debris processing project before being loaded in a basket. The process can capsule is placed in a failed fuel can, which has been placed in any one of the four oversized storage cells in the basket.

l l

l 5-2 November 25,1996

)

l l s Troian Independent Spent Fuel Storave Installation Safety Analysis Report

()

Cl l

The GTCC Can effectively contains GTCC waste currently stored at the Trojan Nuclear Plant .

The cans are filled and placed in the GTCC Basket utilizing a 28 slot alignment grating which is removed afler the GTCC Basket is filled. The cans are open to the GTCC Basket atmosphere to allow drying.

5.1.1.2 Basket Loading and Sealing Ooerations This section describes the sequence of operations and controls necessary to load, seal, and test a PWR Basket or GTCC Basket in the Fuel Building and to control transfer operations to the ISFSI storage pad. The major components described in Chapter 4 are further defined with design and operating characteristics. Test and/or inspection methods demonstrate compliance with design requirements.

The basket and Transfer Cask are brought into the Fuel Building through the crane bay door.

After examination and any needed cleaning, the Transfer Cask is moved by use of the Fuel Building overhead crane (independent dual hook design) and Transfer Cask Lifting Yoke to the cask wash pit area. There, a protective bottom cover (e g., plastic sheeting, plexiglas, plywood,

(]

etc.) is installed on the Transfer Cask lower hydraulic door carrier rails, which will prevent possible contamination from the cask loading pit floor from being imbedded in the cask. The basket is then moved by the same crane and placed into the Transfer Cask. After installation of radiation shielding shims in the gap between the Transfer Cask and basket, a basket retaining ring is bolted to the Transfer Cask wall. The basket is then filled with borated water. The water is filtered, if necessary, to reduce the potential for contamination on the exterior of the basket.

This filling may be done in the cask wash pit area or at the cask loading pit before submergence.

The Transfer Cask (with basket) is then moved by the Fuel Building overhead crane and suspended over the cask loading pit immediately adjacent to the spent fuel pool. Borated water is continuously flushed through the basket / Transfer Cask gap to minimize unnecessary contamination of the basket external surface while the Transfer Cask is in the cask loading pit. ,

After the Transfer Cask is lowered to the pool bottom, the specified basket contents are loaded. l Operations will be conducted in accordance with approved Trojan Nuclear Plant fuel handling l procedures. I i

l l

l

,m 5-3 November 25,1996 f

U)

(

/

O Trojan independent Spent Fuel Storage Installation Saferv A nalnis Report h' fv) contents of the core bafile and bafile former plates are the most limiting. The GTCC waste i gamma source activities are listed in Table 7.2-6.

7.2.1.5 Fuel Debris The PGE fuel debris consists ofindividual fuel pellets and fragments from damaged fuel rods.

For the shielding analysis, fuel debris source terms are conservatively assumed to be the same as for intact fuel. This assumption is conservative because the fuel debris will be stored in fuel debris process can capsules, separate from intact fuel, and the total quantity of fuel debris is only l a few kilograms, as compared to an intact fuel assembly with several hundred kilograms of fuel l material.

7.2.1.6 Non-Fuel Bearing Components In addition to failed fuel, fuel assembly hardware, non-fuel bearing components, and one fuel l skeleton will also be stored. These components are made of 304 stainless steel, zirconium IV, l and Inconel. The source terms from these additional components were not independently considered in the shielding calculations, but the fuel source terms would bound this additional 3 waste.

(G 7.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES j Loading of spent nuclear fuel and other wastes into the basket is carried out under water in the Spent Fuel Pool Cask Loading Pit which prevents the spread of contamination. The baskets are dried and sealed within the controlled environment of the Fuel Building. The gaseous waste from the baskets will be passed through a local HEPA filter.

Once the basket is dried and seal welded, there are no credible off normal events or accidents that will cause breaching of the basket and subsequent release of airborne radioactivity.

Therefore, no airborne releases to the environment from the spent nuclear fuel assemblies or GTCC waste are expected to occur during loading and handling operations.

During normal operation of the ISFSI, the only potential source of airborne radioactivity is from surface contamination on the basket exterior, which would be deposited there from the Spent Fuel Pool water. As discussed in Chapter 5, filtered, borated water is injected into the n

7-8 November 25,1996

/

O Tmian Independent Spent FuelStorare Installation Saferv Analysis Report A basket internal assembly will be loaded with a dummy fuel assembly and a failed fuel can to l check the fit up and satisfactory operation of associated handling tools and equipment. A GTCC loading grate will be checked for fit up with the GTCC basket and ability to load a GTCC can into the GTCC basket. l l

The hydrostatic test and dewatering equipment will be tested to ensure that the hydrostatic l testing and dewatering can be accomplished in the amount of time necessary to prevent boiling of the borated water in the basket as described in Chapter 5. The vacuum drying and helium backfill equipment will be tested to ensure that the vacuum drying and helium backfill can be accomplished within the time limit described in Chapter 5.

9.2.3.1.2 Transfer Cask and Associated Equipment Load testing of the transfer cask and trunnions will be performed at 300% of design load. The Lifling Yoke will be load test to 150% of design load. The crane (s) that lift the loaded transfer ,

cask will also be load tested. Testing will also be performed prior to lifting a transfer cask if the load test has not been performed within the period of time specified in the test procedure, e.g., '

for loading a shipping cask several years after commencing ISFSI operation.

~)

A test load equivalent to the heaviest fully loaded basket will be placed in the transfer cask to j demonstrate the structural capability of the transfer cask bottom doors. The bottom doors will !

then be checked for proper operation afler supporting the test load. ;

The system used to inject water into the annulus between the basket and the transfer cask will be i tested to ensure that sufficient water is injected to minimize surface contamination of the basket external surface.

The load travel path at the site will be checked to ensure that the transfer cask can be safely moved from the Fuel Building bay to the Cask Wash Down pit. From the Cask Wash Down pit, the Transfer Cask will be moved to and lowered into the Cask Loading pit to verify the load O 9 - 10 November 25,1996

.J

0, Troian independent Spent FuelStorare Installation Saferv Analysis Report h O

C' Table 9.2-1 Page1of3 Pre-Operational, Startup, and Other Tests Component Type Test Purpose / Objective (s)

Basket lifting Other 1. Check fit up with shield lid and lining cranes.

equipment (attaches

2. Load test demonstrates ability to safely lift a fully to shield lid) loaded basket.

Basket automated Other 1. Check fit up of shield lid, structural lid, quick connect welding system and valves, and valve access port cover plates.

cutting equipment

2. Demonstrate ability to install and remove the lids and valve access port cover plates.

Basket shield lid Other 1. Check fit up of retainers with shield lid.

retainers

2. Demonstrate ability to keep the shield lid on the basket during/afler mishandling event or transfer cask tip over.

Fuel basket internal Other 1. Check fit up with fuel basket.

"S*** Y O 2. Load dummy fuel assembly into basket internal b assembly. I GTCC basket 28-slot Other 1. Check fit up with GTCC basket.

grating

2. Load GTCC can into basket using the grating.

Basket hydrostatic Other 1. Check fit up with basket quick connects.

test, dewatering,

2. Demonstrate ability to pressurize / evacuate basket to acuum ,and required test pressure / vacuum.

systems 3. Demonstrate ability to dewater and evacuate basket in the time required to prevent boiling.

4. Demonstrate ability to vacuum dry and backfill the basket with helium within the required time.

1 .

O November 25.1996 i

O I Troian Independent Spent Fuel Storare in stallation Safety A nalysis Report h Table 9.2-1 Page 2 of 3 Pre-Operational, Startup, and Other Tests I

Transfer cask lining Other Load test demonstrates ability to safely lift a fully loaded crane (s) transfer cask.

l' Transfer cask and Other 300% load test demonstrates ability to safely lin a loaded trunnions transfer cask.

Lining Yoke Other 1. Check fit up with transfer cask and crane.

2. 150% load test demonstrates ability to safely liR a loaded transfer cask.

i Transfer cask bottom Other Demonstrate proper operation of bottom doors aner doors supporting the weight equivalent to a fully loaded basket.

Transfer cask Other Demonstrate the ability to inject suflicient water into the annulus water basket / transfer cask to minimize contamination of basket injection system external surfaces.

Concrete cask air Other Demonstrate ability to lin the weight equivalent of a fully pads loaded storage cask.

Concrete cask air Pre-op Demonstrate proper operation of the temperature outlet temperature monitoring system prior to placing a loaded basket into the monitoring system concrete cask.

Concrete cask shield Other Check fit up.

ring and cask lid Overpack automated Other Check fit up of overpack, structural lid, quick connect, and welding system and quick connect cover and demonstrate the ability to install cutting equipment and remove the lid and quick connect cover.

Overpack shield ring Other Check fit up.

O V

November 25,1996

ML20135B534

Text

1.0 INTRODUCTION

1.1 INTRODUCTION

2.2 DESCRIPTION

SUMMARY

2.8 REFERENCES

SUMMARY

3.7 REFERENCES

SUMMARY

2.7 CRITICALITY EVALUATION

SUMMARY

7.7 REFERENCES

8.4 REFERENCES

2.7 CRITICALITY EVALUATION

Navigation menu

Search