ML20202G451

Rev 0 to PGE-1076, Trojan Reactor Vessel Package Sar

ML20202G451
Person / Time
Site:	07109271, Trojan  File:Portland General Electric icon.png
Issue date:	02/02/1999
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20202G436	List: ML20202G429 ML20202G451 ML20202G463
References
PGE-1076, PGE-1076-R, PGE-1076-R00, NUDOCS 9902050184
Download: ML20202G451 (200)

Text

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k. PGE-1076

O PORTLAND GENERAL ELECTRIC COMPANY TROJAN REACTOR VESSEL PACKAGE i

I SAFETY ANALYSIS REPORT i

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I Revision 0 l

Portland General Electric Company 121 SW Salmon Street Portland, Oregon 97204 O

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9902050184 990202 PDR ADOCK 05000344' H PDR

Trojan Reactor VesselPackagc - Safety Analysis Report n ,

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TABLE OF CONTENTS 1.0 GENERAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 1.1 INTRO D UCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 1.1.1 APPLICATION APPROACH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 2 1.1.1.1 Normal Condition of Transport - RVP Drop . . . . . . . 1 - 3 1.1.1.2 Hvoothetical Accident Condition - RVP Drop . . . . . 1 - 4 1.1.1.2.1 Transportation Route Evaluation . . . . . . . . . . 1 - 4 1.1.1.2_2 Operation Controls and Shipment Specific Considerations . . . . . . . . . . . . . . . . . . . . . . . . 1 - 5 1.1.1.2.3 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 8 1.1.1.3 RVP Brittle Fracture Evaluation Methodolocy . . . . . . 1-9 1.1.1.4 Consecuences of a Non-Credible Reactor Vessel Containment Breach . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 12 1.2 PACKAGE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 14

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~] PACKAGING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 14 1.2.1.1 Reactor Vessel . . . . . . . . . . . . . . . . . . . < . . . . . . . . . 1 - 14 1.2.1.2 Reactor Vessel Internals . . . . . . . . . . . . . . . . . . . . . . 1 - 14 1.2.1.3 Low Density Cellular Concrete . . . . . . . . . . . . . . . . 1 - 15 1.2.1.4 Penetration Covers . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 15 1.2.1.5 S hield ing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 3 5 1.2.1.6 Imnact Lim iters . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 15 1.2.2 OPERATIONAL FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 16 1.2.3 CONTENTS OF PACKAGING . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 16 1.2.3.1 Surface Contamination Activity . . . . . . . . . . . . . . . 1 - 16 1.2.3.2 Activation Activity . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 18 1.3 RE FERENCE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 20 2.0 STRUCTURAL EVAL UATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 1 2.1 STR UCTURA L D E SIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 1 2.1.1 STRUCTURAL DESCRIPTION ... . .. ................2-I s 2.1.1.1 O Reactor Vessel Packace Containment Bcundarv . . . . 2 - 2

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2.1.1.1.1 Reactor Vessel Shell . . . . . . . . . . . . . . . . . . . 2 - 2 2.1.1.1.2 Reactor Vessel Upper Head . . . . . . . . . . . . . . 2 - 3 2.1.1.1.3 Reactor Vessel Head Studs . . . . . . . . . . . . . . 2 - 3 2.1.1.1.4 Reactor Vessel Flange O-Rings . . . . . . . . . . . 2 - 4 2.1.1.1.5 Welded Penetration Closures . . . . . . . . . . . . . 2 - 4 2.1.1.2 Low Density Cellular Concrete . . . . . . . . . . . . . . . . . 2 - 4 2.1.1.3 External Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 6 2.1.1.4 Impact Limiters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 6 2.1.2 DESIGN CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 7 2.1.2.1 Structural Analysis Criteria . . . . . . . . . . . . . . . . . . . . 2 - 7 2.1.2.2 Allowable Stresses . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 8 2.1.2.3 Fracture Toughness . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 9 2.1.3 DESIGN CRITERIA FOR TIEDOWN SYSTEM . . . . . . . 2 - 12 2.1.4 DESIGN CRITERIA FOR LIFTING DEVICES . . . . . . . . . . . . . . 2 - 13 2.2 WEIGHTS AND CENTER OF GRAVITY . . . . . . . . . . . . . . . . . . . . . . . . 2 - 13 2.3 (m\ MECHANICAL PROPERTIES OF MATERI ALS . . . . . . . . . . . . . . . . . . 2 - 13

\m/ 2.4 GENERAL STANDARDS FOR PACKAGES . . . . . . . . . . . . . . . . . . . . . . 2 - 15 2.4.1 MINIMUM PACKAGE SIZE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 15 2.4.2 TAMPER-PROOF FEATURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 15 2.4.3 POSITIVE CLOS URE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 15 2.4.4 CHEMICAL AND GALVANIC REACTIONS . . . . . . . . . . . . . . . 2 - 16 2.4.5 PACKAG E VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 16 2.4.6 ACCESSIBLE SURFACE TEMPERATURE . . . . . . . . . . . . . . . . 2 - 17 2.4.7 CONTINUOUS VENTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 17 2.5 LIFTING AND TIEDOWN STANDARDS FOR ALL PACKAGES . . . . 2 - 17 2.5.1 LIFTING DEVICES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 17 2.5.2 TIEDO WN DEVICES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 17 2.6 NORMAL CONDITIONS OF TRANSPORT . . . . . . . . . . . . . . . . . . . . . . 2 - 20 2.6.1 HEAT................................................2-20 2.6.2 COLD................................................2-22 2.6.3 REDUCED EXTERNAL PRESSURE . . . . . . . . . . . . . . . . . . . . . . 2 - 23 2.6.4 INCREASED EXTERNAL PRESSURE . . . . . . . . . . . . . . . . . . . 2 - 25 O

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2.6.5 VIB RATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 2 7 2.6.6 WATER S PRAY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 2 8 2.6.7 FRE E DRO P . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 2 8 2.6.7.1 Reactor Vessel Free Drop Stress Analysis . . . . . . . . 2 - 29 2.6.7.1.1 Containment Boundary Stress . . . . . . . . . . . 2 - 29 2.6.7.1.2 Upper Head Attachment Studs . . . . . . . . . . . 2 - 30 2.6.7.1.3 Nozzle Cover Stress and Attachment Weld S tre ss . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 3 2 2.6.7.2 Shielding Free Drop Stress Analysis . . . . . . . . . . . . 2 - 33 2.6.7.3 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 33 2.6.8 CORNER D ROP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 3 4 2.6.9 COMPRESSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 3 4 2.6.10 PENETRATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 3 4 2.7 HYPOTHETICAL ACCIDENT CONDITIONS . . . . . . . . . . . . . . . . . . . . . 2 - 35 2.7.1 F REE DRO P . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 3 5 2.7.1.1 Impact Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 3 6

\ 2.7.1.2 Reactor Vessel Free Drop Stress Analysis . . . . . . . . 2 - 40 2.7.1.2.1 Containment Boundary Stress . . . . . . . . . . . 2 - 40 2.7.1.2.2 Upper Head Attachment Studs . . . . . . . . . . . 2 - 41 2.7.1.2.3 Nozzle Cover Stress and Attachment Weld S t ress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 44 2.7.1.2.4 Vessel Shell Buckling Under Side Drop Load s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 4 5 2.7.1.3 Shielding Free Drop Stress Analysis . . . . . . . . . . . . 2 - 45 2.7.1.4 Fracture Toughness Considerations . . . . . . . . . . . . . 2 - 46 2.7.1.5 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 47 2.7.2 C RUS H . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 4 8 2.7.3 P UN CTURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 4 8 2.7.3.1 General Puncture Conditions . . . . . . . . . . . . . . . . . . 2 - 49 2.7.3.2 Puncture Stress Analysis . . . . . . . . . . . . . . . . . . . . . 2 - 50 2.7.3.2.1 Puncture on The Vessel Side in the Region Below Nozzles . . . . . . . . . . . . . . . . . . . . 2 - 50 iii L

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( c 2.7.3.2.2 Puncture on The Lower Head . . . . . . . . . . . 2 - 51  ;

2.7.3.2.3 Puncture on The Upper Head . . . . . . . . . . . . 2 - 52 2.7.3.2.4 Puncture on a Nozzle Cover . . . . . . . . . . . . . 2 - 53  ;

, 2.7.3.2.5 Puncture on The Side ofThe Upper Head  ;

I (Stud Analysis) . . . . . . . . . . . . . . . . . . . . . . . 2 - 54 2.7.3.?.6 Puncture (lateral) On Upper and Lower  !

l Penetration Covers . . . . . . . . . . . . . . . . . . . . 2 - 55  !,

2.7.3.3 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 55 l

2.7.3.4 Fracture Touchness Considerations 3. . . . . . . . . . . . 2 - 55  !

2.7.4 THERMAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 5 6 i 2.7.5 I MMERS I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 5 7 ,

2.7.6

SUMMARY

OF DAMAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 59 ,

2.8 SPE CI AL FO RM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 5 9 2.9 FUE L ROD S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 60 2.10 RE FEREN C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 61 '

3.0 THERMAL EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 1 3.1 DI SC U S SION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 1 3.1.1 PACKAGE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 1 3.1.2 THERMAL ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . 3 - 2 3.1.3 RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 3 3.1.3.1 Normal Conditions of Trangoort . . . . . . . . . . . . . . . . 3 - 3 3.1.3.2 Hypothetical Accident Conditions . . . . . . . . . . . . . . . 3 - 4 3.7 MATERI AL PROPERTIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 4

- 3.3 TECHNICAL SPECIFICATIONS FOR COMPONENTS . . . . . . . . . . . . . . 3 - 4 3.4 THERMAL EVALUATION FOR NORMAL CONDITIONS OF TRAN S PORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 4 3.4.1 THERMAL MODEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 5 3.4.1.1 Analvtical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 5 i 3.4.2 MAXIMUM TEMPERATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 6 l 3.4.3 MINIMUM TEMPERATURES . . . . . . . . . . . . . . . . . . . .. . . . . . . . . 3 - 7 3.4.4 MAXIMUM INTERNAL PRESSURES . . . . . . . . . . . . . . . . . . . . . . 3 - 7 iv I

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Trojan Reactor VesselPackage - Safety Analysis Report 3.4.5 EVALUATION OF PACKAGE PERFORMANCE FOR NORMAL <

CONDITIONS OF TRANSPORT . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 9 l 3.5 HYPOTHETICAL THERMAL ACCIDENT EVALUATION . . . . . . . . . . . 3 - 9 3.5.1 THERMAL MODEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 9 3.5.1.1 Annivtical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 3 - 9 3.5.2 PACKAGE CONDITION AND ENVIRONMENT . . . . . . . . . . . . . 3 - 9 3.5.3 PACKAGE TEMPERATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 10 I l

3.5.4 MAXIMUM INTERNAL PRESSURES . . . . . . . . . . . . . . . . . . . . . 3 - 10 1 3.5.5' EVALUATION OF PACKAGE PERFORMANCE FOR HYPOTHETICAL ACCIDENT THERMAL CONDITIONS . . . . 3 - 10 3.6 . SUM M AR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 10 1

l 4.0 CONTAINM ENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 1 ~ j

- 4.1 RVP CONTAINMENT BOUNDARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 1 I

.4.1.1 RVP CONTAINMENT VESSEL . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 1 l l

.4.1.2 RVP CONTAINMENT PENETRATIONS . . . . . . . . . . . . . . . . . . . . 4 - 1  ;

4.1.3 SEALS AND WELDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 2 4.1.4 CLOS URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 2 4.2 - ' CONTAINMENT REOUIREMENTS FOR NORMAL CONDITIONS OF -

TRAN S PORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 2 l

4.2.1 CONTAINMENT OF RADIOACTIVE MATERIAL . . . . . . . . . . . 4 - 2  ;

1 4.2.2 CONTAINMENT CRITERION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 3 1 4.2.3 PRESSURIZATION OF CONTAINMENT VESSEL . . . . . . . . . . . 4 - 4 4.3 CONTAINMENT REOUIREMENTS FOR HYPOTHETICAL ACCIDENT l CONDITIONS ...............................................4-4 4.3.1 CONTAINMENT OF RADIOACTIVE MATERIALS . . . . . . . . . . 4 - 5 4.3.2 CONTAINMENT CRITERION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 5 4.4 S UM MAR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 5 4.5 RE FERENC ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 6 5.0 SHIELDING EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 1

- 5.1 DISCUSSION AND RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 1 5.2 SOURCE S PECIFICATION . . . . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . 5 - 3 y

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5.2.1 SOURCE REGIONS .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 3 5.2.2 SOURCE TERM NORMALIZATION . . . . . . . . . . . . . . . . . . . . . . . 5 - 4 ]

I 5.2.3 RvP SOURCE DEriNiTiON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 5 5.3 MODEL SPECIFICAT,I.QN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 6 5.3.1. DESCRIPTION OF SHIELDING CONFIGURATION . . . . . . . . . . 5 - 6 j 5.3.2 MATERIALS COMPOSITION . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 5 - 7 )

5.4 SHIFI DING EVALUATION . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . 7

' )

5.4.1 CONTACT SHIELDED RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 8 i 5.4.2 TWO METER SHIELDED RESULTS . . . . . . . . . . . . . . . . . . . . . . . 5 - 8 5.4.3 - NOZZLE RESULTS ' . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 5 - 8

. i 5.4.4 REACTOR VESSEL UPPER HEAD RESULTS . . . . . . . . . . . . . . . 5 - 8 5.4.5 REACTOR VESSEL LOWER HEAD RESULTS . . . . . . . . . . . . . . 5 - 9  ;

1 5.5 SHIELDING ANALYSIS UNCERTAINTIES . . . . . . . . . . . . . . . . . . . . . . . 5 - 9 5.6 RE FE RENC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 1 1 )

6.0 . . CRITICALITY EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 .

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i 7.0 OPERATING PROCEDURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 1 l 7.1 GENERAL TRANSPORT SCENARIO . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 1 i 7.2 PREPARATIONS F'OR TRANSPORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 2 )

7.3 TROJAN SITE TRAN SIT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 7 - 4 1 7.4 COLUMBIA RIVER TRANSIT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 4 7.5 PORT OF BENTON TO US ECOLOGY SITE TRANSIT . . . . . . . . . . . . . 7 - 6 7.6 RE F E RENC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 8 8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM . . . . . . . . . . . . . . 8 - 1 8.1 A C C E PTAN C E TE STS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 1 8.2 PRFI IMINARY DETERMINATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 1 8.2.1 VISUAL INSPECTION . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 8 - 1 8.2.2 STRUCTURAL INSPECTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 2 8.2.3 L EAK TEST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 2 i 8.2.4 COMPONENT TESTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 3 y;

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8.2.4.I' Reactor Vessel Flance O-Rinos . . . . . . . . . . . . . . . . . . . . . . . 8 - 3 8.2.4.2 Imnact Limiter Foam . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 3 8.2.4.3 Low Dencity Cellular Concrete . . . . . . . . . . . . . . . . . . . . . . . 8 - 5 8.2.5 TESTS FOR SHIELDING INTEGRITY . . . . . . . . . . . . . . . . . . . . . 8 - 6 8 2.6 . THERMAL ACCEPTANCE TESTS . . . . . . . . . . . . . . . . . . . . . . . . 8 - 6 8.3 PRE S S URE TESTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 7 8.4 P ACKAGE MARKING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 7 8.5 ' ROUTINE DETERMINATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 7 8.5.1 PROPER PACKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 8 8.5.2 UNIMPAIRED PACKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 8 8.5.3 CLOSURE DEVICES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 8 8.5.4 LIQUID CONTAINMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 9 8.5.5 PRESSURE RELIEF DEVICES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 9 8.5.6 LOADING AND CLOSURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 9 8.5.7 NEUTRON ABSORBER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 9 p 8.5.8 LIFTING OR TIEDOWN ATTACHMENTS . . . . . . . . . . . . . . . . . 8 - 10 8.5.9 SURFACE CONTAMINATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 10 8.5.10 EXTERNAL RADIATION LEVELS . . . . . . . . . . . . . . . . . . . . . . . 8 - 10 8.5.11 PACKAGE SURFACE TEMPERATURE . . . . . . . . . . . . . . . . . . . 8 - 11

' 8.6 S UM M AR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 1 1

\. vii y l

i Trojan Reactor VesselPackage - Safety Analysis Report i  :

LJ LIST OF TABLES Table Title 1-1 Reactor Vessel Package Closure Materials 1-2 Reactor Vessel Package Radioactivity and A2 Fractions 2-1 Summary of Results for Normal Conditions of Transport 2-2 Sununary of Results for Hypothetical Accident Conditions 2-3 Reactor Vessel Containment Boundary Components 2-4 Reactor Vessel Package Calculated Weights 2-5 Range or Minimum Mechanical Properties of Steel Materials (-40"F -200*F) 2-6 Summary of Design Stress Intensity Values, S.

2-7 Material Properties of Polyurethane Foam (20 pcf density) 2-8 Design Accelerations for Tiedown System Design

, 2-9 Surranary of Margins of Tiedown Induced Stresses (m) , 2-10 Thermal Stress Analysis Results 2-11 Summary of Results for Reduced External Pressure 2-12 Summary ofNB-3133 External Pressure Analysis 2-13 Summary of Results for Increased External Pressure 2-14 Summary of FCT Free Drop Margins of Safety 2-15 Free Drop Impact Analysis Results 2-16 Summary of HAC Free Drop Margins of Safety 2-17 Summary of HAC Puncture Margins of Safety 2-18 Summary of Results for Immersion 3-1 Intemal Component Decay Heat 3-2 Environmental Conditions for Normal Conditions of Transport 3-3 Maximum NCT Temperatures for the Trojan RVP 3-4 Maximum Temperatures for the Trojan RVP During HAC Fire Case 5-1 Summary of RVP Maximum Dose Rates (mrem /hr)

,a Vill

Trojan Reactor YesselPackage - Safety Analysis Report .

o Sb Calculated Dose Rates for the Unshielded Reactor Vessel (mrem /hr) 5-3 ~ April 1996 Reactor Vessel Survey Results Measured @ 8% inches from Vessel  :

Wall (mrem /hr) ..

5-4 ANISN Radial Flux Results for Reactor Vessel Wall l

5-5 ~ Cobalt-60 RVP Normalized Source Strength as of11/97 (Curies) 5-6 - Material Inputs Elemental Partial Densities (g/cc) 5-7 Shielded RVP NCT Dose Rates (mrem /hr)

'5-8' Shielded RVP HAC Dose Rates (mrem /hr)  :

5-9 Calculated Two-Meter Dose Rates for the Shielded Reactor Vessel & Intemals

_ 5-10 Trojan Operating History I 5-11. .. Trojan Vessel and Intemals ANISN Radial Flux Results l5-12 UpperInternals Surface Contamination Source Term -

1 Foam Impact Limiter Required Static Crush Values i

'O I

C ix

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

Trojan Reactor VesselPackage . Safety Analysis Report i

LIST OF FIGURES .

~

Figure Shact Title 1-1 1 Reactor Pressure Vessel Package on Transporter 4

2 Reactor Vessel Package on Barge 1-2 I General Arrangement Integral Shielding, Cradle, Tie Down Assembly and Foam Ring Impact Limiters '

l-3 1 Reactor Vessel and Internals 1-3A 1 Reactor Vessel Package 2 Reactor Vessel Package 1-4 1 Reactor Vessel Penetrations 2 ReactorInletNozzle 3 Reactor Outlet Nozzle 4 Drain Penetration 5 Incore Instrumentation Penetration 6 Head Vent Penetration 7 CRDM Penetration 8 Flange Monitoring Tube 5 1 Reactor Vessel Main Shielding Details ,

2 Reactor Vessel Supplemental Shielding Detail )

3 Reactor Vessel Supplemental Shielding Attachment Detail at Main Shielding Closure Plates 4 Reactor Vessel Shielding Detail Notes 5 General Notes for Figures 1-4 and 1-5  ;

l .

l-6 1 Impact Limiter Casing Assembly View From End 2 Impact Limiter Casing Assembly Reactor Head End (Elevation) x

l I

Trojan Reactor VesselPackage- Safety Analysis Report @

L) 3 l Impact Limiter Casing Assembly Reactor Head End (Elevation) 4 Impact Limiter Casing Assembly Reactor Bottom End (Elevation) 5 Impact Limiter Casing Assembly Reactor Bottom End (Elevation) 6 Impact Limiter Casing Assembly Assembly Details 7 Impact Limiter Casing Assembly Assembly Details 1 8 Impact Limiter Casing Assembly Notes l l

2-1 RVP Transport Configuration l 2-2 RVP Puncture Event Orientations l

2-3 RVP to Transporter Lor d Path (Longitudinal) i 2-4 RVP to Transporter Load Path (Vertical) 2-5 RVP to Barge Load Path (Longitudinal) l 2-6 RVP to Barge Load Path (Vertical) l 2-7 RVP to Transporter and Barge Load Path (Transverse) 1 pq 5-1 Reactor Vessel Model C/ 5-2 Reactor Vessel Model Source Regions 5-3 Reactor Vessel Conveyance Dose Points 7-1 On-site Transport Route Trojan Nuclear Plant 7-2 River Transport Route Trojan to Port of Benton 7-3 Overland Transport Route Port of Benton to US Ecology l l

,rx X

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Trojan Reactor Vesse! Package - Safety Analysis Report i

LIST OF APPENDICES Appendix Title  !

4 1-1 Probabilistic Safety Study for the Transportation of the Reactor Vessel Package (RVP) by Barge on the Columbia River from the Trojan Site to the Port of Benton  !

l-2 Probabilistic Safety Evaluation of the Overland Movement of the Trojan Reactor Vessel Package for Disposal as Part of Plant Decommissioning f

j 1-3 Assessment of External Events (

t 1-4 RPC 97-007 - Dose Projections Due to Hypothetical Drops of the Reactor Vessel  :

1-5 RVAIR Transport Temperature Analysis l

d 2-1 RVAIR - Weight and C.G.

2-2 . Longitudinal Restraint and Tiedown Structural Analysis 2-3 RVAIR - Hot and Cold Environment t 2-4 RVAIR- Reduced External Pressure i

2-5 RVAIR-Increased External Pressure I 2-6 RVAIR - Vibration 2 RVAIR-Impact Analysis 2-8 RVAIR - Penetration 2-9 RVAIR Free Drop Stress Analysis 2-10 RPV External Shielding StructuralIntegrity Analysis 2-11 RVAIR HAC Puncture Analysis 2-12 Trojan Reactor Pressure Vessel Side Impact and Puncture Load Brittle Fracture Analyses with Impact Limiters j

i 2-13 RVAIR -Immersion 3-1 RVAIRTnermal Analysis i

~

_ l 4-1 RPC 97-009 - Calculation of H 2Generation in the Reactor Vessel 4-2 RPC 97-006 - Calculation of Allowable Test Leakage Rate for Reactor Vessel l

Package per ANSI 14.5 1 4-3 RPC 96-008 - Reactor Vessel & Internals Surface Area Activity xii

- -- =

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

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Trojan Reactor Verse! Package - Safety Analysis Report O

t.j '

LIST OF EFFECTIVE PAGES Page Number - Revision 1

Table of Contents i through xiii 0 Pages 1-1 through 1-20 O Tables 1-1 and 1-2 0 Figures 1-1 through 1-6 0 Pages 2-1 through 2-62 0 Tables 2-1 through 2-18 0 Figures 2-1 through 2-7 0

. Pages 3-1 through 3-11 0 s Tables 3-1 through 3-4 0 i

Pages 4-1 through 4-6 0

' Pages 5-1 through 5-11 0 Tables 5-1 through 5-12 0 Figures 5-1 through 5-3 0 Page 6-1 0 Pages 7-1 thorugh 7-8 0 Figures 7-1 through 7-3 0 Pages 8-1 through 8-11 0 Table 8-1 0 9

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5.

- Trolan Reactor VesselPackare- Safety Analysis Report i

1.0 GENERAL INFORMATION j

This chapter of the Trojan Reactor Vessel Package Safety Analysis Report (SAR) presents a general description of the package. General arrangement and package design drawings are also  !

provided in this section.

l i

1.1 INTRODUCTION

The Trojato eclear Plant (TNP) is located approximately 4 miles SSE of the town of Rainier,

' Oregon at Ic. .er Mile 72.5 on the west bank of the Columbia River. The reactor was a Westinghouse pressurized water reactor licensed for 3411 megawatts thermal. The plant is jointly owned by Portland General Electric (PGE), Eugene Water and Electric Board (EWEB),

. and Pacific Power and Light (PacifiCorp). TNP received an operating license from the U.S.

Nuclear Regulatory Commission (NRC) on November 21,'1975 and began commercial operation  !

on May 20,1976. The TNP permanently ceased operations on January 27,1993 and was granted a Possession Only License by the NRC on May 5,1993. Subsequently, the decision was made to decommission the plant resulting in the Trojan Nuclear Plant Decommissioning Plan (Ref.1-1).

? The reactor vessel and associated reactor internals are the subject of this safety analysis report.

.e

'( The reactor vessel and the internals will be packaged as a single entity, the Reactor Vessel

• Package (RVP). This safety analysis report serves as the basis for PGE's application for a Certificate of Compliance to allow a single shipment of the TNP RVP as an exclusive use shipping packag- in accordance with the requirements of 10 CFR 71, " Packaging and Transportation of Radioactive Material." The RVP will be one-time shipped as an exclusive use -

Type B (as exempted) shipping package, as defined by 10 CFR 71.4, for disposal at the US

- Ecology licensed radioactive waste facility on the Hanford Nuclear Reservation near Richland,

. Washington.

To prepare the reactor vessel and associated internals as a shipping package the following actions will be performed:

1. The vessel and upper head sealing surfaces will be cleaned, inspected, and repaired (as necessary), new metallic O-rings installed, and the head studs tensioned to 720,000 lbs.

2. The internal volume, with the internals in place, will be filled with low density cellular concrete (LDCC).

3. The reactor vessel external surfaces will be decontaminated, as necessary, to

. ensure compliance with 10 CFR 71.87(i).

1-1

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F Trojan Reactor VesselPackare- Safety Analysis Report O

4. The reactor vessel extemal attachments will be removed and all penetrations will be sealed with welded closures. )

i

5. L Shielding will be installed on'the exterior surface of the reactor vessel, as- i necessary, to ensure compliance with the dose rate limits of 10 CFR 71.

>l

6. Impact limiters will be attached to the RVP to limit stresses to values well below l yield for the impact (drop) loads.

1 After preparation as a Type B (as exempted) shipping package, the RVP will be loaded onto a transporter and tied down and transported as an exclusive use shipment. The transporter is a l hydraulically leveled platform designed for transporting large, heavy loads. Once loaded onto i the transporter and tied down, the RVP will remain attached to the transporter until the RVP is >

off-loaded at the disposal site. The RVP on the transporter is shown on Figure 1-1.

i The loaded transporter will be moved from the Trojan Industrial Area to the barge slip on the l TNP site where it will be moved onto the barge and secured. The loaded transponer will be  ;

barged to the Port of Benton in Washington. The loaded transporter on the barge is shown on Figure 1-1. The loaded transporter will be moved off of the barge and transported by road (less than 30 miles) to the disposal facility operated by US Ecology, near Richland, Washington. The  ;

RVP will be then be off-loaded from the transporter at the disposal facility.

The quality assurance requirements of 10 CFR 71, Subpart H, applicable to the design,  ;

fabrication, and use of packaging for radioactive materials are covered by TNP's NRC-approved 10 CFR 50, Appendix B quality assurance program (PGE-8010). PGE-8010," Trojan Nuclear Plant Nuclear Quality Assurance Program," was approved by the NRC for application to design, fabrication, assembly, and modification of transportation packages by NRC letter dated December 22,1994, " Quality Assurance Program Approval for Radioactive Material Packages No. 0327, Revision No. 7."

1.1.1 APPLICATION APPROACH The PGE application is based on alternative transport conditions due to the uniqueness of the RVP and its one-time shipment. Therefore, this PGE request requires NRC approval pursuant to 10 CFR 71.41(c) or alternatively,10 CFR 71.8. The 10 CFR 71 regulations provide the NRC with authorization to approve packaging and shipments based on alternative transport conditions.

These regulations recognize that special controls imposed by the shipper may provide equivalent safety as those specified in 10 CFR 71.

1-2

Trolan Reactor Vessel Package - Safety Analysis Report h

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The following provides a description of the alternative and itsjustification. In addition, a section is provided which discusses the consequences of two non-credible accidents in which the RVP containment boundary is breached. The conclusion reached in this section is that the breach of the containment boundary would not result in a significant radiological consequence to the public.

1.1.1.1 Normal Condition of Tramport - RVP Drop The normal conditions of transport specified by 10 CFR 71(c)(7) requires a drop of the specimen through a distance of I ft onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. PGE is taking special precautions to ensure the safe shipment of the RVP which are described in Chapter 7 of the SAR. Based on the SAR, given the specified transportation route, method of shipment, and special controls, the 1 foot drop should not be considered a normal condition of transport. Therefore, PGE requests N'tC acceptance of the conditions and controls for this shipment under 10 CFR 71.41(c) or an exemption under 10 CFR 71.8 based on the demonstration of adequate safety of the shipment.

Prior to shipment, the RVP will be prepared as a shipping package and will be loaded and tied down onto a specially designed transporter. The loaded transporter will be moved onto a

(] specially selected barge and secured utilizing an engineered tie down system. The barge will be

(/ grounded during this evolution. The RVP loaded transporter will be barged up the Columbia River to the Port of Benton where a prime mover will connect to the transporter and move it off the barge and overland to the disposal facility. The RVP will be off-loaded at the disposal facility.

The RVP will be rotated to a horizontal position (i.e., the centerline longitudinal axis of the package will be horizontal) during preparation in the TNP industrial area. During transport, the RVP will remain oriented in the horizontal position. Because of the unique size and mass of the package and the method ofsupport of the package, no other orientation is reasonable during RVP transport. Once loaded onto the transporter, the RVP will not be removed from the transporter at anytime during transport.

Based on above conditions and the special handling and operational controls to be exercised, the one foot drop should not be considered a normal condition of transport. However, PGE has designed and analyzed the RVP with impact limiters to withstand the effects of a 1 foot horizontal drop and I foot oblique pivot drops.

O V 1-3

Trojan Reactor Vessel Packare - Safety Analysis Retsort p-1.1.1.2 Hvoothetical Accident Condition - RVP Drop The hypothetical accident condition specified by 10 CFR 71.73(c)(1) requires a drop of the specimen through a distance of 30 feet on a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. PGE is taking special precautions to ensure the safe shipment of the RVP which are described in Chapter 7 of the SAR.

Based on the SAR specified transportation route, method of shipment, and special controls, the 30 foot drop should not be considered a hypothetical accident condition. Therefore, PGE requests NRC acceptance of the conditions and controls for this shipment under 10 CFR 71.41(c) or an exemption under 10 CFR 71.8 based on the demonstration of adequate safety of the shipment.

10 CFR 71.73 specifies the hypothetical accident conditions to which Type B radioactive material packages are to be designed for unrestricted use. According to a NRC-sponsored modal study, packages designed to 10 CFR 71.73 are capable of withstanding greater than 99% of all credible accidents, rail and highway, without functional failure. As a consequence,10 CFR 71 does not require in-transit precautions such as escorts, routing restrictions, speed controls, moving safety zones, etc.

f PGE has conducted a safety evaluation of the specific route, method of transport, and operational

{} controls and shipment specific conditions to be utilized during the transportation of the package.

1.1.1.2.1 Transportation Route Evaluation The overland rcute for the RVP from the Port of Benton to the US Ecology disposal site can be considered as very benign in that: 1) the road is in good condition,2) there are no obstacles such as heavy traffic, bridges or overpasses to cross,3) there are no "hard targets" or surfaces,4) the l area is essentially unpopulated,5) there are no hazardous terrain features, and 6) it is an area where nuclear activities are well known and understood. The haul route was used for disposal of the TNP Steam Generator and Pressurizer and a portion of the route is routinely used for the transport of decommissioned defueled naval submarine reactor plants.

/'N V i~4

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Trolan Reactor VesselPackage- Safety Analysis Retrort t

l i O V

The discrete probabilities of the significant accidents evaluated on both the Columbia River and overland are:

i i r

Event Discrete Probability i

Impact of the RVP on the Barge 3.2E-07 l Separation of the RVP/ Barge 1.0E-06 l Severe Fire / Explosion - 6.0E-10 (during barging) l Land Transport Accident 1.3E-07 i

As noted in the RVP river transit study, because of the unique nature of the RVP shipment, and the importance of safety in all aspects ofits planning, the accident rates determined from historical records (and utilized in the study) will tend to overstate the accident probabilities for the RVP river shipment. Appendix l-1 and 1-2 provide additional details on the transportation route evaluation.

1.1.1.2.2 Operation Controls and Shipment Specific Considerations

]

i l The following operational controls and shipment-specific considerations will be utilized during  ;

l the transport of the RVP.

i A Transportation Safety Plan (TSP) will be prepared that identifies the responsibilities and interfaces of PGE, PGE contractors, Federal Agencies, State Agencies, and Local Agencies. The

! TSP will address the operating controls and procedures, radiological controls, and contingency actions to be taken in the event of a problem during transport. The requirements of the TSP will be implemented by detailed procedures and coordinated with state and local agencies responsible for emergency response along the route. The TSP will be approved by the Oregon Office of Energy (OOE) prior to shipping the Reactor Vessel Package. The TSP will be made available to the United States Coast Guard (USCG) Captain of the Port (COTP) for review prior to shipment.

Changes to the plan will be approved by OOE and reviewed by the USCG. Appropriate advance notifications, per 10 CFR 71.97, will be made prior to the shipment.

The shipment will comply with the applicable specifications of ANSI N14.24-1985, "American

l. National Standard for Highway Route Controlled Quantities of Radioactive Materials - Domestic

[ Barge Transport," the applicable requirements of 10 CFR 71 - Packaging and Transportation of i -

i

1-5 0

l l

Troian Reactor VesselPackace- Saferv Analysis Report l

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'J Radioactive Material,33 CFR - Navigation and Navigable Waters,46 CFR - Shipping, and 49 CFR - Transportation.

l The transporter, specially designed for transporting large, heavy loads, will be limited to a  !

maximum overland transport speed of 5 mph. The transporter and prime mover will be inspected  !

prior to shipment to ensure the vehicles are working properly and to ensure conformance with applicable state and federal standards. The structural adequacy of the transporter will be demonstrated by analysis and the transporter will be loaded in accordance with the manufacturer's specifications. The entire land transportation route, onsite and offsite, will be evaluated to confirm that it is structurally capable of withstanding the load. The Oregon l Department of Transportation may, at its option, perform Commercial Vehicle Safety Alliance (CVSA) inspections on the loaded transporter and prime mover prior to leaving the Industrial Area. In addition, the RVP will be inspected for acceptance by US Ecology prior to leaving the Industrial Area. ,

The tug and barge chosen for this shipment will be required to have high standards of I maintenance. Two well-qualified and well-informed crews with local knowledge of the i Columbia River will be provided (one on the primary tug and one on the backup tug). The barge loading procedure will be specified by a naval architect. The prime mover will not be n transported on the barge. The barge will have a current classification by the American Bureau of (s_) Shipping and a Certificate ofInspection by the U.S. Coast Guard (USCG), and assigned for sole use. The barge selected will have a high level of subdivision (i.e., multiple watertight compartments). Intact and damage stability calculations will be performed by a naval architect l to verify compliance with 46 CFR 172 and will be reviewed and approved by the USCO. Tie l

down design and calculations will be reviewed and approved by the National Cargo Bureau (NCB).

To ensure the structural integrity of the RVP, it will be transported only if the initial vessel l temperature is greater than 50 F equilibrium, and a minimum average daily temperature (average l

of forecasted high and low daily temperatures) of 50 F and a minimum daily temperature of 40 F l are forecasted along each day for the transportation route and the expected shipping duration.

Prior to departing the Trojan barge slip, the following requirements will be met:

1. A backup tug is present to accompany the primary tug.

2. Communication is established between the tugs and a base station to monitor progress of the barge transport.

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Trolan Reactor YesselPackaee- Safety Analysis Report O

3. The primary and backup tug and the barge are equipped with navigation and  ;

emergency equipment appropriate for river navigation (per ANSI N14.24-1985) ,

and approved by the USCO. t t

4. In addition to the forecasted temperature requirements noted above, there is no  :

adverse weather forecast between the TNP and the Port of Benton that may threaten the safety of the barge and package. '

5. There are no mechanical problems with the tug, backup tug, or the barge that may i

- affect the capability to safely transport the package. .

6. The primary and backup tug captains are licensed per 46 CFR Subchapter B

" Merchant Marine Officers and Seamen."

7. The U.S. Coast Guard will establish a safety zone per 33 CFR 165, if required, to '

ensure appropriate safety and security measures are met. l t

i

8. A PGE Radiation Protection (RP) representative and a PGE transportation coordinator will accompany the shipment. The RP representative will be trained 4

in the principles of health physics and equipped with appropriate radiation ,

protection instruments to provida radiological support by performing inspections and/or surveys and maintain personnel exposure ALARA.

9. Arrangements have been made with the U.S. Army Corps of Engineers to provide priority passage and exclusive use through the locks en route to the Port of

)

Benton.

l l

10. The NCB has evaluated the package to transporter and transporter to barge tie '

down systems and has certified that the two tie down systems comply with the applicable regulations.

. I 1. A marine surveyor, the NCB, and the USCG have inspected the condition of the barge and the stowage of the package on the barge to ensure integrity of the barge and that the final stowage configuration complies with the approved designs.

12. . Notifications of the pending shipment to appropriate authorities have been made.

13. The tugs will meet the applicable requirements of 46 CFR Subchapter C, "Uninspected Vessels." The tugs will be inspected by a marine surveyor prior to departure.

O l-7

l Trojan Reactor vessel Packare - Safety Analysis Ret, ort h gy l

-J During river transport, maximum speed will be 10 knots. Transit of the barge will be halted to  ;

avoid hazardous conditions or to await passage through a lock or to await safe navigation

{

conditions. When moored, appropriate measures will be taken to restrict un .orized access to j the barge. During the barge transport phase, the tug will check in with the base station at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

After arrival at the Port of Benton barge slip, the barge will be grounded, the transporter-to-barge tie downs will be removed and the transporter will be moved off the barge onto the landing.

Prior to depaning the Port of Benton Propeny, the Washington State Patrol may, at its option, i perform a Commercial Vehicle Safety Analysis (CVSA) inspection of the RVP loaded I transporter and the prime mover.

The RVP loaded transporter will travel to the US Ecology site, less than 30 miles from the barge slip, with a maximum overland transport speed of 5 mph. An overload permit will be obtained from local authorities for the overland travel. Escons will control road traffic in the vicinity of l

the transporter during the entire overland transport. All railroad traffic in the area will be stopped j during transit.

Prior to entry at the disposal site, the RVP loaded transporter will be re-inspected for acceptance rm by US Ecology. Once accepted, the transporter will be moved to a prearranged location at the Q US Ecology site. The prime mover and transporter will be driven into a prepared trench and the RVP will be off-loaded.

1.1.1.2.3 Conclusion l

The transportation environment (i.e., route and operational controls) are significantly different  !

from that contemplated in 10 CFR 71 for unrestricted packages. There are no man-made or  !

natural terrain features (e.g., bridges, cliffs, overpasses) which could produce impacts which are l equivalent to the 10 CFR 71.73(c)(1) described 30 foot drop onto an unyielding surface. The 30 I foot drop in the regulations is equivalent to a 44 fps (30 mph) vebcity. The overland travel speed of the RVP transporter is limited by procedure to 7.3 fps (5 mph), a factor of 6 less. The river travel speed of the barge is limited by procedure to 16.9 fps (10 knots), a factor of 2.6 less.

In addition, there are extensive in-transit precautions such as escorts, routing restrictions, inspections, etc. that are not typically required of a standard Type B package.

As previously stated, the RVP will be moved to a horizontal position (i.e., the centerline longitudinal axis of the package will be horizontal) during preparation in the TNP industrial area.

During transpon to the US Ecology disposal facility, the RVP will remain oriented in the horizontal position. Because of the unique size and mass of the package, no other orientation is reasonable during RVP transport.

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0 Trojan Reactor Vessel Packaee - Safety Analysis Ret, ort

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L)

Once loaded onto the transporter, the RVP will remain tied down to the transporter until the RVP is unloaded at the disposal site. Therefore, based on the Transportation Evaluation and the operational controls and specific considerations, the hypothetical accident condition of a 30 foot i drop should not be considered a condition of transport for the RVP.

Based on the specific transportation route and controls established, a maximum non-mechanistic drop configuration has been determined. The maximum postulated distance that the RVP could drop, based on the design of the impact limiters, transportation system, and route, during a hypothetical transport accident is 11 feet. This free drop height and horizontal orientation were used as a design basis for the RVP. Appendix 2-7A, "RVAIR Impact Analysis," analyzes two '

drop scenarios which begin with an 11 foot drop height. The first is a simple horizontal drop with both impact limiters striking the ground at the same time. The second is an oblique drop with one end dropping 11 feet and the other remaining in place until the initial impact occurs. In both drops, the total energy absorbed by the impact limiters includes the potential energy due to the limiter crush distance in addition to the initial drop height.

1.1.1.3 RVP Brittle Fracture Evaluation Methodology The Trojan reactor vessel was a Safety Class 1 vessel, and was originally designed in accordance with the requirements of the ASME Code,Section III (1968 edition with Addenda through (a) winter 1968). The design pressure for the vessel and the reactor coolant system was 2485 psig, and the maximum design temperature was 650 F.

Part 71 of Title 10 of the Code of Federal Regulations, " Packaging and Transportation of Radioactive Materials," requires that packages used to transport radioactive material be designed with consideration of normal transport and hypothetical accident events that might occur at temperatures as low as -20 F (-29'C). In order to assess the potential for brittle fracture of the Trojan vessel under transport accident conditions, the hypothetical side drop and puncture conditions were analyzed to determine the appropriate stresses. The concerns for brittle fracture of the vessel are addressed using an ASME Code Section XI, Appendix A fracture mechanics approach.

Linear elastic fracture mechanics analyses are normally used to evaluate brittle fracture concerns in reactor vessels during normal operation and accident (i.e., emergency and faulted) conditions.

The approach defined in ASME Section XI, Appendix G specifies the assumptions to be used in the brittle fracture analysis for determining the allowable operating pressure-temperature limits for heatup and cooldown. The Appendix G procedure establishes safety margins to prevent crack initiation during normal heatup, cooldown, and hydrostatic test operation. For severe emergency and faulted conditions, more detailed fracture mechanics analyses are required and attemative acceptance criteria are available in ASME Section XI, Appendix A. These ash.E Code rh

() 1-9

l Trojan Reactor Vessel Package - Saferv Analvsis Retwrt C i LJ requirements assure adequate margins against brittle fracture of the reactor vessel for operating plants.

l No specific design criteria are provided in the regulations, however, for protecting a shipping i container against brittle fracture initiation and unstable growth. Studies conducted by Lawrence Livermore National Laboratory (LLNL) in the early 1980s were published in NUREG/CR-3826, j and the recommendations for brittle fracture prevention contained in that document led to the

)

publication of Regulatory Guide 7.12. The guidance in Regulatory Guide 7.12 is applicable only 3 to ferritic steels with extremely low nil-ductility transition temperatures; however, NUREG/CR-3826 provides recommendations for an alternative approach based on ASME Section XI brittle 1 fracture evaluation methods. The criteria utilized in the analysis center around the concept of no I fracture initiation based on the allowable flaw sizes specified in IWB-3510-1 of Section XI of the ASME Boiler and Pressure Vessel Code.

l 1

Other methods were investigated to benchmark the adequacy of this approach. In particular, the l IAEA "Guidelinesfor Safe Design ofShipping Packages Against Brittle Fracture " describes three acceptable methods for preventing brittle fracture of radioactive material transport I packagings. One method can be referred to as a fracture mechanics approach. This method is used to demonstrate, by analysis or by a combination of analysis and test, that sufficient margin

()

p is available (considering applied stress, flaw size, and/or material fracture toughness) to preclude crack initiation and brittle fracture during design-basis events. This is equivalent to the method utilized in Appendix 2-12. Furthermore, this method provides three acceptable options for determining material fracture toughness. One of these options defines a lower-bound or near (statistically) lower-bound value for the fracture toughness at the lowest service temperature.

This definition of material fracture toughness is inclusive of the dynamic (K d i reference material fracture toughness curve for ferritic steels assumed in this analysis, and for this reason this option is noted to coincide with the present analysis approach.

With respect to safety factors used in this analysis, no specific safety factors are given in NUREG/CR-3826. However, the IAEA guidelines suggest that, when using the applied method, an overall minimum safety factor (KdK )i of approximately 3 is appropriate for normal loading conditions during shipping, with a suggested safety factor of 1.4 for unexpected loading events such as hypothetical accident conditions.

The results from the fracture mechanics calculations using stresses derived from the finite element analysis show significant margins against brittle crack initiation (i.e., KJKi > 1.0) under all evaluated conditions. In general, stress levels in the vessel are sufficiently low so that no crack initiation is predicted to occur for the analyzed side drop loading condition. This is due to the ability of the impact limiters to reduce the amplitude of the longitudinal beam bending stresses relative to ovalization and localized bending stresses. The membrane stress contribution n

{) 1 - 10 I

Trojan Reactor VesselPackage - Safety Analysis Report g

t t V

is small regardless of the assumption for intemal pressure (< 200 psi) near the impact locations and at geometric discontinuities (e.g., at the comer of the outlet nozzle). At an ambient temperature of -20 F, large margins against crack initiation (> 5.65) are calculated for all locations along the shell with the smallest margins (2.0) occurring close to the outlet nozzle. The other points of high stress concentration are at the contact locations with the impact limiters where the calculated margins are 2.51 and 3.17, respectively.

The axial stresses throughout much of the beltline region are very low (less than 10 ksi tension) or even compressive. Ovalization effects at the mid-core region are present and some tensile hoop stresses develop in these regions. However, the safety factor against brittle crack iwtiation at the most embrittled region of the vessel is greater than 7.0. Because of this, the beltline region of the Trojan vessel is not considered to be a limiting concern for the analyzed drop condition.

A more limiting condition is the hypothetical puncture loading. A safety factor of 1.01 and 1.0 was calculated at the upper and lower heads, respectively, at an ambient temperature of-20*F.

Although these values are relatively low,it should be noted that the calculations are based on very conservative assumptions. For example, use oflarge assumed full-circumference or quarter-circular surface flaws at the worst stress locations and use oflower bound dynamic reference toughness instead of using the static reference toughness add to the conservatism of the (N analysis results. In addition, the Inconel welds at the head penetrations have not been modeled V separately, and the toughness properties of the adjacent base metal were assumed for the fracture mechanics analyses. The Inconel nickel-base alloys are known to have excellent toughness at low temperatures because of their strength and ductility. Furthermore, Inconel alloys do not undergo a ductile-to-brittle transition in toughness, even at very low temperatures. A flaw remaining within the Inconel overlay would not experience brittle crack initiation. Therefore',

assuming dynamic toughness consistent with that of the ferritic base metal properties is very conservative.

Based on utiliziag worst case assumptions, the transportation route's lack of hard targets, the b limited orientations for the analyzed drop conditions, and the overall extremely low probability of an accident, this is considered acceptable. However, using the conservative fracture mechanics analysis techniques described in the SAR, a minimum safety factor of 1.4 has been demonstrated for the evaluated RVP drop and hypothetical puncture loading conditions at a minimum vessel wall temperature of 45 F.

Given the large number of variables necessary to be met prior to shipment, such as water level, weather (i.e., temperature, wind), and other uses of the Port of Benton facility, and the relatively short transportation window of three months each year, a limit which meets the package design criteria as described above and provides shipment period flexibility is highly desirable. A separate thermal analysis (provided as Appendix l-5) was completed, which shows that the A

l l 1 - II l

Trojan Reactor Vessel Package - Saferv Analysis Report o i vessel exhibits a large thermal inertia. The calculation shows that if the vessel initial temperature is 50 F and the daily average temperature has a minimum value of 50 F and varies plus or minus 20 F, the vessel wall temperature will remain above 45 F for the duration of the shipment.

Therefore, administrative controls will be implemented to ensure that the RVP will be transported only if the initial vessel temperature is greater than 50*F equilibrium, and a minimum average daily temperature (average of forecasted high and low daily temperatures) of 50 F and a minimum daily temperature of 40 F are forecasted for each day along the transportation route and the expected shipping duration.

The brittle ' vture analysis in Appendix 2-12 was provided to the Electric Power Research Institute (EPRI) Major Component Reliability group for their review. Their review determined that the input assumptions and the fracture mechanics methodology have a well founded technical bases and the results demonstrate adequate safety factors for the loading conditions postulated. They also noted that the overall conclusion of the analysis is conservative.

1.1.1.4 Consecuences of a Non-Credible Reactor Vessel Containment Breach PGE evaluated the radiological consequences to the public from two non-credible accidents which result in a non-credible breach of the package. The following conditions were evaluated:

1.

During overland shipment, the reactor vessel package falls off the transporter, and is breached.

2.

During river shipment, the reactor vessel package falls off the barge near the intake to a water treatment plant that draws drinking water from the river, and is breached.

The evaluation concluded that calculated doses for individuals due to either non-credible accident are well below the doses used to establish the Type A transportation package limits (5000 mrem TEDE,50,000 mrem TODE, and 15,000 mrem eye dose equivalent).

Type A package limits are also used for several other purposes in the regulations, such as specifying Type B activity leakage limits, LSA, and excepted package content limits. In establishing the Type A limits, the regulations assume that a person is unlikely to remain at a distance of 1 meter from the damaged package for more than 30 minutes.

PGE Radiation Protection personnel will accompany the RVP shipment, and therefore, will be available to establish boundaries and perform emergency response fimetions. In addition, traffic escorts will control / restrict road traffic in the vicinity of the transporter during the overland transport. As a result, the evaluation for the overland shipment accident assumed exposure O

d 1 - 12 l

1

Trolan Reactor Vessel Packaee - Soferv Analysis Report p .

V conditions for an individual standing 100 meters from the scene of the accident for 30 minutes.

The evaluation determined the extemal exposure (DDE) to an individual is approximately 25 mrem and the internal (CEDE) is approximately 12 mrem.

The evaluation for the river shipment accident determined that the external exposure (DDE) to an individual is 0 mrem and the internal exposure (CEDE) is approximately 1.25 mrem.

Therefore, it is concluded that the above non-credible accident conditions will not result in a significant radiological consequence to the public. Details of the evaluation are included in Appendix l-4.

Based on the accident study presented in Appendix l-2, the probability of a land accident is 1.3 E-07, and the worst case accident, a drop of the package from the transport height of the RVP, is mitigated by impact limiters. As discussed in Section 2.7.1 the package is not breached by this accident. However, to establish the hypothetical worst case consequence to the public and to the recovery workers, a package breach of a 1-inch crack half way around the circumference of the belt line area of the vessel was assumed. This package breach provides a bounding case for all other potential damage locations, based on its location in the highest dose rate area. It is large enough to provide for the potential release of the surface corrosion layer.

/7 This assumed package breach was also used to establish the hypothetical doses for the public and V recovery workers in the water transport accident analysis.

Details of the recovery operations following a non-credible breach of the Trojan RVP overland or underwater are provided in Section 3 of the Environmental Report. For land recovery operations, doses to response personnel were estimated to be 3.18 person-rem to repair dislodged closure plates,5.18 person-rem to perform shielding repairs, and an additional 8.35 person-rem to repair a vessel wall breach, were that to occur. For water recovery operations, doses to recovery personnel were estimated to be 0.66 person-rem to repair dislodged closure plates,0.96 person-rem to perform shielding repairs, and an additional 7.68 person-rem to repair a vessel wall breach, were that to occur. In all cases administrative controls would be implemented to keep exposures As Low As Reasonably Achievable (ALARA) and individual worker exposures within 10 CFR 20 limits.

r0 C/ 1 - 13 v

0 Trojan Reactor Vessel Package- Safety Analysis Report L,I 1.2 PACKAGE DESCRIPTIOh 1.2.1 PACKAGING The RVP will be prepared as a Type B (as exempted) shipping package. The containment boundary of the RVP is defined as the reactor vessel shell, the reactor vessel upper head with flange O-rings installed and tensioned by 54 reactor vessel head studs, and the welded penetration closures. The interior void space of the vessel will be filled with LDCC. Extemal shielding will be installed on the reactor vessel to meet the requirements of 10 CFR 71.47 and 10 CFR 71.51. Impact limiters will be installed to ensure the RVP is able to withstand stresses associated with the analyzed drop scenarios for the Normal Conditions of Transport (NCT) of 10 CFR 71.71 and the Hypothetical Accident Conditions (HAC) of 10 CFR 71.73.

Figures 1-1 and 1-2 provide an overview of the RVP. Reactor vessel intemal components are shown on Figure 1-3. The maximum gross weight of the RVP has been calculated to be 2,025,898 lbs. However, analyses to demonstrate compliance with the requirements of 10 CFR 71 were conservatively based on a weight of 2,040,000 lbs. Detailed information on the RVP weight and center of gravity (CG) is contained in Section 2.2.

<m 1.2.1.1 Reactor Vessel

! )

\_/

The reactor vessel is a cylindrical shell with an integral lower head and a removable upper head.

The upper head is attached to the vessel by 54 tensioned head studs and sealed using dual metallic O-rings at the flange surface between the reactor vessel and the upper head.

The overall length of the reactor vessel is 42' 6" with a diameter of 17' 1" (excluding the nozzles). The vessel thickness (excluding the stainless steel cladding) varies from 5%" at the lower head,6%" at the upper head,8%" at the core ring to 10%" at the nozzle ring. The dimensions, configuration, and penetrations are shown on Figures 1-3 and 1-4. The vessel is primarily composed of SA-533 Gr B Class I steel. The materials of construction are detailed in Section 2.1.

1.2.1.2 Reactor Vessel Internals The reactor vessel intemals consist of an upper intemals assembly and a lower intemals assembly. 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 inteinals assembly consists of a core barrel, core baffle plates, core former plates, neutron shield pads, lower core plate, lower core support columns, and the lower core support

!g) 1 - 14

W Trojan Reactor VesselPackage - Safety Analysis Ressort (3

V plate. The internals are constructed ofstainless steel. The configuration of the internals within the reactor vessel is shown on Figure 1-3.

1.2.1.3 Low Density Cellular Concrete The void space inside the reactor vessel will be filled with LDCC. The density of the LDCC will be between 45 lbs/ft) and 65 lbs/ft'. The LDCC will provide shielding of the radioactive components, fix the intemal surface contamination in place, and form an additional barrier to prevent the escape of contaminants. By preventing movement of the radioactive material inside the vessel, no changes in external dose rates will occur.

1.2.1.4 Penetra1hm Covers Reactor vessel penetrations will be covered using closures as shown on Figure 1-4. The reactor vessel penetrations include the reactor vessel inlet (4) and outlet (4) nozzles, a drain penetration, incore instrumentation penetrations (58), a head vent, Control Rod Drive Mechanism (CRDM) penetrations (78), and flange monitoring tube penetrations (2). These 148 reactor vessel penetrations will be sealed by welding steel plates over the openings. The plates vary in thickness from %" to 2%". Figure 1-4 and Table 1-1 provide additional penetration closure p details.

1.2.1.5 Shielding Chapter 5 discusses the dose rate limits specified in 10 CFR 71.47 and 10 CFR 71.51 applicable to the RVP. Dose calculations in Section 5.4 determined that extemal shielding is required to i maintain dose rates within these limits. Steel shielding (SA-516 Gr 70) will be installed on the exterior surface of the reactor vessel between the flange and the lower head region. The nominal shielding configuration is shown on Figures 1-4 and 1-5. Supplemental shielding will be installed, as necessary, based on the radiation surveys conducted prior to shipment. This will ensure the dose rate requirements of 10 CFR 71.47 are met. The shielding required to meet the dose rate requirements of 10 CFR 71.51 are designed to stay in place for the HAC. The remaining shielding (including the supplemental shielding) is designed to stay in place during the NCT (10 CFR 71.71). Section 2.1.1.3 includes a discussion of the shielding installation on the reactor vessel.

1.2.1.6 Imoact Limiters Although the size, weight, and administrative controls placed on the handling of the RVP, coupled with the extraordinary measures being taken for the short duration, one-time shipment make any drop scenario non-credible, the design of the RVP includes impact limiters installed on

(

1 - 15 l

. _ _ . - _ . _ . _ . . .__ . - . _ . . _ _~_ ___. -._ _ . _ _ _ . _ .._ _ _ _..

Troian Reactor Vessel Package - Safety Analysis Report ' >

0 t

- both ends of the vessel as shown on Figure 1-2. They provide an additional margin of safety by limiting the impact loads experienced by the RVP during the analyzed normal and accident condition drops. The limiters are constructed ofclosed cell polyurethane foam encased in sheet ,

steel. The impact limiters are each approximately 58" in length and 28'in diameter. They are  !

attached to the outer surface of the RVP by a system of brackets and tie bolts. The impact j limiters are discussed in Section 2.1.1.4 and details of the impact limiters are shown on >

' Figure 1-6.

1.2.2 OPERATIONAL FEATURES ~

' The reactor vessel will be modified to function as a self-contained shipping container.

Penetration closures will be welded in place and the interior void space will be filled with LDCC.

Steel shielding will be installed on the vessel as required to comply with the dose limit requirements of 10 CFR 71. Therefore, there are no operational requirements associated with the RVP.

1.2.3. CONTENTS OF PACKAGING

{

The radionuclide inventory in the RVP is presented in Table 1-2. The table provides values for (J "as'"ed b on radioactive decay through November 1,1997. Since shipmen summer of 1999, these activities are conservative and bounding. The RVP is estimated to contain 155 curies ofinner surface activity and 2,010,000 non-releasable curies of activated metal.' The surface contamination activity is contained on the interior surfaces of the reactor  !

vessel and on the surfaces of the reactor internals. The RVP will be filled with LDCC to provide i additional shielding, prevent contaminant movement thereby preventing changes in extemal dose .

rates, and provide an additional barrier against release in the unlikely event of a package  !

containment breach.

1.2.3.1 Surface Contamination Activity Approximately 155 curies of material is associated with activity distributed on the surface of the reactor vessel intemals as well as the interior surface of the reactor vessel. The total surface activity in the reactor vessel and intemals was calculated based on the isotopic distribution of the material, actual radiation dose rates, and calculated surface areas.

The isotopic distribution for the Steam Generators (S/Gs) was utilized in this evaluation. As presented in the Safety Analysis Report for Steam Generator Packages (Reference 1-2),

previously removed U-tube segments from the S/Gs were analyzed by two independent off-site laboratories to determine the fractional percentages of the radionuchdes. Radiation dose rates 1 - 16

Trolan Reactor Vessel Packaze- Safety Analysis Rervort A

V were measured on a 9' length of Reactor Coolant System (RCS) piping. These dose rates and the steam generator isotopic distributions were then used in a Microshield 3.13 computer model to determine the total radioactivity for each square meter of surface area. The surface areas of.

the reactor vessel and internals were calculated and then multiplied by the activity per square meter to determine the total surface contamination activity. The calculation by which the total surface contamination activity was determined is provided as Appendix 4-3.

Westinghouse Report WCAP-12932, Revision 1, " Full Reactor Coolant System Decontamination Program Generic Topical Program Report," Volume 2, January 1992, indicates that for typical Pressurized Water Reactors (PWRs) the isotopic distribution is similar throughout the RCS, including the S/G tubes, out of core surfaces and the surfaces in the reactor vessel. The report is based on a broad range ofanalyses performed on samples obtained from the RCS of a number of PWR plants. The types of samples include Inconel S/G tube surfaces, Inconel S/G tubes, crud from S/G manway inserts, S/G channel head smears, S/G channel head decontamination products, reactor vessel stainless steel surfaces, and out of vessel RCS surfaces.

According to WCAP-12932 the magnitude of corrosion product surface contamination is greater on stainless steel than on Inconel (Trojan S/G tube material) surfaces by a factor of approximately 15. 'lle report also discusses the deposition of fission products and transuranics (TRU) and concludes that these radionuclides would be deposited similarly on Inconel and g stainless steel. The WCAP indicates that approximately 15 times more activity is deposited per unit area or, stainless steel than on Inconel. The primary contributor to the public dose from a transportation incident is Co-60. Co-60 is also the primary gamma emitter used to calculate the areal distribution based on the dose rate at a given distance from the RCS pipe. The actual levels of Co-60 deposited on the pipe are calculated as part of the pipe model. The fission product and TRU nuclides are then calculated based on the Co-60 activity. The limiting nuclides for public dose during a transportation accident are Co-60 and TRU. Since the level of Co-60 is accurately modeled based on RCS pipe dose data and the TRU nuclides are estimated based on the ratio of Co-60 to TRU from the Inconel tubes, the Trojan RVP inner surface contamination estimate is valid.

It is noted that the use of the S/G tube sample isotopic data as representative ofrelative distributions throughout the RCS is not impacted by the S/G tube cracking that contributed to the permanent shutdown of the Trojan Nuclear Plant. That cracking was on the outside (secondary side) surface of the S/G tubes and did not penetrate to the inside (primary side) of the tubes.

Trojan calculations assume that the deposition of radioactivity on the stainless steel RCS pipe is the same as on the reactor vessel stainless steel internal surfaces. The basis for this assumption is that:

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t

t i

Trolan Reactor VesselPackaee- Safety Analysis Report O

1. The same radioactive fluid is circulating at high flow rates (low resident time at  !

any location)in the RCS system,

2. The chemical environment is the same throughout the RCS system,

3. The temperatures of the stainless steel components are similar, and

4. No phase change of the RCS fluid occurs.

" Individual hot spots due to flow changes or crud traps may be found in the RCS; however, the average contamination level is reasonably used for purposes of this calculation.

The S/G isotopic distribution used for the estimation ofreactor vessel corrosion layer activity did l

not include a measured uranium value. In order to determine the amount of uranium present, results from a steam generator smear sample were analyzed. Although U-235 and U-238 were below the detection limits, U-234 was detected. The uranium activity on the S/G tube was  !

estimated by determining the ratio of U-234 to Pu-239/240 in the smear sample, and applying the l same ratio to the measured Pu-239/240 activity in the S/G tube analysis. Th: result was an

estimated 4.54E-04 microCi of U-234 of the tube sample. Based on this result, the amount of '

U-234 contained in the reactor vessel corrosion layer was calculated to be a total of 7.02E-03 j Os curies, using the same methodology described previously in this section The 7.02E-03 Ci surface contamination value for U-234 is small compared to the total values listed in Table 1-2 of 155.2 curies for surface contamination and 2.01E+06 curies for activation l

[ activities. The addition of U-234 has a negligible effect on the thermal, containment and  !

shielding analyses. The calculation which includes determination of the U-234 in the corrosion 4' layer is in Appendix l-4. '

1.2.3.2 Activation Activity '

Approximately 2,010,000 curies of activated metal is present in the RVP. These activation i products do not present a risk to the public in the form of radioactive release since the radioactive material is contained within the metallic structure of the components and is, therefore, ,

non-releasable.  !

L The activation radioactivity was determined through calculations performed by TLG Services, Inc. (TLG) in support of the Trojan Nuclear Plant Radiological Site Characterization Report (Ref.1-3). These calculations consisted of one-dimensional neutron transport and point neutron activation analyses of the reactor vessel and its internals. These calculations were performed using the FISSPEC and 02 FLUX computer codes written by TLG and the ANISN O 1 - 18

}

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

Troian Reactor VesselPackage- Safety Analysis Rer> ort O

and ORIGEN2 computer codes, obtained through the Oak Ridge National Laboratory's 4 Radiation Shielding Information Center. Reduction of the output from these programs and ancillary calculations were performed using the ANISNOUT and O2 READ computer codes, 4

written by TLG, and the Microsoft EXCEL computer program.

The neutron-induced radionuclide inventories were estimated using a two-step analytical approach. The first step was to determine the magnitude and spectrum of the neutron flux beyond the boundaries of the reactor core. This was accomplished using the ANISN one-dimensional neutron transport computer code with five radial and axial geometric models.

The results of the radial transport calculations were normalized agains' plant-specific neutron flux data obtained from an available reactor vessel neutron fluence surveillance capsule report.

The ANISN outputs were subsequently collapsed into two-energy group formats (fast and thermal) and into a series of ORIGEN2 point activation / depletion calculations. Additional input to the ORIGEN2 calculations included material compositions and historical plant performance data. The results of the calculations are contained in Table 1-2, which lists each isotope and its estimated curie contribution. These values are used as the source terms in determining external dose rates in Chapter 5 and as the source of package decay heat used in the thermal analysis of Chapter 3.

Table 1-2 also provides a comparison of each nuclide to the A 2 limit provided in 10 CFR 71, !

Appendix A and the fractional percentages of these limits. These results show that the sum of fractional percentages is greater than one. Therefore, the package exceeds the Type A quantity of radioactive material as defined by 10 CFR 71.4. Therefore, the package contains a Type B quantity of radioactive material.

k

• Trolan Reactor VesselPackage- Safety Analysis Report O

\J

1.3 REFERENCES

I l-1 Trojan Nuclear Plant Decommissioning Plan (PGE-1061) Rev. 2 - Portland Gen eral Electric, December 19,1996 i l-2 Trojan Nuclear Plant Safety Analysis Report for Steam Generator Packages Rev.1 -

l Portland General Electric, June 15,1995 1-3 . Trojan Nuclear Plant Radiological Site Characterization Report Rev. 0.1 - Portland l General Electric, February 8,1995

.I l

l l

O i 1

l 1

I I

i 1 - 20

Trolan Reactor VesselPackaee - Safety Analysis Report o

Table 1-1 , j 4

Reactor Vessel Package Closure Materials I

Figure Quantity Description Specification ' Penetration l-4 Description Sheet No.

2 4 Plate,2%" x 30" Diameter ASME SA-240 Reactor Inlet Nozzles l Type 304L i

3 4 Plate,2K" x 31%" Diameter ASME SA-240 Reactor Outlet Nozzles l

. Type 304L  !

4 1 Plate, S/8" x 3" Diameter ASME SA-516 Lower Head Drain Hole Gr. 70' l l

5 58 Plate, S/8" x 2%" Diameter ASME SA 516 lacore Instrumentation 1 Gr. 70' Tubes l

l 6 i Plate,5/8" x 3" Diameter ASME SA-516 Head Vent 1 Gr. 70' 7 78 Plate,5/8" x 6%" Diameter ASME SA 516 CRDM Tubes Gr. 705 8 2 Plate,5/8" x 3%" Diameter ASME SA-516 Flange Monitoring Tubes Gr. 70' NA 54 Stud,7" Diameter SA 540 Grade B 24, Seal Upper Head to Reactor Class 3 Vessel NA 2 O-Ring Seal, inner 178.4" J sconel Alloy 718 Seal Between Upper Head OD, outer 181.0" OD (Silver Plated) and Reactor VesselFlange NA NA Reactor Nozzle Closure E308L2 or equivalent Inlet and Outlet Nozzles ,

Welds Weldin Consumables) i NA NA Welds for %" Thick Closure E70xx2 or equivalent) All Other -

Plates Weldin Consumables Reactor Vessel Openings Notes. l

1. ASME SA 516 Gr. 70 material shall be normalized and made to fine grain practice. j

2. Equivalent signifies equal or greater yield and ultimate strengths, and weldability.

Trolan Reactor VesselPackare- Safety Analysis Report Table 1-2 Reactor Vessel Package Radioactivity and A2 Fractions i

Nuclide Activity ' A2 Limit Fraction )

. Surface (Curies) . A 2Value Contamination Activation (Curies) (Curies) {

4 i

H-3 8.llE-02 6.55E+02 1080 6.07E-01  !

C-14 1.15E-01 2.I8E+02 54.1 4.03E+00 Sb-125 1.67E+00 4.04E-01 24.3 8.53E-02 I Ce-144 4.64E-02 5.41 8.58E-03 Mn-54 8.57E-02 2.16E+03 27.0 8.00E+01 Eu-152 2.54E+01 24.3 1.05E+00 Fe-55 2.77E+01 6.97E+05 1080 6.45E+02 Co-60 9.92E+01 1.ISE+06 10.8 1.06E+05 i Ni-59 9.53E+02 1080 8.82E-01 Ni-63 2.00E+01 1.57E+05 8II 1.94E+02  :

Nb-94 3.29E+00 16.2 2.03E-01  ;

i Sr 90 9.24E-01 2.7 3.42E-01 1 Tc.99 7.06E-01 24.3 2.91E-02 Pu-238 8.35E-02 0.00541 1.54E+01 i Pu-239/240 9.33E-02 0.00541 1.72E+01 Pu-241 5.07E+00 0.27 1.88E+01  !

Pu-242 4.70E-04 0.00541 8.69E-02 l Cm-242 1.56E-05 0.27 5.78E-05 Cm 243 3.60E-02 0.00811 4.44E+00 ,

Cm-244 3.41 E-02 0.0108 3.16E+00 Am-241 1.10E-01 0.00541 2.03E+01 Total 2 155.2 2.01 E+06 1.08E+05 Notes: 1. Activity values have been decayed to 11/01/97.

2. Total does not include the contribution from U-234 (7.02E-03 Ci surface contamination and no activation), which is negligible.

O

Trolan Reactor VesselPackare- Safety Analysis Report i O 2.0 STRUCTURAL EVALUATION This chapter presents the structural analyses of the Reactor Vessel Package (RVP) that are required to meet the requirements for approval as a Type B (as exempted), exclusive use shipping package as required by 10 CFR 71.

2.1 - STRUCTURAL DESIGN The components making up the RVP are evaluated and shown to meet the requirements of 10 CFR 71. The analyses address both Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC).

. Ihis section identifies and describes the components that comprise the RVP containment -

boundary and shielding. A discussion of the impact limiters is also included since they are required to ensure containment boundary integrity is maintained during the analyzed drop scenarios. Structural design parameters for these components are discussed with references to drawings provided where appropriate.

Results of the analyses are summarized in Tables 2-1 and 2-2 for NCT and HAC, respectively.

2.1.1 STRUCTURAL DESCRIPTION The principal components of the RVP include:

1. Reactor vessel shell,

2. Reactor vessel upper head,

3. Reactor vessel head studs,

4. , Reactor vessel flange O-rings,

5. Welded penetration closures,

6. Low density cellular concrete (LDCC) injected into reactor vessel, 7.- Shielding,

8. Impact limiters.

2-1

Tro!an Reactor VesselPackam Safety Analysis Report h g

V 2.1.1.1 Reactor Vessel Packnoe Containment Boundarv The containment boundary of the RVP consists of the reactor vessel shell with reactor vessel upper head and flange O-rings installed and tensioned by 54 reactor vessel head studs, and the welded penetration closures.

The reactor vessel shell and upper heads were constructed in accordance with ASME Boiler and Pressure Vessel Code Section III Class I (1968 edition with Addenda through winter of 1968) and ma.'ntained in accordance with the ASME Boiler and Pressure Vessel Code Section XI. The reactor vessel was designed for a rated pressure and temperature of 2485 psig and 650*F, respectively.

Each of the containment boundary components is discussed in the following subsections.

2.1.1.1.1 Reactor Vessel Shell The reactor vessel shell consists of a hemispherical bottom head, three shell rings, eight nozzle assemblies, and a shell flange assembly. The overall reactor vessel dimensions and configuration are shown on Figure 1-3. A general description of each component of the reactor vessel shell p including typical penetrations and the internals is provided below. I d

The bottom head region is hemispherical in shape and constructed of SA-533 Gr. B Class 1 carbon steel. The interior surface of the bottom head region has a minimum 5/32" thick stainless steel weld overlay. Wall thickness of the bottom head is 5.5" (including the overlay). The bottom head contains 58 penetrations for instrument tubes. One additional penetration will be added at the bottom of this region to facilitate water drainage prior to filling the interior void space with LDCC. See Figure 1-4 for penetration closure details. The bottom head is welded to the first shell ring.

The beltline region of the reactor vessel shell is located between the bottom head and nozzle assembly area of the reactor vessel shell. This region consists of the first and second shell rings.

Each of these shell rings is approximately 108" in length,9.5" thick, and has an inside diameter of 173". The shell rings are constructed of SA-533 Gr. B Class I carbon steel and the interior surface have a minimum S/32" thick stainless steel weld overlay. The first shell ring is located between the bottom head and second shell ring. The second shell ring is located above the first shell ring and below the nozzle assembly area. There are no penetrations in these sections of the reactor vessel shell.

The nozzle assembly section of the reactor vessel shell is located between the second and third shell rings. The dimensions and materials of construction for the nozzle assemblies are provided O

V 2-2

- - . - . - - - - . - . . - . . . . .- - - ~ - .- - -.- - . - .. - . -

i i

Troian Reactor vesselPackare- Safery Analysis Reoort A

sg -

in Table 2-3.

1 The third shell ring is located between the nozzle assembly section and the shell 11ange assembly.

This section of the reactor vessel shell has a nominal diameter of 170.9" and is 11.1" thick. The interior surface has a minimum sa2" thick stainless steel weld overlay.

The shell flange assembly is located at the top of the lower rea : tor vessel shell above the third ,

shell ring. The shell flange assembly contains a gasket seating surface and 54 threaded holes.  ;

Two metallic O-rings are installed between the upper head Cange and shell flange on the O-ring seating surface. The shell flange assembly contains two penetrations that allowed leak  ;

monitoring of the O-ring seals during plant operation. One penetration is located between the  !

two 0-rings with the other penetration located outside the outer 0-ring. These penetrations are i referred to as the inner and outer monitoring tubes respectively. The interior surface of the l flange assembly has a minimum sa2" thick stainless steel weld overlay. See Figure 1-4 for details of the penetration closures.

2.1.1.1.2 Reactor Vessel Upper Head i

The reactor vessel upper head consists of the upper head flange assembly and a hemispherical 3

upper head. The reactor vessel upper head components are constructed of SA-533 Gr. B Class 1 g, material and the interior surfaces have a minimum sn2" thick stainless steel weld overlay. The shell flange and upper head flange are keyed to ensure proper orientation and alignment of the 54 stud holes. Two 0-ring grooves are provided in the upper head flange to maintain 0-ring placement and ensure proper sealing of the flange surfaces. The upper head flange is approximately 19" thick in the area of the head bolt holes and 7.2" nominal thickness where it joins the upper head. The head stud holes are approximately 7.5" in d4 meter. The upper head has a nominal thickness of 7.2" and contains 79 penetrations. There are 78 control rod drive mechanisms (CRDM) housing penetrations and one upper head vent. See Figure 1-4 for details of the penetration closures.

2.1.1.1.3 Reactor Vessel Head Studs Fifty-four head studs will be used to tension the upper head to the reactor vessel shell. These head studs are approximately 57.5" in length and have a nominal diameter of 7". The studs are constmeted of SA 540 Gr B 24, Class 3 steel.

To provide tensioning of the upper head to the reactor vessel shell,8.25" tall cylindrical nuts and spherical washers are used. The cylindrical nuts and spherical washers are constructed of SA-540 Gr B 24, Class 3 material.

2-3

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

t

. Troian Reactor vesselPackare- Safety Analysis Report

o The head studs will be tensioned to 720,000 lbs in accordance with a procedure.

2.1.1.1.4 Reactor Vessel Flange O-Rings 1 I

.Two O-rings are installed in the O-ring grooves of the upper head flange assembly. These O-rings are constructed ofInconel Alloy 718 with a silver plating. I 2.1.1.1.5 Welded Penetration Closures i

' A total of 148 welded closures is required to establish the containment boundary of the reactor vessel package. The majority of the closures are located on the upper and lower heads. These closures consist of round steel plates, %" thick, which are located over the existing penetrations and are welded to the heads. Eight large closures seal the inlet and outlet nozzle openings.

These 2%" plates are welded to the nozzle walls as shown in Figure 1-4 Closure welds will be performed in accordance with the methods and guidance of ASME Section III Subsection ND. ,

These welds will be examined by magnetic particle or liquid penetrant method in accordance with the methods and guidance of ASME Section III, Subsection ND.

Figure 1-4 provides details of the reactor vessel penetrations and closure raethods.

2.1.1.2 Low Density Cellular Concrete

)

Prior to removing the reactor vessel from containment, it will be injected with LDCC. The LDCC is composed of Portland cement, and a foaming agent to give it a low density. No aggregate will be used. LDCC will fill the interior void space of the entire vessel including the upper head; encapsulating the reactor internals and providing additional shielding inside the reactor vessel. The LDCC will also fix the contaminants in place on the vess el interior surfaces and will prevent shifting of contaminants within the package thereby stabilizing dose rates during transport.

The LDCC will be injected into the reactor vessel utilizing a combination of cold and hot leg I nozzles and Control Rod Drive Mechanism (CRDM) penetrations and vented nrom the upper l head of the reactor vessel using CRDM penetrations and the head vent. The reactor vessel will be filled in its installed orientation, using a flow path for the lower section of tl e vessel similar to

. the flow path of reactor coolant during normal plant operations. A cooling system will be installed prior to grouting of the vessel to lower the temperature of the vessel to ensure proper curing of the LDCC.

A concrete batch plant will be set up on site to mix the LDCC for injection into the reactor vessel. Prior to injection, the density of the LDCC will be measured to ensure it is between 45 2-4

Troian Reactor Vessel Package . Saptv Analysis Rer> ort @

and 65 lb/ft). This density was selected based on its flow characteristics and its stable properties during the injection and curing process. The LDCC injection will be completed in two lifts. The first lift will be injected through a vessel inlet / outlet nozzle until the grout is observed at the vent installed on another inlet / outlet nozzle. This will fill the vessel from the bottom up to the level of the inlet nozzles. The second lift is delayed approximately four days to ensure the first lift has i

properly cured and to ensure its density will not be affected by the second lift. The second lift will be injected into an inlet nozzle, an outlet nozzle, and a spare CRDM penetration until there is a solid and continuous vent of LDCC observed from another CRDM vent penetration and the head vent. This will ensure the vessel is filled with LDCC. The vessel will remain vented for 28 days to allow the excess moisture from the concrete to escape and the vessel and concrete to reach equilibrium conditions. The bottom penetration on the vessel will then be opened to ensure there is no free water in the vessel. Following removal of the injection and vent temporary covers, the penetrations will be inspected to ensure there are no voids. Any noted voids will be filled with LDCC prior to installation of the permanent transportation closures.

The LDCC contractor for the Trojan RVP project was selected based on his participation in the LDCC injection of the Trojan steam generators and pressurizer. All five Trojan vessels were successfully filled with LDCC with tighter density tolerances and more difficult process restrictions. The LDCC contractor has extensive tunnel and Slipline experience involving strict density, curing, and flow characteristic management. The Slipline and tumiel experience has involved large volume pours of extremely small annular spaces with tight density, weight, compressive strength, and cure time restrictions. The LDCC contractor holds many patents on processes and LDCC chemistry. These provide additional confidence that the vessel will be I filled as required and that the density will be in the required target range.  !

l The contractor recommended the density range of 45 to 65 lbs/ft' based on his experience and resultant confidence with the predictability of the material's rheological behavior (i.e., density and viscosity). In this density range, the material behaves very much like water, and it flows i naturally around internal supports to fill void spaces without pressurization.

The LDCC contractor and Portland General Electric Company (PGE) have participated in ajoint testing program to prepare for the LDCC injection. Based on the testing program results, a two-lift approach to the process was established, the density changes expected during the process were verified to be acceptable, and a maximum lift height of 30 feet was established. In addition, an evaluation of the thermal response of the reactor vessel was performed to determine the effects of the decay heat and of the heat of hydration on the grouting process and the reactor vessel. The evaluation results were utilized to establish the cooling system requirements, the time period length between the lifts, and the time peried length for the final vent. In addition, it also determined that the maximum temperaturea developed during the grouting process until equilibrium conditions were established were less than the reactor vessel design temperature.

A b 2-5

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Trolan Reactor VesselPackare - Safety Analysis Report l

l 2.1.1.3 External Shieldino 10 CFR 71.47 and 10 CFR 71.51 establish external radiation standards for all packages. The  ;

AVP will be a Type B (as exempted), exclusive use package. Chapter 5 discusses the anslyses >

that were performed to determine the shielding requirements.  !

4-  :

The dose rate evaluations determined that shielding is required on the RVP. The shielding L consists of eight 2" thick plates with penetrations to allow fitting over the reactor vessel inlet and

! outlet nozzles, three 5" thick plates between the nozzles and the lower head, and a 1" by 3' lower

[ skirt located around the upper portion of the lower head. Figure 1-5 provides details of the

{'

shielding installation. Additional shielding may be necessary to address localized areas of high .

dose rates. This determination will be b'ased on surveys performed in a low radiation background

after the vessel is filled with LDCC. Shielding welds will be performed in accordance with the i methods and guidance of ASME Section VIII. The shielding welds will be inspected in

, accordance with the methods and guidance of ASME Section VIII. -

The shielding has been structurally designed to ensure the RVP meets the requirements for the ,

NCT and the HAC. The main section of shielding (2" and 5" shielding section) is designed to remain attached for both conditions. The remaining shielding, the lower head skirt and any q localized shielding, is designed to stay attached for the NCT. The analysis for the structural b attachment of the shielding (Appendix 2-10) addresses both the nominal shielding configuration and additional shielding which may be added based on the survey results. The actual shielding configuration will be verified to ensure the package does not exceed the design weight limit or result in a shift of the center of gravity (CG) beyond the design range specified in Section 2.2.

The Manager, Radiation Protection will ensure that the final configuration meets the applicable dose rate requirements of 10 CFR 71.47 and 10 CFR 71.51, and the Manager, Engineering will j ensure that the final configuration meets the over-all package design weight and center of gravity l requirements.

2.1.1.4 Imoact Limiters Protection of the vessel during the specified impact events will be provided by a cylindrical impact limiter at each end of the RVP. Each limiter surrounds a 58-inch length of the vessel and has an outer diameter of 336" as shown on Figure 1-2. The limiters will be constructed from closed cell polyurethane foam, Last-a-Foam FR-3720 manufactured by General Plastics,Inc. A nominal density of 20 lb/ft' is required. The external surfaces of the foam are encased in a sheet steel envelope. This arrangement of energy absorbing material limits the impact accelerations to acceptable values to meet the requirements of 10 CFR 71. The impact limiters are secured to the reactor vessel by brackets welded to the external shield plates. The impact liraiters are attached 2-6

Trolan Reactor VesselPackare- Safety Analysis Report

(}

to the brackets by through bolts which extend completely through the limiters. No attachments are welded to the reactor vessel. The impact limiter casing material is %" SA-516 Gr 70. Impact limiter details are shown on Figure 1-6.

2.1.2 DESIGN CRITERIA Package approval standards are identified in 10 CFR 71 Subpart E. Demonstration of compliance with these standards is specified in 10 CFR 71.41. Since the RVP will be used to transport Type B quantity, it must be evaluated for both the NCT (10 CFR 71.71) and the HAC (10 CFR 71.73). Environmental and test conditions different from those specified by 10 CFR 71.71 and 10 CFR 71.73 may be approved by the Commission if the controls proposed by the shipper are demonstrated to be adequate to provide equivalent safety of the shipment (10 CFR 71.41(c)). Because of the unique nature of this shipment, specific conditions or exemptions from 10 CFR 71 are being sought and are discussed in Section 1.1.1.

Compliance with the " General standards for all packages" specified in 10 CFR 71.43 is discussed in Section 2.4. Section 2.5 describes requirements associated with lifting and tiedown standards.

The NCT is discussed in Section 2.6 and Section 2.7 discusses the HAC.

7 The structural analyses used in determining compliance with the requirements of 10 CFR 71 are

( performed in accordance with the guidance of Regulatory Guide 7.6, " Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels." The load combinations used in these analyses are treated in accordance with the guidance of Regulatory Guide 7.8, " Load Combinations for the Stmetural Analysis of Shipping Casks for Radioactive Material."

2.1.2.1 Structural Analysis Criteria Analyses were performed on the RVP to ensure that during the NCT and the HAC, that containment boundary integrity is maintained. The general structural criteria for the NCT are provided in 10 CFR 71.43(f). These criteria state:

"A package must be designed, constructed, and prepared for shipment so that under the tests specified in s71.71 (Normal Conditions of Transport) there would be no loss or dispersal of radioactive contents, no significant increase in external radiation levels, and no substantial reduction in the effectiveness of the packaging."

For the HAC,10 CFR 71.51 allows for a limited release of radioactive contents and a slight increase in package dose rates. For conservatism, the RVP design has been analyzed to demonstrate that even under the HAC, there would be no loss or dispersal of the radioactive 2-7 l

l

2 Troian Reactor vesselPackage - Safety Analysis Report contents.

The limits for stresses and fracture toughness'for the RVP containment boundary are discussed in

. the subsequent sections.

2.1.2.2 Allowable Stresses Stresses in the RVP satisfy' Regulatory Guide 7.6, Design Criteriafor the Structural Analysis of Shipping Cask Containment Vessels. In addition, for NCT load cases, stresses satisfy the ASME B&PV Code,Section III, NB-3000, design and service level A. The following component-specific stress limits are satisfied.

j i

For the NCT:

Reactor Vessel Shell and Closure Plates Allowable primary membrane stress intensity (P ) is S,.

Allowable primary membrane plus bending stress intensity (P. + P3 ) is 1.5S.

Allowable range of primary plus secondary stress intensity (P + P + Q) is 3S.

1 Upner Head Closure Studs

. Allowable average stress intensity is 2S, per Section NB-3232.1.

' Allowable maximum stress intensity is 3S per Section NB-3232.2.

Closure Plate Fillet Welds Allowable stress values for closure plate fillet welds are 0.49S per Section ND-3356.l(c) and ND-3359(b). S is the allowable base metal stress in tension oer ND-3321-1 and is taken from ASME B&PV Code,Section II, Part D.

Shielding Comoonents Allowable stresses for the primary membrane stress and primary membrane plus primary bending are as defined in ASME Section VIII, Figure 4-130.1, Stress Categories and Limits of Stress Intensity. These values are S., = S and S . = 1.5 x S. for the primary membrane stress and primary membrane plus primary bending respectively. These allowables are also consistent with Regulatory Guide 7.6, position 2.

Shielding Welds Allowable stress for welds is defined in accordance with ASME Section VIII, ,

Division 2, paragraph AD-920. This allowable is defined as S . = (Joint Factor)x S for the material in question, and as defined in ASME Section II D.

2-8

' Troian Reactor VesselPackare- Safety Analysis Recort i O l ' ;%)  :

l 1

For fillet welds, thejoin: factor is 0.5 and for grove welds it is 1.0. For example, i

, for SA-516 Gr. 70 material, the S.no, = 0.5 x 23300 = 11.65 ksi for a fillet weld. '

Stmsses resulting from thermal effects or other self-limiting stress are secondary stresses and per  !

Regulatory Guide 7.6 are considered only for NCT.

For HAC:

L <

- Reactor Vessel Shell. Closure Plates. and Unner Head Closure Studs -

~

t Allowable primary membrane stress intensity (P.) is the lesser of 2.4S, or 0.7S,. 'i Allowable primary membrane plus bending stress intensity (P,, + P.) is the lesser I

- of 3.6S, or S,. i Closure Plata Fillet Welds f

Allowable stress in filler welds is based on minimum weld throat area, and is treated as a primary stress. . >

Allowable stress intensity (P ) is the lesser of(0.6)2.4S or (0.6)0.7S, and ND-3359(b).  !

Shieldino Comnonents  !

Stress allowables are based on the guidance provided in Regulatory Guide 7.6, position 6 for accident conditions. This guidance states that primary membrane ,

stress and primary membrane plus primary bending allowables are based on the lesser of 2.4S, and 0.7S, for primary membrane; and 3.6 S, and S, primary L

membrane plus primary bending. For the materials used in this analysis, the allowables based on S, are always the " lesser" value and controlling for the ,

evaluation.

Shieldina Welds I For HAC conditions, the allowable stress for welds under shear load with no bending component is 0.6 x 0.7 x S,. The allowable stress of a weld subject to '

shear and bending under these conditions is 0.6 x S,.

Table 2-1 provides a summary of the allowables for NCT and Table 2-2 for the HAC.

2.1.2.3 Fracture Touchnen Part 71 of Title 10 of the Code of Federal Regulations, " Packaging and Transportation of Radioactive Materials," requires that packages used to transport radioactive material be designed l . with consideration of normal transport and hypothetical accident events that might occur at i

i i O 2-0 i

Trolan Reactor VesselPackage- Safety Analysis Report i

temperatures as low as -20 F (-29'C). In order to assess the potential for brittle fracture of the Trojan vessel under transport accident conditions, hypothetical side drop and puncture load conditions were analyzed to determine the appropriate stresses. These analyses are described in I Sections 2.7.1 and 2.7.3. The concerns for brittle fracture of the vessel are addressed in  !

Appendix 2-12 using an ASME Code Section XI, Appendix A fracture mechanics approach. I l

Linear clastic fracture mechanics analyses are normally used to evaluate brittle fracture concerns l

in reactor vessels during normal operation and accident (i.e., emergency and faulted) conditions.  !

The approach defined in ASME Section XI, Appendix G specifies the assumptions to be used in the brittle fracture analysis for detennining the allowable operating pressure-temperature limits i for heatup and cooldown. The Appendix G procedure establi.ches safety margins to prevent crack  !

initiation during normal heatup, cooldown, and hydrostatic test operation. For severe emergency  ;

and faulted conditions, more detailed fracture mechanics analyses are required and altemative j

acceptance criteria are available in ASME Section XI, Appendix A. These ASME Code requirements assure adequate margins against brittle fracture of the reactor vessel for operating plants. I 1

l No specific design criteria are provided in the regulations, however, for protecting a shipping container against brittle fracture initiation and unstable growth. Studies conducted by Lawrence O and the recommendations for brittle fracture prevention contained in th '

publication of Regulatory Guide 7.12. The guidance in Regulatory Guide 7.12 is applicable only >

to ferritic steels with extremely low nil-ductility transition temperatures; however, NUREG/CR-3826 provides recommendations for an alternative approach based on ASME Section XI brittle fracture evaluation methods. The criteria utilized in Appendix 2-12 center around the concept of no fracture initiation based on the allowable flaw sizes specified in IWB-3510-1 of Section XI of the ASME Boiler and Pressure Vessel Code. The calculated stresses resulting from the analyzed scenarios are used in this analysis. The impact stress levels are less than yield for these

~

conditions.

Other methods were investigated to benchmark the adequacy of this approach. In particular, the IAEA "Guidelinesfor Safe Design ofShipping Packages Against Brittle Fracture " describes three acceptable methods for preventing brittle fracture of radioactive material transport packagings. One method can be referred to as a fracture mechanics approach. This method is used to demonstrate, by analysis or by a combination of analysis and test, that sufficient margin is available (considering applied stress, flaw size, and/or material fracture toughness) to preclude crack initiation and brittle fracture during design-basis events. This is equivalent to the method utilized in Appendix 2-12. Furthermore, this method provides three acceptable options for determining material fracture toughness. One of these options defines a lower-bound or near (statistically) lower-bound value for the fracture toughness at the lowest service temperature.

2 - 10 e r- w*- ,-. -- 3 - , w w - . - -

. r.- . --, .-,,--,,-.-.

1 i

Trojan Reactor Vessel Packaee - Safety Analysis Report A

V This definition of material fracture toughness is inclusive of the dynamic (Ku) reference material

)

fracture toughness curve for ferritic steels assumed in this analysis, and for this reason this option ,

'is noted to coincide with the present analysis approach. l With respect to safety factors used in this analysis, no specific safety factors are given in ,

- NUREG/CR-3826. However, the IAEA guidelines suggest that, when using the applied method, j an overall minimum safety factor (K i /K )i of approximately 3 is appropriate for normal loading '

conditions during shipping, with a suggested safety factor of 1.4 for unexpected loading events such as hypothetical accident conditions.

The results from the fracture mechanics calculations using stresses derived from the finite element analysis show significant margins against brittle crack initiation (i.e., iK /K i > 1.0) under all evaluated conditions. In general, stress levels in the vessel are sufficiently low so that no crack initiation is predicted to occur for the hypothetical side drop loading condition. This is due to the ability of the impact limiters to reduce the amplitude of the longitudinal beam bending stresses relative to ovalization and localized bending stresses. The membrane stress contribution is small regardless of the assumption for internal pressure (< 200 psi) near the impact locations and at geometric discontinuities (e.g., at the comer of the outlet nozzle). At an ambient temperature of -20*F, large margins against crack initiation (> 5.65) are calculated for all locations along the shell with the smallest margins (2.0) occurring close to the outlet nozzle. The other points of high stress concentration are at the contact locations with the impact limiters where the calculated margins are 2.51 and 3.17, respectively. ,

The axial stresses throughout much of the beltline region are very low (less than 10 ksi tension) or even compressive. Ovalization effects at the mid-core region are present and some tensile l hoop stresses develop in these regions. However, the safety factor against brittle crack initiation at the most embrittled region of the vessel is greater than 7.0. Because of this, the beltline region l of the Reactor vessel is not considered to be a limiting concern for a hypothetical drop condition. i A more limiting condition is the hypothetical puncture loading. A safety factor of 1.01 and 1.0 was calculated at the upper and lower heads, respectively, at an ambient temperature of-20*F.

Although these values are relatively low, it should be noted that the calculations are based on very conservative assumptions. For example, use oflarge assumed full-circumference or quarter-circular surface flaws at the worst stress locations and use oflower bound dynamic reference toughness instead of using the static reference toughness add to the conservatism of the analysis results. In addition, the inconel welds at the head penetrations have not been modeled

~ separately, and the toughness properties of the adjacent base metal were assumed for the fracture mechanics analyses. The Inconel nickel base alloys are known to have excellent toughness at low temperatures because of their strength and ductility. Furthermore, Inconel alloys do not undergo a ductile-to-brittle transition in toughness, even at very low temperatures. A flaw r~

2 - 11

l Trojan Reactor Vessel Packaee- Safety Analysis Report remaining within the Inconel overlay would not experience brittle crack initiation. Therefore, assuming dynamic toughness of the ferritic base metal properties is very conservative.

Based on the worst case assumptions utilized, the transportation route's lack of hard targets, the limited orientations for the analyzed drop conditions, and the overall extremely low probability  ;

of an accident, this is considered acceptable. However, an analysis was completed based on a  !

minimum vessel wall temperature of 45'F, and the safety factor increased by approximately 40%. This resulted in a safety margin of at least 1.4 against crack initiation.

This brittle fracture analysis has been provided to the Electric Power Research Institute (EPRI) {

Major Component Reliability group for their review. Their review determined that the input assumptions and the fracture mechanics methodology have a well founded technical bases and 3

the results demonstrate adequate safety factors for the loading conditions postulated.' They also '

noted that the overall conclusion of the analysis is conservative.

Additional information is provided in Sections 2.7.1.4 and 2.7.3.4.

2.1.3 DESIGN CRITERIA FOR TIEDOWN SYSTEM

(

10 CFR 71.45(b)(1) provides design requirements for tiedown devices which are a structural part of the package. These design requirements are:

"If there is a system of tiedown devices which is a structural part of the package, the system must be capable of withstanding, without i

generating stress in any material of the package in excess ofits yield strength, a static force applied to the CG of the package having a <

{

venical component of two times the weight of the package with its contents, a horizontal component along the direction in which the vehicle travels of 10 times the weight of the package with its contents,  :

and a horizontal component in the transverse direction of five times the weight of the package with its contents."

The tiedown system has been designed based on the above criteria. The nozzles are considered to be part of the tiedown system in that they are used to prevent longitudinal motion of the RVP.

The description of the tiedown analyses and results are given in Section 2.5.2.

2 - 12

i Troian Reactor vesselPackaee- Safety Analysis Report o 2.1.4 ' DESIGN CRITERIA FOR LIFTING DEVICES .

i The requirements of 10 CFR 71.45(a) are as' follows:

... Any other structural part of the package which could be used to liA

' the package must be capable of being rendered inoperable for lining the package during transport or must be designed with a strength i equivalent to that required for liAing attachments."

In accordance with the requirements of 10 CFR 71.45(a), other structural parts of the package -

that could be used for liAing the package during transport will be rendered inoperable prior to  :'

shipment. The reactor vessel head liAing lugs (see Figure 1-3A, Sheet 1 of 2) will be disabled prior to shipment. The vessel nozzles and studs are not considered structures which could - '

normally be considered liR points but could provide attachment locations. This is an exclusive i use shipment and use of these locations will be prevented by administrative control for all liAing '

operations.

2.2 WEIGHTS AND CENTER OF GRAVITY i

. ' The design weight of the RVP, including vessel intemais, maximum shielding, LDCC I

, (assuming 65 lbs/A2), and impact limiters is 2,040,000 lbs. Itemized weights of the various  ;

. components are provided in Table 2-4.

.d Results of a sensitivity analysis described in Appendix 2-1 indicate the total package weight will .

be bounded by a minimum of 1,904,438 lbs. and a maximum of 2,025,898 lbs. A conservative upper bound for the RVP weight of 2,040,000 lbs has been used as the design weight for the

. analyses described in this chapter. The CG axial location will be 12' 11" +/- 8" from the lower flange sealing surface. As described in Section 2.1.1.3, it will be verified that any additional shielding added to the package will not cause the weight or CG location to fall outside of the range given above. The location of the CG is shown on Figure 1-3A. Calculation summaries for the weight of the RVP and location of the CG are provided in Appendix 2-1. >

c c  : 2.3 - _ MECHANICAL PROPERTIES OF MATERIALS The materials of construction for the steel components of the RVP are listed in Table 2-3 and the minimum specified mechanical properties for these materials are listed in Tables 2-5 and 2-6.

1 Values for these mechanical properties of the steel were taken from ASME Code Section II, Part

' D and mechanical properties for the welding electrodes were taken from ANSI /AWS DI.1,

" Structural Welding Code - Steel."

2 - 13 5

, y- ,, -,- ,, . --- -

A Trolan Reactor VesselPackage- Safety Analvsis Report O

The design stress intensities (S.) for the component materials, listed in Table 2-6, were taken from ASME Code,Section II, Part D, Subpart 1. '

It was determined from a literature search (References 2-17,2-18, and 2-21) that the Young's  ;

modulus, yield strength, and tensile strength for the materials of construction irradiated during plant operation would increase during this period due to irradiation effects. Except for the brittle  :

fracture analysis of Appendix 2-12, the calculations assume the materials to be in the unirradiated state so that the allowable values (as listed in Table 2-6) derived from yield strength and tensile strength were lower than would have been the case for irradiated materials. Since Young's modulus increases only slightly for the materials of construction when they are subjected to long-term exposure to radiation, the effect ofinadiation on this property was considered to be insignificant.

{

The ductile to brittle transition temperature also increases for the materials of construction when  ;

they are subject to long-term exposure to radiation. However, this increase has a negative effect on the margin to brittle fracture. Therefore, this effect ofirradiation was accounted for in applicable analyses related to brittle fracture. Specifically, Appendix 2-12 used adjusted material propeny values to account for the irradiation effects on the materials.

The ductile to brittle transition temperature was measured previously for each of the Trojan .

i reactor vessel beltline materials that could potentially be influenced by irradiation during the l Trojan plant operating period. Surveillance capsules for monitoring the effects of neutron

{

exposure on the Trojan reactor vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the ,

neutron shield pads and the vessel wall. The vertical center of the capsules was opposite the i venical center of the core. Data from analyses (References 2-19,2-20, and 2-21) of the capsule specimens was used to make adjustments to the ductile to brittle transition temperature using the Regulatory Guide 1.99 (Reference 2-16) approach to account for irradiation damage.

The potential effect of corrosion on the vessel material properties was also reviewed. Based on the 40-year design life of the reactor vessel for plant operation, the use of minimum material thickness values in the calculations, and the low corrosion rate of the materials of construction, this effect was considered insignificant.

Impact limiter foam mechanical properties were used in the impact analysis of the RVP.

Table 2-7 gives the bounding stress-strain curve values at the extreme temperatures analyzed.

Appendix 2-7A provides additional information concerning foam properties relevant to the 1 analysis.

2 - 14

Trolan Reactor VesselPackaer- Safety Analysis Reoort o . 2.4 - GENERAL STANDARDS FOR PACKAGES 10 CFR 71.43 establishes the general standards for packages. This section identifies those standards and provides the bases for the RVP compliance.

' 2.4.1 ' MINIMUM PACKAGE SIZE-10 CFR 71.43(a) requires that:

"The smallest overall dimension of a package must not be less than 10 cm (4 inches)." 1 1

Figure 1-2 shows that the smallest overall dimension is greater than 10 cm; therefore, minimum

. package size requirements are met.'

2.4.2 TAMPER-PROOF FEATURE 10 CFR 71.43(b) requires that:

"The outside of the package must incorporate a feature, such as a seal, .

f)' which is not readily breakable, and which, while intact, would be b evidence that the package has not been opened by unauthorized persons."

The reactor vessel containment boundary consists of the reactor vessel shell with the reactor head tensioned in place by 54 reactor head studs and the welded penetration closures. Although removal of the reactor vessel head stud nuts requires special equipment, a tamper proofindicator i will be provided on' one of these studs. Since the RVP will have welded closures, unauthorized j access into the package would be evidenced by cutting of the welds or the package itself. i Therefore, the requirements for a tamper-proof feature are met.

2.4.3 POSITIVE CLOSURE '

10 CFR 71.43(c) requires:

"Each package must include a containment system securely closed by a positive fastening device which cannot be opened unintentionally or

- by a pressure that may arise within the package."

Since removal of 54 head stud nuts requires special equipment and the RVP penetrations will 2 - 15

Trolan Reactor VesselPackaee- Safety Analysis Report

\

l I

have welded closures, unintentional opening will be precluded thereby satisfying the positive closure requirements. The intemal pressures produced in this package are shown in Chapte be much lower than the closure design pressure and, therefore, the pressure concern of this requirement is satisfied.

2.4.4 CHEMICAL AND GALVANIC REACTIONS 10 CFR 71.43(d) specifies the following:

"A package must be ofmaterials and construction which assure that there will be no significant chemical, galvanic, or other reaction among the packaging components or between the packaging components and the package contents, including possible reaction resulting from in-leakage of water to the maximum credible extent."

The RVP boundary is fabricated from ferrous metals as listed in Table 2-3. Penetration ope will have welded closures and nondestmetive examinations will be performed to verify their integrity. The integrity of the package will preclude the in-leakage of water to the package interior. Due to the similarity in base composition of these ferrous metals, the potential for galvanic corrosion either through direct contact or coupling is minimal.

The interior void space of the RVP will be filled with LDCC. This LDCC is a Portland cement-based mixture. There is no significant chemical reaction between Portland cement mixtures and ferrous metals.

Since the RVP is constructed of similar materials oflow galvanic potential and its construction prevents environments conducive to corrosion, the RVP will provide adequate assurance that there will be no significant chemical, galvanic, or other reaction including possible reaction resulting from in-leakage of water. Therefore, the requirements of 10 CFR 71.43(d) are satisfied.

2.4.5 PACKAGE VALVES 10 CFR 71.43(e) requires that:

"A package valve or other device, the failure of which would allow radioactive contents to escape, must be protected against unauthorized operation and, except for a pressure relief device, must be provided with an enclosure to retain any leakage."

Figures 1-3 and 1-4 show that the RVP does not contain valves or other such devices.

2 - 16

Trojan Reactor Vessel Packare - Safety Analysis Report

. D.

V Containment penetrations will be welded closed, thus, precluding escape of the contents. The design of the RVP meets the requirement of 10 CFR 71.43 (e).

2.4.6 ACCESSIBLE SURFACE TEMPERATURE

' 10 CFR 71.43(g) specifies the following:

"A package must be designed, constructed, and prepared for transport so that in still air at 38'C (100'F) and in the shade, no accessible surface of a package would have a temperature exceeding 50 C (122*F)in a nonexclusive use shipment or 85 C (185*F)in an exclusive use shipment."

Chapter 3 provides a discussion of the analysis performed to demonstrate compliance with this requirement. The maximum surface temperature of the package in still air at 38 C(100 F)in the shade is calculated to be 147"F which is below the 185 *F limit for an exclusive use package.

Therefore, the requirement on accessible surface temperature is met.

2.4.7 CONTINUOUS VENTING 10 CFR 71.43(h) requires that:

"A package must not incorporate a feature which is intended to allow continuous venting during transport."

As shown on Figures 1-3 and 1-4, the RVP does not incorporate features which are intended to or would allow venting of the package. Therefore, the requirement for continuous venting is met.

2.5 LIFTING AND TIEDOWN STANDARDS FOR ALL PACKAGES 2.5.1 LIFTING DEVICES The RVP does not include lifting attachments. The reactor vessel will be lifted from its operating location with a system oflifting straps and lift fixtures. After assembly of the package, it will be lifted using a system of straps which encircle the vessel. Therefore, no analysis of lifting devices is required.

2.5.2 TIEDOWN DEVICES An engineered tiedown system is used to attach the transporter and barge. Tiedowns will not be O

g 2 - 17

)

l

Trojan Reactor Veswl Packaee - Safety Analysis Ret > ort '

/3 V

added as a structural part of the RVP. Lugs or other attachments will not be welded to the package outer surface.

The RVP to transporter tiedown system consists of two cradles mounted on a cradle support structure, a longitudinal restraint assembly, slings / straps, and attachment brackets. During transport, the RVP will be restrained on two shipping cradles with sling / straps over the top of the package. The cradles and straps restrain the package against vertical and transverse loading, and are bolted to a cradle support structure. Two RVP nozzles are used to restrain the package in the longitudinal direction. The nozzles are captured in a longitudinal restraint assembly that is bolted to the cradle support structure. The nozzles are analyzed for stress as tiedowns in Appendix 2-2. The longitudinal restraint assembly is segmented so that it may be clamped around the nozzles. The cradic support structure is attached with brackets to the deck of the transporter. The cradle support structure is restrained during overland transit against longitudinal movement by eight shear lugs that engage slots in the deck of the transporter and against transverse movement by four brackets that fit against the outside of the transporter deck.

During barge transport the center two of four transverse beams in the cradle support structure connect to four bolsters that are welded to the barge. Vertical, transverse, and longitudinal movements are restrained during barging through the connection of these beams to the bolsters.

-s Once positioned on the barge, the transporter hydraulics will be " bottomed out" such that the

( )

deadweight of the package and transporter is transmitted through the cradle support beams to the bolsters located above the barge's two watertight longitudinal bulkheads. The tops of the bolsters are recessed such that the cradle support beams are captured in the longitudinal and transverse direction. The sides of this recessed area provide the longitudinal and transverse restraint of the cradle support beams. Vertical motion of the cradle support beams is limited by bolting the cradle support beams to the top of the bolsters.

The tiedown arrangement load paths are illustrated in Figures 2-3 through 2-7. Figure 2-3 shows the RVP on the transporter subjected to a longitudinal acceleration. The longitudinal restraint assembly restrains movement in the horizontal direction. One cradle is placed in compression and the strap on the other cradle is placed in tension. The shear lugs restrain the cradle support structure from horizontal movement on the transporter. Figure 2-4 shows the RVP on the transporter subjected to a vertical acceleration. The straps hold the RVP in the cradles. Load is transferred through the cradle support structure and the attaclunent brackets to the transporter.

Figure 2-5 shows the RVP and transporter on the barge subjected to a longitudinal acceleration.

One cradle is placed in compression and the strap on the other cradle is placed in tension. Loads are transferred through two beams in the cradle support structure and bolsters to the barge.

Figure 2-6 shows the RVP and transporter on the barge subjected to a vertical acceleration. The straps hold the RVP in the cradles. Loads are transferred through two beams in the cradle support structure and bolsters to the barge. Figure 2-7 shows the RVP on the transporter and the A

Q 2 - 18 l

Trojan Reactor VesselPackare - Safety Analysis Report l

A barge subjected to a transverse acceleration. The RVP is restrained by the cradles and straps.

One side of the cradle support structure is in compression and the other side is in tension. Loads are transferred through two beams in the cradle support structure and bolsters to the barge.

A detailed tiedown analysis was submitted to the National Cargo Bureau and US Coast Guard for review and has been approved.

The RVP to transporter tiedown system was designed to meet the requirements of the March 1993 draft of ANSI N14.2 (Reference 2-10). The transporter to barge tiedown system was designed to ANSI N14.24 - 1985 (Reference 2-5). The allowable stresses in the transporter to barge tiedown system will be based on AISC standards (American Institute of Steel Construction Manual,9* Edition), which are more conservative than the corresponding allowables in N14.24 - 1985.

Analyses have been performed to demonstrate that RVP shell and nozzle stresses resulting from tiedown imposed loads remain within required allowables. The tiedown system design accelerations are summarized in Table 2-8. The analyses are contained in Appendix 2-2. The stresses in the RVP shell due to the transport induced loads reacted by the tiedown system are calculated using a static force applied to the CG of the package having a vertical component of 2g, a horizontal component along the direction in which the vehicle travels of 10g, and a

' horizontal component in the transverse direction of Sg. The stresses produced by these loadings are calculated in a detailed finite element analysis. The results of the finite element analysis are summarized in Table 2-9 and are shown to be within allowable limits, based on Code values for RVP material yield strengths.

An alternative method (Appendix 2-2B) was also used to calculate the stresses induced on the RVP shell and nozzles by the longitudinal restraint system. The results of tha evaluation compare well with the finite element analysis and also demonstrate that nozzle stresses under a 10g longitudinal load to the RVP do not exceed material yield strengths. The evaluation compares the calculated stresses to the Code minimum yield strength (S, = 50 ksi) and to the actual yield strengths for the RVP materials. The actual material yield strengths, taken from test reports, exceed 60 ksi. Thus, a comparison to the actual yield strengths shows there is an even greater margin to yield.

I v 2 - 19 l

Troian Reactor Vessel Packnee - Safety Analysis Rer> ort

() .

2.6 NORMAL CONDITIONS OF TRANSPORT This section demonstrates that the RVP is structurally adequate to meet the performance requirements of Subpart E of 10 CFR 71 when subjected to the NCT as defined in 10 CFR 71.71.

The acceptance criteria for meeting these transport conditions are that there will be " . . . no loss or dispersal of radioactive contents, no significant increase in extemal radiation, and no substantial reduction in the effectiveness of the packaging"[10 CFR 71.43 (f)].

The initial conditions for evaluating packages for NCT are specified in 10 CFR 71.71(b). These initial conditions specify a temperature range and internal pressure to be used. The ambient temperature for the evaluations is specified to remain constant preceding and following the test and will remain between -20'F and 100 F. The internal package pressure must be considered to be the maximum normal operating pressure, unless a lower internal pressure consistent with the ambient temperature considered to precede and follow the tests are more unfavorable.

A summary of the analysis results for the NCT is provided in Table 2-1.

2.6.1 HEAT For NCT, hot environment,10 CFR 71.7)(c)(1) specifies exposure to an ambient temperature of f)

U 2

100*F in still air, with insolation of 400 g cal /cm (for curved surfaces) for a period of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The decay heat load is approximately 60,635 Btu /hr generated by the activated components inside the vessel and within the vessel wall in the core region. Calculations for the maximum temperature are described in Section 3.4.2. These calculations result in a maximum expected surface temperature of 147 F which is below the 185 *F limit specified by 10 CFR 71.43(g) for an exclusive use shipment. A summary of calculated temperatures is given in Table 3-3.

The increased ambient temperature and insolation heat load also cause the pressure inside the RVP to increase. The initial condition for the pressure calculation is considered to be the point just prior to sealing the vessel after the LDCC has been injected and has cured for several days.

The initial conditions are then a minimum 40*F ambient temperature, no insolation, and atmospheric pressure. The maximum pressure occurs at 100"F ambient temperature with insolation. The gas temperature inside the vessel is assumed to vary with the bulk average temperature of the LDCC, and its pressure varies as a function of the ideal gas law. In addition, l the internal pressure exerted on the vessel wall is affected by the vapor pressure from the moisture within the concrete matrix. A small amount of water vapor may remain in the concrete void spaces following the venting period and will tend to migrate toward the outer surface of the l LDCC due to the higher temperatures in the core region of the vessel. At the outer surface of the LDCC (inner vessel wall) the vapor will also tend to migrate to the lowest temperature point.

O (V 2 - 20

Trolan Reactor Vessel Packare- Safety Analysis Report G

V This will limit the water vapor pressure exerted on the vessel to that corresponding to the lowest temperature present on the inner vessel surface. However, a value of 125 F (instead of the lowest temperature) was selected based on a review of the thermal calculation, which showed that approximately 40 percent of the inner vessel surface (surface above the Nzzles, including the upper head, and the surface of the lower head) remains at or below this temperature throughout transport. Lastly, the internal pressure exerted on the vessel wall is affected by the gas pressure from radiolysis.

Using the ideal gas law to calculate the change in pressure between the two conditions 460 +LDCC P,on.r

• 460 +LDCC,gso.yeg soo r ,14,9 psia The bulk average temperatures for the two conditions are:

LDCCag40.r = 218'F LDCCastoo r = 289"F (D

%) The resulting pressure is P 3 on.7 = 460 + 289 x 14.7 = 16.2ps/a 460 + 218 Using steam tables, the vapor pressure for a vessel wall temperature of 125 F i',2.0 psi. From Section 3.4.4, the gas pressure from radiolysis for the duration of transport (90 days)is 0.8 psi.

The differential pressure for the duration of transport is:

P, =(P ino.7 -P,,) +P,,y,+P,,, =(16.2 -14.7) +2.0 +0.8 =4.3pst Since the pressure increase from gas production by radiolysis is approximately linear (0.28psihr.onth) over the time periods ofinterest, it is determined that the time period from sealing the vessel to completion of actual transport must be less than 160 days to maintain the vessel internal pressure less than 5 psig. Ilowever, as discussed in Section 4.2.3, the stdpment time period must be less than 90 days based on hydrogen concentration limitations.

(O 2 - 21 i

0 Trojan Reactor VesselPackare- Safety Analysis Retsort C/

Because the pressure due to the change in gas temperature is so small, it was neglected in the differential thermal stress calculations given below.

The effect of temperature changes on the RVP is also evaluated in this sectica. Three types of thermally induced stresses are calculated. The first is due to a temperature difference suddenly applied across the RVP contaimnent boundary. This results in increased stresses in the vessel wall. This case is evaluated in Section VIIA of Appendix 2-3. The results are summarized in Table 5 of the Appendix. The second case is the evaluation of stresses due to differential thermal expansion between the steel vessel and the LDCC contained inside. This is evaluated in Section VIIB of Appendix 2-3 with the results tabulated in Tables 6,7,8, and 9. Because the coefficient of thermal expansion for the vessel steel is greater than that of the LDCC, the vessel is loaded by the LDCC at the cold condition. The load is considered as a hydrostatic pressure acting on the entire inner surface of the vessel and has a magnitude of 43.3 psi. Because of the magnitude of this loading, the design pressure in all structural evaluations where the maximum internal pressure is to be considered is assumed to be 100 psig. This is included in both hot and cold conditions even though the loading does not occur at the high temperature. The third type of thermal stress considered is that due to differential thermal expansion between various components of the package. These include the closure plates covering the nozzle openings and closure studs. These are evaluated in Section VIIC of Appendix 2-3 with the results tabulated in ,

Tables 10 and 11.

[v] l The results of the thermal stress calculations are combined and evaluated against allowable Primary plus Secondary stress intensities. This combination is very conservative in that all stresses are assurned to occur at the same time and the thermal properties are selected to maximize the stress over a given temperature range. The resulting combined stress intensities and allowables are summarized in Table 13 of Appendix 2-3. They are also repeated in Table 2-10 of this chapter. Selected components are also included in the summary of all NCT margins presented in Table 2-1 of this chapter.

2.6.2 COLD For the cold condition, a -40 F steady state ambient temperature is utilized, with zero insolation.

Section 3.4.3 provides a discussion of the analysis and Table 3-3 summarizes the resulting temperatures found for the various regions of the RVP. The pressure calculated for the cold condition follows the same method as used above for the hot condition but uses -40 F as the minimum pressure point and conservatively neglects the effects of vapor pressure and the pressure from gases produced by radiolysis. The vessel is assumed to be closed at 100 F after installation of the penetration closures.

/7 , . 22 L)

Trolan Reactor YesselPackaee- Safety Analysis Recort

.O ,

~ \.J '

J \

P-so*F " 460 +LDCC,esoo.1460 +LDCC,eer ,

The bulk averige temperatures for the two conditions are:

LDCCagioo., = 289 F LDCC,o.r = 144 F  ;

I The resulting pressure is -

P_ , = 460 + 144 , 3 4,7 , 3 3,9 p,f, 460 + 289 O The results above for the heat condition are considered to bound the cold condition when evaluating the stresses due to thermal gradients across the wall of various portions of the RVP.

For a given temperature gradient these stresses are identical but shift from inside to outside or Vice versa as the direction of the teniperature gradient changes. Note that thermal properties of

- the materials were selected to maximize stresses over the entire temperature range ofinterest and

thus the results are conservative for either hot or cold conditions The stress summary noted above for the Heat analysis can also be considered bounding for the Cold case and shows all

. stresses to be within allowables.  !

2.6.3 ' . REDUCED EXTERNAL PRESSURE 10 CFR 71.71(c)(3) requires the application of a reduced external pressure of 3.5 psia to the RVP. The effect of this pressure is small, as demonstrated by the following calculations.  ;

. The pressure internal to the RVP is conservatively evaluated at its maximum value, determined by considering the greatest rise in temperature from that at which the vessel is sealed. The

, minimum ambient temperature at which the vessel will be scaled is 40 *F, and the maximum ambient temperature is 100 'F. The temperature of the gas within the RVP is taken as equal to the bulk average temperature of the LDCC which fills the interior. From Chapter 3, the bulk i

'l 2 - 23

l Troian Reactor vesselPackage - Safety Analysis Report average LDCC temperature corresponding to an ambient of 40 *F and no insolation is 218 'F, and that corresponding to an ambient of 100 'F and full insolation is 289 'F. The internal pressure at  ;

. the time of sealing is 14.7 psia. Using the ideal gas law and considering the contributions of l vapor pressure and the pressure from gases produced by radiolysis, the pressure change within i the vessel is 4.3 psi as calculated in Section 2.6.1 above. Assuming a conservative value of 5.0 I psi for the internal pressure change, the maximum differential pressure under a reduced external

. pressure of 3.5 psis, therefore, is:

j P, = 14.7 +5.0 -3.5 = 16.2pst

\

The RVP is analyzed to demonstrate that Regulatory Guide 7.6 allowables are met for the j reduced external pressure case, namely a limit of S for primary membrane stress and a limit of j 1.5S for membrane plus bending stress. Stresses in the upper head, lower head, flanges, and cylindrical shells are determined by means of the finite element model described in Section VI of Appendix 2-4. Stresses in other components are calculated in the same appendix and summarized as follows. Stresses in the inlet and outlet nozzle shells and safe ends are determined using ASME B&PV Code,Section III, Appendices, Subarticle A-2220. The average I tensile stress in the upper head attachment studs is calculated by apportioning the pressure load i on the upper head equally to all 54 studs. The maximum stress intensity in the studs is A determined using the aforementioned finite element model. Stresses in the circular closure plates  ;

'() are determined assuming simply supported edges, using Case 10a, Table 24, of Reference 2-11.  !

The closure plate weld stresses are determined by applying the differential pressure to the inside i surface of the plates and evaluating the stress in accordance with ASME B&PV Code, Section  :

III, Subsection ND, Article ND-3356.l(c) and ND-3359(b). The load on the welds is equal to the applied pressure, q, times the plate area (where the area is based on the plate outer diameter, d),

or F = (q)xd2/4. The weld area is equal to xds, where s is equal to 0.5 for a %" fillet weld. The weld shear stress, z, is equal to the total load divided by weld area, or z = SA 2

A bounding temperature of 200 'F is assumed for all components. The results of the analyses are given in Table 4 of Appendix 2-4 and summarized in Table 2-11 below. The maximum membrane stress intensity occurs in the lower head, and is equal to 0.20 ksi. As shown in Table 2-11, the value of S,,, for the material of the lower head, ASTM SA-533, Grade B, Class 1, is 26.7 ksi at 200 'F. The minimum margin of safety on membrane stress is MS=

- 1 = +large 0.20 2 - 24

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

i Trolan Reactor VesselPackaer- Safety Analysis keoort a . .

s The maximum membrane plus bending stress intensity occurs in the vessel flange, and is equal to 0.34 ksi.' As shown in Table 2.6.3-1, the value of 1.5S for the material of the lower head,  ;

ASTM SA-533, Grade B, Class 1, is 40.05 ksi at 200 'F. The minimum margin of safety on  !

membrane plus bending stress is i MS = 40'00 - l = +1m '

O.34  !

The goveming nozzle closure plate is for the outlet nozzle, where the membrane plus bending i stress intensity is equal to 0.82 ksi. As shown in Table 2-11, the value of 1.5S for the nozzle  ;

closure plate material, ASTM SA-240, Type 304L, is 25.05 ksi at 200 'F. The minimum margin of safety on membrane stress is i i

MS = * '* -1 = +1arge 0.82 All other closure plates have both lower stress and higher allowables. The governing closure

. plate weld stress occurs again for the outlet nozzle, and is equal to 0.26 ksi. The value of 1.5S.

l for the nozzle closure plate welds, E308L weld rod, is 7.01 ksi at 200 'F. The minimum margin i

- - of safety on membrane stress is ys = 7*0' -1 = +large 0.26

• All other closure plate welds have both lower stresses and higher allowables. These calculations .

demonstrate that the RVP has positive margins of safety for reduced external pmssure.

]

l 2.6.4 INCREASED EXTERNAL PRESSURE i

a 10 CFR 71.71(cX4) requires the application of an increased extemal pressure of 20 psia to the j RVP. The effect of this pressure is small, as demonstrated by the following calculations.

The pressure internal to the RVP is conservatively evaluated at its minimum value, and no support from the LDCC or internal structures are assumed and the contributions from vapor pressure and the pressure from gases produced by radiolysis are neglected. ' The minimum pressure in the vessel is determined by considering the greatest fall in temperature from that at which the vessel is sealed. The maximum ambient temperature at which the vessel can be sealed is 100 *F, and the minimum ambient temperature is -20 'F. The temperature of the gas within the RVP is taken as equal to the bulk average temperature of the LDCC which fills the interior.

From Chapter 3, the bulk average LDCC temperature corresponding to an ambient of 100 'F and 2 - 25 4

t Troian Reactor vesselPe%e- Safe' ty Analysis Report full insolation is 289 7, and the lowest corresponding to an ambient of-20 T and no insolation is the reduced decay heat case with a resulting 107 T. The intemal pressure at the time of  !

sealing is 14.7 psia. Using the ideal gas law, the pressure within the vessel at -20 T is t

P_g, = 460 +2894 0 + m , 3 4,7 , 3 3,3 p,f,  ;

The maximum differential pressure under increased extemal pressure equal to 20 psia, therefore,  ;  ;

is:

P, = ;0 -P_g, = 8.9 psi Allowable extemal pressures for the cylindrical shell, spherical heads and cylindrical nozzles of i the reactor vessel are determined in accordance with ASME B&PV Code,Section III, Division I, ,

Subsection NB, Class 1, Article NB-3133. Allowable nozzle pressures are conservatively based on the minimum wall thickness at the end of the nozzle. Details of these calculations are given in Section VIIIA of Appendix 2-5, and the results are summarized in Table 2-12. The minimum

~ allowable external pressure is for the segment of cylindrical shell located between the nozzles and the lower head, and is 1,327 psi. The margin ofsafety is: '

ys = ' '3 2' - 1 = +large 8.9 Therefore, buckling of the RVP due to increased extemal pressure is not of concem.

The RVP is further analyzed to demonstrate that Regulatory Guide 7.6 allowables are met for the increased external pressure case, namely a limit of S,,, for primary membrane stress and a limit of  ;

1.5S. for membrane plus bending stress. Stresses in the upper head, lower head, flanges, and cylindrical shells are determined by means of the finite element model described in Section  ;

VIIIB of Appendix 2-5. Stresses in other components are calculated in the same appendix and summarized as follows. Stresses in the inlet and outlet nozzle shells and safe ends are determined assuming thick walled cylinders with free ends, using Case 1, Table 32, of Reference 2-11. Stresses in the circular closure plates are determined assuming simply supported edges, using Case 10a, Table 24, of Reference 11. The closure plate weld stresses are determined by conservatively assuming that the welds must support the plates in position without aid from the vessel surface.' This is equivalent to an assumption ofintemal, rather than extemal pressure in the case of the closure plates.' The load on the welds is equal to the applied pressure, q, times the plate area (where the area is based on the plate outer diameter, d), or F = (q)nd2/4. The weld area 2 - 26

-+w , , . , . - - -

Trolan Reactor vesselPackage- Safety Analvsis Reoort is equal to xds, where s is equal to 0.5 for a %" fillet weld.- The weld shear stress is equal to the totalload divided by weld area, or z qd 2

These results are summarized in Table 2-13. A bounding temperature of 200 'F is assumed for all components. The maximum membrane stress intensity occurs in the lower head, and is equal to 0.12 ksi. As shown in Table 2-13, the value of S,,, for the material of the lower head, ASTM '

SA-533, Grade B, Class 1, is 26.7 ksi at -20 'F. The minimum margin of safety on membrane stress is 4

MS= - 1 = +large 0.12 The maximum membrane plus bending stress intensity occurs in the vessel flange, and is equal to i 0.23 ksi. As shown in Table 2-13, the value of 1.5S,,, for the material of the lower head, ASTM SA-533, Grade B, Class 1, is 40.05 ksi at -20 'F. The minimum margin ofsafety on membrane plus bending stress is:

t O - ys = 40.05 _3 , .

0.23

, [

The governing nozzle closure plate is for the outlet nozzle, where the membrane plus bending .!

stress intensity is equal to 0.45 ksi. As shown in Table 2-13, the value of 1.5S,, for the nozzle closure plate material, ASTM SA-240, Type 304L, is 25.05 ksi at -20 "F. The minimum margin ofsafety on membrane stress is: l 0'

MS= - 1 = +large i 0.45 '

All other closure plates and all closure plate welds have both lower stress and higher allowables. ,

These calculations demonstrate that the RVP has positive margins of safety for increased I external pressure.

2.6.5 VIBRATION 110 CFR 71.71(c)(5) requires that the package be evaluated for the effects of vibration "normally incident to transport." Transport of the RVP will involve roadway thipment while supported by-hydraulic suspension trailers and river barge transport. The effects of the vibration loads imposed O

U 2 - 27

Trolan Reactor Vessel Packare - Saktv Analysis Reoort o on the package during the transport are evaluated in Appendix 2-6. The vibration accelerations 1

considered in the analysis of Appendix 2-6 are taken from ANSI 14.23, " Proposed American -

National Standard Tiedowns for Tmck Transport of Radioactive Materials," March 5,1993.- The  !

vibration loads can be considered bounded by assuming a light load and a 0-5 Hz natural -

frequency range for the package and tiedown system. These upper bounds for the peak '

accelerations are as follows:

Vertical: 2.0g Transverse: 0.lg ,

Longitudinal: 0.lg -  ;

These loads are applied simultaneously at the CG of the RVP.

Detailed stress evaluations were performed using analytical hand computations and finite element analysis using the ALGOR computer program. The calculations and finite element analyses results are provided in Appendix 2-6. The acceleration loadings produced highest  :

. stresses in the vessel wall at the pedestal support locations and in the upper head closure bolts.

The results are compared with membrane plus bending allowables and are shown to meet these  :

steady state limits. Alternating stress intensities are also calculated and are shown to be less than ,

the limit for unlimited stress cycles. Results of these calculations are provided in Table 2-1 and l

'show stresses are all within allowable values. l l

2.6.6 WATER SPRAY l l

10 CFR 71.71(c)(6) requires a determination of the impact of a water spray that simulates .

j exposure to iainfall of approximately two inches per hour for at least one hour. The RVP will be i

. constructed of steel. The minimum thickness, excluding penetration covers, is 5.5". l Penetrations will have steel cover plates of at least s/s" thick welded in place. Chapter 8 provides discussion on the testing performed to demonstrate the package does not contain cracks, pinholes, uncontrolled voids or other defects. The interior void space will be filled with LDCC providing assurance against the escape ofcontaminants and minimizing volume available for intrusion of external material such as water. The design and construction of the RVP are adequate to ensure the specified water spray condition would have no adverse effect on the RVP.

2.6.7 FREE DROP 10 CFR 71.71(c)(7) requires a free drop through a distance of one foot for packages weighing more than 33,100 lb. Since the RVP is transported solely in a horizontal orientation, the package is analyzed for the NCT free drop in only a horizontal orientation to satisfy the intent of the  !

regulation. The free drop analysis is described in detail in Appendix 2-7a with the resulting 2 - 28

Troian Reactor vesselPackare - Safety Analysis Report G

V stresses calculated in Appendix 2-9. In Appendix 2-7a, the maximum impact resulting trom a l' free drop assuming the maximum impact limiter stiffness under conditions of-20 "F ambient is 8g. The RVP is analyzed to show that, when exposed to an impact level conservatively rounded up to 9g, the acceptance criteria established in Section 2.1.2 are satisfied for the containment boundary. These criteria are, that stress remains clastic in the sealing region of the vessel bod and upper head; that attachment stud preload will not be significantly affected, and that, in accordance with Regulatory Guide 7.6, stresses in the vessel components satisfy the following:

P < S, P + P, < l .5S.

P + P 3+ Q (range) < 3S,..

The RVP structural shell, including upper and lower heads, flanges, and main body section, is made of SA-5'33, Grade B, Class I carbon steel. The inlet and outlet nozzles are made from ASTM SA-508, Class 2. The upper head attachment studs are made from ASTM SA-540, Grade B24, Class 3. Material properties are evaluated at a conservative temperature of 175 'F for the vessel as a whole, except for the core region, defined as located between the nozzles and the join to the lower head, where a conservative temperature of 200 'F is used. To simplify the analysis, the maximum loading conditions (-20 *F, maximum impact) are conservatively used with the minimum strength values.

2.6.7.1 Reactor Vessel Free Dron Stress Analysis The free drop analysis evaluates stresses in several areas of the RVP in order to demonstrate the adequacy of the design under these conditions. The description and results for each of these analyses are given below.

2.6.7.1.1 Containment Boundary Stress Stresses in the RVP shell due to the NCT free drops are determined using the finite element model described in Appendix 2-9. Since the pressure applied to the model (100 psi) includes the effects of differential thermal expansion (see Section 2.6.1), the resulting stresses include secondary stress, Q. The resulting maximum stress intensity in the vessel shell due to the 9g NCT free drops impact, P + 3 P + Q, is 10,396 psi, located near the bottom of the vessel, on the outside surface, beneath the lower impact limiter. Even though this stress includes a mondary component, the smaller allowable stress of 1.5S is conservatively used. For the vessel wall material of SA-533 Grade B, Class 1, S, is 26,700 psi at the core region bounding temperature of CT U 2 - 29

l Trolan Reactor VesselPackaze- Sat l sty Analysis Report o 200 'F. The margin ofsafetyis gg , (1.5)26,700 -1 = + 2.85 10,396 The maximum membrane stress intensity, P + Q, is 6,787 psi, located adjacent to a nozzle.

Again, due to the application of the 100 psi pressure, this value also conservatively includes secondary stress. In this case, the allowable stress is S,. For SA-533 S is again 26,700 psi.

The margin ofsafetyis:

MS= -1 = +2.93 6,787 The maximum stress in the radius, P + P, + Q, is 14,016 psi. Due to the location of this stress in a region of stress concentration, the value necessarily contains peak stress, which, per Regulatory Guide 7.6, does not need to be included. Conservatively, however, peak stress is included in the computation of margin ofsafety. As before, the allowable stress is 1.5S , where S is again 26,700 psi. The margin ofsafety is:

p us (1.5)26,700 h 14,016

-1 = + 1.86 Thus, the margin of safety on vessel body stress is positive during the NCT free drop event.

2.6.7.1.2 Upper Head Attachment Studs As determined in the upper head attachment stud analysis for HAC, Section 2.7.1.2.2, the stud load due to an intemal pressure of 100 psi alone is F, = 40,728 lb each. This load is added to the load arising from the NCT side drop impact. A length of the head outer flange equal to 12.75" is loaded by the impact limiter in the upward vertical direction, with an additional inertia load downward equal to the head weight times the impact load of 9g, located at the head CG. Thus, there are two opposing forces on the upper head; the impact limiter transmitted load upward, and the head inertia load downward. First, the impact limiter transmitted force is determined. The load on the head flange is 12.75/58 = 22% of the total impact limiter load, for a 58-inch wide 2 -30

Troian Reactor VesselPackase - Safety Analvsis Report O 6 limiter. For a total package weight of 2.04 x 10 lb and an impact of 9g, the load on the head flange is, therefore:

Fu

=

2 0.22 = 2.02 (10')lb The moment on the head due to the impact limiter transmitted force is, therefore,12.75/2(Fu) =

12.9 x 10' in-lb, which is applied in a clockwise sense about the vessel sealing surface. Next, the inertia load is determined. The location of the upper head CG is 35.2" above the vessel sealing surface, and has a weight which is bounded by 150,000 lb. The LDCC material located within the head is assumed to break free of the remainder within~ the vessel and further load the he  ;

For an internal radius of r = 83.67" (radius to the base metal, ignoring cladding), the volume '

within the head is 710 fP. Since the density of the LDCC material is 65 lb/fP, the total weight of head and contents is W3 = 150,000 + 710(65) = 196,150 lb. For a 9g impact, the moment is t equal to W3(35.2)(9) = 62.1 x 10' in-lb, in the counterclockwise sense. Note that the CG location of 35.2" conservatively neglectr the effect of the LDCC, which would decrease the moment arm.

The net moment on the stud pattern is, therefore, M = 62.1 x 106- 12.9 x 10' = 49.2 x 10'in-lb.

A conservative estimate of the maximum stud force is found from O F* * =

nR

+F' = 59,7211b I

Since the tensile load of 59,721 lb is small relative to the stud preload force of 720,000 lb, there is no added tensile load on the stud due to the NCT free drop event, and the flange seal l compression is unaffected.

{

If friction between the upper head and the main vessel is neglected, the average shear load on a stud is a function of the difference between the impact limiter transmitted force, Fow, and the downward shear load due to the inertia force, since these loads are in opposite directions. The average load per stud is:

F, =

F"-9W" = 4,716 lb n

The area, A,, of each stud, which has a 0.75 inch central hole, is 35.9 in 2. The shear stress in each stud due to the drop load is, therefore:

2 - 31

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

r._.._ .

Trojan Reactor Vessel Packare - Safety Analvsis Report

[

F T = '- = 131 pst -

y A, ,

- The shear stress of 131 psi is clearly insignificant, and the margin of safety is very large.

Therefore, the head studs remain completely clastic during the NCT free drop event, and the upper head seals retain full effectiveness.

2.6.7.1.3 Nozzle Cover Stress and Attachment Weld Stress The nozzle covers are made from 2.5" thick plate stock using Type 304L stainless steel material -

and welded using a %" fillet weld to the nozzle ends. The governing nozzle cover is the largest, j for the outlet nozzle, and is 31.63" in diameter. In the NCT side drop, the governing case is for the nozzle pointing directly toward the ground. The nozzles are loaded by the internal pressure of 100 psi, their self-weight, and a portion of the LDCC material located above the plate. In a

- side drop event, the maximum amount of concrete material which could break loose from the internal monolith is equal to the concrete actually inside the nozzle, and a small " fracture cone" above that. This material is conservatively upper bounded by a cylinder, d = 36" in diameter and L = 62.5" long. The volume of this cylinder is 36.8 ft'. For a concrete density of 65 lb/ft', the total weight ofconcrete loading the cover plate is 2,392 lb. The weight of the cover plate is 570 lb. The total weight of the steel and concrete load is, therefore,2,962 lb. Under an inertia load  !

of 9g, the inertia load in the side drop event is Fi = 2,962(9) = 26,658 lb. Next, the inner area is '

2 2

. computed as A = x/4(31.63 ) = 785.8 in The inertia load may then be applied as a pressure on the inside of the plate, smd including the internal pressure, the total pressure load is 133.9 psi.

The stress in the plate is found from Table 24, Case 10a, of Reference 11. The maximum stress is a bending stress located at the center of the plate. The center moment is M= #

= 6,907 in-lblin 16 where a is the plate outer radius of 31.63/2 = 15.815", and v is 0.3. The stress is o= 6M* = 6,631 psi t2 where t is the thickness of 2.5". For NCT, the allowable stress is 1.5S . For the Type 304L plate material at 175 *F, S = 16,700 psi. The margin of safety is 2 - 32

kolan Reactor VesulPackage - Safety Analysis Rer> ort h

n U

yg (1.5)l6,700

-1 = + 2.78 6,631 2

The area of the attachment weld is A,= 49.68 in . This weld must support a load equal to F,, =

qA = 105,219 lb. ' Die stress in the weld is:

F*

T- - 2,118 psi A,

For an allowable weld stre.ss S of 7,007 psi, the margin ofsafety is:

MS = 7,007 -

= +2.31 2,118 Thus, the nozzle cover plate and attachment weld design margins are positive for the worst case analyzed free drop event.

2.6.7.2 Shidhg Free Dron Stress Analysis The external shielding installed around the reactor vessel is necessary for the RVP to meet the dose rate limits defined in 10 CFR 71. The main shielding (2" and 5" plate) shown on Figure 1-5 must remain in place during both NCT and HAC events. The HAC events are considered bounding and, therefore, the analysis described in Section 2.7.1.3 below demonstrates that the -

main shielding will remain intact during these events. However, the lower skirt and supplemental shielding that may be added as needed to meet the NCT dose rate requirements must remain in place during all NCT events.Section VI E of Appendix 2-10 provides the analysis to demonstrate this. The welds which attach the supplemental shielding to the main shielding have positive margins of safety.

2.6.7.3 Summarv of Resuhs The preceding analyses demonstrate that the RVP has positive margins of safety for the analyzed NCT free drop events. Table 2-14 sununarizes the results for the containment components.

i 2 - 33 t

s Troian Reactor Vessel Packare - Safety Analysis Report

(

q 2.6.8 CORNER DROP 10 CFR 71.71(c)(8) states:

... This test applies only to fiberboard or wood rectangular packages not exceeding 50 kg (110 pounds) and fiberboard or wood cylindrical packages not exceeding 100 kg (220 pounds)."

This test is not applicable since the RVP is not constructed of either fiberboard or wood.

2.6.9 COMPRESSION 10 CFR 71.71(c)(9) states:

"For packages weighing up to 5000 Kg (11,000 lbs), the package must be subjected, for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to a compressive load applied uniformly to the top and bottom of the package . . . "

This test is not applicable to the RVP since it weighs approximately 2,040,000 lbs.

.V 2.610 PENETRATION 10 CN . 71.71(c)(10) requires a determination of the impact of " . . . a vertical steel cylinder of 3.2 cr 41 1/4 in) diameter and 6 kg (13 lb) mass, dropped from a height of 1 m (40 in.) onto the exposed surface of the package which is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface.'"

The penetration analysis is described in Appendix 2-8. The analysis uses th method developed by the Ballistics Research Laboratories (BRL) where a missile of known diameter, mass, and velocity impacts a steel plate target. Using this method, a minimum thickness can be calculated which will not be penetrated. The analysis of Appendix 2-8 assumes the most vulnerable location for perforation is the 0.625" steel plate used to cover the penetrations. For the specified 40" drop, the calculated impact velocity is 14.6 ft/sec. However, the lower bound of the BRL data is 70 ft/sec. Therefore, the impact velocity is conservatively assumed to be 70 ft/sec. Using the BRL equation, the maximum thickness ofplate that would be perforated is 0.118". Since the minimum thickness of the RVP wall is 5..i", and the penetration closures are a minimum of 0.625" thick, the bar will not penetrate the package.

(m) 2 - 34

Trolan Reactor vesselPackaee- Safety Analysis Rer> ort p

\'

2.7 HYPOTHETICAL ACCIDENTCONDITIONS The HAC specified by 10 CFR 71.73 are:

1.

Free drop of 30' onto a flat essentially unyielding horizontal surface 2.

Crush (not required for packages with mass greater than 1100 lbs.)

3. Puncture from a 40" drop onto a vertical steel bar

4. Thermal exposure to 1475'F for 30 minutes

5. 4 Immersion under 3' of water (required for fis ;ile material)

6. Immersion under 50' ofwater Evaluation of these accident conditions is to be based on a sequential application in the order indicated above to determine the cumulative effect on a package. An undamaged package may be used for test requiring immersion under 50' of water. ,

The initial conditions for evaluating packages for HAC, excluding the water immersion test, are specified in 10 CFR 71.73(b). These iaitial conditions specify an ambient air temperature O between -20"F and 100*F which is most unfavorable for the feature under consideration. The internal pressure of the package must be assumed to be the maximum normal operating pressure, unless a lower intemal pressure, consistent with the ambient temperature assumed to precede and follow the fests, is more unfavorable.

A discussion ofeach of the six test conditions is provided in the following sections. A summary of the margins of safety for the HAC is provided in Table 2-2.

2.7.1 FREE DROP Per 10 CFR 71.73(c)(1), a free drop is to be performed from a height of 30' onto a flat, essentially unyielding surface, rtriking the surface in a position for which maximum damage is expected. The potential energy of the drop is to be completely absorbed by impact limiting structures which remain in place throughout the drop event. An analysis of the impact event results in deceleration values which then may be used to perform quasi-static stress analyses of the RVP, demonstrating that the RVP satisfies Regulatory Guide 7.6 stress allowables.

Containment integrity of the RVP subsequent to the drop event is demonstrated by assuring thrt stresses are clastic in the region of the head closure seals, that attachment studs retain sufficient preload, and that vessel stresses, including closure plates and associated attachment welds, VO 2 - 35

f Trojan Reactor Vescel Package- Safety Analysis Report t

remain below allowable levels. Shielding integrity is maintained by demonstrating that the external shield structure remains intact and essentially in place following the free drop event.

2.7.1.1 Impact Analysis The RVP is transported in essentially two modes. First, on a barge on the Columbia River, and second by transporter (less than 30 miles) over land between the Port of Benton and the U.S.

Ecology disposal facility on the Hanford Reservation. IAEA Safety Series 37 (Reference 12),

Paragraph A-618, suggests that an essentially unyielding surface should be one which has an effective weight of at least 10 times that of the package, and possessing a steel impact surface.

The RVP, including vessel shell, contents, and impact limiters, weighsjust over 2,000,000 lb.

In light of this magnitude of weight, neither the barge nor the land portions of the voyage (which is over loose gravel soil), can expose the RVP to an impact surface which is " essentially unyielding." However, the impact analyses performed conservatively assume the ground surface absorbs no energy and satisfies the requirements of an essentially unyielding surface.

The analysis of the impact is detailed in Appendix 2-7A.

During transport, the RVP rests horizontally in two cradles which are part of the transportation support structure as shown on Figures 1-1 and 1-2. Due to the large weight and overall size of n the package, due to the transport route, due to the fact that at no time during the less than 72 V hour transport will the RVP be lifted from the transporter and due to the extraordinary controls placed on the transport of the RVP (see Chapter 7), a 30' free drop is not considered a condition of transport. Because of the above considerations, the maximum speed during ground transport is limited to 5 mph. This is much less than the typical 55 mph speed assumed possible during shipment of smaller packages. Although no credible drop scenario exists, for analysis purposes, a conservative free drop height was used. The analyzed drop scenario is one in which the package unmechanistically fr.ils from the cradle structure to the ground. Therefore, the largest free drop height of the rackage CG is obtained from conservatively taking the largest package diameter (205") as resting on the top of the cradle (16.5' above the ground) as the initial position of the package. From there, the maximum free drop height is the distance between the lowest point of the impact limiters and the ground. The impact limiter outer diameter is 28', or 336". The depth of the impact limiters in a radial direction is (335 - 205)/2

= 65.5". The distance between the lowest point of the impact limiters and the ground is, therefore, 16.5(12) - 65.5 = 132.5", which may be rounded to 1l' without significant loss of accuracy. A free drop of I l' is, therefore, the analyzed free drop distance for the RVP, and this value is used in subsequent calculations. The I l' distance is considered the drop distance at the time the impact limiter contacts the ground. The limiter then crushes, allowing the package to continue falling several more inches. This additional potential energy is included in the subsequent analyses of the drop event. The velocity of the package at impact is approximately O

(v 2 - 36 1

[

Troian Reactor VesselPackage - Saferv Analysis Rer> ort

_ \

I

(/)

27 fVsec. This introduces another conservatism in that the maximum allowed ground speed is 7.3 fVsec (5 mph).

10 CFR 71.73(c)(1) requires that the package strike the ground "in a position for which the maximum damage is expected." However, again due to the special nature of this shipment, not all orientations are credible. The RVP will be moved to a horizontal position on the Trojan site, and henceforth will remain in a horizontal position until it reaches the final disposal site. In all barge and transporter operations, the package axis will remain horizontal. Therefore, end drop orientations or steep angle oblique drops from any height are not credible. Two free drop orientations are considered for the RVP: a Dorizontal drop from the transport cradles, in which the package rolls ofTof the top of both cradles simultaneously, and falls Il' to the ground as discussed above; and a case in which only one end of the package falls from the transport structure at one time, causing the package to strike the ground at an oblique angle.

In the oblique drop event, one end of the package is assumed to be held and act as a pivot point until the other end has struck the ground. Then, the pivot end is assumed to become free and fall to the ground. Since the initial condition height of the package is Il',just as for the horizontal drop, the package CG falls through a distance which is less than I l' before the first end strikes the ground. The determination of the oblique impact orientation is detailed in g Appendix 2-7A. Since impact severity is a function of the drop height of the CG, this distance is conservatively maximized for purposes of calculation. This is done by assuming the maximum possible distance between the pivot point of the package and the CG of the package.

In this analysis, the pivot point is, therefore, chosen at the apex of the hemispherical head at the extreme end of the package. Even though such a pivot point is, by itself, not credible, such a choice maximizes the drop distance of the package CG, and thus conservatively maximizes impact severity for the oblique drop event. The choice of which end of the package is allowed to fall first is made based on the package protrusions or "hard points" which would exist on the striking end. The upper end of the RVP is the location of the largest exposed diameter (the head flange, at 205"), as well as the head attachment stud ends. No comparable features exist at the lower end of the package. Therefore, the pivot point is chosen at the apex of the lower hemispherical head, and first impact at the upper end is considered. The resulting geometry and dimensions are given in Appendix 2-7A. The package CG falls 85.3 ", or essentially 7', before the upper impact limiter strikes the ground. The package continues to fall a few more inches as the impact limiter crushes. The secondary drop is then assumed to follow, but is r.u vverning and is not specifically analyzed. This is because the remaining drop height of the package CG is less than Il'- 7' = 4', orjust over half of the first drop height. Therefore, the secondary impact level is not governing. Second, when the second impact limiter strikes the ground, the package axis will be again essentially horizontal. This means that the second impact limiter will strike the ground in a non-oblique orientation, one in which the cmsh distance will be less Q

v 2 - 37

Troian Reactor Vessel Package - Safety Analysis Report

\ )

than for an oblique orientation. Therefore, the secondary crush distance is also not governing.

Impact calculations are performed using initial and environmental conditions which result in the maximum impact severity for the RVP. Subsequent evaluations of the ability of the package to withstand these loads without compromise of the containment boundary utilize minimum material properties. Thus, all conditions of applied loading and structural strength are conservatively bounded. The maximum impact conditions correspond to the maximum mechanical strength of the polyurethane foam energy absorbing material, and occur at the ]

l minimum ambient temperature of-20 *F and without insolation, in accordance with Regulatory Guide 7.8. Additional calculations are performed to determine the maximum deformation

)

l under wann conditions, in order to demonstrate that the maximum impact level obtained under l

cold conditions is not exceeded by excessive crush of the polyurethane foam or by contact of a l

"hard point"(i.e., uncushioned contact) with the ground.  ;

l l

Since impacts on the ends of the RVP are not credible, as discussed above, the impact limiters  ;

are required to protect the package in only side and relatively shallow oblique impacts. The l impact limiters are, therefore, constructed in the shape of annular rings and fastened to the RVP as shown in Figure 1-6. The outer diameter is 336", and the width is 58", dimensions which do not include the %" thick structural angles on the outer corner seams. The inner diameter of the n lower limiter is based on the outer diameter of the shield, approximately 202". The inner

() diameter of the upper limiter is based on the outer diameter of the upper flange, or 205". These two inner dir. meters,202" and 205", differ by an insignificant amount, and a value of 205" is used conservatively for both. The outside edge of the upper impact limiter is located 99" above I the nozzle centerline, and the outside edge of the lower impact limiter is located 246" below the i nozzle centerline. The inside face of each impact limiter is buttressed by a steel structure to prevent the dislodging of the limiter in a drop event. The impact limiters are retained against I the inboard buttresses by means of tie rods through their thickness. No we ding is performed to the material of the vessel. The impact limiters are encased in %" thick steel shells of ASTM A516 Gr 70, but due to the ease with which buckling may occur, this material does not add to impact severity. See Appendices 2-7A and 2-9 for further description of the impact limiter structure and impact stress analysis.

l The maximum RVP protrusion is the inlet nozzle, which is conservatively taken as extending 41.13" from the vessel wall. Since the vessel wall outer diameter is 192", the maximum protrusion diameter is 192 + 2(41.13) = 274.3". However, a distance of 275" is conservatively used. Since the impact limiter outer diameter is 336", the minimum distance from the limiter o.d. to the nearest protrusion is (336 - 275)/2 = 30.5". In the oblique drop case, the upper head attachment stud is nearest the ground. The studs are nominally 7" in diameter and located on a 95.94" radius, and the outer end extends above the sealing surface a maximum of 47.5" The (O

U) 2 - 38

l Troian Reactor VesselPackage- Safety Analysis Report 1

(3 V

design weight of the RVP, including the impact limiters, is 2.04 x 106 lb. The CG is located approximately 155" below the sealing surface, which is essential!y equidistant between the two impact limiters, at the geometric center of the package.

The free drop analysis is performed using the proprietary code CASKDROP, which is described in detail in Appendix 2-7B. In brief, the crush area of the impact limiter at each deformation step is calculated. The area is subdivided into equilateral cells, and for each cell, the strain is calculated. The force corresponding to each cell at the calculated strain is found using the foam stress-strain data, and added up to give the total force of the impact limiter at each deformation.

CASKDROP uses a quasi-static energy balance approach, in which the amount of energy consumed is the cumulative sum of crush force times deformation increment. When the total energy absorbed equals the total potential energy of the drop (including cmsh distance), the solution was complete. The impact force is equal to the total maximum crush force divided by the package weight. l The results of the free drop impact analyses are given in Table 2-15. All impact values are given for the package CG, defined in units of g, and are normal to the ground. For the horizontal drops, the package remains in a horizontal position throughout the drop event. The crush distance is defined as the total deformation of each limiter in a direction normal to the o package axis. Maximum strain is the maximum value of the ratio: crush distance / original Cl distance for the limiter. Strain has no absolute upper limit, but the values reached in this analysis are well below any strain hardening limit. The resulting maximum impact is 20.lg for the cold, -20 'F case, and the minimum clearance over the inlet nozzle is 6.4" in the warm case.

These values are conservatively approximated for use in the free drop stress analysis as 22g and 6.0". Since the impact limiter forces are well balanced and the package remains horizontal, no i forces are developed which would tend to dislodge the impact limiters from the package. Note l also that, due to the relatively small strain, the global deformations of the impact limiters are small, and the overall integrity of the impact limiter structure is not significantly degraded. In l the oblique drop case, the initial pivot point on the end of the lower hemispherical head is assumed to exist throughout the primary drop event. For a free drop height of Il' for the distant limiter, the CG drops 85.3", or 7', and has an initial impact orientation of 19 to the horizontal. The package continues to rotate throughout the impact until it comes to rest, but the added rotation is relatively small. The package can, therefore, be assumed to undergo the drop event at a constant orientation to the ground equal to the average of the initial and final crush angle, or an average of 22.5'(warm case). The results of the analysis are also given in Table 2-

15. The maximum crush distance and corresponding strain are given for the edge of the limiter which is crushed the most (i.e., toward the upper head) and are measured normal to the ground.

Relative to both impact severity at the package CG and to minimum ground clearance over uncushioned structure, the oblique drop event is not governing compared to the horizontal drop.

2 - 39

{}

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

Troian Reactor Vessel Packaer - Saktv Analysis Report i

w 4 2.7.1.2 Reactor Vennel Free Drop Stress Amlvsis The RVP is analyzed to show that, when exposed to the impact levels determined in Section 2.7.1.1 (22g for HAC horizontal drop), the acceptance criteria established in Section 2.1.2 are .

satisfied for the containment boundary. These criteria are, that stress remains clastic in the

. sealing region of the vessel body and upper head; that attachment stud preload is not i significantly affected; and that, in accordance with Regulatory Guide 7.6, stresses in the vessel l

components satisfy the following: .

P < 2.4S or 0.7S , whichever is less P + P. < 3.6S, or S , whichever is less '

In addition, Section 2.7.1.4 demonstrates that the maximum possible hypothetical flaw in the vessel containment boundary remains stable under the bounding conditions of ambient l temperature, material toughness, and applied stress. '

i The following analyses are performed and described in Appendix 2-9 to demonstrate the adequacy of the RVP design under free drop conditions.

( 2.7.1.2.1 Containment Boundary Stress Stresses in the RVP shell due to the goveming 22g side drop impact are determined by means l of the finite element model described in Appendix 2-9. The resulting maximum stress intendty in the vessel shell, P. + P., is 26,438 psi, located near the bottom of the vessel, on the outside surface, beneath the lower impact limiter. Since this stress includes bending components, the allowable stress is the lesser of 3.6S, or S,. For the vessel wall material of SA-533 Grade B, Class 1, S, is goveming and is 80,000 psi at the core region bounding temperature of 200 *F.

. The margin of safety is MS= -1 = + 2.03 26,438 The maximum membrane stress intensity, P., is 14,700 psi, located in the region just below the -

sealing flange, above the nozzles. In this case, the allowable stress is the lesser of 2.4S, or 0.7S,. For SA-533, at 175 *F (outside the core region),0.7S, is governing, where S, is again i

2 - 40

1 Troian Reactor VesselPacLaec- Safety Analysis Report o - 80,000 psi.- The margin of safety is:

1 l

MS=

-l = +2.81 14,700

' Stresses in the nozzle region are detennined by means of a submodel, based on the main finite element model, as described in Appendix 2-9. The maximum stress in the radius is 24,548 psi, ,

and, since it is less than the maximum stress intensity of 26,438 psi, is not governing. Thus, the j

- margin of safety on vessel body stress is positive during the free drop event. Fu ther, all material in the region of the upper head seals remains completely clastic.  ;

i 2.7.1.2.2 Upper Head Attachment Studs The upper head is attached using n = 54, nominally 7" diameter studs on a radius of R = . .

95.94", pretensioned to 720,000 lb. Since the internal cross sectional area of the head is A = l 2

x(83.672) = 21,993 in and the internal pressure is p = 100 psi, a tensile load is applied to each  ;

stud equal to '

F' = N = 40,72816 i O " '

In the horizontal side drop, a length of the head outer flange equal to 12.75" is loaded by the impact limiter in the upward vertical direction, with an additional inertia load downward equal to the head weight times the impact load of 22g, located at the head CG. Thus, there are two opposing forces on the upper head; the impact limiter transmitted load upward, and the head inertia load downward.

First, the impact limiter transmitted force is determined. The load on the head flange is 12.75/58 = 22% of the total impact limiter load, for a 58-inch wide limiter. For a total package weight of 2.04 x 10' lb and an impact of 22g, the load on the head flange, Fu , is 4.94(106) lb.

The moment on the head due to the impact limiter transmitted force is, therefore, i 12.75/2(Fon) = 31.5 (106) in-lb, which is applied in a clockwise sense about the vessel sealing surface. j l

Next, the inertia load is determined. The location of the upper head CG is 35.2" above the l

. vessel sealing surface, and has a weight which is bounded by 150,000 lb. The LDCC material located within the head is assumed to break free of the remainder within the vessel and further load the head. - For an intemal radius of r = 83.67", the volume of the head is 710 ft). Since the maximum density of the LDCC material is 65 lb/ft), the total weight of head and contents is W, 2 - 41

. Trolan Reactor Vessel Packan- Mov Analysis Reoort .

n- .

i U .

150,000 + 710(65) = 196,150 lb.' For a 22g impact, the moment is equal to W,(35.2)(22)

152 x 10' in-lb, in the counterclockwise sense. Note that the CG location of 35.2" i conserystively neglects the effect of the LDCC, which would decrease the moment arm.

i

-\

The net moment on the stud pattem is, therefore, M .152 x 10'- 31.5 x 10' = 121 x 10' in-lb. i

' A conservative estimate of the maximum stud force is found from  !

F,=2

+ F, = s7,43916 ,

If friction between the upper head and the main vessel is neglected, the average shear load on a stud is a function of the difference between the impact limiter transmitted force, Fu,, and the .

. downward shear load due to the inertia force, since these loads are in opposite directions. The l

. average load per stud is:

. Fu-22 W, F, = = 11,569lb

, n O . These tensile and shear stud loads are now compared to those resulting from the oblique drop case. in this case, a vertical impact of 16.3g at the vessel CG is used (see Table 2-15). The upper head CG is a total of 177.9 + 240.4 = 418.3" from the pivot point, measured horizontally with the package axis 22.5' from the aorizontal. The impact at the location of the head CG is, therefore:

Gu,, = I6.3 = 28.4g 240.4

Again, the force on the head flange due to an impact limiter transmitted force is determined first. 'lhe component of total impact limiter force which is parallel to the scaling surface (and normal to the head flange OD) is 23.1 x 10' lb. As before, only 22% of this load is carried by the head flange, and, therefore, Ft n = (0.22)23.1 x 10' = 5.08 x 10' lb. The moment Benerated is-equal to 12.75/2(Fu) = 32.4 x 10' in-lb, clockwise.

The inertia load is determined next. The impact load, Gw, is resolved into components parallel and normal to the sealing surface. The parallel inertia force on the head is Fy = 5.14 x 10' lb The resulting raoment is 35.2F, = 181 x 10' in-lb in the counterclockwise direction.

The net moment applied to the studs is, therefore, M = 181 x 106 - 32.4 x 10' = 149 x 106 in-lb.

For the normal force, the weight of the entire internal material (core materials plus LDCC),

2 - 42

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

Trolan Reactor Vessel Packame - Safety Analysis Report o rounded up to a value of 700,000 lb, is conservatively used in addition to the weight of the f upper head itself. This force is Fw = Gw(700,000 + 150,000) = 9.27 x 10' lb, which acts in parallel to the pressure load on the studs. The total maximum bolt force is.

j F- = + "" +F = 269,915 lb -

a , r i

If friction between the upper head and the main vessel is neglected, the average shear load on a ,

' stud is a function of the difference between the impact limiter transmitted force, Fa, and the

]

' downward shear load due to the inertia force since these loads are in opposite directions. The I average load per stud is:

)

i F, = =1,111/b n

The governing stud loads are, therefore,269,915 lb tensile from the oblique drop event, and 11,569 lb shear from the horizontal side drop event. Since the tensile load of 269,915 lb is small relative to the stud preload force of 720,000 lb, there is no added tensile load on the stud due to the free drop event, and the flange seal compression is unaffected.

' If the coefficient of friction were as low as 0.1, the frictional shear resistance force of (0.1)720,000 = 72,000 lb would still be greatly in excess of the maximum shear load of 11,569 j lb. However, this is conservatively ignored and the stud stress due to both the maximum shear force and the preload forces are conservatively combined as follows. The area of the stud, .

2 excluding the 0.75" central hole is 35.8 in . This results in a shear stress due to the drop load of

~ 322 psi and a tensile stress due to the preload of 20,067 psi. These stresses may be combined to ,

.give SI=/0 24 7 =20,070pst 2

. For the ASTM SA-540 Grade B24 Class 3 studs, the value of S,,, at a temperature of 175 'F is 41,875 psi. The margin ofsafety is:  !

MS= '

-1 = + 1.09 20,070

' Therefore, the head stud margin of safety is positive, and the upper head seals retain full ,

effectiveness, during the worst case free drop event. l

Trolan Reactor YesselPackaee- Safety Analysis Reoort O

O 2.7.1.2.3 Nozzle Cover Stress and Attachment Weld Stress '

The nozzle covers are made from 2.5" thick plate stock using Type 304L stainless steel material t and welded using a %" fillet weld to the nozzle ends. The outlet nozzle cover is the largest, 31.63" in diameter, and is assumed to be governing. In the side drop, the governing case is for  ;

the nozzle pointing directly toward the ground. The nozzles are loaded by the internal pressure '

of 100 psi, their self-weight, and a ponion of the LDCC material located above the plate. In a side drop event, the maximum amount of concrete material which could break loose from the internal monolith is equal to the concrete actually inside the nozzle, and a small " fracture cone" above that. This material is conservatively upper bounded by a cylinder, d = 36" in diameter 3

and L = 62.5" long. For a maximum concrete density of 65 lb/ft), the total weight ofconcrete  !

loading the cover plate is 2,392 lb. The weight of the cover plate is (x/4)(31.632)(2.5)(0.29) = <

570 lb, where the density of the Type 304L is taken as 0.29 lb/in3 . The total weight of the steel l and concrete load is, therefore,2,392 + 570 = 2,962 lb. Under an inertia load of 22g, the inertia load in the side drop event is Fi = 2,962(22) = 65,164 lb. Next, the inner area is computed as 2

A = (x/4)(31.632) = 785.8 in . The inertia load may then be applied as a pressure on the inside of the plate, and including the internal pressure, the total pressure load is F

q = J+100 =l82.9 psi A

l The stress in the plate is found from Table 24, Case 10a, of Reference 11. The maximum stress is a bending stress located at the center of the plate. The center moment is:

M* = 9" = 9,435 in -lblin 16 where a is the plate outer radius of 31.63/2 = 15.815", and v is 0.3. The stress is:

o= 6 M' = 9,058 psi t2 where t is the thickness of 2.5". For HAC, the allowable stress is the lesser of 3.6S, or S,. For 2 - 44

0 Trojan Reactor VesselPackaee - Safety Analysis Report

/

u%

the Type 304L plate material, the minimum is 3.6S, = 60,120 psi, where S, = 16,700 psi. The margin of safety is:

MS= -1 = +5.64 9,058 2

The area of the attachment weld is A, = 35.13 in . This weld must support a load equal to F. = qA = 182.9(785.8) = 143,723 lb. The stress in the weld is then 4,091 psi. Of the two metals (carbon and stainless steel) making up the weld, the lowest strength is possessed by the Type 304L, and is (0.6)2.4S = 24,048 psi. The margin of safety is:

gp 24,04 8 _, , 4,gg 4,091 Thus, the nozzle cover plate and attachment weld design margins are positive for the worst case free drop event.

2.'7.1.2.4 Vessel Shell Buckling Under Side Drop Loads q Although well supported by the internal LDCC material, the vessel shell is checked for buckling

. U

' due to the HAC side drop using ASME Code Case N-284. Hoop, axial, and shear stresses are obtained using the finite element model discussed in Appendix 2-9, taken at the upper, compressive side of the vessel, at the axial center and mid-plane of the shell. Consistent with Reg. Guide 7.6 philosophy, a factor of safety corresponding to ASME Code, Level D Service conditions are employed, equal to 1.34 as specified in Code Case N-284. The shell length is found by taking the total inside length and subtracting the two head inner radii, and is equal to 324.5". The wall thickness over this entire length is conservatively taken equally to the minimum wall thickness of 8.5". Therefore, the stiffening effects of the flange, the thicker material above the nozzles, and the nozzles themselves, are conservatively ignored. The component stresses are taken at the location of the central girth weld, essentially at the center of the core region, in the area ofminimum wall thickness. The buckling analysis conservatively uses the bounding temperature for the core region,200 *F. All buckling checks are much less than unity, as required, and buckling in the worst case side drop will not occur.

2.7.1.3 Shielding Free Droo Stress Analysis

.The magnitude of the internal activation of the reactor vessel requires that an external layer of shielding be installed around the barrel section of the vessel. In order that the external dose rate requirements of 10 CFR 71.51 be met after all HAC events, the shielding must be showr. to b

v 2 - 45

Trojan Reactor VesselPackage - Safety Analysis Report remain intact after the free drop event described in Section 2.7.1.1 above.

The shielding is not attached directly to the reactor vessel. Instead, the shielding is installed around the vessel as separate plates and these plates are welded together. The analysis of the shielding welds is contained in Appendix 2-10 and demonstrates that the shielding will remain in place during all HAC events. Several loading conditions and orientations were considered. In addition, the loads imposed on the shielding are shown not to damage either the reactor vessel or the shielding plates.

An analysis is also included in Appendix 2-10 to determine the weld size necessary to retain the supplemental shielding during NCT events. This shielding, which will be installed as necessary to meet NCT dose rate requirements, does not need to be in place during or afler the HAC events since the maximum allowable HAC dose rate levels are higher than the allowable NCT levels.

In summary, the shielding is shown to be adequately restrained such that it will remain in place as necessary in the HAC events.

2.7.1.4 Fracture Toughness Considerations The RVP is analyzed to show that, when exposed to the impact levels determined in Section

/_T V 2.7.1.1, the maximum possible hypothetical flaw remains stable. The analysis uses crack initiation methodology and conservative, lower bound values for fracture toughness to demonstrate flaw stability. The basis for the methodology used is discussed in Section 2.1.2.3.

Worst-case values are chosen for all relevant evaluation parameters to ensure a conservative result.

As a result of applied loadings, a concentrated stress is developed at the tip of any flaw which may exist in the RVP containment boundary. This stress intensity, denoted iK , is defined as a function of the flaw size and shape, and the applied stress distribution in the region of the flaw tip, and is unrelated to the stress intensity as used elsewhere in this report. It represents a driving force for flaw initiation (growth). The resistance of the material to flaw initiation is denoted critical stress intensity, oricK , and is a function of the material, its temperature, and other factors, such as copper content and total neutron fluence. In the evaluations of Appendix 2-12, the crack arrest critical stress Intensity, Ku, which is lower than Kic, is conservatively used as the flaw stability criterion. It is demonstrated that the resistance to flaw initiation, Ka, is greater than the driving force, K i, for all of the governing locations in the RVP.

The stress intensity, i K , is a function of the flaw size and shape, and the applied stress distribution in the region of the flaw tip. The larger the flaw size, the greater the stress intensity.

The maximum hypothetical flaw size is taken from Table IWB-3510-1 of the ASME B&PV O

O 2 - 46

\/

1 ^

Trojan Reactor VesselPackare- Safety Analysis Report

@J O

Code,Section XI, Article IWB 3000. A full circumference crack geometry is conservatively assumed, located in a region of maximum local tensile stress. It is thus oriented normally to the axial (o,) component stresses, and is bounding since the axial stresses are larger than the hoop (o y) stresses. The stress intensity is found from a relation of the form K, = 5 f(a,0,F) where a is maximum crack depth, o represents the applied stress distribution local to the crack, and F represents geometric dependence (Reference 2-15).

The resistance of the material to flaw initiation, Ku, or conservatively, Ka, is considered next.

This value is known as fracture toughness, and is a function of the nil-ductility transition temperature of the material, or RTm. This temperature is measured and recorded for the vessel material at the time of fabrication, and is adjusted for the effects ofneutron bombardment using Regulatory Guide 1.99 (Reference 2-16). Calculated values of neutron fluence are used in this evaluation. For the RVP, only the core beltline region received a significant neutron dose during the reactor's service life. Given the adjusted RTm value and the temperature of tae material under HAC, the minimum material fracture toughness is taken from Figure 0-2210-1 of the ASME Code,Section XI. The factor of safety on brittle fracture for the free drop event is then determined from

• O FS= K" K,

I Critical locations for fracture evaluation are determined from the finite element model described in Appendix 2-9. These locations are: near the center of the upperimpact limiter location, near the center of the lower impact limiter location, at the upper bound, middle, and lower bound .

locations of the core region, and in the radius between a nozzle and the vessel shell. The minimum HAC temperature of-20 *F is assumed, conservatively ignoring decay heat. The minimum factor of safety on flaw initiation for the RVP is 2.0, for the nozzle-body radius. The fracture evaluation is described in detail in Appendix 2-12.

2.7.1.5 Summarv of Results l

The preceding analyses demonstrate that the RVP has positive margins of safety for all HAC free drop events. Table 2-16 summarizes the margins of safety for the containment components of the RVP.

2 - 47

. . . . .---U

w Troian Reactor VesselPackare- Safety Anahsis Report O 2.7.2 CRUSH 10 CFR 71.73(c)(2) requires that a crush tu be performed only for packages that have a mass not greater than 1100 lbs, an overall density not greater than 62.4 lbs/fP, and radioactive contents greater than 1000 A2 not as special form radioactive material.

The RVP will weigh nearly 2 x 10'lbs (without impact limiters). Based on this value the crush test of 10 CFR 71.73(c)(2) is not required.

2.7.3 PUNCTURE The RVP is evaluated for puncture resistance under HAC as defined in 10 CFR 71.73(c)(3). The puncture event is defined as the ability to withstand a 40-inch drop onto a vertical, cylindrical mild steel bar,6" in diameter in a position and location for which the maximum damage is expected. The load used in the analysis is based on the maximum load that the mild steel puncture bar could impart under the defined accident conditions. It is shown that the acceptance criteria established in Section 2.1.2 are satisfied for the containment boundary. These criteria are that stress remains elastic in the sealing region of the vessel body and upper head; that 7 attachment stud preload not be significantly affected; and in accordance with Regulatory Guide

( 7.6, stresses in the vessel components satisfy the following:

P. < 2.4S or 0.7S , whichever is less P + P < 3.6S, or S,, whichever is less In addition, Section 2.7.1.4 demonstrates that the maximum possible hypothetical flaw in the vessel containment boundary remains stable under the bounding conditions of ambient temperature, material toughness, and applied stress.

10 CFR 71.73(c)(3) requires that the package strike the puncture bar "in a position for which maximum damage is expected." However, as discussed in Section 2.7.1, no credible drop scenarios exist. However, given the drop scenario chosen for analysis, not all orientations are credible for the RVP. Throughout all transport operations, the vessel is handled and maintained in an essentially horizontal orientation. Therefore, while puncture impacts to the vessel side (in which the bar axis could be aligned with the CG of the package for maximum damage) are reasonable given the drop scenario discussed above, puncture impacts to the upper and lower spherical ends (heads) would be limited to less damaging impact in which the bar axis did not extend through the package CG. However, worst case governing analyses of puncture impact, in which the puncture bar axis is aligned with the package CG, are performed for the upper and lower heads. The following six puncture analyses are performed and are contained in Appendix 2-11, as shown on Figure 2-2.

(

2-48 l

Tro an Reactor VesselPacka e- so e Anal sis Re art t0 G'

1.

Puncture on the vessel side in the cylindrical region below the nozzles, a the package CG is directly above the impact location. This region has th wall thickness of any cylindrical region, and is the area of greatest neutrt fluence, which leads to reductions in material ductility. The puncture foi l applied directly to the vessel wall, and the presence of the extemal shieli I conservatively ignored.

2.

Puncture on the lower head, with the axis of the bar concentrically aligne the largest through wall penetration (1.5" diameter instrument penetratio) aligned with the package CG.

3.

Puncture on the upper head, with the axis of the bar concentrically aligne the largest through wall penetration (4.0" diameter Control Rod Drive Mi (CRDM) penetration), and aligned with the package CG.

4.

Puncture on a Type 304L nozzle cover, oriented normal to the plane of th with the bar conservatively assumed to be directly below the package CG

5. Puncture on the upper head in such a manner as to place the maximum los

/9

'V the head attachment studs. The puncture bar is aligned parallel to the seal surface, and is conservatively assumed to transmit the entire puncture loat head apex point, thus maximizing the moment arm of the load.

6.

Puncture on the upper and lower head penetration covers, with the punctui axis oriented parallel to the plane of the covers, in such a manner as to app worst case shear loads to the cover attachment welds. '(Figure 2-1, inset, s:

only the lower cover).

2.7.3.1 General Puncture Conditions The following general conservative conditions apply to all puncture analyses:

1.

The puncture bar is assumed to be as long as necessary to reach the impact and the potential for any bar buckling response is ignored.

2.

The presence of the 5/32" thick stainless steel weld overlay on the inside si of the vesselis ignored.

3.

The LDCC material is ignored insofar as it may support :bc puncture bar fr D

(u.) 2 - 49

Troian Reactor VesselPackage- Safety Analysis Report O

Table 2-11 Summary of Results for Reduced External Pressure Membrane Membrane + Bending Stress Intensity, ksi Stress Intensity, ksi Component / Material Allowable Calculated Allowable Calculated Location Stress, S,,, Stress Stress,1.5 S. Stress

@ 200 'F @ 200 'F Upper Head SA-533 Gr B 26.7 0.18 40.05 0.22 Cl 1 Upper Head SA-533 Gr B 26.7 0.05 40.05 028 Flange Cl 1 Vessel Flange SA-533 Gr B 26.7 0.09 40.05 0.34 Cl1 Cylindrical Shell SA 533 Gr B 26.7 0.18 40.05 0.19 Cl1 Lower Head SA-533 Gr B 26.7 0.20 40.05 0.23 Cl i Inlet / Outlet Nozzle O Shells Nozzle Safe Ends SA-508 Cl 2 SA-336 Gr F316 26.7 20.0 0.11 0.11 (F8M)

Nozzle Closure SA-240 Type 304L -- -

25.05 0.82 Plates' AllOther Closure SA-516 Gr 70 ---

Plates:

34.65 0.55 Average Stress Intensity, ksi Maximum Stress intensity, ksi Upper Head SA-540 Gr B24 Cl Attachment Studs 3 82.82 20.21 124.24 20.45 Notes:

1. Stress is listed for the governing outlet nozzle closure plate.

2. Stress is listed for the next-highest governing CRDM closure plate.

3. The allowable value for average stress intensity is 2S .

4. The allowable value for maximum stress intensity is 3Sm .

5. Welding consumables:

Allowable stresses / stress intensities for welds are based on lower of the two base metals as referenced in Section 2.1.2.2.

O

O barimpact, 200 'F, 0.7

' fety Analysis Report a

f Therefore,I Table 2-12 2.7.3.2.2 try of NB-3133 External Pressure Analysis An axisymr Appendix 2 All wable RVP Region assumed tol Extemal Pressure, hole is cons are 2.75" iniper Head 1,747 bar will esse I

region of welindrical Shell 1,632 puncture barove nozzles) of the puncthindrical Shell G Puncture barlow nozzles) solid 6" diar 1,327 sornewhat g uer Head 1,376 The resulting head, near th t Nozzles 2,276 allowable str Class 1, S,is.let Nozzles 2,202 margin ofsal The maximur hole, at the se For SA-533 a O

O

noor VesselPackaee- Safety Analysis Report MS= *

- 1 = + 0.18 68,028 imum membrane stress intensity, P,,,, is 43,280 psi, located on the inside edge of the te section midpoint. In this case, the allowable stress is the lesser of 2.45,,, or 0.7S,.

33 at 175 'F,0.7S,is goveming, where S,is again 80,000 psi. The margin of safety is:

MS= '

' 1 = + 0.29 43,280 Enal andysis was performed assuming the absence of the CRDM penetration hole and n resulting maximum stress intensity, P,,, + P., is 39,226 psi. The allowable stress is S,,

agin of safety is:

MS= *

- 1 = + 1.04 39,226 stum membrane stress intensity, P,,,, is 25,763 psi at the section midpoint. In this case, kle stress is 0.7S , and the margin of safety is:

MS= ' *

- 1 = + 1.17 25,763 the upper head is not penetrated by a puncture event.

Puncture on a Nozzle Cover Id outlet nozzle covers are flat circular plates,2.5" in thickness, and made from 240, Type 304L. The puncture bar is assumed to contact the cover normal to its fevelop the full load of 1.4 x 10'lb. The shear perimeter on the outside of the plate is the inside, is (6 + 2t)x, due to the spreading of the load through the plate thickness, t =

Dvtrage cylindrical shear perimeter is, therefore, (6 + t)x. The shear area is 66.76 in 2, mess is, therefore,20,971 psi. The allowable shear stress is 60% of the lesser of 2.4S,,,

or th e Type 304L plate material at 175 'F,2.4S,,, is governing, the margin of safety is:

2 - 53

i Trojan Reactor Vessel Packare - Saferv Analysis Report I

'J MS= '

- 1 = + 0.15 20,971 Therefore, a nozzle cover is not penetrated by a puncture event.

2.7.3.2.5 Puncture on The Side of The Upper Head (Stud Analysis)

The upper head is attached using n = 54, nominally 7" diameter studs (minimum diameter of 6.8") on a radius of R = 95.94", pretensioned to 720,000 lb. As determined in the upper head attaclunent stud analysis performed in Section 2.7.1.2, the stud load due to an internal pressure of 100 psi alone is F, = 40,728 lb each. This load is added to the load arising from the puncture impact. A puncture impact on the upper head causes a load to be developed in the attachment studs. The analysis may be simplified and rendered conservative by assuming that the puncture bar load is applied to the head at its apex, in a direction parallel to the sealing surface. The outside radius of the upper head is 90.17", and the distance from the spherical center to the sealing face is approximately 0.8", so that the moment arm of the puncture force is 90.17 + 0.8 = 90.97". Since the maximum imr force is 1.4 x 10'lb, the maximum moment on (l the stud attachments is M = 1.4 x 10'(90.97) = 1.27 x 108 in-lb. As shown in Section 7 V Appendix 2-9, a conservative estimate of the maximum stud force, including the pressure load is 89,756 lb. Since this load is small relative to the preload force of 720,000 lb, there is no added tensile load on the stud due to the puncture event, and the flange seal compression is unaffected.

If friction between the upper head and the main vessel is neglected, the average shear load on a stud is 25,926 lb. The shear stress in the studs due to the puncture load is 723 psi. The tensile stress due to the preload is 20,067 psi. These stresses may be combined to give SI = /02,472 = 20,119pst For the ASTM SA-540 Grade B24 Class 3 studs, the value of S at a temperature of 175 'F is 41,875 psi. The margin ofsafety is:

MS= '

- 1 = + 1.0 8 20,119 Therefore, neither the head stud nor the upper head seals are damaged by the puncture event.

l (

) 2 - 54

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

Trojan Reactor VesselPackage - Safety Analysis Report O 2.7.3.2.6 Puncture (lateral) On Upper and Lower Penetration Covers ,

The penetration covers protrude from the outer surface of the RVP by an amount equal to their thickness, %". If contacted by the puncture bar, the puncture load will develop a shear stress in the cover attachment weld. The smallest penetration covers are located on the lower head, and are %" thick and 2%" in diameter. They are attached to the vessel with a %" fillet weld all around. The shear area of this minimum weld is then 3.05 in 2. The maximum force which can be applied by the puncture bar is equal to the projected edge area of the cover times the maximum .

flow stress of the bar, or 86,500 lb and results in a shear stress, t, of 28,360 psi. For accident conditions, the allowable stress is the lesser of 2.4S or 0.7S,. Of the two materials making up the weld (vessel head and penetration cover), the one having the minimum S, is the SA-516 Gr 70 cover, where S, = 70,000 psi. The margin of safety is:

y s = (0. q 0.U S" - 1 = + 0.04 T

The CRDM peretration covers on the upper head are 6.5" in diameter, and are also %" thick and ,

attached with a %" fillet weld all around. The calculated weld area of the upper covers is A = (6.5/2.75)(3.05) = 7.21 in .2The maximum applied force is F, = (6.5/2.75)(86,500) = 204,4 lb. The stress in the attachment weld is 28,360 psi which is the same value as for the lower covers. Therefore, the upper and lower penetration covers cannot be dislodged by a puncture event.

2.7.3.3 Summary of Results The preceding analyses demonstrate that the RVP has positive margins of safety for all analyzed l HAC puncture events, as summarized in Table 2-17. I 1

2.7.3.4 Fracture Toughness Considerations The RVP is analyzed to show that, when exposed to the stresses resulting from puncture bar impact, the maximum possible hypothetical flaw remains stable. The analysis uses crack initiation methodology and conservative, lower bound values for fracture toughness to l demonstrate flaw stability. Worst-case values are chosen for all relevant evaluation parameters j 1

to ensure a conservative result.

The fracture evaluation for the case of HAC puncture is presented in Appendix 2-12. As discussed in Section 2.1.2.3, Appendix 2-12 utilizes the approach given in Section XI of the ASME B&PV Code (Reference 2-14) to show that brittle fracture will not occur. A quarter-circular crack is assumed to exist in the upper and lower heads, with one edge of the crack on the inside surface of the head, and the other edge on the inside surface of the penetration hole. The 2 - 55

0 Troian Reactor Vessel Packare - Safety Analysis Report O

( V-crack face is oriented normal to the maximum tensile stress generated in the puncture event. The crack is located completely within a weld deposit ofInconel metal. As this material has relatively high fracture toughness at low temperatures, it is conservatively ignored, and the properties of the ferritic carbon steel base metal are assumed. An aspect ratio of 0.5 is used, and the maximum hypothetical crack size selected from Table IWB-3510-1 of the ASME B&PV Code,Section XI, Article IWB 3000. The stress intensity is found from a relation of the form

~

Kf =[n'a f(a,0,F) where a is maximum crack depth, o represents the applied stress distribution local to the crack, and F represents geometric dependence (Reference 2-15).

The nil-ductility transition temperature, RTwr, of the upper and lower heads was measured and recorded at the time of vessel fabrication. Since the heads did not experience a significant neutron fluence, no modification of the as-built RTer values is necessary. From this information and from the temperature of the material under HAC, the minimum material fracture toughness is taken from Figure G-2210-1 of the ASME Code,Section XI. The factor of safety on brittle fracture for the puncture event is then determined from h K FS= 2

(& K, For a minimum material temperature under HAC of-20 F, the minimum factor of safety on flaw initiation is 1.0, for the upper head. Based on the worst case assumptions utilized, the transponation route's lack of hard targets, and the overall extremely low probability of an accident, this is considered acceptable. However, an analysis was completed based on a .

minimum vessel wall temperature of 45 F, and the safety factor increased by approximately 40%. This resulted in a safety factor of at least 1.4 against crack initiation. The complete fracture evaluation is described in Appendix 2-12.

2.7.4 THERMAL The HAC 10 CFR 71.73(c)(4) requires that the effects of a fully engulfing fire of 1,475 'F and 30 minutes in duration be assessed. This is assumed to occur after the free drop and puncture events and to consider any damage which might be expected as a result. The temperatures in the components of the RVP as a result of the fire are calculated in Chapter 3. The maximum temperature of the vessel containment boundary increases from a steady-state pre-fire temperature of 196 *F to a fire peak temperature of 844 'F. The bulk average temperature of the LDCC material intemal to the RVP increases from a steady-state pre-fire value of 289 'F to a fire peak of 310 *F. As shown in Section 2.6.1, there is no interference between the vessel 2 - 56 I

Trolan Reactor VesselPackaee- Safety Analysis Report containment boundary and the LDCC as a result of thermal expansion under 100 *F ambient, full

, solar conditions. In the HAC fire, the vessel containment boundary experiences a much larger increase in temperature (greater than 600 'F) than does the LDCC (310 - 289 = 21 'F).

Therefore, the containment boundary expands more than the LDCC, and no thern al stresses are i induced due to differential thermal expansion in the HAC fire event.  !

The peak intemal pressure of the RVP during the HAC fire event can conservatively be determined by considering the rise in temperature between the temperature at which the vessel is ,

sealed after injection of the LDCC and the peak temperature during the fire. The temperature of ,

the gas within the RVP is taken as equal to the bulk average temperature of the LDCC which fills i the interior. The minimum ambient temperature at which the vessel can be sealed is 40 'F, at l

which point the bulk average temperature of the LDCC is 218 'F. As stated above, the maximum  ;

bulk average temperature of the LDCC during the fire is 310 'F. This increase in temperature is i small enough that the imernal pressure increase is insignificant. Instead of using the calculated pressure for the HAC analyses, a bounding value of 100 psi was used in all the analyses l

described in Section 2.7.1.2. The resulting margins of safety are positive and demonstrate the i ability of the RVP to withstand the HAC fire event. >

2.7.5 IMMERSION I 10 CFR 71.73(c)(6) requires that the package be subjected to water pressure equivalent to immersion under a head of water of 50'. An external pressure of 21.7 psig is considered to meet .

these conditions.

~

The following section describes the analysis documented in Appendix 2-13. The pressure  ;

- intemal to the RVP is conservatively evaluated at its minimum value, and no support from the  !

' LDCC or intemal structures is assumed. In Section 2.6.4, the minimum pressure within the RVP  ;

under conditions of-20 'F and no insolation is calculated to be 11.1 psia. The maximum differential pressure in the immersion case, therefore, is:  !

1 P, = 21.7psig + t 4.7 psia -P, = 25.3pst

- Maximum allowable external pressures for the RVP upper head, lower head, and cylindrical body are calculated using ASME B&PV Code,Section III, Division 1, Subsection NB, Class 1,

-Article NB-3133. The RVP structural shell, including upper and lower heads, flanges, and main body section, is made of SA-533, Grade B, Class I carbon steel. The modulus of elasticity at the minimum temperature of-20 'F is 29.6 x 106 psi. Maximum allowable extemal pressures are calculated as described in Section 2.6.4, and listed in Table 2-12. For the components listed, the smallest, goveming value is 1,327 psi for the main body cylindrical shell. For an applied

,, 2 - 57 I

l Troian Reactor VesselPac.*we- Safety Analysis Report O extemal pressure of 25.3 psi, the margin of safety is very large. Therefore, buckling of the RVP l'

in the case ofimmersion is not of concem.

The RVP is further analyzed to demonstrate that Regulatory Guide 7.6 allowables are met for the I increased extemal pressure case, namely a limit of 2.4S, or 0.7S , whichever is less, for primary ,

membrane stress, and a limit of 3.6S, or S , whichever is less, for membrane plus bending stress.  !

Stresses in all RVP components are determined for the case of net external pressure in Section  !

2.6.4. The only difference between the stresses reported in that section and the stresses which result from immersion is an increase in the net external pressure. Therefore, the stresses in the RVP components for the case ofimmersion are obtained by scaling the values given in Table 2-13. The external pressure used in generating the stresses in Section 2.6.4 is 8.9 psi, and the i scaling factor used to obtain immersion stresses is, therefore,25.3/8.9 = 2.843. A temperature of  !

-20 'F is assumed for all components. The results of the scaling and appropriate allowables are given in Table 2-18.

The maximum membrane stress intensity occurs in the lower head, and is equal to 0.34 ksi. As j shown in Table 2-18, the value of 0.7S,is governing for the material of the lower head, made from ASTM SA-533, Grade B, Class 1, and is 56 ksi at -20 F. The minimum margin of safety l on membrane stress is:

i D MS= 6 _, , ,,,,, l 0.34 )

The maximum membrane plus bending stress intensity occurs in the vessel flange, and is equal to 0.65 ksi. As shown in Table 2-18, the value of S, is governing for the material of the lower head, made from ASTM SA-533, Grade B, Class 1, and is 80 ksi at -20 *F. The minimum margin of safety on membrane plus bending stress is: .

'O MS= -1 = +1arge 0.65 The governing nozzle closure plate is for the outlet nozzle, where the membrane plus bending stress intensity is equal to 1.27 ksi. As shown in Table 2-18, the value of 3.6S, is governing for the nozzle closure plate material, made from ASTM SA-240, Type 304L, and is 60.2 ksi at

-20 F. The minimum margin of safety on membrane stress is:

MS= - 1 = + 4 6.0 1.27 All other closure plates have both lower stress and higher allowables. The governing closure O 2 - 58

Trolan Reactor YesselPackare- Safety Analysis Reoort o plate weld stress occurs again for the outfet nozzle, and is equal to 0.57 ksi. As shown ic Table 2-18, the nozzle closure plate welds, made from E308L weld rod, have an allowable stress of

24.05 ksi at -20 'F. The minimum margin of safety on membrane stress is

,

MS= - 1 = + 41.2 0.57 All other closure plate welds have both lower stress and higher allowables. These calculations demonstrate that the RVP has positive margins of safety for increased external pressure.

2.7.6

SUMMARY

OF DAMAGE I

From the analyses presented in Section 2.7.1 through 2.7.5, it is shown that the hypothetical i accident sequence does not result in any significant structural damage to the RVP. All criteria I established for HAC in Section 2.1.2 are satisfied. Permanent damage occurs to the impact l limiters as a result of the free drop, and to the vessel as a result of the puncture bar impact, and is acceptable as discussed in Sections 2.7.1 and 2.7.2, respectively. Further, the maximum hypothetical flaw remains stable when subjected to the stresses resulting from the HAC free drop and puncture events, as detailed in Sections 2.7.1.4 and 2.7.2.5, respectively. Thus, the j requireinents of 10 CFR 71 are satisfied. -

2.8 SPECIAL FORM 10 CFR 71.4 defines special form radioactive material as material which meets the following conditions:

- - i

1. It is either a single solid piece or is contained in a sealed capsule that can i be opened only by destroying the capsule

2. The piece or capsule has at least one dimension not less than 5 millimeters 1 (0.197 inch)

3. It satisfies the test requirements of 10 CFR 71.75 The RVP does not meet the def'mition requirements specified; therefore, it is not considered special from radioactive material.

( This application is submitted for approval of the RVP as a Type B (as exempted), exclusive use shipping package and, therefore, the testing requirements of 10 CFR 71.75 are not applicable.

2 - 59

4. 4sg 4. te .aNa,i guWu-m4_be.w.-a.ma---.*ag.u a e dii d Wa4=4 Ed---'4-6WQJo eg Me a N en ei . 6ElimM-d #@e J4AhadA%Wwh o amJo 4-N 4 s %%m. . M- m-em e aes.6h4 4- d ee8MA md4 44 s JM J -W o. 4 4*rehm. me A(J, rJ .6 24a h- _J M4ae I l e ,

Trojan }lextor Vessel Packare - Safety Analysis Reoort O - 2.9 FUEL RODS

- The RVP does not contain fuel rods. Therefore, this section does not apply.

I l

I 1

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I 1

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2 - 60

=

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L i l

Trolan Reactor Vessel Packaee - Safety Analysis Report

1. 2.10 REFERENCES 2-1 Regulatory Guide 7.6, " Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels," Revision 1, March 1978. ,

1 2-2 Regulatory Guide 7.11, " Fracture Toughness Criteria of Base Material for Ferritic l Steel Shipping Cask Containment Vessels with a Maximum Wall thickness of 4 l . inches (0.lm)," June 1991.

2-3 ASME Boiler & Pressure Vessel Code,"Section II, Materials,1992 Edition.

L 2-4 ANSI /AWS DI.1-83," Structural Welding Code - Steel," American Welding Society.  !

l l

2-5 ANSI N14.24-1985, "American National Standard for Highway Route Controlled  !

Quantities of Radioactive Materials - Domestic Barge Transport." l 2-6 "ANSYS Engineering Analysis System User's Manual," DeSalvo, G.J. and f

Gorman, R.W., Swanson Analysis Systems, Inc., May 1990.

2-7 "ASME Boiler & Pressure Vessel Code,"Section III, Subsection NB,1992 Edition.

2-8 " Design of Structures for Missiic Impact," Topical Report BC-TOP-9-A, Rev. 2, ,

Linderman, R.B., Rotz, J.V., Yeh G.C.K., Bechtel Power Corp.,1974. '

7 2-9 Metals Handbook, American Socit.,y for Metals, page 4-91, April 1992

• 2-10 ANSI 14.2, " Proposed American National Standard Tiedowns for Truck Transport of Radioactive Materials," March 1993.

. 2-11 Warren C. Young, Roark's Formulas for Stress and Strain, Sixth Edition, McGraw-Hill, Inc.,1989.

L 12 International Atomic Energy Agency, Advisory Material for the IAEA

! Regulations for the Safe Transport of Radioactive Material, Third Edition, Vienna,1990.

I i

Lo L

e 2 - 61

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

Trolan Reactor Vessel Packare - Safety Analysis Report l

\

U 2-13 M. W. Schwartz, " Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping containers Greater than 4 Inches Thick,"  ;

NUREG/CR-3826, Lawrence Livermore National Laboratory, Livermore, CA,

' July 1984 l i

2-14 ASME Boiler and Fressure Vessel Code,Section XI, Appendix A " Analysis of  ;

Flaws", American Society ofMechanical Engineers,1995. '

2-15 C. B. Buchalet and W. H. Bamford, " Stress Intensity Factor Solutions for .

Contmuous Surface Flaws in Reactor Pressure Vessels," Mechanics of Crack l Growth, ASTM STP-590, American Society for Testing and Materials,1976, pp.

385-402.  ;

2-16 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessels Materials," U.S. Nuclear Regulatory Commission, May 1988.  !

2-17 Courtney, T. H., Mechanical Behavior of Materials, pp 667-668.  :

i 2-18 . Thompson, T. J., and J. G. Beckerley, The Technology of Nuclear Reactor Safety, Volume 2, pp 132-134.

O -

2-19 Analysis of Capsule U from Portland General Electric Company Trojan Reactor Vessel Radiation Surveillance Program, WCAP-9469, Westinghouse Electric  ;

Corporation, May 1979. l 2-20 Analysis of Capsule X from Portland General Electric Company Trojan Reactor Vessel Radiation Surveillance Program, WCAP-10861, Westinghouse Electric )

Corporation, June 1985.

i 2-21 Analysis of Capsule V from Portland General Electric Company Trojan Reactor l Vessel Radiation Surveillance Program, WCAP-12868, Revision 2, Westinghouse )

Electric Corporation, December 1991.

1 1

O 2. e2 a .'

f'

0 Trolan Reactor VesselPackare- Scfety Analysis Report

(~)\

Table 2-1 Summary of Results for Normal Conditions of Transport Section Condition Calculated Value Allowable Value  % of (KSI)' (KSI)' Allowabic Value' 2.6.1 Heat j

- Max Primary Plus Secondary Stress 17.1 3S. = 60.0 29 2.6.2 Cold

- Max Membrane Plus Bending Stress 2.2 1.5S. =25.05 8.7

- Max Primary Plus Secondary Stress 17.1 3S. = 60.0 29 2.6.3 Reduced External Pressure

- Max Membrane Stress 0.20 S. = 26.7 0.7 O - Max Membrane Plus Bending Stress 0.82 1.5S,,, = 25.05 3.3 2.6.4 Increased External Pressure

- Max Membrane Stress 0.12 S. = 26.7 0.5

- Max Membrane Plus Bending Stress 0.45 1.5S. = 25.05 1.8 2.6.5 Vibration

- Max Membrane Plus Bending Stress 17.5 1.5S. '= 40.05 44

-Max Altemating Stress 8.8 12.5 70.2  ;

I 2.6.7 Free Drop (one foot) l

- Max Membrane Plus Bending Stress See Note 2 See Note 2 See Note 2 2.6.10 Penetration

- Max Penetration Depth 0.118 inches 0.625 inches 18.9 Ho1CL 1 Values provided are for tha limiting package component 2 See summary of results in Section 2.6.7.

C

.n+

Trolan Reactor VesselPackare - Safety Analysis Report O

Table 2-2 Summary of Results for Hypothetical Accident Conditions Section Condition Calculated Value Allowable Value  % of (KSI)' (KSI)' Allowable Value' 2.7.1 Free Drop

- Max Membrane Stress 14.7 0.7 S = $6 26

- Max Membrane Plus Bending Stress 26.4 S,-80 33

- Max Stud Stress Intensity 20.1 S, =41.9 4g 2.7.3 Puncture

- Max Membrane Plus Bending Stress 68.0 S, = 80 85

- Max Shear Stress 21.0 (.6)2.4S, - 24.0 88

-Max Weld Shear 28.4 (.6)(.7)S,=29.4 97 2.7.4 Thermal See Nete 2 See Note 2 See Note 2 2.7.5 Immersion

- Max Membrane Stress 0.34 S. = $6.0 1

- Max Membrane Plus Bending Stress 1.27 1.5S = 60.2 -

2 Notes.

I Values provided are for the limiting package component 2 See summary of results in Section 2.7.4.

O

. ..- -.- . . - - . -- .- . . . - - - . . . . .- . - - -.~.- -

Trolan Reactor VesselPackaee- Safety Analysis Report Table 2-3 Reactor Vessel Containment Boundary Components RYP Component Dimensions Material of Construction (inches) . .

Bottom Head inside radius 88.2 SA 533 Grade B Class 1 Min. Thickness 5.5' Interior Surface 5/32 inch S. Steel overlay First Shell hg & Insit.e Diameter _ 173 SA-533 Grade B Class i Second Shell hg Norainalnickness 9.5 Interior Surface 5/32 inch S. Steel '

Mraimum Thickness 8.7' overlay Ht:ight 108 i Inlet Nozzles hside Diameter 36.2 x 27.2 SA-508 Class 2 Nominalnickness 12.3  ;

Outlet Nozzles inside Diameter 35.6 x 28.8 SA-508 Class 2 Nominalnickness 10.8 Third Shell hg Inside Diameter 170.9 SA-533 Grade B Class 1 Nominalnickness 11.1 Interior Surface 5/32 inch Minimum Thickness 10.78 S. Steel overlay i Shell Flange Inside Diameter SA 508 Class 2 O At nird ShellRing Core Support Ledge Gasket Seal Ledge 170.9 167 182.7 Minimum nickness8 At Third Shell hg - 10.7' Core Support Ledge 19.0 Gasket Seal Ledge 17.6 Studs Length 57.5 SA 540 Grade B 24 Class 3 Diameter (min) 6.8 Metallic O nngs Inner: Inconel Alloy 718 178.4 Silver plated OD i Outer: )'

181.0 OD l Head Flange Inside Diameter 172 SA 508 Class 2  ;

Minimum Thickness ' 6.7 i Top Head Inner Hemispherical Radius 83.5 SA-533 Grade B Class 1 Minimum nickness 6.78 Notes:

1. Minimum thickness includes internal stainless steel cladding.

2. Thickness includes 7% inch diameter stud holes located in shell flange assembly area above core support ledge O

Trojan Reactor VesselPackare - Safety Analysis Report 9 Table 2-4 Reactor Vessel Package Calculated Weights Component Calculated Weight,lbs Reactor Vessel and Internals (includes closure plates) 1,295,712 Shielding 241,292 LDCC2 344,435 Impact Limiter 144,460 Total Weight of Package 2,025,8995 Notes:

1. This is maximum installed weight.

2. Assumes the maximum concrete density of 65 lbs/fP. .

3. Weight of 2,040,000 lb used in all calculations.

O

Trojan Reactor VesselPackare - Saferv Analysis Report

. Table 2-5 Range of Minimum Mechanical Properties of Steel Materials (-40 *F - 200*F)'

Young's Yield Tensile Material Application Modulus Specification Strength Strength E (x 103 ksi) Sy (ksi) S, (ksi)

SA-533 Gr.B Reactor Vessel Shell, 29.6-28.5 Class 1 50-47.2 80 Top Head SA-508 Inlet and Outlet Nozzles 28.2 - 27.1 Class 2 50-47.1 80 SA-240 Type 304L Nozzle Closure Plates 28.8-27.6 25 - 21.3 70 - 66.2 SA-336 Nozzles Safe Ends 28.8 -27.6 30 -25.8 70 Grade F316(F8m)

SA-540 Grade B24 Reactor Vessel Head 28.2 -27.1 Class 3 130 - 121.5 145 Studs

[ SA-516 Grade 702 Shieldin 29.9-28.8 38 - 34.6 70 5/8 inches Penetration Closures Notes:

1.

HAC thermal (fire) metal temperature values are not included. See Appendix 3-1.

2. Normalized, to fine grain practice.

Welding consumables: Allowable stresses / stress intensities for welds are based on lower of the two base metals as referenced in Section 2.1.2.2.

1 1

1

1 Trolan Reactor Vessel Package - Safety Anahsis Reoort Table 2-6 Summary of Design Stress Intensity Values, S.

i STRESS INTENSITY (KSI) FOR METAL TEMPERATURE, 'F I MATERIAL -20 to 200 300 400 500 600 650 700 100 SA-533 26.7 26.7 26.7 26.7 26.7 26.7 26.7 26.7 Gr. B Class l' SA-240 16.7 16.7 16.7 15.8 14.8 14.0 13.7 13.5 Type 304L' SA-508 26.7 26.7 26.7 26.7 26.7 26.7 26.7 26.7.

Class 25 j

SA-336 20.0 20.0 20.0 19.3 18.0 17.0 16.7 16.3  !

Grade F316(F8m)' i SA-516 23.3 23.1 22.5 21.7 20.5 18.7 18.4 18.3 Gr. 705 l SA-540 Grade B24 43.3 41.4 40.2- 38.8 37.6 35.9 -

33.7 Class 3' Notes;

1. Stress intensity values are taken from ASME Code,Section II, Part D, Subpart 1, Tables 2A, end 4.

2-~ ~ Welding consumables:

Allowable stresses / stress intensities for welds are based on lower of the two base metals as referenced in Section 2.1.2.2.

l O

Trolan Reactor Vessel Packare - Safety Analysis Report 4

Table 2-7 Material Prop 20(erties pcfdensity)of Polyurethane Foam 4

Strain Dynamic Crush Strength Dynamic Crush Strength j (%) (at -20 *F), psi (at 160 'F), psi  ;

upper bound 1

lower bound '

Parallel Perpendicular Parallel Perpendicular 3 to nse to nse to nse to nse j

10 2736 2889 999 1060 20 2904 3028 1085 1149 30 3134 3251 1207 1255 40 3700 3758 1421 1469 50 4826 4873 1842 1881 O 60 7891 7871 2924 2948 l

1 i

O

Troian Reactor VesselPackare- Safety Analysis Report Table 2-8 Design Accelerations forTiedown System Design Tiedown Direction Acceleration Basis for Design -

Component (g) Criteria Longitudinal *10 10CFR71.45(b)(1)

Transverse *5 10CFR71.45(b)(1)

Vertical *2 10CFR71.45(b)(1)

Longitudinal 1.5 ANSI NIC TransporterTiedowns Transverse

• 1.6 ANSI N14.2i Vertical 1.5 ANSI N14.2 Longitudinal *l .5 ANSI N14.24 BargeTiedowns Transverse *l .6 ANSI N14.24' Vertical 1.25 ANSI N14.24 Notes

1. Transverse accelerations based on probabilistic safety study.

2. Vertical acceleration for wave motion (downward) and for capsize (upward) in normal orientation.

O

I 0 -

Trolan Reactor VesselPackare- Safety Analysis Report h' O

Table 2-9 Summary of Margins of Tiedown Induced Stresses l

Section Condition Location Calculated Allowable  % of I Value Value Allowable (ksi) (ksi) Value 2.5.2 Vertica', Restraint Head .

Flange

- Maximum Stress 29.6 S, = 3 8 78 2.5.2 Longitudinal Restraint Nozzle Transition

- Maximum Stress 43.5 S, = 50 87 2.5.2 Lateral Restraint Vessel O

\,J Wall

- Maximum Stress 48.5 S, = 50 97 l

I

- . . ~ . . . . . . ..

i  !

Trolan Reactor VesselPackare - Safety Anahsis Ret > ort ,

O Table 2-10 l' ThermalStress Analysis Results

~

S Allowable Calculate Margin Component Material (ksi) Value d Value of -

(3.0 S ) (ksi) Safety RCS Nozzle SA-240 16.7 50.1 9.82 4.1 Closure Plates Type 304L Head Closure SA-516 23.1 69.3 14.4 3.8 Plates Gr70 Vessel SA-533 GR B 26.7 80.1 9.74 7.2 Cl 1 Upper Head SA-533 GR B 26.7 80.1 10.02 7.0 Cl1 Lower Head 533 GR B 26.7 80.1 9.49 7.4 Nozzles SA-508 26.7 80.1 15.87 4.0 Cl2 Nozzle Safe Ends SA-336 Gr 20.0 60.0 17.14 2.5 F316 (F8m)

Closure Stu'ds SA-540 Gr 41.4 124.2 4.62 25.9 B24 C13 Note:

Welding consumables:

Allowable stresses / stress intensities for welds are based on lower of the two base metals as referenced in Section 2.1.2.2.

O

Trolan Reactor VesselPackare - Safety Analysis Report 3 .h (J

Tabic 2-11 Summary of Re.sults for Reduced External Pressure Membrane Membrane + Bending Stress Intensity, ksi Stress Intensity, ksi Component / Material Allowable Calculated Allowable Calculated Location Stress, S,,, Stress Stress,1.5 S. Stress

@ 200 'F @ 200 'F Upper Head SA-533 Gr B 26.7 0.18 40.05 0.22 Cl 1 Upper Head SA-533 Gr B 26.7 0.05 40.05 0.28 Flange Cli Vessel Flange SA-533 Gr B 26.7 0.09 40.05 0.34 Cl1 Cylindrical Shell SA-533 Gr B 26.7 0.18 40.05 0.19 Cl I Lower Head SA-533 Gr B 26.7 0.20 40.05 0.23 Cl 1 p Inlet / Outlet Nozzle Shells SA-508 C12 26.7 0.11 -- --

Nozzle Safe Ends SA-336 Gr F316 20.0 0.11 --- -

(F8M)

Nozzle Closure SA-240 Type 304L -- --

25.05 0.82 Plates' All Other Closure SA-516 Gr 70 ---

Plates 2 34.65 0.55 Average Stress intensity, ksi Maximum Stress intensity, ksi Upper Head SA-540 Gr B24 Cl Attachment Studs 3 82.85 20.21 124.24 20.45 Notes:

1. Stress is listed for the governing outlet nozzle closure plate.

2. Stress is listed for the next-highest governing, CRDM closure plate.

3. The allowable value for average stress intensity is 2S .

4. The allowable value for maximum stress intensity is 3S .

5. Welding consumables:

Allowable stresses / stress intensities for welds are based on lower of the two base metals as referenced in Section 2.1.2.2.

O

Trolan Reactor VesselPackaec- Safety Analysis Report f

Table 2-12 Summary of NB-3133 External Pressure Analysis RVP Region All wable External Pressure, Upper Head 1,747 Cylindrical Shell 1,632 (above nozzles)

Cylindrical Shell 1,327 (below nozzles)

Lower Head 1,376 Inlet Nozzles 2,276 Outlet Nozzles 2,202 O

1 Trojan Reactor Vessel Packare . Safety Analysis Report h  ;

O Table 2-13 Summary of Results for Increased External Pressure Membrane Stress Intensity, Membrane + Bending Stress ksi Intensity, ksi Component / Material Allowable Calculated Allowable Calculate Location Stress, S,,, Stress Stress,1.5 d Stress

@ 200 'F S. 200 Upper Head SA-533 Gr B 26.7 0.10 40.05 0.11 l Cl 1 Upper Head Flange SA-533 Gr B 26.7 0.03 40.05 0.18 Cl1 Vessel Flange SA-533 Gr B 26.7 0.05 40.05 0.23 Cl I p Cylindrical Shell SA-533 Gr B 26.7 0.09 40.05 0.10 Q Cl 1 Lower Head SA-533 Gr B 26.7 0.12 40.05 0.13 Cl 1 Inlet / Outlet Nozzle SA-508 26.7 0.06 --- ---

Shells Cl 2 ,

Nozzle Safe Ends SA-336 Gr 20.0 0.06 --- --

F316 (F8m)

Nozzle Closure SA-240 Type --- ---

25.05 0.45 Plates' 304L All Other Closure SA-516 Gr 70 --- ---

34.65 0.30 Plates 2 j

Notes: 1

1. Stress is listed for the governing outlet nozzle closure plate.

2. Stress is listed for the next-highest governing CRDM closure plate.

3. Welding consumables: ,

Allowable stresses / stress intensities for welds are based on lower of the two base metals as referenced in Section 2.1.2.2.

tD v

l Trojan Reactor Vessel Packace - Saferv Analysis Report h

Table 2-14 Summary of NCT Free Drop Margins of Safety Analysis Description Margin of Safety a Containment boundary stress -

Bending +2.85 Membrane +2.93 Nozzle Radius +1.86 b Upper head attachment studs +1arge c Nozzle -

cover stress +2.78 cover attachment weld stress +2.31 O

Trolan Reactor VesselPackare. Safety Analysis Report Table 2-15 Free Drop Impact Analysis Results Foam Dro Crush Foam impact, Clear

• mP, Het t, Defonn.,

Condition 5 mn, 8 I

,p NCT -20 1.0 5.08 7.76 8.0 Nc 160 1.0 8.53 13.02 5.3 Nc HAC -20 11.0 13.64 20.81 20.1 Nc 160 11.0 24.55 37.45 11.5 Nc HAC -20 7.0 26.10 36.69 16.3 No i 160 7.0 40.43 56.68 10.0 No Note: The clearance in these cases is not governing.

O

- - - - - - - - -~ '"~"" ~ ~

Troian Reactor Vesse1 Package- Safety Analysis Report O

determined a permissible leak rate of 21.56 cm3 /sec for the NCT. This is above the limit of 10 cm)/sec given in ANSI N14.5. Therefore, no leak testing is required.

Based on the above discussion, the containment requirements for 10 CFR 71.43(f) for NCT are m et.

4.2.3 PRESSURIZATION OF CONTAINMENT VESSEL Title 10 CFR 71.71(c)(1) requires evaluation of the RVP for the effect of heat with an ambient temperature of 38 C (100 F) in still air, and with solar insolation for a 12-hour period of 2

400 g cal /cm for curved surfaces. The results of this evaluation are provided in Sections 3.4.2, with the maximum corresponding differential pressure calculated in Section 3.4.4 to be 4.3 psi.

It is shown in Section 2.6 that this increased pressure will have a negligible effect on the containment structure of the RVP.

An assessment ofpotential hydrogen generation in the packaged reactor vessel using the guidance contained in Electric Power Research Institute (EPRI) Publication NP-4938,

" Methodology for Calculating Combustible Gas Concentration in Radwaste Containers," and GEND-052, " Hydrogen Control in the Handling, Shipping, and Storage of Wet Radioactive O bead resin solidified with concrete. The use of the resin / concrete is conserva beads are a source of hydrogen that does not exist in the reactor. The LDCC with the foaming agents and water of hydration are clearly included in the model since the solidification agent used is concrete. The calculation assumed an essentially uniform distribution of hydrogen due to the ability of hydrogen to migrate throughout the interior of the vessel. The calculation determined the reactor vessel will reach 5% by volume of hydrogen 142 days following closure of the vessel. The 5% by volume of hydrogen was chosen as the limit for flammability and detonation based on the guidance in NRC Information Notice 84-72, " Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas Generation."

The shipment time period is expected to be less than 45 days including the time period from sealing the package until the actual transportation is completed. In accordance with Information Notice 84-72, the shipment must be completed within 90 days.

4.3 CONTAINMENT REOUIREMENTS FOR HYPOTHETICAL ACCIDENT CONDITIONS The radiological characterization analysis described in Section 1.2.3 demonstrates

• hat the RVP contains Type B quantity ofradioactive material. The RVP will be designated as a Type B (as 4-4

O rry Analysis Report

• O i in Section 1.1, and will be transported as an exclusive use

• l.4. In addition, the RVP will be a single use package in that Am w ill be disposed of at the licensed disposal facility of US

""= Washington.

ted) package hence, per 10 CFR 71.51 its containment should 171.51(a)(2) relative to package radiological leak-tightness and 73, Hypothetical Accident Conditions. Section 2.7 discusses 10 CFR 71.73. Table 2-2 provides a summary of the results of 1 strate compliance with these requirements.

4DIOACTIVE MATERIALS S I and thermal evaluations presented in Sections 2.7 and 3.5, ere is no release of radioactive materials under any of the HAC ERION e 'was completed as above for the NCT (Section 4.2.2) case but available for dispersal. The HAC calculation (Ref. 4-1),

14.5 for accident conditions, determined an allowable leak rate the limit of 10 cm%ec set in ANSI N14.5, Section 6.1.

ired.

RVP is specified in Sections 4.2.2 and 4.3.2 above. It has been

>ersal of contaminants is prevented by the welded containment de interior of the containment barrier to retain the contaminants e

f O

4-5

er VesselPackare- Safety Analysis Retsort 205" OD

.lVP I

336" OD l Impact Limiter l*

1

.l

! Cradle l _ *,/

Transporter

/

'fffffMMM/ff/#/fff/Mf/4 Figure 2-1 RVP Transport Configuration -

O Troian Reactor VesselPackace- safety Analysis Report w/ Ll- -

.,,/~,--__~-,_,,,,,,,,,,.

.l I*

.. - I, .

O e

~

,l g'  % ,y. M - - - _ _ _ , , _ , , , , , .

'I.J l

e O '@

@ LOWER PENETRATON COVER g

{ 1.50

-- e2.rs I

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Figure 2-2 RVP Puncture Event Orientations O

Trojan Reactor Vessel Packare - Safety Anah sis Report i O,

3.0 TIIERMAL EVALUATION This chapter identifies and describes the principal thermal engineering design aspects of the Reactor Vessel Package (RVP) which are important to safety and are required for compliance with the performance requirements of 10 CFR 71.43(g),10 CFR 71.71(c) and 10 CFR 71.73(c).

A detailed description of the thermal analyses conducted is contained in Appendix 3-1.

3.1 DISCUSSION 3.1.1 PACKAGE DESCRIPTION The RVP has the following principal physical characteristics:

1. The pressure containing boundary consists of the reactor vessel, reactor vessel upper head, reactor vessel flange O'-rings and welded closures which seal nozzle and penetration openings.

2. The nominal base metal wall thickness of the steel reactor vessel is 8 %" in the core barrel region and 10 %" in the nozzle region. The nominal base metal wall thickness

( of the reactor vessel upper head is 6 %". The nominal base metal wall thickness of the reactor vessel lower head is 5 %" (excluding stainless steel internal cladding).

3. The reactor vessel is to be filled with Low Density Cellular Concrete (LDCC) with a density between 45-65 lb/ft'.

4. A steel radiation shield is installed around the circumference of the reactor vessel shell. The nominal thickness of the radiation shield around the nozzle area is 2". The nominal thickness of the radiation shield below the nozzles is 5". In addition to the shielding mentioned above, a 1" thick by 3'long band of shielding is installed below the junction between the lower head and the reactor vessel barrel.

The passive thermal system of the RVP operates as follows:

1. Normal Transport Conditions

a. Decay heat produced by the internal components is transferred to the inner surface of the vessel by conduction through the LDCC.

b. Heat is then transferred by radiation and conduction to the outer surface of the vessel shell or radiation shielding.

a-1

l Trojan Reactor VesselPackare - Safety Analysis Rer> ort 3 '

(V

c. Heat is transferred from the outside of the package to the surrounding air by radiation and natural convection.

2. HypotheticalThermal Accident

a. The heat of a hypothetical fire is imposed upon the exterior of the entire package by both radiant heat transfer and convection, as summarize:1 in Section 3.5.1.

b. Heat is transferred directly into the unshielded upper and lower reactor vessel heads from the fire by radiation and convection.

c. Heat is transferred directly onto the radiation shielding surrounding the vessel shell area by radiation and convection. ,

d. Conductive and radiative heat transfer conveys the heat from the radiation shield to the reactor vessel shell wall.

e. Heat is transferred through the reactor vessel wall and heads by conduction into the LDCC.

O f. Subsequent to the fire, the stored heat is dissipated to the ambient environment using the same heat transfer mechanisms described for the normal transport conditions.

An internal heat generation rate due to radioactive decay of the package contents has been determined to be approximately 60,635 Btu /hr (17,766 Watts). Table 3-1 provides a summary of decay heat production based on package contents.

3.1.2 THERMAL ACCEPTANCE CRITERIA Acceptance criteria for the Normal Conditions of Transport (NCT) analysis are as follows:

1. 10 CFR 71.43(g) stipulates that the maximum accessible surface temperature for a package in exclusive use is 185 F (85 C) in the shade with an ambient environment of100 F.

2. The metallic containment vessel 0-ring seal shall be shown not to exceed its design temperature of 650 F.

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Troian Reactor Vessel Package- Safety Analysis Report O

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3. The bulk impact limiter foam temperature cannot exceed 160 F so that it may retain  !

sufHeient impact absorbing capability to withstand the worst case hypothetical drop

. accident as described in Section 2.7.1.

i Acceptance criteria for the Hypothetical Acciden! Conditions (HAC) analysis are as follows:

i

1. 10 CFR 71.51 stipulates that subsequent to the accident conditions there cannot be the  ;

escape of greater than 10 A2 of Kr" per week, the escape of greater than 1 A2 of all  ;

other radioisotopes per week, and the external dose rate at I meter from the package i cannot exceed I rem /hr. .

l

2. To this end, the thermal analysis must demonstrate that the vessel containment 0-ring l seal does not exceed its design temperature of 650*F and, in conjunction with the structural analysis, show that the thermal stresses within the vessel does not compromise the containment boundary. I 3.1.3 RESULTS The important maximum package temperatures and thermal analysis conclusions for both NCT and HAC are summarized below.

3.1.3.1 Normal Conditions of Transoort 10 CFR 71.43(g) specifies the maximum temperature for the design and construction requirements for the reactor vessel package. These requirements state the maximum accessible surface temperature of the package in still air at 38 C (100 F) and in the shade, shall not exceed a temperature of 50 C (122*F) in a nonexclusive use' shipment, or 85*C (185 *F) in an exclusive use shipment. The evaluation of the package design must also account for internal heat generation rate from decay heat. A maximum surface temperature of 147 F for this thermal loading condition is determined in Section 3.4.2. This maximum temperature is below the 185*F limit specified by 10 CFR 71.43(g). The temperatures are summarized in Table 3-3.

The maximum surface temperature of the package in still air at 38"C (100 F) including the effects ofinsolation is determined for use in stress calculations to satisfy the requirements of 10 CFR 71.71(c)(1). The evaluation of the package design must account for internal heat generation rate from decay heat and the insolation rates specified in 10 CFR 71.71(c)(1). The 2 2 reactor vessel package is a curved surface, so a total insnlation of 400 g cal /cm (1,475 BTU /ft )

for a 12-hour period is specified. A maximum surface temperature of 160 F for this thermal loading condition is determined in Section 3.4.2.

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Trolan Reactor Vessel Packaee- Safety Analysis Report '

(-_ s

- \. l 3.1.3.2 Hvnothetical Accident Conditions

~

I The maximum temperatures in the package components recorded during the thermal accident scenario are summarized in Table 3-4. The maximum outer surface temperatures occur 30 minutes after the start of the accident and cools rapidly after removal of the applied external heat source. The package inner surfaces reach a maximum temperature after removal of the applied external heat source. The maximum O-ring seal temperature of 197'F is less than the'650 F design temperature of the seal. Therefore, the metallic O-ring seals remain functional for the duration of the hypothetical accident.

3.2 MATERIAL PROPERTIES Material properties used in this analysis are shown in Section 5.0 of Appendix 3-1-. The RVP materials included in the thermal model were carbon and stainless steel, LDCC, and polyurethane foam.  !

I 3.3 TECHNICAL SPECIFICATIONS FOR COMPONENTS The reactor vessel package containment bounda y consists of the reactor vess:1 shell, upper head, and the welded penetration closures. The upper head and the reactor vessel have a design temperature of 650*F and a design pressure of 2,485 psig. The penetration covers and welds, as shown on Figure 1-4, are designed to withstand the NCT and HAC as described in 10 CFR 71.71  !

and 10 CFR 71.73.

The majority of the decay heat is located in the baffles, baffle formers, upper and lower core support plates and core barrel. To simplify the model, it was conservatively assumed that all of the 60,635 Btu /hr (17,766 watt) decay heat that will be present November 1,1997 is distributed within these five components. Table 3-1 provides a breakdown of the activity and decay heat for each of these components. To investigate a lower bound temperature for detennining the material properties of the package during accident conclitions, an additional case was run with the thermal source term being 75% ofits maximum value.

3.4 THERMAL EVALUATION FOR NORMAL CONDITIONS OF TRANSPORT This section describes the models and analyses used in the thermal evaluation of the reactor vessel package for NCT.

Trojan Reactor Vessel Package - Safety Analysis Rer> ort @

(L,')

3.4.1 THERMAL MODEL The thermal loading conditions of 10 CFR 71.71(c)(1) and (c)(2) are used to determine the maximum and minimum reactor vessel package temperatures. Additionally, the maximum temperature for accessible surfaces is evaluated using the criteria specified in 10 CFR 71.43(g).

For determination of the accessible surface temperature, the environmental conditions consist of a 100*F (38*C) ambient temperature with no insolation evaluated at steady state. For determining the performance of the package during NCT per 10 CFR 71.71, the package is evaluated at a minimum ambient temperature of-40 F (-40 C) without insolation and a maximum ambient temperature of 100 F (38 C), with insolation. A 24-hour average insolation value was used, which is justified given the large thermal mass of the package. An additional case was evaluated with an ambient temperature of-20 F (-29*C) and no insolation to provide the minimum temperatures for evaluation of the HAC. The decay heat source term of 60,635 i Btu /hr (17,766 watts) was used for these cases as discussed in Section 3.3.

To provide an additional estimate of the miniraum temperature for the vessel wall, an additional

, case was run. This case assumed an ambient temperature of-20 F, no solar radiation, a reduced f, ,) heat load of 45.477 Btu /hr (13,325 watts) and negligible air gaps between the radiation shield,

the impact limiters and the vessel.

Additional cases are evaluated at 40 F and 70 F without insolation to represent the low and high temperatures for filling the reactor vessel with LDCC. These temperatures provide the initial temperatures for determining the differential thermal expansion of the package components for high and low temperature shipping conditions.

Table 3-2 provides a summary of the environmental conditions used in the NCT evaluation.

3.4.1.1 Analvtical Model The thermal analysis of the RPV for NCT was conducted using the HEATING 7.21 computer code. HEATING 7.21 is a fm' ite difference thermal analysis code capable of solving steady-state and transient thermal analysis problems in one, two, or three dimensions in a rectangular, cylindrical, or spherical coordinate system. It is capable of modeling heat transfer via a combination of conduction, radiation, and both natural and forced convection. HEATING 7.21 was developed by Oak Ridge National Laboratories as part of the SCALE 4.3 package and has been used extensively in the packaging industry for thermal evaluations of packages for both onsite transfer and storage.

m

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0 Trojan Reacter Vessel Package Safew Anahsis Report h 7

V l 1

The RVP was simulated using an axisymmetric model. Intemal features, such as the baffles l were approximated using cylindrical geometry to produce a feature ofidentical mass and l

volume. In cases, such as the upper and lower support plates, where the components had a large '

amount of void volume, the conductivity and density of the components were adjusted by the ratio of the model volume to the component volume. External features were simplified. In the case of the nozzles, the thermal mass of the nozzles was conservatively neglected, and for the NCT case, the nozzle thermal mass and surface area was neglected. For the HAC case, the surface area of the model region that contains the nozzles was increased by a factor of two.

l The vessel walls are constructed of SA-533 Grade B Class I carbon steel varying in thickness I from 5 %" to over 20" thick at the flange. Most of the interior of the vessel is lined with 5/32" I thick SA-304 stainless steel. This lining has a negligible effect on the thermal performance of l the vessel and is therefore neglected for this analysis. Similarly, the metal penetration closures were also neglected. l l

To insure that the vessel complies with surface dose requirements, additional shielding made of SA-516 Grade 70 carbon steel varying from 2 to 5" thick is added to the vessel. This shielding is intended to have a relatively tight fit with the vessel. Given the limitations of such a fit, a uniform air gap of %" is assumed between the vessel and shielding. Heat is transferred across fl V

this gap via conduction and radiation. For the purposes of this model, all internal components, such as the baffles, baffle formers, and core support plates were assumed to be made of 304L stainless steel. The void spaces between the components are filled with LDCC that varies in density from 45 lb/ft) to 65 lb/ft). The lower value of 45 lb/ft) was used in the model to minimize the LDCC conductivity and thermal mass. The lower conductivity maximizes internal temperatures for the steady state NCT analysis. The transient vessel wall temperatures reached during the HAC fire are maximized because minimum heat is transferred into the LDCC due to its lower conductivity and lower thermal mass.

3.4.2 MAXIMUM TEMPERATURES The worst case high temperature ambient condition specified by 10 CFR 71.71(c)(1)is 100 F (38 C) with full insolation. Table 3-3 presents the temperatures of selected package components for this high temperature condition. All components are within the acceptance criteria.

10 CFR 71.43(g) requires that the maximum accessible surface temperature for an ambient condition of 100 F and no insolation be less than 185 F. For this condition, the maximum accessible surface temperature was determined to be 147 F therefore, this condition is satisfied.

Additional results are contained in Appendix 3-1.

p

\j 36

I O l Trojan Reactor Vessel Packare - Safety Analysis Rer> ort 3.4.3 MINIMUM TEMPERATURES 10 CFR 71.71(c)(2) requires that packages be analyzed for an ambient temperature of-40 F in )

still air and shade to ensure there will be no loss or dispersal of radioactive contents, no 1 significant increase in external radiation, and no substantial reduction in the effectiveness of the l packaging.

Table 3-3 contains temperatures of selected package components. The lowest internal temperatures for the package occur with -20 F ambient temperature, no insolation and minimum -

decay heat. The lowest vessel wall temperatures occur at the top and bottom of the vessel with a

-40'F ambient temperature rad are slightly higher than the ambient temperature.

3.4.4 MAXIMUM INTERNAL PRESSURES The initial condition for the maximum pressure calculation is considered to be the point just prior to sealing the vessel after the LDCC has been injected and has cured for several days. The initial conditions are then a minimum 40 F ambient temperature, no insolation, and atmospheric pressure. The maximum pressure occurs at 100 F ambient temperature with insolation. The gas temperatme inside the vessel is assumed to vary with the bulk average temperature of the LDCC,

,q and its pressure varies as a function of the ideal gas law. In addition, the internal pressure l

() exerted on the vessel wall is affected by the vapor pressure from the moisture within the concrete i matrix. A small amount of water vapor may remain in the concrete void spaces fo!!owing the j venting period and will tend to migrate toward the outer surface of the LDCC due to the higher  ;

temperatures in the core region of the vessel. At the outer surface of the LDCC (inner vessel l wall) the vapor will also tend to migrate to the lowest temperature point. This will limit the l

water vapor pressure exerted on the vessel to that corresponding to the lowest temperature present on the inner vessel surface. However, a value of 125 F (instead of the lowest -

temperature) was selected based on a review of the thermal calcuation which showed that i approximately 40 percent of the inner vessel surface (surface above the nozzles, including the upper head, and the surface of the lower head) remains at or below this temperature throughout transport. Lastly, the internal pressure exerted on the vessel wall is affected by the gas pressure from radiolysis.

Using the ideal gas law to calculate the. change in pressure between the two conditions 460 +LDCCg oo.r P go.y = x 14,7 psia 460 +LDCC g o.g V 3-7

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

l Trolan Reactor VesselPackare'- Safety Ana!vsis Report Y.

1 The bulk average temperatures for the two conditions are  !

~ LDCCg ., = 218'F

- LDCC,gion., = 289*F The resulting pressure is: i 4

460 + 289 .

Pg.,. = x 14.7 = 16.2 psia 460 + 218  ;

The hydrogen concentration at one year from the time the vessel is closed (Appendix 4-1) is 11.5 :  !

. percent. The resultant pressure contribution from the gases (hydrogen and oxygen) produced l from radiolysis is:

100 P8 ,, = ( 100.- (11.5)(1.5) x 16.2)-16.2 = 3.4pst  !

?

4

/m ' Using steam tables,'the vapor pressure for a vessel wall temperature of 125'F is 2.0 psi. i

'& l The differential pressure at the end of one year is then:

P, =( P 3 ,,,,.-P,,,) +P,,, +P,,,,, =( 16.2 - 14.7) +3.4 +2.0 = 6.9pst The shipment time period must be less than 90 days as described in Section 4.2.3 to meet the

. hydrogen gas concentration limits. Pressure from gas for the duration of transport is:

P,,, =

E x 90 days = 0.8 psi (lycar)(365 ###)

yr l

Therefore, the differential pressure for the duration of transport (90 days) is:

i P, =( P , .,. -P,,,) +P,,, +Py,, =(l 6.2 -14.7)+0.8 +2.0 = 4.3 psi

'I l

3-8

'\,

l 1

l

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Troian Reactor VesselPackace- Saferr Analysis Rer> ort h/

C 3.4.5 EVALUATION OF PACKAGE PERFORMANCE FOR NORMAL CONDITIONS >0F TRANSPORT Under the NCT as described in 10 CFR 71.71, all components of the RVP are within their acceptable temperature ranges. Specifically, the metallic vessel 0-ring seal reaches 125 F, which is well below the 650 F seal design temperature. The worst case bulk foam temperature, which occurs in the impact limiter near the bottom of the vessel is 131'F, which is well below the 160 F temperature that was used to determine the impact limiter crush strength. '

10 CFR 71.43(g) is satisfied by demonstrating that the maximum accessible surface temperature of the package will be 147'F in a 100*F ambient environment without solar radiation.

3.5 HYPOTHETICAL THERMAL ACCIDENT EVALUATION This section describes the models and analyses used in the hypothetical thermal accident of the RVP.

3.5.1 THERMAL MODEL The thermal loading condition of 10 CFR 71.73(c)(4) is used to evaluate the RVP. The package rq

(.) is fully engulfed in a hydrocarbon fuel / air fire with an average flame temperature of at least 1475*F for a period of 30 minutes.

3.5.1.1 Analytical Model The model used is virtually identical to that used for the NCT analysis. Damage resulting from the drop and puncture accidents is considered minimal to the vessel and of no significant effect.

However, the diameter of the impact limiters was decreased for the HAC analysis in order to approximate the damage occurring as a result of the drop.

3.5.2 PACKAGE CONDITION AND ENVIRONMENT Puncture damage does not measurably alter the thermal behavior of the reactor vessel package.

Puncture deformations imposed upon the vessel or radiation shield only effect the package configuration in the immediate vicinity of the mild steel bar strike. This area represents a small '

fraction of the total package surface area. The direct heat input to this small area of the package wall could increase because an existing gap between the shielding and vessel wall could collapse allowing direct conduction between the two surfaces, thus increasing the local wall temperature.

Conduction to cooler adjacent areas in the radial, axial, and longitudinal directions will limit maximum temperatures.

O b 3-9

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Troian Reactor Vessel Package - Saferv Anahsis Rer> ort h

I ws 3.5.3 PACKAGE TEMPERATURES The highest temperature HAC case occurs when pre- and post-fire conditions are 100 F ambient with maximum insolation. Table 3-4 presents the maximum temperatures of various components at the initiation of the fire event, the end of the fire, at steady state after the fire, and the maximum temperature reached during the transient analysis. Note that the package temperatures are slightly cooler for the post fire steady-state than they were at the initiation of the transient.

This is due to the compression of the impact limiters slightly decreasing the heat path from the surface of the radiation shield to the environment. No component exceeds its allowable temperature during the fire event. Additional results are presented in Appendix 3-1.

3.5.4 MAXIMUM INTERNAL PRESSURES When the package undergoes the hypothetical thermal accident, the intemal pressure will increase due to the increased temperature of the air inside the vessel based on the ideal gas law and the increase in vapor pressure, and the pressure from gases produced by radiolysis. This intemal pressure was calculated using the same methodology as Section 3.4.4. This internal pressure condition has been bounded using a value of 100 psi in the analysis of the HAC.

p 3.5.5

() EVALUATION OF PACKAGE PERFORMANCE FOR HYPOTHETICAL ACCIDENT THERMAL CONDITIONS Results of the thermal analysis demonstrate that the RVP will survive the hypothetical accident scenario. The maximum seal temperature of 197 F is less than the 650 F design temperature of the seal. Therefore, the metallic O-ring seals remain functional during and after the hypothetical accident.

3.6

SUMMARY

A thermal analysis has been performed to demonstrate that the RVP meets the maximum accessible surface temperature limit criteria set forth in 10 CFR 71.43(g). The NCT and HAC thermal analyses have demonstrated compliance with the requirements of 10 CFR 71.71(c) and 10 CFR 71.73(c).

Under the NCT as described in 10 CFR 71.71, all components of the RVP are within their acceptable temperature ranges. Specifically, the metallic vessel 0-ring reaches 125 F, which is well below the 650 F seal design temperature. The worst case bulk foam temperature, which occurs in the impact limiter near the bottom of the vessel is 131 F. This temperature is well below the 160 F temperature that was used to determine the impact limiter crush strength.

10 CFR 71.43(g) is satisfied by determining that the maximum accessible surface temperature of A

() 3 - 10

l Trolan Reactor VesselPackaec- Saferv Anahsis Report h

O the package will be 147 F in a 100 F ambient environment without solar radiation.

Under the HAC described in 10 CFR 71.73, all of the components of the package will remain within their acceptable temperature range. The seal will remain well below its 650*F design temperature and the shielding effectiveness will be unchanged, so the package will be able to satisfy the requirements set forth in 10 CFR 71.51(a)(2).

O O 3 11

Trojan Reactor Vessel Package - Safety Analysis Report f

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Table 3-1 I

Internal Component Decay Heat Component Co* Activity ' Component Decay Heat 2 (curies) (Btu /hr)

Upper Core Plate 4,040 169 1 Lower Core Barrel 124,000 5,193 l Baffle Formers 350,000 14,658 Core Baffle 911,000 38,152 l

Lower Core Plate 58,800 2,463 Total 1,447,840 60,635 O

G

1. As of 4/1/96 for listed components only.

2. The total decay heat (60,635 BTU /hr) was calculated using the radionuclides listed in Table 1-2, which are decayed to 11/1/97. For the thermal calculations, it was assumed that all of the heat load (60,635 BTU /hr) was due to Co-60 decay of the components listed above in Table 3-1. The relative contribution from each component was calculated using the proportional contribution of the component to the total Co-60 activity as of 4/1/96. See Appendix 3-1 for additional details.

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Trojan Reactor VesselPackage- Safety Analysis Report h O

Table 3-2 Environmental Conditions for Normal Conditions of Transport Case Ambient Insolation Decay Heat Temperature NCT Maximum 100 F Maximum 60,635 Btu /hr Temperature (17,766 watts)

NCT Accessible 100 F None 60,635 Btu /hr Surface (17,766 watts)

LDCC Filling 70 F None 60,635 Btu /hr Maximum (17,766 watts)

LDCC Filling and 40*F None 60,635 Btu /hr Sealing Minimum (17,766 watts)

Structural Accident -20 F None 60,635 Btu /hr O Analysis, Design Decay Heat (17,766 watts)

Structural Accident -20 F None 45,476 Bru/hr Ana!ysis, Reduced (13,325 watts)

Decay Heat NGT, Low -40 F None -

60,635 Btu /hr Temperature (17,766 watts)

O

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. Troian Reactor VesselPackare- Safety Analysis Report O

Table 3-3 Maximum NCT Temperatures for the Trojan RVP Component 100*F, 100 F 70 F, 40*F, -20

• F, -40

• F, -20

• F, Max Solar, '98 No No No No Low Solar Source Solar Solar Solar Solar Heat LDCC Max 595*F 548"F 557 F 532"F 482"F 465*F 342*F LDCC Bulk 289*F 272*F 244*F 218*F 163"F 144*F 107*F Vessel Seal 125*F- 125 F 98*F 70 F -9*F -29 'F -13 *F Accessible 160*F 156 F 119*F 92*F 35*F 15"F 23*F Surface 147 F*

Impact Foam 131 F 130 F 88*F 59"F 0*F -20 F -5*F O Bulk VessellD 149 F 143 F 104*F 77 F -

1*F 2*F Average Vessel Max 196"F 191 F 156 F 129"F 74*F 55*F 37"F Vessel Min 120*F 120 F 71*F 43"F -16 F -36 *F -18 F

• Maximum without insolation O

m Trojan Reactor Vessel Packaee - Safety Analysis Report O

Table 3-4

, Maximum Temperatures for the Trojan RVP During HAC Fire Case Cornponent Fire Initiation Fire End Post-Fire SS HAC Max I

P-dage 160*F 864 F 168*F 864"F

.. . face l

Seal 125*F 129 F 133 F 197'F

• l l

LDCC Max 595'F 595 F 601'F 601"F LDCC Bulk 289*F 304*F 296"F 310 F Vessel Max 196 F 844 F 203 F 844"F Vessel Min 120 F 127'F 130 F NA h

V

b Trotan Reactor Vessel Package - Safety Analysis Rer> ort h LJ 4.0 CONTAINMENT This chapter describes how the Reactor Vessel Package (RVP) satisfies the requirements stated in Subpart E," Package Approval Standards," of 10 CFR 71. More specifically, this chapter addresses the RVP containment and discusses how the requirements specified in 10 CFR 71.43 and 10 CFR 71.51 are satisfied.

4.1 RVP CONTAINMENT BOUNDARY I The containment boundary of the RVP consists of the reactor vessel shell, upper head, flange O-rings, and the welded penetration closures. The containment boundary of the RVP is shown on Figure 1-4. Section 1.2.1.4 discusses the penetrations, Table 1-1 provides a listing of the penetrations, and Figure 1-4 provides the closure details.

The upper head is attached to the reactor vessel by 54 pre-tensioned 7 inch diameter studs and I scaled using two metallic O-ring seals between the two flanges (reactor vessel flange and upper head flange). The studs are each pre-tensioned to 720,000 lbs.  !

4.1.1 RVP CONTAINMENT VESSEL g

\  ! The vessel is constructed of SA-533 carbon steel and the nozzles are SA-508 carbon steel.

Section 2.1.1.1 provides a discussion of the RVP containment boundary and Tables 2-3 and 2-5 detail the materials of construction. The O-rings are made of silver plated Inconel Alloy 718.

The design specifications for the reactor vessel and the upper head are the same. The reactor vessel was designed for a pressure and temperature of 2,485 psig and 650"F, respectively.

4.1.2 RVP CONTAINMENT PENETRATIONS Table 1-1 and Figure 1-4 provide the location, size, and closure method for each of the reactor vessel penetrations. The materials used are given in Table 1-1. The reactor vessel penetrations will be welded closed prior to transport and there will be no vents, valves, or other penetrations in the RVP. The penetration closures are designed to withstand the Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) as required in 10 CFR 71.71 and 10 CFR 71.73, respectively. Weld closure details are provided on Figure 1-4 and discussed in Section 2.1.1.1.5.

[ \

() 4-1

0 Trojan Reactor VesselPackare - Saferv Anah' sis Retwrt h 4.1.3 SEALS AND WELDS Containment closures will be welded in accordance with the guidance provided in American Society of Mechanical Engineers (ASME) Code,Section III, Subsection ND, and Section VIII.

Weld sizes and types are specified on Figure 1-4. Weld inspections will include nondestructive examination and will be performed per the guidance in ASME Code Section III, Sub-Section ND. Section 8.2.2 provides additional discussion on weld examinations.

4.1.4 CLOSURES I During final preparations for transport, the RVP containment penetrations, which are listed in Table 1-1, will be closed by welding cover plates over each opening. There will be no access to the interior of the package once these penetration closures are welded. The upper head to reactor vessel closure will be achieved by 54 7- inch diameter studs that have been pre-tensioned to 720,000 lbs. As discussed in Sections 2.6 and 2.7, this pre-tensioning of the bolts is sufficient to maintain a positive package seal for both the NCT and HAC.

4.2 CONTAINMENT REOUIREMENTS FOR NORMAL CONDITIONS OF l TRANSPORT I

( l s_) The radiological characterization analysis described in Section 1.2.3 demonstrates that the RVP contains Type B quantity of radioactive material. The RVP is designated as a Type B (as exempted) package, as discussed in Section 1.1, and will be transported as an exclusive use package as defined by 10 CFR 71.4. In addition, the RVP will be a single use package. Upon completion of transport, the RVP will be disposed of at the licensed disposal facility of US Ecology located near Richland, Washington.

Item (f) of 10 CFR 71.43 stipulates that "A package must be designed, constructed, and prepared for shipment so that under the tests specified in {71.71 (Normal Conditions of Transport) there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no rubstantial reduction in the effectiveness of the packaging." Section 2.6 discusses each of the NCT as defined by 10 CFR 71.71 and provides a summary of the results of the analyses performed to demonsuate compliance with these requirements.

4.2.1 CONTAINMENT OF RADIOACTIVE MATERIAL The radioactive contents of the RVP are primarily contained within the internal components of the vessel and thus are not considered releasable. A small amount of surface contamination is present on the inner surfaces of the vessel and internals. These are the areas that were directly in contact with the primary coolant during operations. Two engineered barriers will be used to 4-2

Troian Reactor VesselPackare- Safety Anahsis Report O

effect:yely prevent release of the contaminants to the environment. The first barrier includes the original design pressure boundary of the reactor vessel (vessel shell, nozzles, metallic O-rings) ,

with modifications made to seal the vessel after removal of the cooling system piping and

~ .

' instrumentation (nozzle closure plates, penetration closures). The second barrier is the Low l

' Density Cellular Concrete (LDCC). i 1

As discussed in Section 2.1.1.1, the RVP is a welded steel enclosure. Section 8.1 discusses the acceptance testing that will be performed to ensure the package construction is in accordance

. with the design and appropriate regulatory requirements. Furthermore, the RVP is analyzed for the NCT and HAC during transport, per 10 CFR 71 requirements, as described in Chapters 2 and  ;

3, where it is shown that the RVP has adequate margins of safety against breach of containment. l i

LDCC will be injected into the interior of the reactor vessel. This LDCC will fill the intenor 1 void space of the entire vessel, including the upper head, encapsulating the reactor internals and

. providing additional shielding. The LDCC will fix the contaminants in place on the vessel interior surfaces. The scouring ofinternal surface contamination during the injection process will not increase the releasable source term. The scouring that takes place will result in dispersing the radioactive material into the volume of LDCC. The LDCC will trap the contamination more securely in its volume than just on the surface.

n b In addition to preventing dispersal of contaminants, the LDCC will prevent shifting of the contaminants within the package during transport. Although the external dose rates are primarily 3

i due to activation products within the vessel wall and internals, the surface contamination contributes a small amount to the calculated dose rete. It is conceivable that movement of some of this material within the vessel could cause an increase in the external dose rate. Therefore, the ability of the LDCC to prevent any movement inside the vessel provides additional confidence that the assumptions used in the shielding calculations remain valid throughout the transportation l process.

4.2.2 CONTAINMENT CRITERION The leak tightness of a package is normally demonstrated by leak tests of the package containment boundary following the methods given in American National Standards Institute (ANSI) N14.5. These methods include a determination of the maximurn allowable leak rate for a given package, if the payload of a package contains a limited value of releasable activity or the concentration of the activity is very low, all leak testing requirements are exempt per ANSI N14.5, Section 6.1. A calculation (Reference 4-1), provided as Appendix 4-2, was performed to determine the reference air leakage rate and determine ifleak testing was required. This calculation assumed that a portion of the 155 Curie surface activity can become dislodged in the form of particulates and a fraction of these particulates become airborne. The calculation

! 4-3 j

Trojan Reactor VesselPackage- Safiw Analysis Report n

V determined a permissible leak rate of 21.56 cm'/see for the NCT. This is above the limit of 10 cm'/sec given in ANSI N14.5. Therefore, no leak testing is required.

Based on the above discussion, the containment requirements for 10 CFR 71.43(f) for NCT are met.

4.2.3 PRESSURIZATION OF CONTAINMENT VESSEL Title 10 CFR 71.71(c)(1) requires evaluation of the RVP for the effect of heat with an ambient temperature of 38'C (100*F) in still air, and with solar insolation for a 12-hour period of 2

400 g cal /cm for curved surfaces. The results of this evaluation are provided in Sections 3.4.2, with the maximum corresponding differential pressure calculated in Section 3.4.4 to be 4.3 psi.

It is shown in Section 2.6 that this increased pressure will have a negligible effect on the containment structure of the RVP.

An assessment of potential hydrogen generation in the packaged reactor vessel using the guidance contained in Electric Power Research Institute (EPRI) Publication NP-4938,

" Methodology for Calculating Combustible Gas Concentration in Radwaste Containers," and GEND-052, " Hydrogen Control in the Handling, Shipping, and Storage of Wet Radioactive p Waste," has been completed (Appendix 4-1). The reactor vessel was modeled using contents of V bead resin solidified with concrete. The use of the resin / concrete is conservative since the resin beads are a source of hydrogen that does not exist in the reactor. The LDCC with the foaming egents and water of hydration are clearly included in the model since the solidification agent used is concrete. The calculation assumed an essentially uniform distribution of hydrogen due to the ability of hydrogen to migrate throughout the interior of the vessel. The calculation determined the reactor vessel will reach 5% by volume of hydrogen 142 days following closure of the vessel. The 5% by volume of hydrogen was chosen as the limit for flammability and detonation based on the guidance in NRC Information Notice 84-72, " Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas Generation."

The shipment time period is expected to be less than 45 days including the time period from sealing the package until the actual transportation is completed. In accordance with Information Notice 84-72, the shipment must be completed within 90 days.

4.3 CONTAINMENT REOUIREMENTS FOR HYPOTHETICAL ACCIDENT CONDITIONS The radiological characterization analysis described in Section 1.2.3 demonstrates + hat the RVP contains Type B quantity of radioactive material. The RVP will be designated as a Type B (as O

h 4-4

Trolan Reactor VesselPackage- Safety Analvsh Ret > ort I

fD g

exempted) package, as discussed in Section 1.1, and will be transported as an exclusive use package as defined by 10 CFR 71.4. In addition, the RVP will be a single use package in that upon completion of transport, it will be disposed of at the licensed disposal facility of US Ecology located near Richland, Washington.

The RVP is a Type B (as exempted) package hence, per 10 CFR 71.51 its containment should meet the requirements of 10 CFR 71.51(a)(2) relative to package radiological leak-tightness and the requirements of10 CFR 71.73, Hypothetical Accident Conditions. Section 2.7 discusses each of the HAC as defined by 10 CFR 71.73. Table 2-2 provides a summary of the results of the analyses performed to demonstrate compliance with these requirements.

4.3.1 CONTAINMENT OF RADIOACTIVE MATERIALS The results of the HAC structural and thermal evaluations presented in Sections 2.7 and 3.5, respectively, demonstrate that there is no release of radioactive materials under any of the HAC described in 10 CFR 71.73.

4.3.2 CONTAINMENT CRITERION D A leak rate analysis for the HAC was completed as above for the NCT (Section 4.2.2) case but I

with a higher amount of activity available for dispersal. The HAC calculation (Ref. 4-1),

following the method of ANSI N14.5 for accident conditions, determined an allowable leak rate of 16,744 cm'/sec. This is above the limit of 10 cm 2/sec set in ANSI N14.5, Section 6.1.

Therefore, no leak testing is required.

4.4

SUMMARY

The containment criterion for the RVP is specified in Sections 4.2.2 and 4.3.2 above. It has been shown in this chapter that the dispersal of contaminants is prevented by the welded containment barrier and the LDCC added to the interior of the containment barrier to retain the contaminants in place.

4-5

Trojan Reactor VesselPackare - Safety Analysis Report h O '

4.5 REFERENCES

4-1 Calculation of Allowable Test Leakage Rate for Reactor Vessel Package per ANSI N14.5, Calculation 97-006, Re.. O, Portland General Electric. This calculation is provided as Appendix 4-2.

4-6

Trojan Reactor VesselPackace- Safety Analysis Report O

5.0 SHIELDING EVALUATION This chapter discusses the shielding design features of the Reactor Vessel Package (RVP) and the adequacy of these features in meeting the requirements of 10 CFR 71.47 and the external dose rate requirements of 10 CFR 71.51(a)(2).

The RVP is an exclusive use package transported by water from the Trojan Nuclear Plant (TNP) site to the Port of Benton, Washington. The RVP is then transported by highway to the licensed disposal facility of US Ecology near Richland, Washington. The RVP is not transported in a closed transport vehicle, therefore, the higher dose rate exemptions of 10 CFR 71.47(a) are not applicable. Based on these conditions of transport, the RVP external dose rates delineated by 10 CFR 71.47 must not exceed the following limits:

1. 200 mrem /hr on the external surface of the package,

2. 200 mrem /hr at any point on the vertical planes projected from the outer edges of the vehicle, on the upper surface of the load, and on the lower external surface of the vehicle,

(

\

3. 10 mrem /hr at any point two meters from the vertical planes projected from the outer edges of the vehicle,

4. 2 mrem /hr in any normally occupied spaces.

Prior to shipment, actual dose rate measurements are obtained to demonstrate compliance with 10 CFR 71.47, as required by 10 CFR 71.87(j).

The extemal radiation requirements of 10 CFR 71.51(a)(2) for Hypothetical Accident Conditions (HAC) of transport include a requirement that the dose rate not exceed I rem /hr at any point one meter from an outer surface of the package.

5.1 DISCUSSION AND RESULTS The components of the RVP design which are included in the shielding evaluation include:

1. The carbon steel wall of the reactor vessel and the stainless steel clad on the vessel wall.

5-1

I Trolan Reactor Vessel Packare - Safety Analysis Rer> ort O

d

2. Low density cellular concrete (LDCC) which is added to the interior volume of the reactor vessel. The LDCC is a Portland cement composition and water, without aggregate.

3. Carbon steel shielding around the exterior surface of the reactor vessel. This shielding is 2" thick above the core barrel and surrounding the nozzles,5" thick around the core barrel section, and I" thick exter. ding 3' below the transition to the lower head.

4. Carbon steel closure plates over penetrations in the upper and lower heads and the nozzle openings.

After the RVP has been moved to a low dose rate area, a field survey of the RVP will be performed, and additional %" thick carbon steel shielding plate (s) may also be added to localized areas which exceed the dose rate limits.

During the HAC, the RVP is conservatively assumed to be separated from the cradle (s) on the transporter. Additionally, the 1" thick x 3' steel shielding below the lower head transition and any localized steel shielding (if used) are conservatively ignored in the HAC shielding analyses.

All radiation shielding analyses are performed using the computer code QAD-CGGP. For gamma-ray calculations, the QAD-CGGP computer program uses the point kemel ray-tracing technique. In this method, the point kemel representing the transfer energy by the uncollided flux along a line-of sight path is combined with an appropriate buildup factor to account for the contribution from the scattered photons. With a distributed (i.e., disk or volume) source, the point kemel is integrated over the source area or volume for each source energy considered to arrive at a total dose rate at the designated receptor point.

Conservatisms present in the shielding analyses are:

1. Minimum wall thicknesses of the reactor vessel wall and heads were used.

2. The density of the LDCC was assumed to be the minimum value of 45 lbs/ft 2 The actual density range for the LDCC is between 45 and 65 lbs/ft8.

3. The impact limiters were ignored for all shielding evaluations.

The reactor vessel wall is constructed of carbon steel and has minimum wall thicknesses which range from 5 %" to 8 %". Figures 5-1 and 5-2 illustrate the model of the RVP used in the shielding evaluation.

5-2 l

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l V Trolan Reactor VesselPackare - Saferv Analysis Report 1

i The results of the shielding evaluation performed for the RVP under Normal Conditions of f Transpon (NCT) and HAC are summarized in Table 5-1. Since the dose rates at two meters l

l from the ends of the conveyance are less than 0.1 mrem /hr, and the cab will be located more  !

than 2 meters from the end of the transporter, the radiation level in normally occupied positions of the transport vehicles is less than 2 mrem /hr.

- 5.2 - SQURCE SPECIFICATION 5.2.1 SOURCE REGIONS u

l The source term consists of the irradiated reactor vessel and reactor vessel intemals. The l . activation product radionuclide inventory was taken from the Trojan Nuclear Plant Radiological Site Characterization Repon (Reference 5-1). The activation analysis methodology involved using the ANISN discrete ordinates neutron transport code to estimate the flux levels and energy spectra at the locations ofinterest. The ANISN flux results were then used as input to the ORIGEN2 activation program along with the plant operating history summarized in Table 5-10.

The only gamma emitter which will be present in significant abundance at the time of shipment is Cobalt-60. The effects of any other activated isotopes on the overall calculated dose rate is minimal.

.J Eight source regions were considered for contribution to the measured dose rate. These regions include:

1. Vessel side wall

2. Core bafile

3. Core formers

4. Core barrel

5. Thermalshields

6. - Vessel inner wall cladding

7. Upper core plate

8. Lower core plate L

L -The upper and lower core plates were determined to have a negligible contribution to the total

! dose rate at the locations ofinterest. Therefore, these source terms were eliminated once the l normalization cases were completed and subsequent cases considered only six source regions:

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.1. Vessel side wall i .

1

2. Core baffle

3. Core formers "4.~ Core barrel-

• I L . 5. Thermalshields ,

L 6. Vesselinr uall cladding

'ANISN radial flux results for these six regions are summarized in Table 5-11. I

.< 5.2.2 SOURCE TERM NORMALIZATION Radiation surveys 'of the unshielded reactor vessel were taken in April 1996. The source term l from the Radiological Site Characterization Report (Reference 5-1) was decayed to April 1996 -

!' and normalized to the measured dose rates at the peak azimuthal locations. The radiation

detector readings were taken with the vessel in place and the detector in the space between the reactor vessel and the biological shield wall. As such, the readings included contributions frora y the activated reactor vessel stainless steel insulation and concrete biological shield wall, and V backscatter from'the biological shield wall. A summary of the April 1996 measured dose rates is l presented in Table 5-3. The reactor vessel and internals were mcdeled with water occupying the l~  ; void spaces up to the bottom of the nozzles. The calculated total exposure rate at a distance of 8.5 inches from the vessel wall at the core centerline was adjusted for axial and azimuthal L

peaking to yield a maximum calculated' exposure rate of 5640 mrem /hr. This result was l compared to the maximum measured dose rate of 4400 mrem /hr found at the 21 foot location at L ' 225 degrees. These two value were used to calculate a multiplicative normalization factor, which

, . was base don t eh max mum i measure dexposure rate divided by the calculated maximum. The normalization factor of 0.78 was then applied to the Co-60 source term for all subsequent shielding cases which incorporated LDCC in all reactor vessel voids. The normalized source terms'were then decay-conected to November 1997 to conservatively establish a source term.

The normalization factor is conservative since it uses the absolute maximum measured dose rate, E neglects the dose contribution _s from the activated stainless steel insulation and concrete l , biological shield wall, and neglects the effects of backscatter from the biological shield wall.

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• Trojan Reactor VesselPackare'. Safety Analysis Report )

s 5.2.3 RVP SOURCE DEFINITION

, The RVP shielding analyses use a Cobalt-60 source. Cobalt-60 is a pure gamma emitter producing two significant gamma energies.of 1.173 and 1.333 MeV, both with essentially 100%

. intensity. These values are used in all analyses. .

)

- The source is distributed among six different regions:

I

1. Vessel side wall

' 2. Core baffle

3. Core formers

4. Core barrel

5. Thermalshields 6.? Vesselinner wall cladding

- The curie contents of each region'are provided in Table 5-5, which shows both the LANISN/ORIGEN2 calculated activities and the normalized activities. For the shielding analysis e of the nozzles, the dose rates resulting from neutron activation and surface contamination were quantified separately. The source term for the surface contaminants of the upper internals is

' shown in Table 5-12. i

The QAD-CGGP models consisted of detailed three-dimensional representations for each of the eight source regions ofinterest. The activities in the source regions were assumed to be ,

uniformly distributed except within the reactor vessel wall.' The distribution of activity within  !

f, - the internal components varies based on their position relative to the active core region. The  !

~

- level of detail incorporated into the model was based on the impact each component had on the extemal dose rates. Detailed two-meter dose rates for the shielded reactor vessel are summarized l L

in Table 5-9. These results are provided to illustrate the contribution from each of the source regions to the total dose rate at axial locations relative to the reactor vessel flange. As shown in L this table, the majority of the dose comes from the reactor vessel itself, particularly at the ,

j- elevations above the core region. j l'

! The' distribution of the activity within each of the internal components varies radially, axially, L and azimuthally. The internal component with the largest contribution to the external dose rates is the core barrel. The core barrel accounted for about 9 percent of the two-meter exposure at the e ' maximum axial location; The core barrel activity distribution varies axially with a peaking factor of about 1.5 (maximum to average) and has an azimuthal peaking factor of about 1.25. The radiation variation is the most significant, with the concentrations varying by a factor 2.4 across 5-5 ll l-I I

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Trolan Reactor VesselPackare- Safety Analysis Report O

the core barrel thickness of 2.4 inches. The radial variation was conservatively neglected and i cancels out the effects of the relatively small axial and azimuthal variations. The activity distribution within the thermal shields is similar to that of the core barrel with the exception that the azimuthal variation is minimal. Neglecting the radial variation of a factor of 2.9 across the thermal shield thickness of 2.75 inches cancels out the effects of the relatively small axial and azimuthal variations. The remaining internal components and vessel cladding had relatively small contributions to the external dose rates, and the activity distributions were, therefore, neglected.

The reactor vessel wall contributed the majority of the extemal exposure as shown in Table 5-9.

l The source term within the wall was distributed radially and varied as a function oflocation -

across the vessel wall thickness. ANISN output from the original reactor vessel activation analyses was used to determine the source radial distribution. The mesh intervals from the ANISN results were used as the QAD-CGGP input. The ANISN thermal flux results were input in the source weighting factor array, thus yielding the appropriate source distribution as a function of the radial location. All QAD-CGGP cases using the reactor vessel wall as the source region used this weighting factor array. The ANISN source radial weighting factors are provided in Table 5-4.

The correlation was necessary to obtain realistic results when the reactor vessel was the source region, since the activities at the inner reactor vessel walls are several orders of magnitude higher i than the activity at the reactor vessel outer wall. This reactor vessel source term model was  !

further refined by incorporating the survey results as axial weighting factors and used to estimate j the dose rates at two meters.

I 5.3 MODEL SPECIFICATION The computer model used for the QAD-CGGP cases is illustrated on Figures 5-1 and 5-2.

5.

3.1 DESCRIPTION

OF SHIELDING CONFIGURATION The shielding configuration for the RVP consists of the LDCC installed in the interior void

spaces, the carbon steel reactor vessel, and the carbon steel shielding of various thicknesses installed around the exterior of the vessel. In addition, %" thick carbon steel plates may be required for localized areas identified in a field survey of the package. An exploded view of the L carbon steel shielding plates is illustrated in Figure 1-5.

To ensure that results were conservative, minimum vessel wall thicknesses of 6%" (upper head),

L 8%" (shell), and 5%"(lower head) were used. Additional conservatism was added by j neglecting shielding provided by the thermal shields for the exposure from the core baffle, 5-6 4

l

'; - , _- - ~ _- _ _ _ . , ,_ _ - . _ -__ - _ _-

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O Troian Reactor Vessel Packare - Safety Analysis Report O i formers, and core barrel. Since the neutron shield pads do not provide full circumferential l

coverage, they were modeled as void, water, or LDCC grout, depending on the case. l 1

1 Results for the unshielded vessel are presented in Table 5-2. The unshielded model (no external shielding) results were used to identify locations which would require shielding to satisfy the regulatory contact radiation level limit of 200 mrem /hr. A series of contact shielding cases were run to estimate the contact dose rates as a function of shielding thickness and axial position along the reactor vessel outer wall. A minimum of 3" of steel is required around the active fuel region to satisfy the contact dose rate limit. The actual steel shielding thickness utilized is 5" in this reg on. Between the 5" thick shield and the nozzles, a minimum of 2" of steel shielding is required. Although the 2" minimum thickness could be reduced above the nozzles to 1", the steel shielding thickness was maintained at 2" from the top edge of the 5" thick shield to the transition below the vessel flange.

5.3.2 MATERIALS COMPOSITION i The QAD-CGGP models included elemental partial densities for four material types. These l types include: 1

1. Water (for the benchmark cases)

2. Type 304 stainless steel  ;

3. LDCC j

4. Carbon steel (vessel and exterior steel shielding)

These compositions are summarized in Table 5-6. The composition data for the Type 304 ,

stainless steel was taken from NUREG/CR-3474, and the LDCC composition was taken from ANSI /ANS-6.6.1 for concrete. Both carbon steel regions were assumed to be iron at a density of 7.86 g/ce. These material inputs were used by the QAD-CGGP program to determine attenuation coefficients. Geometric progression buildup factors for carbon steel were utilized to estimate buildup in all models, since the majority of the mean free paths were typically in the reactor vessel steel wall or the exterior steel shields.

5.4 SHIELDING EVALUATION The RVP dose rates for both the NCT and HAC are presented in Tables 5-7 and 5-8,

. respectively. The location of the dose points is shown on Figure 5-3. As demonstrated by these

. results, the RVP satisfies the radiation dose requirements of 10 CFR 71.47 and 10 CFR 71.51(a)(2).

5-7

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

Trolan Reactor VesselPackare- Saferv Analysis Report h O During transport of the RVP, the dose rate for normally occupied positions of the transport -

. vehicle may not exceed 2 mrem /hr. The normally occupied positions during transport are the -

truck cab attached to tne end of the transporter and the pilot house on the tug during river transport. Of these two vehicles, the closest nonnally occupied position to the RVP is the truck cab. The two-meter dose rate calculations are provided in Table 5-7. Since the dose rates at 2 meters from the ends of the conveyance are less than 0.1 mrem /hr, and the cab will be located more than 2 meters from the end of the transporter, the radiation level in normally cccupied positions of the transport vehicle is less than 2 mrem /hr.

5.4.1 CONTACT SHIELDED RESULTS The calculated contact dose rates for the RVP are within the regulatory limit of 200 mrem /hr at all locations of the package. As expected, the maximum contact dose rate,95 mrem /hr, is calculated to occur at the junction of the 2" thick steel shielding and the 5" thick steel shielding.

5.4.2 TWO METER SHIELDED RESULTS' The maximum 2 meter dose rate was calculated to be 5.5 mrem /hr at a point in line with approximately the center of the core barrel. This maximum value is below the regulatory limit of O 10 mrem /hr.

5.4.3 NOZZLE RESULTS One inlet nozzle and one outict nozzle were included in the model to estimate the dose rates at these locations. In addition, the upper core barrel was neglected in the model to minimize shielding from the internal sources at the nozzle locations.

The maximum contact dose rate at the face of the reactor vessel nozzles was calculated to be 5 mrem /hr. This dose rate results from a combination of dose rate due to neutron activation of the base material and the surface contamination of the upper reactor vessel intemals. These results include the shielding contribution of the 2 %" thick nozzle closure plates.

5.4.4 REACTOR VESSEL UPPER HEAD RESULTS The analysis performed include'd the source terms from neutron activation of the reactor vessel and intemals components. While the neutron activation in the upper head is negligible, yielding 1 mrem /hr on contact, crud traps from the control rod drive mechanisms (CRDMs) penetrations may contribute to hot spots. Depending upon the results of the additional surveys, some

~

5-8

Troian Reactor VesselPackare- Sa6% Analysis Rer> ort

)

~J localized shielding may be required in this region.

5.4.5 REACTOR VESSEL LOWER HEAD RESULTS A separate evaluation was performed to estimate the maximum contact reading of the reactor I vessel lower head where itjoins the lower shell. This area is significant since it is relatively close to the core region and has a nominal wall thickness (5%"). The ANISN/ORIGEN2 results for the core support plate were reviewed to approximate the reactor vessel activation in this area.

Because of the activity in this region, a 1" thick skirt will be placed around the RVP just below I the transition to the lower head as shown on Figures 1-5 and 5-1. The calculated dose rates with this shielding configuration are 35 mrem /hr on contact and 15 mrem /hr at 1 meter. Depending upon the results of the final surveys, some additional localized shielding may be required in this region.

5.5 SHIELDING ANALYSIS UNCERTAINTIES The estimated errors associated with the shielding analysis, source term and source normalization come from several sources. Major areas of uncertainty include the ANISN/ORIGEN2 activation methodology, cobalt impurity levels in the base metal, the point kernel shielding methodology (y and the detector response and location. Normalization of the source terms relative to the

( / measured dose rates negates the uncertainty in the activation methodology and in the cobalt impurity levels in the base metal. This limits the uncertainty to the point kemel modeling and radiation detector, which are discussed below.

The potential errors for the measured survey results include instrument calibration and variances in the detector distance to the vessel wall. The survey instrument used was an Eberline RO-7 with a high range probe. The manufacturer's literature indicates that the detector accuracy with the high range probe is +/- 10%. The radiation survey results were taken at a distance of 8.5 inches +/- 1 inch. This 1-inch variance in detector location would cause an additional 2 percent variation in the detector response based on the model geometry. Assuming these errors are additive, the total error associated with the measured survey results is 12 percent.

Detailed point kernel models were developed with QAD-CGGP to model the reactor vessel and internals as accurately as possible including local variations within the source region. However, there is error inherent in the use of the point kernel methodology itself. A study regarding uncertainties in irradiated hardware characterization, NUREG/CR-4968, estimated the uncertainty for point kernel shielding techniques at +/- 25%, which is reasonable for this analysis.

As discussed above in Section 5.2.2, the analysis results were normalized to the measured dose L) 5-9

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1

' Trolan Reactor vessel Pack.we - Safety Analysis Report h U

O rate. The errors associated with the normalization factor used can be directly applied to the shielding analysis results. The normalization factor (XNF) was calculated as follows:

o o j

XNF = Measured dose / Calculated dose ' j The two parameters in this equation each have a random error associated'with them as discussed above. The relative errors for the measured and calculated dose rates are 12% and 25% <

respectively. .

Measured dose = 4,400 +/- 12%

j

. Calculated dose = 5,640 +/- 25% _

The total relative error for the calculated normalization factor is given by the root sum of the i squares as 28%. The shielding was designed assuming that the error in the analysis could be as -

high as +/- 50%, which is why the design basis contact and two-meter dose rates were maintained below 100 mrem /hr and 6 mrem /hr, respectively. In addition, it is noted that the flux to dose conversion factors used for the analysis were in unites of Roentgen per hour as opposed 4 . to the dose equivalent rate in rem per hour. The efore, the reported dose rates are conservative by an additional 11 percent (approximately) for Co-60 energies.

The shielding design includes provisions to account for the above uncertainties and also hot spots  !

that occur. Any hot spots would only impact th' econtact dose rates and have a negligible effect at 2 meters from the conveyance. The maximum contact dose rate given in Table 5-7 is 95 mrem /hr, which is well below the 10 CFR 71.47 limit of 200 mrem /hr. The dose rate would have to more than double due to hot spots and uncertainties before any supplemental shielding would

- be required.

In the event that hot spots do occur, which exceed the conservatively designed shieldmg I configuration, provisions have been made to add multiple layers of supplemental shielding (also '

referred to as " spot shielding"). SAR Figure 1-5, Sheet 2 of 4 notes that additional 0.5-inch

. layers of spot shielding will be added, as required. The addition of 0.5 inches of steel shielding will reduce the exposure rate by approximately one-third for Co-60 gamma energies. This would

~ '

. reduce a hot spot of about 300 mrem /hr to the 200 mrem /hr limit. Spot shielding can also be

' installed in other areas, such as the top of the vessel, where the control rod drive mechanisms

. ' will be removed.

i Allowances have been made in the package design calculations to account for the additional supplemental shielding. The reactor vessel package will be surveyed prior to leaving the site to ensure that the installed shielding is sufficient to meet the external dose rate limit requirements of 10 CFR 71.47 ~

5-10

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_. . _ ~ . . . . . . _ _ _. . _

f Trolan Reactor Vessel Packaee - afety Analysis Report h uj O '

i l5.6' REFERENCES

!- 5 Trojan Nuclear Plant Radiological Site Characterization Report, Rev. 0.1 - Portland General Electric, February 8,1995.

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0 Troian Reactor VesselPackage. Saferv Analysis Retrort ,

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Table 5-1 Summary of RVP Maximum Dose Rates I (mrem /hr) (

l

' Normal Package Surface 2 Meters from Conveyance Conditions of (Transporter)

Transport Side Nozzle End Side Nozzle End I (NCT)

Total Dose 95 5 35 5.5 4.2 <0.1 (Location) (10' below (Lower (16' below flange) head) flange) 10 CFR 71.47 200 200 200 10 10 10 Limit ,

Hypothetical 1 Meter from Package Surface i f-~3 V Accident Conditions Nozzle End I

Side (HAC) )

I Total Dose 14.2 8.4 15 (Location) (8' below flange) (from nozzle end) (from lower head) 10 CFR 71.51 1,000 1,000 1,000 Limit l'

l

0 Troian Reactor Vessel Packare - Safety Analysis Report Table 5-2 Calculated Dose Rates for the Unshielded Reactor Vessel (mrem /hr)

Distance From Contact One Meter Two Meters Vessel Flange (ft)* from Surface from Conveyance 2 40 110 120 4 200 160 150 6 430 250 190 8 770 360 260 10 1,300 580 340 12 2,140 970 440 14 2,970 1,280 520 16 3,610 1,430 580 18 3,930 1,510 600

,q 20 4,030 1,520 590 Q 22 3,760 2,920 1,430 1,150 540 460 24 26 2,000 720 370 28 1,200 440 280 Maximum 4,030 1,520 600

• Note: The nozzle centerline is at 7' from vessel flange and the core region runs between 12' to 24'.

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Trojan Reactor vesset l'ackaze- safety Analysis Rernort .

Table 5-3 April 1996 Reactor Vessel Survey Results Measured @ 8% inches from Vessel Wall (mrem /hr)

Axial Azimuthal Degrees Azimuthal Azimuthal Ratio 45* 90* 135* 180* 225* 270** 315' I 25 22 29 27 29 26 24 26 29 1.115 2 30 26 36 35 37 26 31 32 37 1.172 3 65 65 83 94 85 26 74 70 94 1.337 4 180 145 170 170 180 200 159 172 200 1.163 5 320 300 340 325 360 300 316 323 360 1.115 6 430 340 390 370 410 300 350 370 430 1.162 7 600 480 520 510 540 500 485 519 ' 600 1.155 3 I

8 725 665 770 740 720 700 640 703 770 1.095 9 940 895 990 925 965 900 880 928 990 1.067 10 1,300 1,160 1,300 1,200 1,300 1,100 1,250 1,230 1,300 1.057 Ii 1,600 1,360 1,600 1,400 1,620 1,300 1,540 1,489 1,620 1.088 12 2,000 1,630 2,000 1,700 2,000 1,600 1,960 1,841 2,000 1.086 13 2,500 1,900 2,600 --- 2,600 1,800 2,400 2,300 2,600 1.130 14 2,800 2,300 3,000 --- 3,000 2,100 2,800 2,667 3,000 1.125 15 3,350 2,500 3,500 --- 3,400 2,400 3,200 3,058 3,500 1.144 16 3,600 2,600 3,700 --- 3,700 2,500 3,400 3,250 3,700 1.138 17 3,800 3,200 3,900 --- 3,900 2,800 3,600 3,533 3,900 1.104 I8 3,900 3,500 4,000 --- 4,100 3,000 3,700 3,700 4,100 1.108 19 3,900 3,500 4,100 --- 4,200 3,000 3,800 3,750 4,200 1.120 20 3,900 3,600 4,100 --- 4,300 3,000 3,800 3,783 4,300 1.137 21 3,900 3,500 4,000 --- 4,400 2,600 3,800 3,700 4,400 1.189 22 3,600 3,100 3,800 --- 4,100 2,500 3,500 3,433 4,100 1.194 23 3,400 2,700 3,400 --- 3,800 2,300 3,200 3,133 3,800 1.213 24 2,900 2,800 2,800 --- 3,100 2,100 2,800 2,750 3,100 1.127 25 2.600 2.200 2.600 --- 3,100 1,800 2,700 2.500 3,100 1.240 Note:

• Survey results for the 270 degree location taken from survey no. 40702 dated 12/1 I/95 l

f r~

. Trojan Reactor VesselPackake- Safety Analysis Retrort i O

Table 5-4 ,

ANISN Radial Flux Results for Reactor Vessel Wall Radial Distance Thermal Flux Normalized W**'"*) ^*

(in) (cm) 86.50 219.71 7.36E+10 0.3110

-86.54 219.80 4.39E+10 0.1860 l 86.65 220.10 3.63E+10 0.1530 86.81 220.50 2.77E+10 0.1170 87.01 221.00 192E+10 0.0811 87.28 221.70 1.17E+10 0.0492 87.68 222.70 5.72E+09 0.0242 88.07 223.70 2.90E+09 0.0123 88.46 224.70 1.54E+09 - 0.0065 88.86 225.70 8.53 E+08 0.0036 89.25 226.70 5.02E+08 0.0021 -!

89.65 227.70 3.14E+08 - 0.0013 90.04 228.70 2.10E+08 0.0009 l 90.43 229.70 1.50E+08 0.0006 90.83 230.70 1.114E+08 0.0005 91.22 231.70 9.30E+07 0.0004 .

91.61 232.70 8.09E+07 0.0003 91.97 233.60 7.65 E+07 0.0003 92.36 234.60 8.06E+07 0.0003  !

92.76 235.60 9.67E+07 0.0004 93.15 236.60 1.34 E+08 0.0006 93.54 237.60 2.12E+08 0.0009 93.94 238.60 3.72E+08 0.0016

- 94.33 239.60 7.05E+08 0.0030 94.65 240.40 1.12E+09 0.0047 94.84 240.90 1.59E+09 0.0067 94.96 241.20 2.06E+09 0.0088 95.08 241.50 2.56E+09 0.0108 95.13 241.62 2.84E+09 0.0120 w - - _ - , - - - - ~ , , - - - , e- , - - , e- , e - -w- e.

O Trojan Reactor VesselPackage- Safety Analysis Report .

O Table 5-5 Cobalt-60 RVP Source Strength as of11/97 (Curies)

RVP Region ANISN/ORIGEN2 Normalized to Results Measured Dose Rates Core Baffle 7.25E+05 5.66E+05 Core Formers 2.41E+05 1,88E+05 Core Barrel 8.56E+04 6.68E+04 Vessel Cladding 6.19E+03 4.83E+03 Pressure Vessel 3.93E+02 3.07E+02 Thermal Shields 5.15E+03 4.02E+03 O

O

0 Troian Reactor VesselPackage- Safety Analysis Report l

C~

Table 5-6 Material Inputs Elemental Partial Densities (g/cc)

\

Atomic Element LDCC Type 304 Water Carbon No. SST Steel I Hydrogen 0.0041 --

0.1111 --

8 Oxygen 0.3593 --

0.8888 --

11 Sodium 0.0124 -- -- -

12 Magnesium 0.0017 - -- --

13 Aluminum 0.0331 -- -- --

14 Silico:: 0.2276 -- -- --

19 Potassium 0.0138 -- -- --

20 Calcium 0.0600 -- -- --

24 Chromium --

1.4775 - --

26 Iron 0.0089 5.6692 - 7.860 m/ 28 Nickel --

0.8030 -- --

Totals 0.7208 7.9497 0.9999 7.860 O

l t -

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

)

'i Trolan Reactor VesselPackaee - Safety Analysis Report ,h O l l

l l

Table 5-7 j Shielded RVP NCT Dose Rates (mrem /hr)

Location - Contact Dose 2 Meter Dose Distance From Vessel Flange (ft) i 1 30 --

2 40 3.1 3 7 --

-4 15 3.8 26  !

5 --

6 31 4.0  ;

7 44 --

l 8 56 4.4 O 9 10 72 95 4.8 l l

l 12 7 5.2 14 9 5.4 16 11 5.5 18 12 5.4 l- 20 12 5.1 l ~22 12 4.4 24 9 3.5

! 26 6 2.5

! 28 4 1.5 Nozzle 5 4.2

. Lower Head 35 0.5" l Upper He'ad 1 1.1" l

l

• Dose rate is calculated from edge of 23' 7.5" wide x 105' long conveyance

• • Tabulated dose rates are for 2 meters from the package. At 2 meters from the conveyance the calculated dose rates are <0.1 mrem /hr.

i rO 4

i i

Trojan Reactor VesselPackage- Safety Analysis Report O

Table 5-8 Shielded RVP HAC Dose Rates (mrem /hr)

Location 1 Meter Distance From Dose hte Vessel Flange (ft) 0 1.1 2 3.3 4 7.6 6 13.4 8 14.2 10 9.5 12 7.4 14 7.5 Ci 16 18 7.9 8.5 20 8.6 22 8.1 24 6.3 26 3.5 28 1.5 Nozzle 8.4 Lower Head 15 Upper Head 1 i

O

v

) h%J h V jo ,

Troian Reactor Vessel Package- Safety Analysis Report Table 5-9 Calculated Two-Meter Dose Rates for the Shicided Reactor Vessel & Internals Distance From Total Vessel Baffle Barrel Clad Former Pad Vessel Flange (ft)* mrem /hr rem /hr rem /hr rem /hr rem /hr rem /hr rem /hr 0 2.29 2.29E-03 8.33E-09 7.65E-08 4.64E-08 1.37E-08 2.65E-08 2 3.14 3.14E-03 1.51E-07 6.76E-07 5.53E-07 1.81E-07 3.17E-07 4 3.76 3.75E-03 1.99E-06 6.12E-06 3.04E-06 2.09E-06 1.82E-06 6 4.02 3.95E-03 1.09E-05 2.59E-05 1.27E-05 1.11E-05 8.24E-06 8 4.42 4.19E-03 3.93E-05 8.08E-05 4.07E-05 3.80E-05 3.14E-05 10 4.82 4.26E-03 9.96E-05 1.88E-04 9.63E-05 9.16E-05 8.89E-05 12 5.18 4.17E-03 1.76E-04 3.25E-04 1.69E-04 1.62E-04 1.83E-04 14 5.43 4.02E-03 2.40E-04 4.41 E-04 2.29E-04 2.20E-04 2.82E-04 16 5.50 3.87E-03 2.74E-04 5.04E-04 2.61E-04 2.51E-04 3.46E-04 18 5.37 3.68E-03 2.83E-04 5.03E-04 2.70E-04 2.59E-04 3.70E-04 20 5.05 3.42E-03 2.71E-04 5.00E-04 2.59E-04 2.49E-04 3.58E-04 22 4.44 3.03E-03 2.34E-04 4.31 E-04 2.24E-04 2.15E-04 3.08E-04 24 3.52 2.5 iE-03 1.68E-04 3.10E-04 1.61E-04 1.54E-04 2.18E-04 26 2.46 1.91E-03 9.22E-05 1.73E-04 8.86E-05 8.42E-05 1.17E-04 28 1.53 1.30E-03 3.36E-05 7.20E-05 3.62E-05 3.36E-05 4.60E-05 Percent Contribution 70.26 % 4.98 % 9.16% 4.75 % 4.56% 6.29 %

at 16 feet

• Note: The nozzle centerline is at 7 feet from vessel flange, and the core region runs between 12 to 24 feet.-

1 l

Trojan Reactor VesselPackage- Safety Analysis Rer> ort I

(

Table 5-10 Trojan Operating History Date Date Cycle  ;

Cycle Startup Shutdown Days EFPD 1 1 12/15/75 3/17/78 823 420.5 2 1/1/79 4/11/80 466 290.7 l

3 7/16/80 5/1/81 289 259.6 4 7/16/81 3/27/82 254 227.2 5 8/20/82 1/22/83 155 136.8 1

6 7/14/83 4/27/84 288 255.3 I 7 9/18/84 5/2/85 226 189.5 8 7/4/85 4/16/86 286 261.3 r

t j%

., 9 6/16/86 4/1/87 289 267.9 10 8/25/87 4/13/88 232 181.7

m. ,

11 7/10/88 4/6/89 270 238.7 12 7/26/89 3/19/90 236 194.7 13 7/7/90 3/4/91 240 212.8 14 2/24/92 11/9/92 259 178.2 O

i l

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Trolan Reactor VesselPackare- Safety Analysis Report

\

r"\

U Table 5-11 Trojan Vessel and Internals '

ANISN Radial Flux Results I

Thermal Flux Total Flux 2 2 (n/cm -sec) (n/cm -sec)  :

Baffle 9.31E+12 1.22E+14 Former 1.87E+13 4.43E+13 .

Barrel 5.83E+11 5.42E+12 >

t Thermal Shields 1.26E+11 1.82E+12 Vessel Clad 3.34E+10 9.95E+10 '

Vessel Wall 1.90E+09 3.55E+10 V ,

1 l

i l

O

Trojan Reactor VesselPackare- Safety Analysis Report Table 5-12 Upper Internals Surface Contamination Source Term Radionuclide Activity in curies Am-241 5.18E-02 C-14 5.42E-02 Ce-144 - 2.19E-02 Cm-242 7.35E-06 Cm-243 1.70E-02 Cm-244 1.61E-02 Co-60 4.67E+01 Fe-55 1.30E+01 m

(j H-3 3.82E-02 Mn-54 4.04E-02 Ni-63 9.42E+00 Pu-238 3.93E-02 Pu-239/240 4.39E-02 Pu-241 2.39E+00 Pu-242 2.21E-04 l 1

Sb-125 7.87E-01 Sr-90 4.35E-01 (h

V

F Ji ,l N 16.540 228.s30 UPPER HEAD l

u Il 214.830 ~2" THICK SHIELD G  !!

Q 11 u

ll - II l li 11 7  ;,

I

-r' /

153.289 -

o PRESSUREVESSEL- -- UPPER CORE PLATE L VESSEL CLAD--- + 5" THICK SHIELD 1294.892 7.620 863.219 '

THERMAL SHIELD- ---

373.380 t CORE BARREL-----

CORE BAFFLE -

U LOWER CORE PLATE )

5.060 I

a , 62.484

, 50.000 1* THICK SHIELD 182.655 '

, N N

LOWER HEAD n 14.081 s

, w -

188.820 -

- 171.707 - NOTES:

_ gg7,g _ 1. MODEL DIMENSIONS IN CENTIMETERS

1. " 2. SHIELD DIMENSIONS IN INCHES

3. REACTOR VESSEL FILLED WITH LDCC 106.'M

' 4. CORE BAFFLE MODELED AS A HOMOGENIZED

~ 203.530 RIGHT CIRCULAR CYUNDER

~ 219.710 5. CORE FORMER MODELED WITil SOURCE 220.106 - HOMOGEN12ED IN LOCC REGION 241.696 -

BETWEEN BAFFLE & BARREL  !

FIGURE 5-1 REACTOR VESSEL MODEL

I 0

3 UPPER CORE PLATE l 1

=

hRESSURE VESSEL lH .

u l-VESSEL CLAD l: '

j

/

$l

Ei 3 i THERMAL SHIELD - =

gi

$l !iE !

CORE BARREL = <

-l 0l wI

' ' $I wI 407.669 CORE BAFFLE lj

=

> i 373.380 s i

- - j

'SEE NOTE 2 '

( ,

=r; j

$5 ,

$$ I ll -

, I i

s: .

l ," -

l  ;

l LOWER CORE PLATE i 5.d 80 168.620 l -l

=

171.707 ,

187.960  :

=

194.005 =I 196.545 =l 203.530 -!

219.710 -'

=

220.106 i 241.696  :

NOTES:

1. ALL DIMENSIONS IN CENTIMETERS

b, 2. REACTOR VESSEL FILLED WITH LDCC i

FIGURE 5-2 REACTOR VESSEL MODEL SOURCE REGIONS

O \

Trojan Reactor vesset package. safety Analvsis Rer> ort h

O

6.0 CRITICALITY EVALUATION

The fissile material content of the Reactor Vessel Package (RVP) consists of 3.56 g of Plutonium. The RVP is exempt from fissile material classification, and from the fissile material standards of 10 CFR 71.59, per 10 CFR 71.53(a), smee it contains no more than 15 g of fissile material. Therefore, a criticality evaluation, per 10 CFR 71.55(b), (d)(1), and (e), of the RVP is not required.

O 6-1

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Trojan Reactor VesselPackage- Safety Analysis Report r%

U l 7.0 OPERATING PROCEDURES The reactor vessel will be modified to function as a self-contained shipping package (RVP).

Penetration closures will be welded on all vessel openings and the interior volume will be filled with Low Density Cellular Concrete (LDCC). Steel shielding will be installed as required to comply with the dose limit requirements of 10 CFR 71.47 and 10 CFR 71.51. In addition, the package will conform to the contamination control requirements of 49 CFR 173.443.

Based on the guidance of Regulatory Guide 7.9, this chapter is intended to describe the operating features of the shipping package ano the corresponding operating procedures. Since there are no operational requirements associated with the RVP, this chapter will discuss the method of transport and demonstrate the planning and safety that will be exercised for the transport of the package. In addition, all plant procedures for fabricating and transporting the RVP will adhere to the ALARA requirements of 10 CFR 20.

7.1 GENERAL TRANSPORT SCENARIO The following sections describe the special controls which will be placed on the shipment of the RVP.

(3 V The RVP will be assembled and shipped approximately 300 miles as a one-time shipment from the Trojan Nuclear Plant (TNP) to the US Ecology disposal site located on the Hanford Nuclear Reservation near Richland, Washington. The expected duration of the land transport will be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This includes the portions at the TNP and at Hanford. The duration of the river transit is expected to be less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Thus the total expected time the RVP will be outside Trojan and disposal site boundaries will be less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. To ensure the structural integrity of the RVP, it will be transported only if the initial vessel temperature is greater than 50 F equilibrium, and a minimum average daily temperature (average of forecasted high and low daily temperatures) of 50 F and a minimum daily temperature of 40 F are forecasted for each day along the transportation route for the expected shipping duration.

After preparation as a shipping package as discussed in Chapter 2, the RVP will be loaded onto a transporter at the TNP site and transported as an exclusive use shipment. The transporter is a hydraulically leveled platform designed for transporting large and heavy loads. It is equipped with an air pressure brake system acting on all wheels with spring set air release brakes for parking. The RVP will be secured to the transporter by an engineered tiedown system. This tiedown system is designed to meet the requirements of ANSI N14.2," Proposed American National Standard Tiedowns for Truck Transport of Radioactive Materials,"(Ref. 7-1) and ANSI N14.24-1985,"American National Standard for Highway Route Controlled Quantities of Radioactive Materials - Domestic Barge Transport. "(Ref 7-2) as described in Section 2.5.2.

O d 7-1 1

1 1

Trojan Reactor VesselPackaee. Safety Anahsis Ret, ort

(

Once loaded onto the transporter, the RVP will not be removed until it is off-loaded into the disposal trench at the disposal site (US Ecology).

The loaded transporter will be moved from the packap preparation area in the Trojan Industrial Area to the barge slip on the TNP site. Transporter speed will be limited to 5 mph. It will then be moved onto the barge and secured by an engineered tiedown system. This tiedown system is designed to meet the requirements of ANSI N14.24-1985, except that the transverse collision acceleration loading was increased from 1.5g to 1.6g based on the probabilistic safety study for river transport (Appendix l-1). Figure 7-1 shows the on-site transport route. The barge will then travel up the Columbia River approximately 270 miles to the Port of Benton in Washington where the loading process will be reversed (i.e., the loaded transporter will be moved off of the barge). Figure 7-2 shows the river transport route. The loaded transporter will be transported less than 30 miles by road to the disposal facility operated by US Ecology near Richland, Washington. Figure 7-3 shows the overland transport route from the Port of Benton to the US Ecology disposal facility. The RVP will ien be removed from the transporter for disposal.

The shipment will comply with the specifications of ANSI N14.24-1985, "American National Standard for liighway Route Controlled Quantities of Radioactive Materials - Domestic Barge Transport," and with the applicable requirements of 10 CFR 71 - Packaging and Transportation p of Radioactive Material,33 CFR - Navigation and Navigable Waters,46 CFR - Shipping, and C/ 49 CFR - Transportation.

7.2 PREPARATIONS FOR TRANSPORT The RVP will be prepared as a Type B (as exempted) shipping package that meets 10 CFR 71 requirements prior to transport from the TNP Industrial Area. A discussion of the preparations required to achieve compliance with these requirements is provided in Chapter 2. Chapter 8 describes the inspections and tests that wiu be performed to verify the package has been properly constructed.

Sections 8.5.9 and 8.5.10 discuss the radiation surveys that will be performed to ensure compliance with 10 CFR 71 requirements prior to shipment. Additional surveys required by 49 CFR 173.443 will also be performed. Package markings will meet the requirements as stated in Section 8.4.

The transporter and prime mover will be inspected to ensure the vehicles are working properly and to ensure conformance with applicable state and federal standards. The structural adequacy of the transporter will be demonstrated by analysis and the transporter will be loaded in accordance with the manufacturer's specifications. Prior to transport of the RVP, the entire transportation route, onsite and offsite, will be evaluated to confirm that it is structurally capable

7-2 l

Troian Reactor Vessel Packaee - Safety Analysis Retwt o

b of withstanding the load.

The barge slips at Trojan and at the Port of Benton will be inspected and silt, rock, and debris will be removed, as necessary, to permit safe access by the barge.

The barge will have current classification by the American Bureau of Shipping (ABS) and Certificate ofInspection by the U.S. Coast Guard (USCO), and assigned for sole use. Intact and damage stability calculations will be performed by a naval architect and reviewed and approved by the USCG. Tiedown design and calculations will be reviewed and approved by the National Cargo Bureau (NCB). The barge will be surveyed by a marine surveyor and the NCB to ensure the as-built configuration is in accordance with the design upon which the calculations are based.

The barge will be inspected to ensure integrity of the barge by a marine surveyor prior to ballasting and after deballasting at the Trojan barge slip. The barge loading and unloading procedures will be specified by a naval architect.

The " Reactor Vessel and Internals Removal Project Transportation Safety Plan" (PGE-1077) has been prepared. The Transportation Safety Plan (TSP) identifies the responsibilities and interfaces of PGE, PGE contractors, Federal Agencies, State Agencies, and Local Agencies. The TSP addresses the operating controls and procedures, radiological controls, and contingency p actions. The shipment of the RVP will be conducted in accordance with the TSP. The

\

requirements of the TSP will be implemented by detailed procedures and coordinated with state and local agencies responsible for emergency response along the route. The TSP will be approved by the Oregon Office of Energy (OOE) prior to shipping the Reactor Vessel Package.

The TSP will be made available to the USCG Captain of the Port (COTP) for review prior to shipment. Changes to the plan will be approved by ODOE and reviewed by the USCG.

Appropriate advance notifications, per 10 CFR 71.97, will be made prior to the shipment.

The transportation of the RVP consists of the following three phases:

1. Trojan Site Transit

2. Columbia River Transit from Trojan Site to Port of Benton, WA

3. Port of Benton to US Ecology Site Transit.

l 7-3

f Trojan Reactor VesselPackare- Safety Anahsis Report V(N 7.3 TROJAN SITE TRANRI As stated in Section 7.1, the RVP will be loaded onto a transporter and will be secured by an engineered tiedown system. The loaded transponer is depicted in Figure 1-1. Once the RVP is prepared for shipment it will not be lifted from the transporter until it is off-loaded at the disposal site and will be maintained in its horizontal orientation. Prior to leaving the Industrial Area, the RVP loaded transporter will be inspected for acceptance by US Ecology and the Oregon Department of Transportation may, at its option, perform Commercial Vehicle Safety Alliance (CVSA) inspections on the transporter and prime mover.

After the loading process is completed, the transit will follow the sequence listed below.

1. The loaded transporter will be driven from the Industrial Area to the barge slip on the Trojan site. The road to the barge slip is entirely on Portland General Electric (PGE) property. This road will be evaluated and tested to ensure it is structurally capable of handling the load prior to hauling the RVP.

2. The barge will be ballasted using a controlled process in order to ground it in the Trojan barge slip.

3. The loaded transporter will be driven onto the barge and will be tied down to the barge using an engineered tiedown system.

4. The barge will be refloated using a controlled deballasting process. The ballasting /deballasting and loading / unloading procedures will be specified by a naval architect. The condition of the barge and the stowage of the package on the barge will be inspected by the NCB and the USCG to ensure integrity of the barge and that the final stowage configuration complies with the approved designs.

7.4 COLUMBIA RJVER TRANSIT Prior to departing the Trojan barge slip, the following requirements will be met:

1. A backup tug is present to accompany the primary tug.

2. The primary and backup tug fuel tanks are full.

3. A base station is established to monitor progress of the barge transport.

4. Com.munication will be established between both tugs and the base station before the tugs depart from the Trojan Site.

bp 7-4

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

i Trolan Reactor VesselPackare- Safety Analysis Report

5. The primary and backup tug and the barge is equipped with navigation and i emergency equipment appropriate for river navigation per ANSI N14.24 '

- (Reference 7-2) and approved by the USCG. '

. 6. In addition to the forecasted temperature requirements noted in Section 7.1, there l

. is no adverse weather foreseen between Trojan and the Port of Benton that may  ;

threaten the safety of the barge and package.  !

l 7.

There are no mechanical problems with the tug, backup tug, or the barge that may .

affect the capability to safely transport the package.

8. The primary and backup tug captains are licensed per 46 CFR Subchapter B i

" Merchant Marine Officers and Seamen."

9. The USCG will establish a safety zone per 33 CFR 165, if required, to ensure appropriate safety and security measures are mct.

10. A PGE Radiation Protection (RP) representative and a PGE transportation  ;

coordinator will accompany the shipment. The RP representative will be trained i in the principles of health physics and equipped with appropriate radiation O protection instruments to provide radiological support by performing inspections and/or surveys and maintain personnel exposure ALARA

11. Arrangements have been made with the U.S. Army Corps of Engineers to provide priority passage and exclusive use through the locks en route to the Port of. 4 Benton.

12. The NCB has evaluated the package to transporter and transporter to barge engineered tiedown systems and has certified that the two'tiedown systems comply with the applicable regulations. .

I

13. The USCG has inspected the condition of the barge and the stowage of the package on the barge.

14. ' Notifications of the pending shipment to appropriate authorities have been made.

15. A trip-in-tow report completed by a marine surveyor.

7-5

i Trojan Reactor Vessel Packaee - Safety Analysis Report h V

16. The tugs will meet the applicable requirements of 46 CFR Subchapter C, "Uninspected Vessels." The tugs will be inspected by a marine surveyor prior to l departure.

During the river transport, the following restrictions apply: l

1. The maximum speed will be 10 knots.

2. The backup tug will escort the barge when the package is on board and the barge is not moored.

3. Transit of the barge will be halted to avoid hazardous conditions or to await passage through a lock or to await safe navigation conditions such as upriver from the Port of Pasco where transit must be made during daylight hours. When moored, appropriate measures will be taken to restrict unauthorized access to the barge.

4. During the barge transport phase, the tugs will check in with the base station at least every four hours.

C 7.5 PORT OF BENTON TO US ECOLOGY SITE TRANSIT After arrival at the Port of Benton barge slip, the loading process will be reversed following the sequence listed below.

1. The barge will be grounded, the transporter-to-barge engineered tiedowns will be removed and the transporter will be moved off the barge onto the landing.

2. Prior to departing the Port of Benton Property, the Washington State Patrol may, at its option, perform a CVSA inspection of the loaded transporter and the prime mover.

3. The loaded transporter will travel to the US Ecology site, approximately 30 miles from the barge slip. A portion of the haul route has been used for the transport of decommissioned defueled naval submarine reactor compartments and the entire route has been used for the disposal of the TNP Steam Generators and Pressurizer (Reference 7-3). The following requirements will be met for the transit:

a. The entire route will be evaluated to ensure it is structurally capable of handling the load prior to the RVP shipment.

g i i V 7-6

Trolan Reactor Vessel Packaec - Safety Analysis Retrort h O  !

b. An overload permit will be obtained from local authorities for the overland travel,

c. Escorts will control and/or restrict road traffic in the vicinity of the transporter during the entire overland transport.

d. All railroad traffic in the area will be stopped during transit.

e. The maximum speed will be 5 mph.

4. Prior to entry at the burial site, the loaded transporter will be inspected for acceptance by US Ecology.

5. Once accepted, the transporter will be moved to a prearranged location at the US Ecology site.

6. The prime mover and transporter will be driven into a prepared trench and the RVP will be off-loaded.

O O 2->

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

Trolan Reactor VesselPackare- Safety Analysis Report @

n V

7.6 REFERENCES

7-1. ANSI N14.2," Proposed American National Standard Tiedowns for Truck Transport of l

. Radioactive Materials," March 1993. <

7-2. ANSI N14.24-1985,"American National Standard for Highway Route Controlled Quantities of Radioactive Materials - Domestic Barge Transport."

7-3. PGE Radwaste Shipment Documentation Numbers 95-29 Through 95-33, Dated 9/28/95 To 10/29/95.

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U 7-8

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Troian Reactor Vessel Package- Safety Analysis Report h

v I 8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM t

The Reactor Vessel Package (RVP) will be shipped as a Type B (as exempted), exclusive use ,

shipping package as defined by 10 CFR 71.4. The package will be transported for the purpose of  :

disposal to the licensed low level radioactive waste disposal facility of US Ecology near

. Richland, Washington.

This chapter describes the acceptance tests and/or inspections that will be performed on the package, before it is transported, to ensure compliance with the requirements of Subpart G of '

10 CFR 71.' The RVP is a single use, steel enclosure and does not contain any instrumentation or i

systems which are relied upon to maintain or monitor the package integrity, both after packaging and during shipment of the package. Therefor:, no maintenance programs or maintenance procedures are required for the safe shipment of the RVP.

8.1 ACCEPTANCE TESTS ,

The requirements for acceptance testing of the packages are given in 10 CFR 71.85, " Preliminary ,

determinations," and 10 CFR 71.87, " Routine determinations." The requirements and the procedures for compliance are discussed in subsequent sections of this chapter. The acceptance tests will be conducted in accordance with the NRC approved 10 CFR 50 Appendix B, PGE Quality Assurance Program (PGE-8010). PGE-8010 includes PGE's NRC approved 10 CFR 71, Subpart H, Quality Assurance Program.

10 CFR 71.85 contains preliminary determinations that must be met prior to the first use of any _

package for the shipment oflicensed material. Although the RVP is a single use package, these requirements are still applicable to the package. The requirements delineated in 10 CFR 71.85 are discussed in Sections 8.2,8.3, and 8.4. The routine determinations required within 10 CFR 71.87 are addressed in Section 8.5.

8.2 PRELIMINARY DETERMINATIONS 8.2.1 VISUAL INSPECTION The requirements for visual inspections specified in 10 CFR 71.85(a) are:

"The licensee shall ascenain that there are no cracks, pinholes, uncontrolled voids, or other defects, that could significantly reduce the effectiveness of the packaging."

The reactor vessel has been used to generate nuclear power for nearly 17 years under high 8-1

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.l Troian Reactor Vessel Package - Saferv Analysis Report O

' temperatuie (= 585'F), pressure (= 2250 psia), and radiation (>l Mev neutrons) conditions. l Defects such as cracks, pinholes, uncontrolled voids or other defcets would have been identified and corrective action taken prior to first use and/or through the routine nondestructive examinations, pressure tests, and leak tests conducted during the service life of the reactor vessel.

Therefore, credit is taken for the integrity of the reactor vessel shell (including existing welds).

The inspection of the RVP, for integrity, will consist of:

1. A visual inspection of the reactor vessel to determine if any damage was

- done to the vessel during its removal. If damage is identified, the damage will be evaluated as to its significance, and remedial action (s) will be reviewed and documented before implementing repair (s). Inspections and repairs will be conducted in accordance with written procedures and appropriately documented.

i 1

2. Visual inspections, prior to shipment, of the new welds and new material (penetration closures and shielding) of the RVP to ensure that there are no

^

cracks, pinholes, uncontrolled voids, or other defects which would significantly reduce the effectiveness of the package. If the inspections reveal any defect (s), the defect (s) will be evaluated to ascertain whether remedial action (s) may be warranted. Inspections and repair (s), ifrequired, j will be appropriately documented.

8.2.2 STRUCTURAL INSPECTIONS i

Structumi inspections will include nondestructive examination of new containment and structural welds on the RVP. These nondestructive examinations will be performed in accordance with guidance contained in NUREG/CR-3019, " Recommended Welding Criteria for use in the

)

Fabrication of Shipping Containers for Radioactive Materials." This guidance specifies the use  !

ofliquid penetrant examination (PT) or magnetic particle examination (MT). The RVP j containment closure welds will be subjected to both visual and liquid penetrant (PT) or magnetic particle examination (MT) which will be performed in accordance with the methods and  ;

guidance contained in ASME Code,Section III, Sub-Section ND. The shielding welds will be

{

inspected in accordance with the methods and guidance of ASME Code,Section VIII. l 8.2.3 LEAK TEST The radioactive contaminants of the RVP are fixed in place by the Low Density Cellular

~

Concrete (LDCC), a nondispersible medium, as described in Chapter 2. The nondispersible j medium removes much of the potential for escape ofcontaminated particles. Regulatory Guide 8-2 e , - . - - - -

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__ _ Troian Reactor VesselPackare- Safety Analysis Report i j

17.4 and ANSI Standard N14.5-1987 provide a method for calculating a maximum allowable leak rate and provide limits to determine'whether or not leak testing is required. A calculation (Reference 4-1) was performed to determine the reference air leakage rate and to determine if leak testing is required. The resulting allowable leak rate for the RVP is above the ANSI N14.5 3

' limit of 10 cm /sec; therefore, no leak test is required. Additional information is included in Sections 4.2.2 and 4.3.2.

8.2.4 COMPONENTTESTS l

8.2.4.1 Reactor Vessel Flanoe O-Rinos The RVP is a steel enclosure of passive design. All vessel penetrations have welded closures.

Valves, rupture discs, gaskets, or similar components are not relied upon to provide package integrity with the exception of the flange O-rings. The metallic O-rings are procured from an approved supplier and visually inspected prior to use. They are installed following established procedures to ensure proper sealing.

8.2.4.2 Imnact Limiter Foam I

The impact limiter foam is tested for conformance with published specifications. Since the accident survivability depends largely on the predicted response of the impact limiters, the actual

- foam must reflect the properties used in the structural and thermal analyses. This is  !

accomplished by testing the foam to be installed within the RVP impact limiters to ensure conformance with the required foam material properties. To ensure that the samples tested are representative of the installed foam, installation will be according to the following requirements: 1 l

1. The foam shall be closed-cell polyurethane material with a density of approximately 20 lb/ft 2

• 10% for the overall density and

• 15% for an individual

- foam test sample. Representative samples of the production pour shall be tested in accordance with General Plastics Procedure TM-9704.

2. Representative samples of the production pour shall be formed using test boxes or containers with material from the actual production pour stream, Test specimens shall be taken from each sample box prepared during production. Test samples shall be tested prior to installation of the next pour. All sample test blocks shall be preserved and be traceable to the installed casing, and shipped to the Trojanjob site with the completed impact limiters.

3. Test specimens from each pour's test block shall be tested at room temperature (72.5

• 2.5 F), in accordance with the requirements of Gene 1 Plastics Procedure 8-3

Trojan Reactor VesselPackaee- Safety Analysis Retrort s

TM-9704, for compressive strength within the parameters presented in Table 8-1.

Stress-strain plots shall be prepared for the perpendicular-to-rise orientation. If the first test specimen falls within

• 10% of the static test correlation values, then no additional static tests are required for that test block. If the first test specime'n falls between

• 10% and 15% of the static test correlation values, then two additional specimens shall be tested. The average of the three specimens shall be within 10% of the requirements or the block and its associated pour shall be rejected.

The test data shall be recorded and reviewed to ensure compliance with the foam structural acceptance criteria. The average of the sample properties from each pour shall be within

• 10% of the specified stress at foam strains of 10%,40% and 60%.

The average of all pours into a single impact limiter casing segment shall be within

• 10%. At the time of foam installation, test samples shall be retained from each pour, as discussed above.

4. Test specimens from each pour shall be tested for flame resistance in accordance with the requirements of 14 CFR 25, Section 853 and Appendix F, Part I(a)(1)(i).

5.

Each production pour, including its properties, shall be completely recorded. Test data and sample pour material shall be traceable back to the steel casing into which 3

(d it is installed. At a minimum, the record shall include pour dates, operator name, impact limiter part number and serial number, Quality Assurance (QA) acceptance, and material tracking identified by batch. A " Pour Plan" will be submitted after contract award. Each pour shall be made and identified in accordance with the

" Pour Plan."

6.

Upon completion of production, a certificate referencing the production record data and testing data pertaining to each impact limiter shall be issued by the contractor.

Test data relevant to the pouring operation shall be included with the certificate.

QA data submittals shall be dated and signed by the foam contractor's designated QA representative.

7. The foam used for filling voids need not be tested, provided that not more than 30 pounds of foam is used for the void filling in any division of any casing segment (i.e.,120 lbs maximum for a 134 segment or 60 lbs for a maximum 70 segment).

8.

Similarly, material that is used for the foam dams inside the casings and all miscellaneous foam shim material supplied with the impact limiters shall be tested as stated above for conformance with the required foam characteristics.

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Troian Reactor VesselPackare- Safety Anah' sis Report 8.2.4.3 Low Density Cellular Concrete I

The critical design characteristic of the Low Density Cellular Concrete (LDCC) is density. The density of LDCC for the RVP has been specified to be between 45 to 65 lbs/ft'. The process tequirement is that the void space inside the reactor be essentially filled with the LDCC. The reactor vessel internals will be initially cooled to ensure the internal conditions support the LDCC injection'and curing process. To ensure that the LDCC density is within the range specified and that the reactor vessel is filled with LDCC, the injection process and the density testing will be performed in accordance with the following requirements:

1

1. The LDCC injection process shall be controlled via'a procedure that has been approved under PGE's Quality Assurance Program. The approved procedure will provide the controls, verification, and documentation to assure that the critical characteristic (density,45-65 lb/ft)) meets the required values and that the process )

requirement (the vessel is filled with LDCC) has been met. j j

2. The basic LDCC design will be a mixture of water, Portland cement, and entrained j air, that uses a proprietary foaming agent and admixtures. The water-to-cement ratio by weight will be approximately 0.6 and will be adjusted as necessary to meet l density requirements.

3. Prior to the first lift,' a chiller system will cool the internal components to ensure proper injection and curing of the LDCC. Air temperature will be less than 170 F prior to LDCC injection.

4. The LDCC will be injected into the vessel in two " lifts" (pours). Each " lift" will l consist of one or more batches. The maximum lift height is 30 feet.

5. The density of each batch shall be tested prior to injection into the vessel. The density shall be tested in accordance with the appropriate American Petroleum Institute (API) Recommended Practice for field testing of drilling fluids and shall be determined to be in the 46-65 lb/ft2 range.

6. The overall density will be calculated as a " batch-weighted average."

7. There shall be at least four days between consecutive lifts to ensure there is adequate time for the previous lift to cure. This ensures that the density of the previous pour is not altered by the weight of the LDCC above.

8. Verification that the vessel has been filled with LDCC shall be based initially on 8-5

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1 Trolan Reactor VesselPackare- Soferv Analysis Report O

the witnessing / documentation that there has been i solid, continuous vent (overflow) of LDCC from a vent point located at the highest part of the vessel. As a final verification that the vessel has been filled with LDCC, the temporary LDCC closure covers on the vessel nozzles, CRDM penetrations, and vent penetrations will be removed after LDCC injection, and the vessel inspected for voids. Any void space shall be filled with LDCC prior to installation of the permanent transportation closures.

9. After the vessel has been filled with LDCC (i.e., after the second lift) the vessel .

will be vented for at least 28 days to permit escape ofexcess water in water vapor form. After this 28-day venting period, the bottom drain of the reactor vessel will be opened to verify that there is no free water in the vessel.

1 8.2.5 TESTS FOR SHIELDING INTEGRITY As discussed in Chapters 2 and 5 the primary shielding for the RVP is provided by the reactor vessel and the LDCC filling the vessel interior. The LDCC will be pumped into the RV in a controlled manner to preclude voids which would compromise its shielding capabilities. An j external layer of steel plate will be installed around much of the reactor vessel to provide the 1 added shielding needed to meet the external radiation dose requirements of 10 CFR 71.47 and i 10 CFR 71.51. The shielding design as shown in Figure 1-5 will ensure that these requirements are satisfied during all operating and accident conditions. The integrity of the reactor vessel will be confirmed by inspection as discussed in Sections 8.2.1 and 8.2.2. Prior to shipment, the i external radiation levels will be measured and, if necessary, additional shielding will be added where needed.

8.2.6 THERMAL ACCEPTANCE TESTS The reactor vessel has been in use as part of an operating reactor and thus is capable of withstanding temperatures produced by the small of amount decay heat present during the required environmental conditions. The LDCC and steel added to the RV are capable of withstanding all temperatures within the design envelope of the RVP.

The thermal analyses performed used conservative thermal properties for the materials present in the RVP. Therefore, thermal acceptance tests are not required.

8-6

0 Trojan Reactor VesselPackaer- Safety Analysis Report J  !

8.3 PRESSURE TESTS The requirements of 10 CFR 71.85(b) are that a pressure test at least 50% higher than normal operati .6 pressure be perfonned,if

".' . .the maximum normal operating pressure will exceed 35 ka (5 lbf/in2 )

gauge..."

.In Section 3.4.4, the maximum normal operating pressure for the duration of transport, including the effects of vapor pressure and the pressure from gases produced by. radiolysis, is calculated to

. be 4.3 psig. Therefore, a pressure test is not required.

8.4 PACKAGE MARKING 1

- 10 CFR 71.85(c) requires that:

i "The licensee shall conspicuously and durably mark the packaging with its model number, serial number, gross weight, and a package identification number assigned by the NRC. Before applying the model number, the licensee shall . 3 determine that the packaging has been fabricated in accordance with the design d approved by the Commission." <

Similar requirements are contained in 49 CFR 172.310 and 49 CFR 173.471(b).

Prior to applying the model number, and in accordance with an approved Quality Assurance Program (PGE-8010), it will be verified and documented that the RVP has been prepared in

~

accordance with the approval issued for the package by the NRC.

The RVP will be marked in accordance with the requirements of 10 CFR 71.85(c),

49 CFR 172.310 and 49 CFR 173.471(b).

8.5 ROUTINE DETERMINATION Prior to shipment of the RVP, routine determinations will be made of the shipment to ensure that the package and its contents satisfy the requirements of 10 CFR 71.87. Sections 8.5.1 through 8.5.11 address each of the determinations enumerated in 10 CFR 71.87.

t N.) 8-7

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l Troian Reactor Vessel Package - Saferv Analysis Rer> ort q

LJ 8.5.1 PROPER PACKAGE 10 CFR 71.87(a) requires the licensee to determine that: .

l "The package is proper for the contents to be shipped." l As explained in Section 8.1, the RVP is a one-time-only package that will be designed and constructed for a Type B quantity of radioactive material as an exclusive use shipment. The package will be designed and constructed in accordance with the requirements contained within this safety analysis report and the NRC issued approval for the package. Testing / inspections to ensure compliance with these requirements are specified in Sections 8.2 and 8.3.

8.5.2 UNIMPAIRED PACKAGE 10 CFR 71.87(b) requires the licensee to determine that:

"The package is in unimpaired physical condition except for superficial defects such as marks or dents."

[]

'o As described in Sections 8.2.1 and 8.2.2 the package (including the new welds) will be inspected visually and the new welds nondestructively examined before transport. These activities will ensure compliance with this requirement.

8.5.3 CLOSURE DEVICES 10 CFR 71.87(c) requires the licensee to determine that:

"Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects."

The upper head is secured to the reactor vessel lower shell with 54 pre-tensioned studs. The flange O-rings provide the seal between the flanges of the shell and head as described in Section 2.1.1. Prior to final installation of the head on the reactor vessel, the studs,0-rings, and sealing surfaces will be visually inspected for defects and repaired, as necessary. The RVP penetrations are sealed by welded closures. As described in Sections 8.2.1 and 8.2.2 the package (including the new welds) will be inspected visually and the new welds nondestructively examined before transport. These activities will ensure compliance with this requirement.

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l Trolan Reactor Vessel Packaste - Safety Analysis Report 0 i 8.5.4 LIQUID CONTAINMENT 10 CFR 71.87(d) requires the licensee to determine that:  !

"Any system for containing liquid is adequately sealed and has adequate space or other specified provision for expansion of the liquid." j l

The RVP will contain no liquid after it has been drained of water and filled with LDCC.

j Therefore, this requirement does not apply. 1 8.5.5 PRESSURE RELIEF DEVICES 10 CFR 71.87(e) requires the licensee to determine that:

"Any pressure relief device is operable and set in accordance with wTitten procedures."

This requirement is not applicable to the RVP since it does not contain any pressure relief device.

O 8.5.6 . LOADING AND CLOSURE U

10 CFR 71.87(f) requires the licensee to determine that:

"The package has been loaded and closed in accordance with written procedures."

The package will not be " loaded" (i.e., filled) with any radioactive material that is not already part of the RVP.' LDCC will be injected and allowed to cure prior to shipment and in accordance with written procedures. The remaining openings will then be closed in accordance with written procedures, design drawings, and specifications, in preparation for shipment.

8.5.7 NEUTRON ABSORBER 10 CFR 71.87(g) requires the licensee to determine that:

"For fissile material, any moderator or neutron absorber, if required, is present and in proper condition."

Neither moderator nor neutron absorber material is required for packaging since the package contains a minimal amount of fissile material as described in Chapter 6. I i

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0' 1 Trolan Reactor Vessel Package - Safety Anahsis Retrort h

/3 V

8.5.8 LIFTING OR TIEDOWN ATTACHMENTS 10 CFR 71.87(h) requires the licensee to determine that:

"Any stmetural part of the package which could be used to lift or tiedown the package during transport is rendered inoperable for that purpose unless it satisfies the design requirements of {71.45."

Any attachments to the package that could be used to lift or tiedown the package during transport will be rendered inoperable prior to shipment. The vessel nozzles and studs are not considered structures which could normally be considered lift or tiedown points but could provide attachment locations. This is an exclusive use shipment and use of these locations will be prevented by administrative control for all lifting and tiedown operations.

8.5.9 SURFACE CONTAMINATION 10 CFR 71.87(i) specifies the maxin um removable contamination permitted on the extemal surfaces of the package at any time during transport.

pb external surfaces. Prior to transport, surface contamination measure ensure that the surfaces of the packages are below the limits specified in 10 CFR 71.87(i).

Section 4.2.1 discusses the design features which prevent the escape of contaminants from the interior of the RVP. Construction of the RVP will be verified in accordance with the acceptance testing described in Sections 8.2.1 and 8.2.2. These design features are adequate to preclude increases in removable surface contamination levels.

8.5.10 EXTERNAL RADIATION LEVELS 10 CFR 71.87(j) requires the licensee to determine that:

" External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in 71.47 at any time during transportation."

10 CFR 71.47 limits are:

1. 200 mrem /hr on the external surface of the package 2.

200 mrem /hr at point on the vertical planes projected from the outer edges bv 8 - 10

0 l Trojan Reacter Vessel Package- Safety Analysis Report V

of the vehicle, on the upper surface of the load, and on the lower external surface of the vehicle

3. 10 mrem /hr at any point two meters from the vertical planes projected from the outer edges of the conveyance

4. 2 rarem/hr in any normally occupied positions of the vehicle As presented in Chapter 5, based on the Normal Conditions of Transport (NCT), the external dose rates delineated by 10 CFR 71.47 are not exceeded.

Prior to shipment, radiation surveys of the package will be performed to ensure that the external dose rates satisfy the requirements of 10 CFR 71.47. Additional local shielding will be added as necessary.

Although some significant portions of the contaminant are assumed to be tightly adhered to the interior surfaces of the RVP, the design of the package provides an additional measure to fix their location. Prior to removing the RVP from containment, the interior void space will be filled with LDCC. In addition to providing shielding and reducing external dose rates, the LDCC will

/

C] prevent the movement of contaminants thereby preventing radiation level changes during transport.

8.5.11 PACKAGE SURFACE TEMPERATURE 10 CFR 71.87(k) requires the licensee to determine that:

" Accessible package surface temperatures will not exceed the limits specified in s71.43(g) at any time during transportation."

Section 3.4.2 shows that the maximum accessible package surface temperature does not exceed the requirements of 10 CFR 7I.43(g).

8.6

SUMMARY

The information provided in this chapter demonstrates that the Trojan Reactor Vessel Package j complies with the packaging criteria of 10 CFR 71.85, " Preliminary determinations," and 10 CFR 71.87, " Routine determinations."

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a I.' Trojan Reactor Vessel Packare - Safety Analysis Report O

Table 8-1 Foam Impact Limiter i Required Static Crush Test Values

• Strain Nominal (psi) minus 10% (psi) plus 10'% (psi) 10 % 1142 ;028 1256 L i

! 40% 1457 -

1311 1603 60 % 2778 2500 3056 1:

• Values to be taken at 72.5 F, perpendicular to foam rise.

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