ML20217A425
| ML20217A425 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 09/11/1990 |
| From: | HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20217A406 | List: |
| References | |
| HI-89407, HI-89407-R03, HI-89407-R3, NUDOCS 9011210056 | |
| Download: ML20217A425 (258) | |
Text
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HOLTEC INTERNATIONAL . REVIEW AND CERTIFICATION LOG DOCUMENT NAME: Licensina ReDort for SDent Fuel Pool Modification of Pool A: Three Mile Island Unit I HOLTEC DOCUMENT I.D. NO. HI-89407 HOLTEC PROJECT NO. 90310 CUSTOMER / CLIENT: PC076295 - l l REVISION BLOCK l l ISSUE NO. AUTHOR REVIEWER Q.A. MANAGER APPROVED
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& DATE & DATE & DATE BY & DATE l
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~l REVISION 5 REVISION 6 j
-l NOTE:' ' Signatures and.. printed names are - required in ,
s the review block. , N Must be Project Manager or the President. ] This document conforms to the requirement of 'the design i specification and the applicable sections of the governing codes.- t
- . ;This document bears'the ink stamp of the professional--engineer who -l
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hQ Professional Engineer ~l SEAL
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t i a i SU*' MARY OF REVISIONS, ! 1 Revision 2 Only Section 5 has been revised in Revision 2 of this report.
.. lU Revision 3 ] <c >
Revision 3 adds Section 11, Environmental Cost / Benefit Assessment, In addition, the following pages have been revised in Revision 3: Table:of Contents .1 1-2 ! 2-1
~
2-8' , 8- l 3-13' 3-14 ; D. 4-3 1
- t. V- .4-10 5 '5-8 5-10 '
5-12 l: 4 5-19. l l 5-21 to 5-26 5-29 to 5-39-(Figures 5.5.3 through 5.5.13) 6-1 ' y .7-2 l~ 7-3 r 7-4 l i r. h
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l TABLE OF CONTENTS
1.0 INTRODUCTION
1-1 2.0 MODULE DATA 2-1 2.1 Synopsis of New Modules 2.2 Multi-Region Storage 2-1 2-2 2.3 Material Considerations 2-4 2.3.1 . Introduction 2-4 2.3.'2 Structural Materials 2-4 2.3.3 Poison Material 2-4 2.3.4 Compatibility with Coolant 2.4- 2-7' Existing Operation Rack Modules and Proposed Reracking 2-7 3.0 RACK FABRICATION AND APPLICABLE CODES 31 3.1 Fabrication-Objective 3-1 3.2 Rack Module for Region I-. 3-2 3.3 Rack Module for Region.II ' 3-4 3.4 Codes, Standards and Practices 3-6 for the TMI-I Spent-Fuel Pool Racks
-3.5 Materials of Construction 3-8
-4.0 CRITICALITY SAFETY' ANALYSES 4-1 4.1 Design Bases 4-1 4.2 Summary of_ Criticality Analyses 4-5 4.2.1 Normal Operating conditions 4-5 4.2.2 4.3' Reference Fuel Abnormal-andStorage Cells Accident Conditions 4-6 4-8 4.3.1 Reference Fuel' Assembly 4-8 4.3.2' Region 1 Fuel Storage Cells 4-8 4.3.3 Region 2 Fuel Storage Cells 4-9 4.4' Analytical Methodology 4-9 4.4.1 Reference Design Calculations 4-9 4.4.2 Fue1~Burnup. Calculations and 4-11 Uncertainties:
4.4.3 Effect of Axial Burnup 4-12 Distribution
'4.4.4 Long-term-Changes in Reactivity 4-13 i
L D: TABLE OF CONTENTS (continued)
.4. 5 Region.I Criticality Analyses and Tolerances 4-14 4.5.1 Nominal Design 4-14 4.5.2 Uncertainties Due to Tolerances 4-14 4.5.2.1 Boron Leading Tolerances 4-14 4.5.2.2 .Boral Width Tolerance 4-15 4.5.2.3 Tolerances in Cell Lattice 4-15 Spacing 4.5.2.4 Stainless Steel Thickness 4-16 Tolerances 4.5.2.5 Fuel Enrichment and 4-16 Density Tolerances 4.5.3 Eccentric. Fuel. Positioning 4-17
- 4. 5. 4 ' ' Reactivity Effects of Boral 4-17 Length 4.6 Region 2 Criticality Analyses
- 4 4'.6.1 Nominal Design Case- 4-19 4.6.2 Uncertainties Due to Tolerances 4-20 4 4.6.2.1 Boron Loading Tolerances 4-20
-4.6.2.2- Boral Width Tolerance- ~4-20 4.6.2.3' Tolerance in Cell Lattice '4-21 Spacing 4.6.2.4 Stainless-Steel 4-21 Thickness' Tolerance
'4.6.2.5 Fuel, Enrichment 4-21 and Density Tolerances 4.6.3 Eccentric Fuel Positioning 4-21 4.6.4 Reactivity'Fffect of Boral Length' 4-22 4.7 Abnormal and Accident-Conditions 4-23.
4 . 7 .1 - Temperature and Water Dansity 4 Effects 4.7.2 Dropped Fuel ~ Assembly 4-23 4.7.3 Lateral-Rack Movement 4-24 4.7.4 Abnormal Locationiof a Fuel Assembly 4-24
-4.7.4.1 Fresh Fuel Assembly 4-22 in Region 2-4.7.4.2 Fuel' Assembly External 4-23 to the' Storage Racks 4.8 References for Section 4 4-26 Appendix A - Benchmark Calculations A-1 11
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$ 1 TABLE OF CONTENTS '
(continued) 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5-1 5.1- Introduction 5-1 5.2 Spent Fuel Cooling. System Description 5-2 5.2.1 System Functions 5-2 ,, 5.2.2 System Description 5-3 5.2.3 Performance Requirements 5-4 5.2.4 . Methods of Operation 5-5 5.2.5 . Leakage Considerations 5-5 5.3 Decay Heat Load Calculations 5-6 ,a 5.4 Discharge-Scenarios 7
, 5.5 Bulk Pool Temperatures 5-8 5.6 Local Pool Water Temperature. 5-13 5 '. 6 .1 Basis 5 5.6.2 Model Description 5-14 '
5.7 Cladding-Temperature 5-16 5.8 Blocked Cell Analysis 5-18. 5.9 References for Section 5 .5-18 Ab/
'6 . 0 RACK STRUCTURAL CONSIDERATIONS 6-1 l
6.1 . Analysis Outline-(for New 6-1
=
-Proposed Rack Modules)
'6.2 Fue1LRack - DynamicqModel' 6-5 6.2.1 Outline of Model for 6-6 .
Computer. Code'DYNARACK '! 6.2.2 Model Description 6-9 6.2.3 Fluid Coupling; .6-9, p 6.2.4 Damping 6-11 1 i 6 ; 2. 5 - Impact 6-11 1 6.3. Assembly of the. Dynamic Model. 6-11 6.4 Time Integration of.the Equations 6-15 of" Motion ~ 6.4.1 Time History Analysis Using 6-15 ' Multi-Degree of Fr3edom Rack Model 6.4.2= Evaluation of Potential 6-17 for Inter-Rack Impact
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@]W fio Wl' TABLE OF CONTENTS 4Ohi, (continued) p '
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Ah 9Ci6'.5 Structural Acceptance Criteria 6-18 J Material Properties 6-20 30}!?'6.6 b 6.7 Stress Limits for Various 6-20 Uf ? Conditions ? < 6.7.1 Normal and Upset Conditions 6-20 L^ (Level A or Level B) [ 6.7.2 Level D Service Limits 6-23 l' 6.8 Results for the Analysis of Spent 6-23 Fuel Racks Under 3-D Seismic Motion j 6.9 Impact Analyses 6-25 l 6.9.1 Impact Loading Between 6-25 Fuel Assembly and Cell Wall 6.9.2 Impacts between Adjacent 6-26 Racks 6.10 Weld Stresses 6-26 6.10.1 Baseplate to Rack Welds 6-26 and Cell-to-Cell Welds C.10.2 Heating of an Isolated Cell 6-27 6.11 Spent Fuel Pool Slab Model 6-28 6.12 Spent Fuel Fool Slab Analysis and Results 6-28 6.13 Definition of Terms Used in Section 6.0 6-28 6.14 References for Section 6 6-29 7.0 ACCIDENT ANALYSIS AND MISCELLANEOUS 7-1 STRUCTURAL EVALUATIONS 7.1 Introduction 7-1 7.2 Results of Accident Evaluation 7-1 7.2.1 Fuel Pool 7-1 7.2.2 Fuel Storage Building 7-2 7.2.3 Refueling Accidents 7-2 7.2.3.1 Dropped Fuel Assembly 7-2 7.2.3.2 Dropped Object 7-3 7.3 Local Buckling of Fuel Cell Walls 7-4 7.4 Analysis of Welded Joints in Rack 7-5 7.5 References for Section 7 7-6 iv
(; . v.- TABLE OF CONTENTS (continued) 6.5 Structural Acceptance Criteria 6-18 6.6 Material Properties 6-20 6.7 Stress Limits for Various 6-20 Conditions 6.7.1 Normal and Upset Conditions 6-20 (Level.A.or Level B) . ' 6.7.2 Level D Service Limits 6-23 6.8 Results for the Analysis of Spent 6-23 Fuel Racks Under 3-D Seismic Motion-6.9 Impact Analyses 6-25 , 6.9.1 Impcct-Loading Between 6-25
-Fuel. Assembly and Cell l Wall 6.9.2 Impacts between Adjacent 6-26
- Racks'-
6.10: Weld Stresses 6-26 f-g 6.10.1 Baseplate to Rack Welds' 6-26 ij and Cell-to-Cell Welds 6 .' 10 ',' 2 . Heat'ing of'an Isolated Cell' 6 6.'111SpentiFuel Pool Slab Model 6-28 6~.12: Spent. Fuel' Pool Slab Analysis-and Results 6 6.13 Definition <of Terms-Used in'Section 6.0 6-28' , 6.14iReferences'for.Section 6, 6-29
~
i 7.0 ACCIDENT. ANALYSIS'AND MISCELLANEOUS 7 -STRUCTURAL EVALUATIONS 7.1. IntroductionL 7-l' . ! '7.2 .Results of Accident Evaluation 7-1 7.2.1 . Fuel Pool 7-1
. 7. 2. 2 ' Fuel Storage Building 7-2 l1 '7.2.3 Refueling Accidents 7-2 7.2.3.1 Dropped _ Fuel Assembly '7-2 7.2.3.2-. Dropped' Object 7 ~
'7.3 'LocalTBuckling of Fuel Cell Walls 7-4 7.4 Analysis of Welded Joints in Rack 7-5
- 7. 5' References-for Section 7 7-6 1
iv
f '. l ( , 1 TABLE OF CONTENTS (continued) ! 8.0 ANALYSIS OF FUEL POOL STRUCTURE 8 H 8.1 Introduction 8-1 - l 8.2 Model 8-2 8.3 Loading Conditions 8-4 8.4 Results of Analyses 8-6 8.5 Pool Liner 8-7 8.6 -Conclusions 8-7 8.7- References for Section 8 8-9 L i 9.0 RADIOLOGICAL EVALUATION 9-1 ) 9 .1 - Fuel Handling Accident 9-1 J 9.1.1- Assumptions and Source Term 9-1 Calculations ' 9.1.2 Results 9-4 9.2- Cask Drop Accident 9-4 9.3 Solid Radwaste 9-5 , 9.4 Gaseous Releases 9-5 !. 9.5 Personnel Exposures 9-6 ' (} 9.6 Anticipated 1 Exposure During Reracking 9-7 10.0-IN-SERVICE SURVEILLANCE PROGRAM 10-1 10.~ 1" Purpos e 10-1 l'.2! O Coupon Surveillance 10-4
.10.2.1 Description-of Test Coupons 4 10.2.2~ Benchmark Data : 10-5 ,
10.2.3- . Coupon Referehce Data 10-5 . 10.2.4- Long Term Surveillance 10-6 i 10 . 2.5 Accelerated Surveillance 10-6
-10.2.6 Inspection Tests to.be Performed 10-7 10.2.7 Methods 10-8 10.3 Acceptance-Criteria 10-9 10.4 Direct Poison Panel Testing 10-10 10.4.1 Benchmark Data 10-10 10.4.2 -Blackness Testing on Irradiated 10-10
, Panels l 10.5 Acceptance Criteria for Blackness Testing 10-11 10.6 References for'Section 10 10-11 11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 11-1 L O v - 1
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LIST OF TABLES TABLE NUMBER- TITLE PAGE 1.1 Discharge Schedule 1-5 , 1.2 Available Storage Capacity 1-6 l 2.1- ' Module Data (Region 1) 2-10 2.2 Module Data (Region 2) 2-10 2.3 Module Data for. Future Region 2-11 2 Racks 2.4 Common Module Data 2-12 2.5 Module Data 2-13 2.6 Boral Experience List 2-14 (Domestic and Foreign) 2.7 1100 Alloy Aluminum Physical 2-16 Properties ,
-2. 8 Chemical Composition - Aluminum 2-17 (1100 Alloy) -
2.9 Boron Carbide-Cremical Composition - 2-18 : L -Weight, % i Boron Carbide Physical Properties
-([ 4.1 4.2 Summary of. Criticality Safety Analyses
-Reactivity' Effects of Abnormal and 4-28 Accident Conditions 4-29 4.3 Design-Basis Fuel Assembly Specifications 4 -i 14 . 4 Allowance for Uncertainties in l Reactivity Due to Depletion Calculations 4-31 !
4.5 Long-Term Changes--in Reactivity _in Storage Rack Calculated by CASMO-2E 4-32 4.6 FuelLBurnup_ Values for. Required 4-33 Reactivities' (km) with Fuel of Various
. Initial Enrichments' 4.7 Effect of Temperature and Void on 4 Calculated Reactivity ofl Storage Rack 5 . 4 '.1 Fuel' Specific Power and Pool Capacity. Data 5-19 5.4.2 Data for Scenarios 1:through 5 5-20 5.4.3 Data for Scenarios 1 through 5 5-21 l5.5.1 Bulk Pool Temperature Results When 5-22 Considering Heat Losses to the Ambient Assuming 90 Tubes Plugged in Each Cooler b
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//: . 1 l LIST OF TABLES i (continued) a 5.5.2 Time-to-Boil Results When Considering 5-23 Heat Losses to the Environment-5.6.1 Peaking Factor Data 5-24 5.6.2 Data for Local Temperature 5-25 5.7.1 Loca11and Cladding Temperature 5-26 i
' output Data for the Maximum Pool Water.
Condition (Case 1) ! i
, 6.1 ' Degrees of Freedom 6-31 6.2- Numbering System for Gap Elements and. 6-32 and Friction Elements o
-6.3 Typical Input Data for Rack Analyses 6-34 .
(lb-inch units) ' 6-35 6.4' Rack Material Data (200*)/ Support Material Data (200*F) () 6.5 Stress Factors and Rack to Fuel Impact Load . 6-36 l 6.6 Rack Displacements and-Support Loads 6-40 'i 8.1 J Safety.Factorsifor-Bending of Pool 8-10 Structure' Regions 8.2 Pool Slab Shear Safety Factors 8-11 4 9 .1n Radionuclide Inventories and Constants. 9 912 Data and Assumptions for~the Evaluation. 9-10 l of the Fuel 1 Handling Accident . 9.3; . Typical. Concentrations of. Radionuclides 9-11 l in the' Spent Fuel > Pool Water- it 9.4 ' Preliminary Estimate of Person-Rem. 9-12 Exposure during Refueling O vil i
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E l J I LIST OF FIGURES. ) FIGURE NUMBER TITLE PAGE 1.1, Module Layout - TMI Unit I (Poal A) 1-7 l 2.1 .Planview of TMI-I Pool System 2-19 2.2 Module Layout - Pool A 2-20 2.3 Module Layout - Pool A 2-21 (Present Reracking Campaign) 3.1 Seam Welding-Precision Formed Channels 3-9 3.2 Lead-In for Region 1' Modules 3-10= 3.3 Composite Box Assembly 11 3.4 Assembling of Region 1 Boxes- 3-12 3.5~ Adjustable Support Leg 3-13 .
- 3.6 Elevation View of a Region _1 Rack 3-14 '
showing Two Storage Cells r' 13.7 Elevation View of Region 2 Rack Module 3-15 (.L 3.8 Array of Region 2 Cells (Non-Flux: LTrap Construction) 3-16 ;
- 4 '.1 Relationship between Initial Enrichment 4-35' and Acceptable Fuel Burnup
- 4. 2. ~ Cross-section of Region 1 Storage Cell 4-36 4.3. Cross-Section of Region 2' Storage Cell 4-37 ,
._~' .i 5.5.1 Pool Bulk Temperature Model for 27 Normal' Discharge Scenario 5.5.2- Cooler Temperature-Effectiveness:p vs. 5-28 Mumber of1 Tubes Plugged 5.5.3 Bulk PooltTemperature for Case 1 5 5.5.4 Bulk-Pool' Temperature-for; Case 2 5-30 5.5.5 Bulk Pool Temperature for Case'3 5-31 5.5.6 Bulk Pool Temperature for Case 4 5-32 5.5.7 Bulk Pool Temperature for Case.5 33 5'.5.8 Maximum Bulk PoolLTemperature of 5 Case I vs. Number of Tubes Plugged L in;the Cooler (One Cooler in Operation)-
5.5.9 Pool Water Inventory During the Loss 5-35 of Cooling Event for Case 1 ; l e 1 l viii .i
b L 'f')-V u LIST OF FIGURES , L (continued) h 5.5.10 Pool Water Inventory During the Loss 5-36 of Cooling Event for Case 2 5.5.11 Pool Water Inventory During the Loss 5-37 of-Cooling Event for Case 3 5.5.12 Pool Water Inventory During the Loss 5-38 of Cooling Event for Case 4 ) L 5.5.13- Pool Water Inventory During the Loss 5-39 ; L of Cooling Event for Case 5 5.6.1 Idealization of Rack Assembly 5-40 5.6.2 Thermal Chimney Flow Model 5-41 5.6.3 Convection Currents in the Pool 5-42 ; 6.0A .TMI-1 Seismic Response 6-43 (Elev. 302-Ft - 6 In.) 6.0B TMI-1 Seismic Response 6-44
-(Elev.s 329 Ft - 0 In.)
.6 ,0C -- Final Spectra 6-45 6.1 TMI H1-SSE Acceleration. Time History 6-46 f rN - 6.2- TMI H2 SSE' Acceleration Time-History 6-47 4 -
- 6. 3. 'TMI VT SSE Acceleration Time History- 6-48
! 6. 4. -TMI.SSE'H1 Spectrum Comparison 6-49 6.5 TMI SSE H2_ Spectrum Comparison ~6 6.6 TMI;SSE VT Spectrum Comparison _ 6-51 6.7 U}U: H1 OBE Acceleration-Time' History 6-52 6.8- TMI H2 OBE Acceleration' Time History 6-53 6.9 TMI VT1OBE' Acceleration Time History 6 .
6.10 TMI OBE H1 Spectrum > Comparison 6-55 6.11- TMI;OBE H2 Spectrum Comparison 6-56
;6.12 .TMI OBE VT Spectrum Comparison 6-57 ~l 6.13 Schematic.Model1for'DYNARACK 6-58 s 6.14 Rack-to-Rack Impact Springs. .
6-59 i 6.15 Impact Spring Arrangement at Node i 6-60 6.16 Degrees of Freedom Modelling Rack Motion- 6-61 6.17- Rack Degrees of Freedom for X-Z Plane 6-62 Bending 6.18 -Rack Degrees of Freedom-for Y-Z Plane 6-63 j Bending
-6.19 J2-D View of Rack Model 6-64 7.1 Loading on Rack Wall '7-7 7.2 Welded Joint in Rack 7-7 8.1 Overall Finite Element Model of 8-12 TMI Spent Fuel Pool ix l
.. .- __ O
1.0 INTRODUCTION
9 Three Mile Island Unit I (TMI-I) is a pressurized water nuclear power reactor owned and operated by GPU Nuclear of Parsippany, New Jersey. TMI-I received its Construction Permit from the AEC in May, 1968, and its Operating License in April, 1974. The plant went into commercial operation in June, 1974. The TMI Unit : fuel storage system is made up of two pools, labelled Pool A and Pool B, which are capable of being separated by a Pool gate in their common wall. Pool A presently contains 253 spent fuel storage locations and Pool B has 496 locations. Thus the total storage capacity of the TMI-I spent fuel at the present time is 749 l locations, of which 440 are presently occupied by previously discharged fuel. Since the TMI-I reactor core has 177 fuel assemblies, maintaining full core offload capability at all times j implies that 572 locations (749 minus 177) are available for normal offload storage. Table 1.1 provides the data on previous l and projected fuel assembly dischargo in the TMI-I Spent Fuel g Pool. Table 1.2, constructed from Table 1.1 data, indicates that TMI-I will lose full core discharge capability after the discharge of the 10th batch into the TMI-I Pool during the scheduled outage in 1993. This projected loss of full core discharge capability prompted the present undertaking to increase spent fuel storage capability in Pool A. The purpose of this licensing submittal is to rerack Pool A and equip it with new poisoned high density storage racks containing 1494 storage locations, of which 846 cells will be installed during fuel cycle 9. The reracking also entails removal of the j existing gate between Pool A and Pool B, and of the load test fixture. The planned '92 reracking campaign will not fill all g 1-1
]
I available pool floor space in Pool A. The remaining 648 storage cells will be installed at a future date. Of the 846 high density storage locations, 195 will be of the so-called Region I type, and the remainder of Region II genre. Region I racks are configured to store new or burned fuel of maximum 4.6% enrichment. Region II racks are- capable of storing 4.6% l l enrichment fuel with 37 MWD /kg-U (min.) burnup and correspondingly ' reduced burnup for . lower enrichment fuel. Included in the 846 , locations are three storage locations for storing failed fuel containers. Figure 1.1 shows the proposed nodule layout for Pool A. c It is noted .that. the proposed reracking, eff art will increase the number of licensed storage ^ locations 111 Pcsl A.to 1494, which,'as
-indicated in Table 1.2, will extend the date of loss of full core discharge capability to the year 2023, which is well past the- !
presently licensed end-of-life date of 2014. Pool B is not contemplated to be raracked at the present' time. The new. spent fuel storage racks are free-standing 'and self supporting. The principal construction materials- for the -new. racksLare ASTM.240-Type-304 stainless steel sheet and' plate stock, ! and M64 (precipitation hardened stainless steel) for the adjust-lable support spindles. The only non-stainless material utilized in. the rack is the neutron absorber material which is boron-carbide and aluminum-composite sandwich available under the-patented. product name "Boral". i O 1-2
F - O The new racks are designed and analyzed in accordance with Section III, Division 1, Subsection NF of the ASME Boiler and Pressure vessel Code. The material procurement and fabrication of the rack I modules conforms to 10CFR 50 Appendix B requirements, ) This Licensing Report documents the design and analyses performed to demonstrate that the new spent fuel racks satisfy all governing requirements of the applicable codes and standards, in particular, "OT Position for Review and Acceptance of Spent Fuel-Storage and Handling Applications", USNRC (1978) and 1979 Addendum thereto.
-The safety assessment of the proposed rack modules involved demonstration of its hydrothermal, criticality and structural adequacy. Hydrothermal adequacy requires that fuel cladding will not fail due to excessive ' thermal stress, and that the steady state Pool bulk temperature will remain within the limits ,fq prescribed for.the spent fuel Pool. Demonstration of structural t-V -adequacy primarily involves analysis showing that the free-standing modules will not impact under the postulated SSE and 1/2 SSE events,.and that the primary stresses in-the module structure will remain below the ASME Code allowables. The structural l qualification also includes analytical demonstration that the l suberiticality of the rtored fuel will be maintained under accident scenarios such as fuel assembly drop, accidental misplacement of fuel outside a rack, etc.
The criticality safety analysis shows that the neutron multiplication factor'for the stored fuel array is bounded by the USNRC limit of 0.95 under assumptions of 95% probability and 95% confidence. Consequences of the inadvertent placement of a fuel n V 1-3
~'
c - assembly are also evaluated as part of the criticality analysis. q Vf The criticality analysis also secs the requirements on the length of the B-10 screen and the areal B-10 density. This Licensing Report contains documentation of the analyses performed to demonstrate the large margins of safety Nith respect to all USNRC sp scified criteria. This report also contains the results.of the analysis performed to demonstrate the integrity of the fuel Pool reinforced concrete structure, and an appraisal of radiological considerations. The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety from all three - thermal-hydraulic, criticality, and structural - vantage points. The No Significant Hazard Consideration evaluation submitted to the Commission along with this Licensing Report -is based on.the j descriptions and analyses synopsized in the subsequent' sections of l
,y this report.
V i f l l I' O 1-4
b i w i s
/ .i 'q) ' ;
5 Table 1.1 DISCHARGE SCHEDULE Ca.clative Number of Inventory . , Assemblies in the TMI-1 Year Discharced Pools 1976 56 56 1977 48- 104 1978 52 156 1979 52 208 1 1986 76 284 1988 76 360 1990 80 440 4 1991* 80 520 1993 80 ~600 1995- 80 680
.I)
'~
1997 80 760- ' 1999' 80 840
'2001 80 920 '
2003 80 1000 2005 80 1080 J2007- 80 1160 > 2009 80 1240 1 2011 80 1320 2013 80 1400 2015' 80 1.d 2017 80 4520 2019 80 1640 E l
* ~
Two year cpcle planned after 1991 outage. l Oi 1-5 l
- - , . .a;
rA Table 1.2 AVAILABLE STORAGE CAPACITY NUMBER OF AVAILABLE LOCATIONS After Proposed Under Increase in Existing Storage Fuel Capacity Capacity , C;' le No . Year (749 cells) 11990 cells) 9 1991 229 1470 10 1993 149* 1390 11 1995 69** 1310 i' 12 1997 --- 1230 13 1999 --- 1150 14 2001 --- 1070 15 2003 --- 990 ! 16 2005 --- 910 17 2007 --- 830 i 18 2009 --- 750*** (' 19 2011 --- 670
-20 2013 ---
590 21 2015 --- 510 22 2017 --- 430 23 2019 --- 350 24 2021 --- 270 25 2023 --- 190 l Loss of Full Core Discharge Capacity Loss of Normal Discharge Capacity Loss of Full' Core Discharge Capacity if 648 cells, planned for future installation, are not installed. () 1-6
I 14'-11 5/8" N
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Figure 1.1 Module Layout - TMI Unit 1 (Pool A) l 1-7
i i f') V 2.0 MODULE DATA 2.1 Synoosis of New Modules i TMI Unit I spent fuel storage system consists of two pools, labelled as Pool A and Pool B respectively. The two pools intercommunicate through an opening in their common wall. This opening is normally open, although the plant has maintained the ' ability to isolate the two pools from each other by manually installing a gasketed " gate". After over a decade of operation of the two pools, GPUN has determined that there has never been a need to isolate the two pools from each other except when raracking Pool B. Therefore, it has been decided to remove the ' gate designed for isolating the two pools from the pool area, and cut-off the wall support brackets. This action removes a potential heavy object drop scenario from the rack safety considerations and avails rooms for the desired rack configuration. The plan view of the two pools is shown in Figure 2.1. It is noted that the fuel storage area of Pool A is narrower in width at-
- the North and compared to its South end. At the present time, Pool-A contains low density unpoisoned racks at 21" nominal pitch.
There is a total of 253 storage l'ocations in Pool A. l Figure 2.2 shows the module layout for Pool A after the proposed
- 1992 raracking campaign. As shown in Figure 2.2 and tabulated in Tables 2.1 and 2.2, there are two racks containing a total of 195 .
l cells in Region I and 4 racks containing 648 cells in Region II. Out of 648 Region II cells provided, one will be used as an '
- " overload test" fixture. Not included in the above cell count are three defective fuel container storage locations. Thus, to 1
l O 2-1 y - .,,r- , _ _, e- . - - - - - - 7 n- e- r v-+- --- '-"
nummarize, there is a total of 843 fuel storage locations b,, (including one for overload fixture), plus three defective fuel container storage locations. l Figure 2.3 also shows the layout schematic of racks which will not be fabricated or installed at this time. As shown in Table 2.3, the so-called " future" racks consist of six modules with a cumulative total of 648 storage locations. These " future" racks are also licensed at this time; however, they will not be fabricated or installed in the 1992 rerackia.; campaign. The essential cell data for all storage cells is given in Table 2.4. The physical size and weight data on the modules may be found in Table 2.5. To summarize, the present reracking application will increase the licensed storage capacity of Pool A from 253 to U94 locations, of which 846 locations will be installed in 1992, and the remainder at a future date. In addition to Pool A, Pool B ! currently contains 496 medium density racks. Thus, the cumulative licensing storage capacity of the TMI-1 spent fuel pool U installation will stand at 1990 locations after approval of this licensing application. ! l 2.2 Multi-Reaion storace The high density spent fuel storage racks in Pool A will provide storage locations for up to 1494 fuel assemblies and will be i designed to maintain the stored fuel, having an initial enrichment of up to 4.6 wt% U-235, in a safe, coolable, and suberitical ' configuration during normal and abnormal conditions.
- i l
All rack modules fer TMI-1 spent fuel pool are of the " free- 1 l standing" type such that the modules are not attached to the pool ; floor nor do they require any lateral braces or restraints. These ; 1 ! ) 2-2 i i
- m
r,) (
's rack modules will be placed in the pool in their designated locations, and the support legs remotely leveled (using a telescopic removable handling tool) by an operator on the fuel handling bridge. No additional lifting equipment is needed to carry the weight of a rack while leveling is being performed.
The racks will be arranged in two regions in the spent fuel pool. Region I will have 195 locations capable of storing unirradiated fuel of up to 4.6 wt% U-235 initial enrichment. Region I has enough locations to store a full core discharge (177 assemblies). Region II will have 1299 locations for storage of fuel which meets enrichment and burnup criteria developed as part of the rack design. Section 4 of this report addresses this in more detail. Included in the count for Region II are three locations for storage of failed fuel canisters. The total number of storage ! locations, as mentioned above, is 1494 for Pool'A. ; O V Each module is supported by four legs which are remotely I adjustable. Thus, the racks can be made vertical and the top of the racks can easily be made co-planar with each other. The rack module support legs are engineered to accommodate variations of the pool floor. The placement of the racks in the spent f'ael pool J has been designed to preclude any support legs from being located l over existing obstructions on the pool floor, i [v 2-3
i 1 2.3 Material considerations O 2.3.1 rntreductien Safe storage of nuclear fuel in the TMI Unit I spent fuel pools i requires that the materials utilized .a the fabrication of racks be of proven durability and be compatible with the pool water environment. This section provides the necessary information on this subject. 2.3.2 Structural Materials j The following structural materials are utilized in the fabrication of the spent fuel racks: l
- a. ASME A240-304 for all sheet metal stock.
- b. Internally thruded support legs: ASME A240-304.
- c. Externally threaded support spindle: ASME A564-630 g precipitation hardened stainless steel.
.g
- d. Weld material -
per the following ASME specification: SFA 5.9 ER308. 2.3.3 Poison Material In addition to the structural and non-structural stainless material, the racks employ BoralTM, a patented product of AAR Brooks and Perkins, as the neutron absorber material. A brief description of Boral, and its fuel pool experience list follows.
Boral is a thermal neutron poison material composed of boron
- O cerbide and m0 e11ov a1mmin . Beton carbide is . -mp-nd having a high boron content in a physically stable and chemical
( inert form. The 1100 alloy aluminum is a light-weight metal with high tensile strength which is protected from corrosion by a _ highly resistant oxide film. The two materials, boron carbide and aluminum, are chemically compatible and ideally suited for long-- term use in the radiation, thermal and chemical environment of a [ l nuclear reactor or the spent fuel pool. Boral's use in the spent fuel pools as the neutron absorbing material can be attributed to the following reasons: (1) The content and placement of boron carbide provides a very high removal cross section for thermal neutrons. (ii) Boron carbide, in the f orm of fine particles, is homogenously dispersed throughout the central layer of the Boral panels. (iii) The boron carbide and aluminum materials in Boral are totally unaffected by long-term exposure to gamma radiation. (iv) The neutron absorbing central layer of Boral is clad with permanently bonded surfaces of aluminum. (v) Boral is stable, strong, durable, and corrosion resistant.
=
Boral is manuf actured by AAR Brooks & Perkins under the control and surveillance of a computer-aided Quality Assurance / Quality Control Program that conforms to the requirements of 10CFR50 > Appendix B, " Quality Assurance Criteria for Nuclear Power Plants". h 2-5
As indicated in Table 2.6, Boral has been licensed by the USNRC U for use in numerous BWR and PWR spt.nt fuel storage racks and has been extensively used in overseas nucAear installations. Boral Material Characteristics Aluminum: Aluminum is a silvery-white, ductile metallic element that is the most abundant in the earth's crust. The 1100 alloy aluminum is used extensively in heat exchangers, pressure and storage tanks, chemical equipment, reflectors and sheet metal work. 'I l 1 l It has high resistance to corrosion in industrial and marine atmospheres. Aluminum has atomic number of 13, atomic weight of I l' 26.98, specific gravity of 2.69 and valence of 3. The physical and mechanical properties of the 1100 alloy aluminum are listed in Table 2.7. I The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the . loss of metal from general corrosion or pitting corrosion and the film remains stable between a pH range of 4.5 to 8.5. I Boron Carbide: The boron carbide contained in Boral is a fine granulated powder that conforms to ASTM C-750-80 nuclear grade i Type III. The particles range-in size between 60 and 200 mesh and the material conforms to the chemical composition and properties l listed in Table 2.9. !
- (J' 2-6
t 4 2.3.4 comnatibility with coolant All materials used in the construction of the TMI-I racks have an established history of in-pool usage. Their physical, chemical and radiological compatibility with the pool environment is an , established fact at this time. As noted in Table 2.6, Boral has been used in both vented and unvented configurations in fuel pools with equal success. Austenitic stainless (304) is perhaps the , most widely used stainless alloy in nuclear power plants. 2.4 EXISTING RACK MODULES AND PROPOSED RERACKIRG OPERATION Pool A currently has low density rack modules containing a total of 253 storage cells. In addition, Pool B has 496 storage , locations. At the time of the proposed raracking operation, 520 of these locations will be occupied with spent fuel.* There is sufficient number of open (unoccupied) cells in the pool to permit. l relocation of a majority of assemblies to Pool B. ' A remotely engagable lift rig, meeting NUREG-0612 stress criteria, will be used to lift the empty modules. The fuel storage building crane will be used for this purpose. 7. module change-out scheme has been developed which ensures that all modules being handled L are empty, and at least four to six feet laterally-from a loaded r module, when the module is more than twelve inches above the pool l floor. In addition, a certain number of storage locations will be filled with miscellaneous elements such as Retainers. It is also. to be noted that certain ?aripheral cells in Pool B are not accessible by the fuel handl:.ng equipment, making them ineffective L for fuel storage. L 2-7
,0 The fuel storage building crane is an overhead unit which rides on rails that traverse the entire Fuel Storage Building. This crane has a 110 ton main hoist, and an auxiliary 15 ton hoist. ,
the defense-in-depth approach of NUREG-0612, the Pursuant to following additional measures of safety will be undertaken for the reracking operation. (i) The crane and hoist will be given a preventive maintenance checkup and inspection within 3 months of the beginning of the reracking operation. (ii) The crane will be used to lift no more than 15% of its rated ::apacity of 110 tons at any time during ,
. the reracking operation. (The maximum weight of any module and its associated handling tool is 14 tons).
(iii) The old fuel racks will be lifted no more than 6 inches above the pool floor and held in that r elevation for approximately 10 minutes before beginning the vertical lift. (iv) The rate of vertical lif t will not exceed 6 feet per minute. (v) The rate of horizontal movement will not exceed 5 feet per minute. L (vi) Safe load paths have been developed. The "old" or [ "new" racks will not be carried over any region of the pool containing fuel. (vil) The rack upending or laying down will be carried I out in an area which is not overlapping to any safety related component. (viii) All crew members involved in the reracking operation will be given training in the use of the lifting and upending equipment. The training seminar will utiize videotapes of the actual lifting and handling rigs on the actual modules to be installed in Pool'A. Every crew member will be
] '
2-8 l
required to pass a written examination in the use ! O or 115ttas =a h =attae er r tu- a t i t r a by the rack designer. i The fuel racks will be brought directly into the Fuel Storage Building through the access door, which is at ground level, I without any maneuvering. This direct access to the building greatly facilitates the rack removal and installation effort. l l The "old" racks will be "hydrolased" while underwater in the pool, ' and approved for shipping per the requirements of 10 CFR71 and 49 CFR 171-178 before being brought to the Fuel Storage Building door. They will be housed in special shipping containers, and transported to a processing facility for volume reduction. Non-decontaminatable portions of the racks will be shipped to a . l
.l licensed radioactive waote burial site. The volume reduction is l expected to reduce the overall volume of the racks to about 1/10th l of their original value.
O ^11 ehises of the raracking activity will be conducted in accordance with written procedures which will be reviewed and approved by GPU Nuclear. ' 2-9 r 1
Table 2.1 Module Data (Reaion 1) Module I.D. Ouantity Array Size Total Cell Count
- A 1 7 x 13 91 B 1 8 x 13 104 Total number of cells for Region 1: 195 Table 2.2 Module Data IRecion 2)
H2dule I.D. Quantity Array Size Total Cell Count C 2 15 x 11 165 "T D 1 15 x 12 180* (O E 1 15 x 12** 138
-7x6
otal number of cells 'for Region 2
. 648 plus 3 failed fuel
~
containers One cell out of 180 cells in this module is lost; it is used as overload fixture. In addition there are three failed fuel container storage locations. , ( }) 2-10
f P 1
,I i O 1 i
Table 2.3
; Module Data for Future Recion 2 Racks i
'l l' Module I.D. Quantity Array Size Total Cell Count F 6 9 x 12 108 Total cell count: 648
!O' .
t 5.fl i= 1 t,- .f a
- l
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I i I I l (~) LJ l l Table 2.4 Common Module Data Storage cell inside dimension: 9 inch .06 inch Storage cell height (above the baseplate): 166" Baseplate thickness: 0.5" Support leg height: 6" (nominal)
-Support leg type Remotely adjustable legs Number of support legs: 4 (minimum)
Remote lifting and handling provision: Yes g o Poison material: Boral Poison length: 138" - Region I l 144" - Region II Poison width: 7.5" l Cell Pitch: (px,p )y 11.07" t .06-inch (Region I) 9.20" i ,06 inch (Region II) 2 0 2-12
?
i ( 1 r ! I Table 2.5 l l Module Data Dimensions * (inches) Weight ** Module I.D. East-West North-South fibs) A 77 144 22,500 i B 88 144 26,000 i l C 140 103 24,000 D 140 112 26,000 E 140 112 20,000 F 84 112 16,000 Dimensions (inches) are bounding rectangular envelopes rounded to the nearest inch. Weights are upper bounds (shipping weights) in pounds. .- 2-13 l
i + l 1 Table 2.6 Beral Experience List (Domestic and Foreign) Pressurized Water Reactors Vented Construc- Mfg. Plant Utility tion Year Bellefont 1, 2 Tennessee Valley Authority No 1981 D.C. Cook 1,2 Indiana & Michigan Electric No 1979 Indian Point 3 NY Power Authority Yes 1987 Maine Yankee Maine Yankee Atomic Power Yes 1977 Salem 1, 2 Public Service Elec & Gas No 1980 Seabrook New Hampshire Yankee No --- Sequoyah 1,2 Tennessee Valley Authority No 1979 Yankee Rowe Yankee Atomic Power Yes 1964/1983 Zion 1,2 Commonwealth Edison Co. Yes 1980 Byron 1,2 Commonwealth Edison Co. Yes 1988 l Braidwood 1,2 Commonwealth Edison Co. Yes 1988 l Yankee Rowe Yankee Atomic Electric Yes 1988 Boiling Water Reactors I l Browns Ferry 1,2,3 Tennessee Valley Authority Yes 1980 l
,- Brunswick 1,2 Carolina Power & Light Yes 1981
(_j Clinton Illinois Power Yes 1981 Cooper Nebraska Public Power Yes 1979 Dresden 2,3 Commonwealth Edison Co. Yes 1981 Duane Arnold Iowa Elec. Light & Power No 1979 J.A. Fitzpatrick NY Power Authority No 1978 E.I. Hatch 1,2 Georgia Power Yes 1981 Hope Creek Public Service Elec & Gas Yes 1985 H mboldt Bay Pacific Gas & Electric Yes 1986 Lacrosse Dairyland Power Yes 1976 Limerick 1,2 Philadelphia Electric No 1980 Monticello Northern States Power Yes 1978 Peachbottom 2,3 Philadelphia Electric No 1980 ! Perry, 1,2 Cleveland Elec. Illuminating No' 1979 Pilgrim Boston Edison No 1978 ! Shoreham Long Island Lighting Yes --- ! Susquehanna 1,2 Penr.sylvania Power & Light No 1979 ; Vermont Yankee Vermont Yankee Atomic Power Yes ! 1978/1986 Hope Creek Public Service Elec & Gas Yes 1989- -l i 2-14 (])
)
4 O Table 2.6 (continued) Foreign Installations Using Boral France 12 PWR Plants Electricite de France South Africa Koeberg 1,2 ESCOM Switzerland Beznau 1,2 Nordostschweizerische Kraftwerke AG
.Gosgen Kernkraftwerk Gosgen-Daniken AG
~
Taiwan Chin-Shan 1,2 Taiwan Power Company
-Kuosheng 1,2 Taiwan. Power Company 1
2-15
t 1
)
r l (. Table 2.7 1100 Alloy Aluminum Physical Properties I Density 0.098 lb/cu. in. ! 2.713 gm/cc j i Melting Range 1190-1215 deg. F 643-657 dog. C j Thermal Conductivity 128 BTU /hr/sq ft/deg. F/ft (77 deg. F) 0.53 cal /sec/sq cm/deg. C/cm Coef. of Thermal 13.1 x 10-6/deg. F Expansion 23.6 x 10 /deg. C (68-212 deg. F)
' Specific heat 0.22 BTU /lb/deg. F (221 deg..F) 0.23 cal /gm/deg. C Modulus of 10x106 p,i Elasticity Tensile Strength 13,000 psi annealed (75 deg. F). 18,000 psi as rolled Yield Strength 5,000 psi annealed (75 deg. F) 17,000 psi as rolled Elongation 35-45% annealed (75 deg. F) 9-20% as rolled Hardness (Brinell) 23 annealed 32 as rolled-Annealing Temperature 650 deg. F 343 deg. C l
h 2-16 o
e O Table 2.8 Chemical Composition - Aluminum (1100 Alloy) 99.00% min. Aluminum 1.00% max. Silicone and Iron l 0.05-0.20% max. Copper
.05% max. Manganese -
.10% max. Zinc
.15% max. others each O
I t P i-' 2-17 r
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.- - .. - . . __. _ _ _ _ _x
Oj t Table 2.9 Boron Carbide Chemical Comeosition, Weicht % Total boron 70.0 min. B10 isotopic content in 18.0 natural boron Boric oxide 3.0 max. Iron 2.0 max. Total boron plus 94.0 min. total carbon Boron Carbide Physical Procerties Chemical formula BC4 r^s Boron content (weight) 78.28% O Carbon content (weight) 21.72% Crystal Structure rombohedral Density 2.51 gm./cc-0.0907 lb/cu. in. Melting Point 24500 0 - 4442 0 p Boiling Point 3500 0 C-6332 0p 1 Microscopic capture 600 barn cross section i 1
- h. 2-18 l
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D O V 3.0 RACK FABRICATION AND APPLICABLE CODES The object of this section is to provide a self-contained description of rack module construction for Pool A of TMI Unit I to enable an independent appraisal of the adequacy of design. A list of applicable codes and standards is also presented. 3.1 FABRICATION OBJECTIVE The requirements in manuf acturing the high density storage racks for Pool A may be stated in four interrelated points: (1) The rack module will be fabricated in such a manner that there is nn weld splatter on the storage cell surfaces p v which would come in contact with the fuel assembly. (2) The storage locations will be constructed so that redundant flow paths for the coolant are available. (3) The f abrication process involves operational sequences which permit immediate verification by the inspection staff. (4) The _ storage cells are connected to each other by austenitic stainless steel corner welds which leads to a honeycomb lattice construction. The extent of welding is selected to "detune" the racks from the seismic input motion (1/2'SSE and SSE). I d 3-1 e
( 3.2 RACK MODULE FOR REGION I This section describes the constituent elements of the TMI Unit I Region I rack modules in the fabrication sequence. The rack module manufacturing begins with fabrication of the box. The " boxes" are fabricated from two precision formed channels by seam weldir.g in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. Figure 3.1 shows the box. The minimum weld penetration will be 80% of the box metal gage which is 0.073- (14 gage). The boxes are manufactured to 9.00" I.D. (nominal inside dimension). A die is used to flare out one end of the box to provide the 300 tapered lead-in (Figuro 3.2). One inch diameter holes are punched p on two sides near the other end of the box to provide the k requisite auxiliary flow holes. Each box constitutes a storage location. Each side of a box facing another box is equipped with a narrow rectangular cavity which houses one integral Boral sheet (poison material). The design objective calls for installing Boral with minimal surface loading. This is accomplished by die forming a " picture frame sheathing" as shown in Figure 3.3. This sheathing is made
'to precise dimensions such that the offset is .010 to .005 inches greater than the poison material thickness.
l l i
'~J 3-2 l
1
i The poison material is placed in the customized flat depression O reston ef the sheathine, which is next 1eid en a side of the l
" box". The precision of the shape of the sheathing obtained by l die forming guarantees that the poison sheet installed in it will not be subject to surface compression. The flanges of the '
sheathing (on all four sides) are attached to the box using skip welds. The sheathing serves to locate and position the poison sheet accurately, and to preclude its movement under seismic conditions. Having fabricated the required number of the composite box assemblies, they are joined together in a fixture using connector elements in the manner shown in Figure 3.4. The pitch between the box centerlines is px in one principal direction and py in the other principal direction. The values of px and py are given in l Section 2 (Table 2.4) of this Licensing Report. The fabrication procedure in either direction is identical, since the channels are fillet welded to make the inter-box connection. Figure 3.6-shows h an elevation view of two storage cells of a Region I rack module. Joining the cells by the interconnection elements results in a well defined shear flow path, and essentially makes the box assemblage into a multi-flanged ba a type structure. In the next step of manufacture, the " base plate" is attached to the bottom edge of the boxes. The base plate is a 1/2" thick austenitic stainless steel plate stock which has a 6" hole burned out in a pitch identical to the box pitch. The base plate is attached to the cell assemblage by fillet welding the box edge to the plate. p 3-3 h
l O i In the final step, adjustable leg supports (shown in Figure 3.5) are welded to the underside of the base plate. The adjustable legs provide a ! 1/2" vertical height adjustment at each leg location. The manufacturing of the Region I rack modules ! culminates with appropriate NDE of welds, which includes visual ! examination of cell longitudinal seam welds and cell-to-cell connection welds and liquid dye penetrant examination of support welds, in accordance with the design drawings. Figure 3.6 shows an elevation view of two cells with a fuel assembly indicated in cross-section in one cell. 3.3 RACK MODULE FOR REGION II Region II storage cell locations have a single poison panel (l between adjacent austenitic steel surfaces. The significant V components (discussed below) of the Region II racks are (1) the storage box subassembly (2) the base plate, (3) the neutron absorber material, (4) picture frame sheathing, and (5) support legs. (1) Storace c. ell box subassembly: As described for Region I, the "coxes" are fabricated from two precision formed channels by seam welding in a machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat input. Figure 3.1 shows the " box". The mininum weld penetration will be- 60% of the box metal gage which is 0.075" (14 gage). The boxes are manufactured to 9 inch inside dimension (nominal). 1 l l 3-4
I l As shown in Figure 3.7, each box has two lateral holes t l punched near its bottom edge to provide auxiliary flow holes. A " picture frame sheathing" similar to Region I I rack construction is attached to each side of the box ; with the poison material installed in the sheathing < cavity. The edges of the sheathing and the box are l welded together to form a smooth edge. The box, with 1 integrally connected sheathing, is referred to as the
" Composite box". J 3
The " composite boxes" are arranged in a checkerboard array to form an assemblage of storage cell locations (Figure 3.8). The inter-box welding and pitch
- adjustment are accomplished by small longitudinal y connectors.
This assemblage of box assemblies is welded edge-to-edge as shown in Figure 3.8, resulting in a honeycomb structure with axial, flexural and torsional rigidity depending on the extent of intercell welding provided. It can be seen from Figure 3.8 that two edges of each interior box are connected to the contiguous boxes resulting in a well defined path for " shear flow". (2) Base Plate: The base plate provides a continuous 1 horizontal surface for supporting the fuel assemblies. '
']k The base plate has a concentric hole in each cell location as described in the preceding section.
The base plate is attached to the cell assemblage by
. fillet welds. The baseplate in each storage cell has a 6" diameter flow hole.
(3) The neutron absorber material: As mentioned in the preceding section, Boral is used as the neutron absorber material. (4)' Picture Frame Sheathina: As described earlier, the sheathing serves as the locator and . retainer of the . poison material. ' Figure 3.3 shows a schematic of the
, -sheathing.
1
)
O 3-5 L
i i-m
;j -
(5): Suenort Leas: As stated earlier, all support legs are the adjustable type (Figure 3.5). The top position is
'made of austenitic steel material. The bottom part is made of 17:4 Ph series stainless steel to avoid galling problems. ;
Each support leg is equipped with a readily accessible i socket to enable remote leveling of the rack after its placement in the pool. _ Lateral holes in the support leg provide ~the requisite coolant' flow path.
<z 3.4 CODES, STANDARDS, AND PRACTICES FOR THE TMI UNIT I SPENT 1 FUEL POOL' RACKS- 1 y The fabrication of the rack modules for TMI Unit I is performed under.a strict quality. assurance system suitable for manufacturing and complying with the provisions of 10CFR50 Appendix B.
1 The following . codes , standards and practices - will be used as j applicable.for the' design, construction, and assembly of the spent:
] fuel storage' racks. ' Additional: . specific . references detailed analyses are given.in each section.
related to
'a.. Desion' Codes (1) .AISCl Manual o
~f Steel Construction, 8th Edition, 1980.
, (2)- . ANSI N210-1976, "DesignLObjectives for Light Wa.ter '
Reactor ' Spent : Fuel Storage - Facilities- at Nuclear
.i , Power Stations'." i
-(3) American Society of. Mechanical ' Engineers- ( ASME );,
Boiler: and Pressure Vessel Code, .Section III, 1986. American 9. (4). ASNT-TC-1AL June, 1980 Society .for Nondestructive ~ Testing (Recommended Practice for u- Personnel Qualifications). 3-6
.n-(), b. Material Codes (1) American Society for Testing and Materials (ASTM)
Standards - A-240. (2) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section II - Parts A and C, 1986. '
- c. Weldino Codes t
ASME Boiler and Pressure Vessel Code, Section IX- ! Welding and-' Brazing Qualifications (1986).
- d. Quality Assurance, Cleanliness, Packaaina, Shippino, Receivino, Storace, and Handlina Reauirements
( l ') LANSI N45.2.2 . Packaging, Shipping, ~ Receiving, Storage and Handling. of Items for Nuclear Power Plants. (2). ANSI. 45.2.1 - Cleaning of , Fluid Systems and. Associated Components during-Construction Phase of t Nuclear Power' Plants. D d (3), ASME: ' Boiler and'- Pressure. Vessel, Section V, Nondestructive
- Examination, 1983 Edition, including-:
. Summer and Winter 1983.
. ANSI -- N16.1-75 Nuclear = Criticality Safety 4 -(4) ~ Operations
. with- Fissionable- Materials- -Outside Reactors.
' (5)' ANSI N16.9 75. Validation of J Calculation ' Methods.
for Nuclear Criticality Safety.
- ('6) ANSI - N45.2.11, 1974 Quality ' Assurance
. Requirements for~ the Design- :of ; Nuclear . Power- j
' Plants. '
[-l ,N
, ?
3-7 y i ~
- - . . . _- .- . . . . . - - . - _ . - . - . .-.- - - ~
4
'O! !
- e. Other References
-(1)- NRC Regulatory Guides 1.13, Rev. 2 (proposed);
1.29, Rev. 3; 1.31, Rev. 3; 1.61, Rev. 0; 1.71, Rev. 0; 1.85, Rev. 22; 1.92, Rev. 1; 1.124, Rev. 1; and 3.41, Rev. 1. (2) General Design Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50, 6 Appendix A (GDC Nos. 1, 2, 61, 62, and 63).. (3) NUREG-0800, Standard Review Plan, Sections o 3. 2.1,- 3.2.2, 3.7.1, 3.7.2, 3.7.3, 3.8.4. (4)- "OT Position for Review and Acceptance of ' Spent i Fuel Storage and Handling Applications," dated' April 14, 1978, and the -modifications. to this-
. document of January 18, 1979.
[ (5). NUREG 0612, " Control of Heavy- Loads at Nuclear ll Power Plants", USNRC, Washington, D.C.- 3.5 MATERIALS OF CONS'TRUCTION Storage Cell: SA240-304 3 , 4 Baseplater- SA240 304 ' ! Support Leg (female) . SA240-304 Support Leg (male):
-Ferritic stainle s s '. . ( anti-galling material)1SA564-630 l Poison:
- Boral
.l
^
l ' Failed Fuel Canister -SA312-304 L > K l-L ~ r%. v 1 3-8 i t
i i 1 i i
.O !
1 1
\
l- . h O . s
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8 l SEAM
=- ' WELD 1
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)
[) k ' N'- I
- s. ~ -
l 1 I (3 V '/ ,, 3 0'o N ~., FIGURE 3.2' LEAD-IN FOR REGION I MODULES =; I !. i
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u i g- I
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4 fl, v a c- ,, y
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~
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. 3== =,4 ,
2- , K j l l l ! ) a FIGURE 3.6
.h 1
ELEV ATION VIEW OF A REGION I RACK SHOWING TWO STORAGE CELLS 1 l 1 3-14 l
T PITCH t n-l, f. f f
, a l , a N > M y
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ONE' INCH FLOW HOLE (TYPICAL) FIGURE 3.7 ELEVATION! VIEW 0F' REGION II RACK MODULE 1 3-15
---mmeumimmmmme----- - .
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1 t r 1. ly . y FIGURE 3.8 ARRAY OF REGION II CELLS (NON-FLUX TRAP CONSTRUCTION) 3-16
~
i 4.O' CRITICALITY SAFETY ANALYSES l
'4 .1 DESIGN BASES i The high density spent fuel storage' racks for Three
' Mile Island Unit 1 are designed to' assure that the effective m neutron multiplication factor (keff) is equal to or less than 0.95 with' the- racks fully loaded with fuel of the highest !
anticipated reactivity, and flooded with unborated water at the ! q temperature within the operating range corresponding to the !
~ highest reactivity. The maximum calculated reactivity includes !
a margin for uncertainty in reactivity calculations including mechanicalL tolerances. All - uncertainties are statistically a j _ combined, _such that the final keff will be equal to or less
'than 0;951with a.95% probability at a 95% confidence-level. :
u Applicable codes, standards, apd regulations or
- l. . pertinent' sections-thereof, include the following:
L o General-Design Criteria 62, Prevention'of Criticality _ in Fuel-Storage'and' Handling.= ti a ll
'o' USNRC' Standard ReviewEPlan, -NUREG-OSOO, Section 9.1.2=, Spent Fuel Storage,_Rev.J3 - July 1981 o' '
'USNRC letter of. April.14, 1978, to_allIPower Reactor i
- Licensees - OT Position for Review andl Acceptance of; 1 Spent Fuel- Storage and Handling . Applications, L including modification letter' dated January 18,- 1979. ,
a L o USNRC. Regulatory Guide-- 1 ; 13,- . Spent Fuel' Storage. Facility Design ~ Basis, Rev. - ' 2. (proposed) , - December ! v 1981. ,
.i . gi.
4-1
- l l:'
t-l L
. . . - - _~
, T o- ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel 'l Outside Reactors. I l
.o; ANSI /ANS-57.2-1983, Design Requirements for Light. l
' Water Reactor Spent Fuel Storage Ft.cilities at I Nuclear Power Plants.
, o ' ANSI N210-1976, Design Objectives for Light Water j Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
'o- ANSI N18.2-1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor
~ Plants.
USNRC guidelines and the applicable ANSI standards-specify that.. the. maximum-. effective multiplication factor,-
"ke'ff", including uncertainties, shall be less than'or equal to O.95.~ The infinite multiplication: factor, "b", is calculated D for an' infinite array,:* neglecting neutron -losses due to . leakage
!: .from,the actual storage rack, and thereforelresults in a higher-
;an'd "more conservative value. In the . present . evaluation of criticality safety in the Three Mile' Island, ' Unit storage 6 racks ,; the ' design basis - criterion was: assumed to be ' ai " A" of .
l
- 0.95 b ' which ' is more -. conservative than> the11imit specified in cthe regulatory guidelines'.-.
\ L . . To assure the. true reactivity; will always : be less i than the calculated Jreactivity, the following conservative !: assumptions were:made: of Moderator is u. . v.ated ' water; at a' temperature that - [
.results-in the highect reactivity-(68'F).
3: l; ' Ly ,
- 4-2 R
?w i i t a s ( L J>~ ' i
i s ,
/? o In all cases (except for the assessment of peripheral and certain abnormal / accident conditions A J effects ,
where ' neutron leakage is inherent), the infinite
~
l multiplication factor, koo , was used rather than the ! ef fective multiplication f actor, . kef f (i.e., neutron i loss from radial and axial leakage neglected). o Neutron absorption in minor ' structural members is neglected, i.e., spacer grids are analytically .i replaced by water. ' The design basis fuel assembly is a 15 x 15-Babcock &- Wilcox MkB8 fuel assembly containing 002 at a maximum initial l enrichmente of 4. 6 : wtt ' U-2 35 by weight, corresponding to 59.4: i grams- U-235 per axial centimeter' of fuel . assembly. Two , separate storage regions are provided in the spent fuel s torage
- pool, with.-independent' criteria - defining . the highest potential reactivity!in each of,the.two_ regions as:follows::
1, o Region:I is-designed to accommodate new fuel with a , i maximum enrichment. of 4. 6 L wt% U-235, or spent fuel
.regardless-of the discharge fuel burnup. ,
': o- Region IILis designed to accommodate fuel ofivarious'
' initial' enrichments which have accumulated minimum
.burnups- within .the: ' acceptable domain depicted in Figure 4.1.
The water - in the ': spent fuel. storage pool - normally T contains soluble boron which would-' result - in large subcriti : ! cality margins)under actual operatingLconditions. 'However,-the
'NRC guidelines, based upon.the_ accident condition inLwhich all: $
soluble poison nis! assumed to have been lost; specify. that. the
-limiting - kegf iofuO.95 be -evaluated in the absence . of . soluble 1 boron.~ . The double contingency principle-of . ANSI N-16.1-1975: !
.and' of, the April 1978_ ; NRC . letter allows credit for: s'oluble 1 boron under other-abnormalLor accident.xconditions-since'only_a
- ' single accident' need be considered at one time. Consequences - '
1 4 . v W v -
.}
of abnormal and accident conditions have also been evaluated,
# where " abnormal" refers to conditions (such as higher water temperatures resulting from full-core discharge) which may reasonably be expected to occur during the lifetime of the !
plant and " accident" refers to conditions which are not i expected to occur but nevertheless must be protected against. i i i
)
l
- 1. \
(', 4-4 l ( 1 l l' l- l ! 1 1 l
x W 4.2- .
SUMMARY
OF CRITICALITY ANALYSES s 4.2'.1 Normal Operatinc Conditions The criticality analyses of each of the two separate regions of the spent fuel storage pool are summarized in Table " 4.1 for the design basis storage conditions which assumes the single accident condition of the loss of all soluble boron. l
-l The , calculated maximum reactivity in Region II includes a burnup-dependent allowance' for uncertainty in depletion calculations and, furthermore, provides an~ additional margin of more than 1 % 6k'below the design basis infinite multiplication
. factor ( km), ' of 0. 9 5. Au ' cooling time increases in long-term storage, decay of Pu-241 results in a significant decrease.in-
. reactivity, which - will provide an i ncreasing subcriticality
. margin'and tends to further compensate for any uncertainty in depletion calculations.
I .l) Region: II- can safely accommodate fuel of various b l initial- enrichments and - discharge fue1L burnups, provided.the combination? falls; within;the acceptable domain illustrated-by, the solid line_in Figure 4.'1. for convenience, ' the" minimum g (limiting)) burnup data Lin : Figure ~ '4.1 for: unrestrict'ed storage : j; ,;l' in' RegionEII can ' be described as a - function .of the' initial' enrichmant, E, in ' weight percent U-235 by~a fitted: polynomial 3 expression as;follows; " $c a L, ,
)
4.- 5 V i
4 i l For Region II Unrestricted Storage I l Minimum Burnup in MWD /KGU = i-
- 36.0 + 27.1 E - 4.0 E2 + .34 E3 4 (for initial enrichments up to 5 wt% U-235)
This polynomial fit is' generally accurate to within 0.002 6k in reactivity, except at the lower enrichment limit (1.75%) where the fit conservatively overpredicts the limiting burnup. l
'The burnup criteria identified above for acceptable
' l storage .inL Region ' II .can be - implemented in appropriate ad-ministrative procedures to assure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2. -Administra-tive procedures will-also be employed to confirm and assure the presence of ' soluble poison in .the pool water during fuel
~
, handling operations, as a further margin pf safety and as a- - Lprecaution. in the event of fuel misplacement during fuel L.- . handling: operations.; o 4 . 2 ^. 2 - Abnormal and Accident' Conditions s'
)
7
.Although credit for- the_ soluble, poison normally.
L present in- the; spent fuel pool -_ water is ' permitted under L4 abnormal Jor accident conditions, most abnormal or accident conditions will not result in~. exceeding-the limiting; reactivity. ' h (kegg of 0.95) even in the absence of soluble poison.* The L Plant! Technical Specification 5.6.1 requires-that 600 ppm / r boron concentration during fuel handling operations. L 4-6 l - l
t 1 effects on reactivity of credible abnormal and accident conditions are presented in detail in Section 4.7 and briefly summarized in Table 4.2. Of these abnormal / accident condi-tions, only one has.the potential for a more than negligible positive reactivity effect. The inadvertent misplacement of a fresh fuel assembly
.l (either into a Region II storage cell or outside and adjacent !
.to'a rack module) has the potential for exceeding the limiting l
-reactivity, =should there be a concurrent and independent accident condition resulting in the loss of all soluble poison.
Administrative procedures to assure the presence of soluble 1 poison during- fuel handling operations will preclude- the possibility!of.the'_ simultaneous occurrence of the two indepen-dent accident conditions. The largest reactivity!-increase would occur if La new fuel assembly of the highest permissible !
. reactivitywere to -be positioned outs'ide and immediately_
~l a'djacent. ~ 'o a . f ully - loaded Region- II storage rack module.
.# LUnder this accident condition, credit for the ' presence _ of a soluble poison ' is. permitted by' NRC ' guidelines *, and Tit 'is 1 i estimated that a minimum boron ; concentration of:- about : 500 ppm boron. would be = adequate to assure that the limiting keyf.of. 0.95'is not exceeded.
.i
- Double- contingency -principle'-of-~_ ANSI N16.'l-1975,
~
as , specified in tha April 1 14, 1978: NRO letter (Section 1. 2 ) .. and - j
~ implied in the _' proposed - revision ' to Reg. Guide 1.13 (Section-l'.4, Appendix A).-
,s 4-- 7 ,
i
i l' , i- 4.3 REFERENCE FUEL STORAGE CELLS i 4'.3.1 Reference Fuel Assembly l~ The design basis fuel assembly, illustrated in Figures'4.2 and 4.3, is a 15 x 15 array of fuel rods with 17 rods replaced by 16 control rod guide tubes and'1 instrument thimble. Table- 4.3 summarizes the fuel assembly design i spec'*ications and the expected range of the significant manufacturing tolerances. l 4.3.2 Recion I Fuel Storace Cells The nominal spent fuel storage cell- used for the l criticality analyses of Region I storage cells is shown in Figure 4.2. s The rack is composed of Boral . absorber material between an -: 9. OO-inch I . D . , 0.075-inch thick inner = stainless steel' box, and a 0.024-inch outer stainless, steel cover plate.
-The fuel assemblies are centrally located In each storage cell on a nominal ~ .latitice spacing: of- 11.07'.1 0.06 inches. Stainless-steel'gapL. channels-connect one storage-cell box.to another in.a rigid-structure!and define an outer' water space between' boxes.
This.' outer-water: space' constitutes a flux-trap between-the'two (thermal-neutron . opaque) Boral absorber panels. The Boral I
~
absorber has. a - thickness 1 of 0.075 i=O.004 inch.and a nominal'
.B-10 areal density of'O.0211 g/cm2 , ' l L
1 4-8 TL
l l l l4.3.3' Recion II Fuel Storace Cells i l
- l. The design basis for Region II storage cells is fuel l of 4.6 wt% U-235 initial enrichment burned to 37 MWD /KgU. In i l Region II, the storage cells are composed of a single Boral l
]
absorber panel between the stainless steel walls of adjacent ! i storage cells. These cells, shown in Figure- 4. 3, are located l on a lattice spacing of 9.20 1 0.06 inches. The Boral absorber 1 has a thickness of 0.089 i O.004 inch and a nominal B-10 areal l density of 0.026 g/cm 2, t 4.4 ANALYTICAL METHODOLOGY 4.4.1 Reference Design Calculations i In the fuel rack analyses, the primary criticality l analyses of the high , density spent fuel storage racks were performed with a two-dimensional multi-gropp transport theory i l' technique,. using the CASMO-2E (Ref. '4.4'.1) c omputer code. Independent verification calculations were made with a Monte i L Carlo technique utilizing the AMPX-KENO computer package (Ref. ' ll 4'. 4. 2 ) ' with the 27-group SCALE
- cross-section library (Ref.
a 4.4.3) and the NITAWL subroutine for U-238 resonance shielding i effects (Nordheim= integral treatment). Benchmark calculations,- ;
. presented in Appendix A, indicate a bias . of 0.0013 with an uncertainty. of i O.0018 for CASMO-2E and 0.0105 i O.0048 l (95%/95%) for NITAWL-KENO. .
L
*" SCALE" is'an acronym for Standardized Computer Analysis' '
for Licensing Evaluation, a standardicross-section set develop-edLby.ORNL for the USNRC. 4-9 4 L
+
.l CASMO-2E was also used both for burnup calculations and for evaluating the small reactivity increments associated l with manufacturing tolerances. In tracking long-term (30-year) reactivity ef fects of spent fuel stored in Region II of the m
fuel storage rack, previous CASMO-2E and EPRI-CELL calculations confirmed a continuous reduction in reactivity with time (after
-Xe decay) due primarily to Pu-241 decay and Am-241 growth.
i Two group dif fusion theory constants, edited in the output of CASMO-2E, were used in PDQO7 (Ref. 4.4.4) for auxiliary--calculations of the small incremental reactivity
- effeet of eccentric fuel positioning: and fuel ' assembly mis- j placement. These constants were also used in a one dimensional
; diffusion theory routine (benchmarked against PDQO7). to evaluate reactivity' effects -of the Boral axial length (Documentation :of Computer Code l SNEID, .Holtec Report No. HI-89411, by S. Turner.(1989)].
In the. geometric model used in the calculations, each fuelL rod and its . cladding. were described explicitly and reflecting-boundary' conditions.(zero' neutron current) were used.
.in the axial directionLand at-the centerline of the Boral and p steel plates between, storage cells.- -These boundary conditions i
have'the=effact of' creating an'infiniteLarray of storage cells. ; in;all directions. AMPX-KENO Monte Carlo-calculationsLinherently includeL o ,a statistical uncertainty due to the random.natureEof. neutron L To' minimize the statistical uncertainty of the KENO-- tracking. , L ' calculated reactivity, a minimum of 50,000 ' neutron histories b in -' 100 generations of 500 neutrons each,sare accumulated - in o I each calculation. 4 - 10
- - - +
4.4.2 Fuel Burnuo Calculations and Uncertainties CASMO-2E was used for burnup calculations in the hot operating condition. CASMO-2E has been extensively benchmarked ( Appendix A and Refs. 4.4.2 and 4.4.6) against cold, clean, critical experiments (including plutonium-bearing fuel), Monte Carlo calculations, reactor operations, and heavy-element concentrations in irradiated fuel. In particular, the analyses
-l (Ref. 4.4.5) Of 11 critical experiments with plutonium-bearing fuel gave an average keff of 1.002 i O.011 (95%/95%), showing adequate treatment of the plutonium nuclides. In addition, l Johansson (Ref. 4.4.6) h as obtained very good agreement in calculations of .close-packed, high-plutonium-content, ex-perimental configurations.
Since critical experiment data with spent fuel is not available for determining the uncertainty in burnup-dependent L reactivity calculations, an allowance for uncertainty in reac-tivity :. was assigned base'd upon other considerations. Over a L considerable portion of the _burnup history in PWRs, the
, reactivity loss rate is-approximately-0.01 6k for each MWD /KgU L burnup, becoming smaller at the higher burnups. Assuming the uncertaintyi in depletion calculations is less than 5% of the l
total reactivity decrement, an uncertainty in reactivity *' equal to-0.0005.-6k for each. MWD /KgU in burnup may be' assigned. -For the:Three Mile Island Unit _1 storage racks at the design basis
.burnup_of 37 MWD /KgU,-the reactivity allowance'for unce-tainty_
is 0.0185 6k. Table 4.4 summarizes results of thr, burnup
*Only that portion of the uncertainty due to burnup..
Other uncertainties are accounted for elsewhere.- 4- 11 l
. 'v.
c
analyses and allowances for uncertainties at other burnups. At O the hieher eurnups, this assumption resu1ts in an uncertainty greater than 5% of the reactivity decrement which provides an
. extra margin to allow for the existence of a small positive reactivity increment from the axial distribution in burnup (see Section 4.4.3). In addition, although the reactivity uncer-tainty due to burnup may be either positive or negative, it is treated as an additive term rather than being combined statis-tically with other uncertainties. Thus, the allowance for uncertainty in burnup calculations is believed to be a conser-vative estimate, particularly in view of the substantial reactivity decrease with aged fuel as discussed in Section 4.4.4.
4.4.3 Effect of Axial Burnuo Distribution Initially, fuel loaded into the reactor will burn-with a slightly skewed cosine power distribution. As burnup O groetesses, the burnug distribution w111 tend to f1atten, L becoming more-highly burned in the central regions than in-the upper and lower ends. This effeet may be clearly seen in the l l curves compiled in Ref. 4.4.6. At high burnup, the more reactive- fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it would be - expected that over .most of the burnup history, distributed burnup fuel-
.assemblins would exhibit a slightly lower reactivity than that
. calculated for the average burnup. - As burnup progresses, the
' distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.
4 - 12 I
.O i Among others, Turner <Ref. 4.4.8) h as provided generic analytic results of the axial burnup effeet based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup. The trends observed, however, suggest the possibility of a small positive reactivity effect at the high burnup values (estimated to be less than about O.007 6k at 37 MWD /KgU) and the uncertainty in koo due to burnup, assigned at the higher burnups (Section 4.4.2) is considered adequate to encompass the potential for a small positive reactivity effect of axial burnup distributions.
Furthermore, reactivity significantly decreases with time in storage (Section 4.4.4 below) providing a continuously increas-ing margin below the 0.95 limit. 4.4.4 Lona-term Chances in Reactivity
. l- Since the fuel racks in -Region II are intended to contain spent fuel for long periods.of time, calculations were made using CASMO-2E to follow:- the long-term changes in reac-tivity of spent fuel over a 30-year' period. Early in the
, decay period, ' Xenon grows from Iodine decay (reducing reac-tivity) and subsequently decays, with the reactivity reaching a ' maximum at 100-200 hours. The decay of Pu-241 (13-year half-life) and growth of Am-241 substantially reduce _ reactivity during long . term storage, as indicated' in Table 4.5. The reference design criticality calculations do not take credit for this long-term reduction in reactivity, other than to l indicate an increasing subcriticality margin in Region II of ! the spent fuel storage pool. l- 4 - 13
=
l l l
1 1 h l'.4.5 REGION I CRITICALITY ANALYSES AND TOLERANCES 4.5.1 Nominal Desien l For the non.inal storage cell design in Region I, the CASMO-2E. calcula'. ion resulted in a bias-corrected b of 0.9217 i O.0018, which, when combined with all' known uncertainties, results in a .naximum b of 0.9281. Independent calculations with AMPX-KENO gave a bias-corrected b of 0.9065 i O.0089, l; including a one-sided tolerance factor (Ref. -4.5.1) f or 95% probability at a 95% confidence level. Combining all known 1
-uncertainty = factors, the calculated A 'becomes 0.908 i O.011 or =i a maximum b value of 0.919. This generally confirms the I reference CASMO-2E. calculations and suggests that the CASMO l
. calculation may be conservatively high.
4'.5.2 Uncertainties Due to Tolerances Q. 4.5.2.1 Boron Loadina Tolerances-The Boral' absorber panels used .in the storage cells are -nominally : 0.075 inch thick, 7.50-inch wide and 136-inch
. long, with ' a nominal B-10 areal' density of : 0.0211. g/cm2. The vendors manufacturing tolerance limit:is i O.0011 g/cm 3 in B-10=
content which assures that at-any point, the minimum B-10 areal
, density ' will not . be . less .than 0.020?g /cm 2. Differential-CASMO-2E calculations indicate- that. these ' tolerance- limits result in incremental reactivity uncertainties of-10.0018 6k.-
? , 4 - 14
i 4.5.2.2 Boral Width Tolerance The reference storage cell design uses a Boral panel with an initial-width of 7.50 t 0.06 inches. For the maximum tolerance of 0.06 inch, the calculated reactivity uncertainty is +0.0007 6k. 4.5.2.3 Tolerances in cell Lattice Spacino i The design storage cell lattice spacing between fuel assemblies is 11.07 t 0.06. A decrease in storage cell lattice spacing may or may not increase reactivity depending upon other dimensional changes-that may be associated with the decrease.in lattice. spacing. Decreasing the water spacing-between the fuel ,
.and the~ inner stainless steel box results in a small decrease.
in reactivitv. However, decreasing the flux-trap water spacing increases . 7tivity and both of these effects have been evaluated for their= independent design tolerances.
.The inner stainless steel box dimencion, 9.00 t 0.06 inches, defines the inner water thickness between,the fuel and the:inside-box. For the tolerance limit =of i O.06 inches, the uncertainty in reactivity is 10.0012 6k as determined by.
differential CASMO-2E calculations. The design flux-trap' water { thickness is 1.70 t 0.06 inches,-which results in an uncertain-ty:'of 20.0036 6k due to the -tolerance in flux-trap water thickness, assuming the water thickness is simultaneously reduced on all~four sides. Since the manufacturing. tolerances ; on each of the four sides are statistically independent, the , actual reactivity uncertainties would be less than t 0.0036 6k, 4 - 15 e _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ . _ - _ _ . _ _ _ _ _ _ _ . _ - _ _ - _ _ _ _ - . - __ __ __ - - _ - - - - - - - - - _ _ _ - - - - _ -_-_--------------l-
although the more conservative value has been used in the l criticality evaluation. 4.5.2.4 Stainless Steel Thickness Tolerances The nominal stainless steel thickness is 0.075 , 0.005 inch for the inner stainless steel box and 0.0235 t 0.0030 inch for the Boral cover plate. The maximum positive reactivity effect of the expected stainless steel thickness tolerance variations, .was calculated (CASMO-2E) to be 10.0008 6k. p 4'.5.2.5 Fuel Enrichment and Density Tolerances
- i. 1 The design maximum enrichment is 4.60 t 0.05 wt%
U-235. Calculations of. the sensitivity to small enrichment
- variations by CASMO-2E yielded a coefficient of 0.0052 6k.per
[ O.1 wt% U-235 at the design enrichment. For the assumed. k tolerance'on U i235-enrichment of.t 0.05 wtt, the uncertainty-on km is i O.0026 6k. Calculations were also made with the UO2 fuel density L incraased to.1 the maximum expected value of 10.44 g/cm3 (stack-L density).- For the reference . design calculations, the uncer - l 3 1 tainty-in reactivity is 1.0.OO25n6k=over the maximum expected j range'of-UO2 densities. L , 4 - 16 'I l , l- i [ O:
, 4.5.3 Eccentric Fuel Positionino tT GJ The fuel assembly is assumed to be normally located in the center of the storage rack cell. Calculations were also made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest l approach). These calculations indicated that, in Region I, the reactivity remains essentially the same (within 0.0001 6k), as x de ta.aiined by differential PDQO7 calculations with diffusion coefficients generated by CASMO-2E. This uncertainty is included in the evaluation of the highest potential reactivity l of the Region I storage cells.
4.5.4 Reactivity Effects of Boral Lencth l Based upon diffusion theory constanta edited in the ( CASMO-2E output (reference design and a special case with water replacing the Boral), one-dimensional- axial calculations were made to evaluate the reactivity effect of reduced Boral l- axial lengths. Reduced length of the Boral leaves small regions of active, fuel without poison at each end of the fuel lt assembly. The unpoisoned region at each end ia referred to as
" cutback".
The axial calculations used a thick ( 30 cm. ) water l reflector, negl acting the higher absorption of the stainless-1 l steel structural material' at the ends of.the fuel assembly. , Results of the calculations showed that the kegg remains less-than the reference kco of the storage. cells until the axial L ' reduction in Boral length substantiallyj exceeds the-design 3 inch cutback top and bottom, corresponding to an overall 4 - 17
- _________________________________.________JL'
Boral length of 136 inches. Thus, the axial neutron leakage O more than compensates for the 3-inch design cutback and the reference km remains a conservative over-estimate of the true heff* O e 4 - 18 O L
r~ l 4.6 REGION II CRITICALITY ANALYSES 4.6.1 Nominal Desien Case l The principal method of analysis in Region II was the CASMO-2E code, using the restart option in CASMO-2E to analyti-cally transfer fuel of a specified burnup into the storage rack configuration at a reference temperature of 20*C (68'F). Calculations were made for fuel of several dif ferent initial enrichments and, at each enrichment, a limiting k. value was established which included an additional factor for uncertainty in the burnup analyses and for the axial burnup distribution. 1 The restart CASMO-2E calculations (cold, no-Xenon, rack geometry) were then interpolated to define the burnup value yielding the limiting k. value for each enrichment, as indi-cated in Table 4.6. These converged burnup values define the boundary of the acceptable domain shown in Figure 4.1. Burnup values calculated with the polynomial function given below are shown in Table 4.6 and on Figure 4.1. l For Region II Unrestricted Storage l Minimum Burnup in MWD /KGU =
- 36.0 + 27.1 E - 4.0 E2 + 0.34 E3 (for initial enrichments up to 5 wtt U-235)
At a burnup of 37 MWD /KgU, the sensitivity to burnup is calculated to be 0.007 6k per MWD /KgU. During long-term l storage, the k. values of the Region II fuel rack will decrease continuously as indicated in Section 4.4.4. 4 - 19 0 .
l g An independent AMPX-KENO calculation was used to l V l provide additional confidence in the reference Region II criticality analyses. Fuel of 1.75 wtt initial enrichment (equivalent to the reference rack design for burned fuel) was analyzed by AMPX-KENO and by the CASMO-2E model used for the l Region II rack analysis. For this case, the CASMO-2E kao (0.9270) at 1.75 wt% enrichment was slightly lower than the bias-corrected KENO value (0.9324 i O.0051, 95%/95%) obtained in the AMPX-KENO calculations. For conservatism, the nominal difference in reactivity (0.0054 6k) was added to the CASMO l result in the final evaluation. 4.6.2 Uncertainties Due to Tolerances 4.6.2.1 Boron Loadina Tolerancet , l The Boral absorber panels used in the Region II q storage cells are 0.089 inch thick with a nominal B-10 areal V density of 0.026 g/cm 2. The manufacturing limit of i O.006 g/cm 3 in B-10 loading assures that at any point the minimum B-10 areal density will not be less than 0.024 g/cm 2, olf. ferential CASMO-2E calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of i .0035 6k. 4.6.2.2 Boral Width Toleransta l The reference storage cell design for Region II (Figure 4.3) uses a Boral absorber width of ' 7.50 i O.06 inches. This tolerance results in a reactivity uncertainty of i O.0009 6k. 4 - 20
_. _ _-.u --.a a._-....__--a . . . . . > . _ . - . _ - _ _ . = + . ~ , _ . _ . _ , _ . _ . . _ O 4.6.2.3 re1erance in ce11 tettice so. cine The manufacturing tolerance on inner box dimension affects the storage cell lattice spacing between fuel assem-l blies in Region II is i O.06 inches. This corresponds to an uncertainty in reactivity of i O.0016 6k. 4.6.2.4 stainless steel Thickness T.olerance The nominal thickness of the stainless steel box wall is 0.075 inch with a tolerance of 10.005 inch, resulting in an uncertainty in reactivity of 10.0001 6k. 4.6.2.5 Fuel Enrichment and Density Tolerances Uncertainties in reactivity due to tolerances on fuel , l enrichment and UO2 density in Region II are assumed to be the
~
O i same as those determined for nesion 1. 4.6.3 Eccentric Fuel Positioning The fuel assembly is assumed to be normally located in the center of.the storage rack cell. Calculations were also made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that eccentric fuel j positioning results in a decrease in reactivity by about O.OO3 6k, as determined by PDQO7 calculations with diffusion I coefficients generated by CASMO-2E. The highest reactivity, 4 - 21 l t
. - . , - ___ _ __ _ ___..__ _ _ _ _ ________ _ d .
l therefore, corresponds to the reference design with the fuel
]
assemblies positioned in the center of the storage cells. i 4.6.4 Reactivity Effect of Boral Lenoth l Because the ends of the fuel assemblies in Region II { have less burnup than the average, and hence are more reactive, - an axial cutback is not used in this region. 1 0 5 4 - 22 P O
_. _ _ .m_... __ _ __.._._...___. ___ _ _ _ _ ___ _. I 4.7 ABNORMAL AND ACCIDENT CONDITIONS { 4.7.1 Temnerature and Water Density Effects The moderator temperature coefficient of reactivity is negative; a moderator temperature of 20'C (68'F) was assumed l for the reference designs, which assures that the true reac- I tivity will always be lower over the expected range of water temperatures. Temperature effects on reactivity have been i calculated and the results are shown in Table 4.7. Introducing voids in the water internal to the storage cell (to simulate ; boiling) decreased reactivity, as shown in the table. Since, , at saturation temperature, there is no significant thermal driving force, voids due to boiling will not occur in the outer l (flux-trap) water region of Region I. I With soluble poison present, the temperature coeffi-cients of reactivity would differ from those inferred from the O data in Tab 1e 4.2. nowever, the reactiviues wou1d aise be substantially lower at all temperatures with soluble boron l , present, and the data in Table 4.7 is pertinent to the higher-reactivity unborated case. ! 4.7.2 Drorced Fuel Assembly For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum
. separation distance from the fuel in the rack of more than 9 1/2- inches, including an estimated allowance for deformation under seismic or-accident conditions (1 inch). calculations show that, at this separation distance, the effect on reac- .
tivity is insignificant (<0.0001 6k). Furthermore, soluble 4 - 23
- _. -- . _ = . - - . . - - _ - _ - - - -. - .-
i boron in the pool water would substantially reduce the reac-O t1vity and assure that the true reactivity is a1 ways 1ess than ; the limiting value for any conceivable dropped fuel accident. 4.7.3 Lateral Rack Movement Lateral motion of the :, ick modules under seismic $ conditions could potentially alter the spacing between rack modules. However, the maximum rack movement has been deter-mined to be less than 0.35 inches under the design basis l seismic event (See Table 6.6, Section 6). In Region I, the water gap between rack modules nominally 3 inches, which, if reduced by 0.35 inches, would still be larger than the cor-responding design water-gap spacing (1.70") internal to the l rack modules. Region II storage cells do not use a flux-trap and the reactivity is therefore insensitive to the spacing l between modules. The spacing between Region I and Region II modules is suf ficiently large to preclude adverse interaction O even with the maxtmum seismica11 7 -induced reduction in spacing. Furthermore, soluble poison would assure that a reactivity less than the design limitation is maintained under all accident or abnormal conditions. , 4.7.4 Abnormal Location of a Fuel Assembly , The abnormal location of a fresh unirradiated fuel assembly of 4.6 wtt enrichment could, in the absence of soluble poison, result in exceeding the design reactivity limitation (k of 0.95). This could occur if a fresh fuel assembly of the highest permissible enrichment were to ' be either positioned l outside and adjacent to a storage rack module or inadvertently ' l 4 - 24 i i I
t i loaded into a Region II storage cell. 9 l sional PDQ) showed that the highest reactivity, including uncertainties, for these postulated accident conditions were as Calculations (2-dimen-followst l o Assembly outside Region I, maximum b = 0.992 l o Assembly outside Region II, maximum b = 1.001 l o Fresh assembly in Region II, maximum b = 0.9744 ! Soluble boron in the spent fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. It is estimated that a soluble l poison concentration of approximately 300 ppm boron uould be ' sufficient to maintain a A less than 0.95 (lacluding uncertainties) under the maximum postulated accident condition. S 9 1 e 4 - 25 v _ - - - - . , . , _ , _ _ ,
O
4.8 REFERENCES
4.4.1 A. Ahlin, M. Edenius, H. Haggblom, "CASHO - A Fuel ' Assembly Burnup Program," AE-RF-76-4158, Studsvik report (proprietary). t A. Ahlin and M. Edenius, "CASMO - A Fast Transport
- Theory Depletion Code for LWR Analysis," Mig Transactiona, Vol. 26, p. 604, 1977. >
M. Edenius et al., "CASMO Benchmark Report," Studsvik/ RF-78-6293, Aktiebolaget Atomenergi, March 1978. 4.4.2 Green, Lucious, Petrie, Ford, White, Wright, "PSR-63/AMPX-1 (code package), AMPX Modular Code System for Generating Coupl d Multigroup Neutron-Gamma ; Libraries from ENDF/B," ORNL-TM-3706, Oak Ridge National Laboratory, March 1976. ; 4.4.3 R.M. Westfall et al., " SCALE: A Modular Code System for ?erforming Standardized Computer Analyses for L:. censing Evaluation," NUREG/CR-02OO, 1979. O. 4.4.4 W.R. Cadwell, PDQO7 Reference Manual, WAPD-TM-678, i Bettis Atomic Power Laboratory, January 1967. 4.4.5 E.E. Pilat, " Methods for the Analysis of Boiling Water Reactors (Lattice Physicis ) , " YAEC-1232, Yankee Atomic Electric Co., December 1980. 4.4.6 E. Johansson, " Reactor Physics Calculations on Close-Packed Pressurized Water Reactor Lattices,"
' Nuclear Technoloav, Vol. 68, pp. 263-268, February
- 1985.
i 4.4.7 H. Rici.! ngs , Some Notes on PWR (H) Power Distribution Probabilities for LOCA Probabilistic Analyses, NRC Memorandum to P.S. Check, dated July
- , 5, 1977.
4 - 26 0
l 1 l REFERENCES (continued) 4.4.8 S. E. Turner, " Uncertainty Analysis - Burnup Distributions", presented at the DOE /SANDIA j Technical Meeting on Fuel Burnup Credit, Special 4
' Session, ANS/ ENS Conference, Washington, D.C., November 2, 1988 4.5.1. M.G. Natrella, Experimental Statistics National i Bureau of Standards, Handbook 91, August 1963. '
n , V 4 - 27 s O. ; 3-
.-_._,.m. .m. . - . . .. ' . ' Table 4.1 O
SUMMARY
OF CRITICALITY SAFETY ANALYSES l Region I Region II Design Basis burnup at O 37 MWD /KgU 4.6% initial enrichment - Temperature for analysis 20'C (68'F) 20*C (68'F) Reference km (CASMO-2E) 0.9204 0.9085 Calculational bias, 6k(l) 0.0013 0.0013 Uncertainties Bias i O.0018(1) i O.0018(1) B-10 loading i O.0018(2) i O.0035(3) Boral width i O.0007(2) g o,0009(3) l Inner box dimension i O.0017(2) i O.0016(3) Water gap thickness t 0.0036(2) NA SS thickness i O.0008(2) t o,00o1(3) Fuel enrichment i O.0026(2) i O.0026(4) Fuel density i O.0025(2) i O.0025(4) Eccentric position t 0.0025(5) Negative (6) O. Statisticalcombinagion i O.0064 i O.0056 of uncertainties ( ) Allowance for Burnup Uncertainty NA + 0.0185 Adjustment from Keno Calc'11ation Negative + 0.0054 Total O.9217 i O.0064 0.9337 i O.0056 Maximum Reactivity (km) 0.9281 0.9393 (1) Section 4.4.1 (2) Section 4.5.2 (3) Section 4.6.2 (4) For fuel tolerances, uncertainties in Region II assumed to be the same as those for Region I. (5) Section 4.5.3 (6) Section 4.6.3 (7) Squa7e root of sum of squares. 4 - 28 9 _ _ _ _ _ _ _ - _ - _ - ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - ' - - - - - - - - - - - - ~ - - - - ' - - - - - - " " ^ ^ - ~ ^ ' ' ' ^ ' ^ ^ ' ^ - ' ^ ^ ~ ^ ' " - ~ " ~ ' ~ ' ~ ~ ~ ^ ~ ~ ~ ^ ^ ^ ' ~ ~ ' ^ ' " ~ ~ ^ ' '
I
) '
Table 4.2 REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS , Accident / Abnormal Conditions Reactivity Effect Temperature increase (above 68'F) Negative (Table 4.7) Void (boiling)- Negative (Table 4.7) 1 Assembly dropped on top of rack Negligible (<0.0001 6k) Lateral rack module movement Negligible (<0.0001 6k) Hisplacement of a fuel assembly Positive (0.062 Max 6k) ' I) s-l.- l 4 - 29 , l i
- ]
p ,i e i O Table 4.3 DESIGN BASIS FUEL ASSEMBLY SPECIFICATIONS FUEL ROD DATA Outside diameter, in. O.430 - Cladding thickness, in. 0.0265 Cladding inside diameter, in. O.377 Cladding material Zr-4 Pellet density, % T.D. 95.0 Stack density, g UO 2/cc 10.225 i O.21 Pellet diameter, in. O.369 Maximum enrichment, wt % U-235 4.60 t 0.05 FUEL ASSEMPLY DATA Fuel rod array 15x15 () Number of fuel rods 208 Fuel rod pitch, in. . 0.568 Number of control rod guide and 17 instrument thimbles Thimble O.O.,'in. (nominal) 0.530 Thimble I.D., in. (nominal) 0.498 , 4 - 30 l L l
- _ jd
O ! Table 4.4 ALLOWANCE FOR UNCERTAINTIES IN REACTIVITY DUE TO DEPLETION CALCULATIONS ' Design Uncertainty Initial Burnup due to Design Enrichment MWD /KgU Burnup, 6k limit k.(1) 1.75 0 0 0.9270 2.0 4.928 0.0025 0.9245 - i 2.5 11.93 0.0060 0.9210 3.0 18.32 0.0092 0.9178 3.5 24.41 0.0122 0.9148 /~T 4.0 30.36 0.0152 0.9118 .V 4.6 37.00 0.0185 0.9085 5.0- 41.58 0.0208 0.9062 , (1) The design limit k. is determined by subtracting the burnup-
-dependent allowance for uncertainty (Column 3) from the '
design basis km (0.9270) for unburned fuel of 1.754 enrich-ment. With all uncertainties added (Table 4.1), the maximum ,
- k. is 0.9393 in all cases.
4 4 - 31 O
I ($) ' J l I 1 Table 4.5 LONG-TERM CHANGES IN REACTIVITY IN STORAGE RACK CTiLCULATED BY CASMO-2E ' Storage 6k from Shutdown (Xenon-free) Time, years at 1.6 wt% E and 37 MWD /KgU 1.0 +0.0012 4.0 -0.0056 10.0 -0.0198 20.0 -0.0371 30.0 -0.0477 uO i 4 - 32 f}
- 1 i
1
,~ ,
(_) l l 1 l Table 4.6 i FUEL BURNUP VALUES FOR REQUIRED REACTIVITIES (km) , WITH FUEL OF VARIOUS INITIAL ENRICHMENTS Calculated Initial Uncertainty (1) Design (2) Burnup limit (3) Enrichment in Burnup, 6k Limit km MWD /KgU 1.75 0 0.9270 0. (0.997) , 2.0 0.0025 o.9245 4.928 (4.920) 2.5 0.0060 0.9210 11.93 (12.06) 3.0 0.0092 0.9178 18.32 (18.48)- . 3.5 -0.0122 0.9148 24.41 (24.43) 4;O' O.0152 0.9118 30.36 (30.16)
-4 . 6 0.0185 0.9085 37.00 (37.11)
L 5.0 0.0208 0.9062 41.58 (42.00) (1) , See Section 4.4.2 8 (2) . See Table 4.4 (3) Parenthetical' values are calculated from the polynomial fit
, l for unrestricted storage in Region II.
4 - 33 - O O
O) k_ Table 4.7 EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF STORAGE RACK Case Incremental Reactivity Change, 6k l Region I Region II 20*C (68'F) Reference Reference 40*C (104*F) -0.003 -0.002 60 0 (140*F) -0.007 -0.005 90*C (194*F) -0.013 -0.010 120'C (248'F) -0.022 -0.016 120*C (248'F) + 20% void -0.083 -0.061 l 1 l L O l l l l : l l 4 - 34 i J dL
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- INmAL ENRICHMENT, WTis U-235 Fig 4.1 Relationship between Initial Enrichment and '
Acceptable Fuel Burnup (Region II) i. LO . 4- 5
+,
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_ _ _ . . _ _ _ _ . . . . - . . _ . - _ _ - _ - . _ _ . . _ . - U
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1 APPENDIY A l 1 l SENCHMARK CALCULATIONS
.O. av 1 ,, Stanley E. Turner, PhD, PE 1
(, HOLTEC INTERNATIONAL s t l l i
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4 m A-1
4 O
1.0 INTRODUCTION
AND
SUMMARY
l The objective of this benchmarking study is to verify both the AMPI (NITAWL)-KENO IV (Refs. 1 and 2) methodology'with the.27-group SCALE cross-section library (Ref. 3 and 4) and the CASHO-2E code (refs. 5,6,7, and 8) for use in criticality l saf ety calculations of high density spent fuel storage racks. Both calculational methods are based upon transport theory and have been benchmarked against critical experiments that slaulate typical spent fuel storage rack designs as realisti-cally as possible. Results of these benchmark calculations with both methodologies are consistent with corresponding calculations reported in the literature. Results of these benchmark calculations show that the 27-group (SCALE) AMPI-KENO calculations consistently under-predict the critical eigenvalue by 0.0106
- 0.0048 ok (with a
() 95% probability at a 95% confidence level) for critical experiments (Ref. 9) selected to be representative of realistic spent- fuel storage rack configurations and poison worths. Slailar calculations by Westinghouse (Ref. 11) suggest a bias of 0.0120 2 0.0023, and the results of ORNL analyses of.54 relatively " clean" critical experiments (Ref. 12) show a bias of 0.0100 a 0.0013. similar calculations with CASH 0-2E for clean critical experiments resulted in a bias of 0.0010 with an uncertainty of 2 0.0018 (95%/95%). , CASMO-2E and AMPI-KENO intercomparison calculations of infinite arrays of poisoned cell configurations show. very good agreement and suggest that a bias of 0.0013
- 0.0018 -is the reasonably expected bias and uncertainty for CASMO-2E calculations. ,
(:) x-2 - (A e
,.-r ..
. ...o.. -
t a U The benchmark calculations reported here indicate that either the 27-group (SCALE) AMPI-KENO or CASM0-2E calcula-tions are acceptable for criticality analysis of high-density spent fuel storage racks. Reference calculations for the rack designs should be performed with both code packages to provide l independent verification. 1 l O i b e 4 eme e v A-3
.~ ._. _ ._ . _ _ _ _ _
O 2.0 AMPY INITAWLi-KENO IV BENCHMARK CALCULATION 1 Analysis of a series of Babcock & Wilco:c critical experiments (Ref. 9), which include some with absorbe;! panels typical of a poisoned spent fuel rack, is summarized i'A Table 1, as calculated with AMPX-KENO using the 27-group 80 ALE cross-section library and the Nordheia resonance integral treatment in NITAWL. The mean for these calculations is 0.9894
, t 0.0019, conservatively assuming the larger standard devia-tion calculated from the k. : '
values. With a one-sided tolerance factor corresponding to 95) probability at a 95% confidence level (Ref. 10), the calculational bias is + 0.0106 with an uncertainty of 2 0.0048. Stailar c11culational deviations reported by Westin- ! ghouse (Ref. 11) are also shown in Table 1 and suggest a bias of 0.0120 2 0.0023 (95%/95%). In addition, ORNL (Ref. 12) has p
} analy:ed some 54 critical logy,. obtaining a mean experiments using the same methodo-bias ** o f 0. 0100
- 0. 0013 ( 9 5 % / 9 5 % ) .
1 .
.These published results are in good agreement with the results j
obtained in the present analysis and lend further credence to ' i tha validity of the 27-group AMPI-KENO calculational model for t use in criticality analysis of high density spent fuel storage racks. Variance analysis of the data in Table i suggest the , possibility that an unknown. factor may be causing a slightly
, larger variance than might be expected f on Monte Carlo .
statistics alone. However, such a factor, if one truly exists, is too small to be resolved on the basis of the critical experiment data presently available. No trends in k st with intra-assembly water gap, with absorber panel reactivity worth, or with, poison concentration were identified. '
'Significantly large trends in ko r t. with water gap and with absorber panel reactivity worth have been reportedtel for AMPI-KENO calculations with the 123-group GAM-THERMOS library.
O ^-* t
- - . - - . . . ~-. -.
l
\'
- 3. CASMO-2E BENCHMARK CALCULATIONS 3.1 OENERAL The CASHO-2E code is a multigroup transport theory code ut111 ring transmission probabilities to accomplish two-dimen-sional calculations of . activity and depletion for BWR and PWR fuel assemblies. As such, CASH 0-2E is well-suited to the criticality analysis of spent fuel storage racks, since general practice is to treat the racks as an infinite medium of storage cells, neglecting leakage effects.
CASHO-2E is closely analogous to the EPRI-CPM code (ref.
- 13) and has been extensively benchmarked against hot and cold critical experiments by Studsvik Energiteknik (Refs. 5, 6, 7 '
and 8). Reported analyses of 26 critical experiments indicate i a mean k.tr of 1.0000 s 0.0037 (la). Yankee Atomic (Ref. 14) has also re-ported results of extensive benchmark calculations () with CASHo-2E. Their analyses of 54 Strawbridge and Barry : critical experiments (Ref. 15) using the reported buck 11ngs indicate a mean of 0.9987 r 0.0009 (la), or a bias of 0.0013 s 0.0018 .(with 95% probability a.c a '951 confidence level). Calculations were repeated for seven of unti Strawbridge and Barry experiments selected at randon, yielding a mean -k.:e of O.9987 2 0.0021-(la), thereby confirming that this crors section library and analytical methodology being used for the.present
- calculations are the same as those used in the Yankee analyses.
Thus,.the expected bias for CASH 0-2E in the analysis of " clean" critical experiments is 0.0013 s 0.0018 (954/954). l l ., i i I [. , A-5 l
/}
I l
- 1. . - , - . . . - - . - - , -
4 O 3.2 BENCHMARK CALCULATIONS CASHO-2E benchmark calculations have also been made for the B&W series of critical experiments with absorber panels simulating high density spent fuel storage racks. However, ! CASH 0-2E. a f, an assembly code, cannot directly represent an entire core configuration
- without introducing uncertainty due to reflector constants and the appropriateness of their spectral weighting. For this reason, the poisoned cell configurations of the central assembly, as calculated by CASHO- J 2E, were benchmarked against corresponding calculations with the 27-group (SCALE) AMPX-KEN 0 IV code package. Results of this comparison are shown in Table 2. Since the differences are well within the normal KINO statistical variation, these calculations confirm the validity of CASHO-2E calculations for the typical high density poisoned spent fuel rack cenfigura-tions. The differences shown in Table 2 are also consistent '
with a bias of 0.0013
- O.0018, determined in Section 3.1 as the expected bias and uncertainty of CASHO-2E calculations.
f n
' Yankee has attempted such calculaH ans(103 -using CASMO-2E ,
- generated constants in a two-dimensions our-group PDQ aodel, obtaining a- mean k rs of 1.005 for 11 oned cases and 1.009 for 5 unpoisoned cases. Thus, Yankt nchmark calculations suggest that' CASH 0-2E tends to'slightly v,erpredict reactivity. O , A-6
- -. ~ ~ .-. . - - - . . - _ - - - .
3' i O-REFERENCES TO APPENDIX A
- 1. Green, Lucious, Petrie, Ford, White, and Wright, "PSR=63-
/AMPI-1 (code package) AMPI Hodular . Code Systen For Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B", ORNL-TH-3706, Oak Ridge National Laboratory, i November 1975.
- 2. L.H. Petrie and N.F. Cross, " Keno-IV. An Improved Monte Carlo Criticality Program", ORNL-4938, Oak Ridge National Laboratory, NovemDer 1975.
I
- 3. R.H. . Westfall et. al., " SCALE: A Modular System for i Performing Standardized Computer Analysis for Licensing l Evaluation", NUREG/CR-0200, 1979.
)
- 4. W. E. Ford,III, et al., " A 218-Neutron Group Haster Cross-section Library for Criticality Safety Studies", ORNL/TM-4, 1976
- 5. A. Ahlin, H. Edenius, and H. Haggblom, "CASHO - A Fuel
(} Assembly Burnup Program", AE-RF-76-4158, Studsvik report.
- 6. A. Rhlin and H. Edenius, "CASHO - A Fast Transport Theory Depletion Code f or LWR Analysis", ANS Transactions, Vol.
26, p. 604, 1977. i
- 7. H. Edenius et al., "CASHO Benchmark Report", Studsvik/RF-78/6293,- Aktiebolaget Atomenergi, March 1978
- 8. "CASHo-2E Nuclear Fuel Assembly Analysis, Applications Users Manual", Rev. A, Control Data- Corporation,1982
- 9. H.N. Baldwin et al., " Critical Experiments Supporting i Close Proximity Water Storage of Power Reactor Fuel", BAW-1484-7, The Babcock & Wilcox Co. , July 1979.
u
- 10. H.G.- Natrella,' Exnerimental Statisties, National ~ Bureau of Standards, Handbook 91, August 1963.
1 1. - B.F. Cooney et al., " Comparisons of Experiments and L , calculations for LWR Storage Geometries", Westinghouse NES, ANS Transactions, Vol. 39, p. 531, November 1981. l-i) i A- 7 i I
I i
- il l
l[) . REFERENCES TO APPENDIX A (Continued) ; l
- 12. R.W. Westfall and J. h. Knight, " SCALE System Cross-section Validation with 3 hipping-eask Critical Experi-ments", ANS Transactions, Vol. 33, p. 368, November 1979
- 13. "The EPRI-CPM Data Library", ARMP Comeute r code Manuals, Part II, Chapter 4, CCH3, Electric Power Research Institute, November 1975 ,
- 14. E.E. Pilat, "Hethods for the Analysis of Boiling Water Reactors (Lattice Physics), YAEC-1232, Yankee Atomic Electric Co., December 1980,
- 15. L.E. Strawbridge and R.F. Barry, " Criticality Calculations for Uniform, Water-moderated Lattices", Nuclear Setence and Enoineerino, Vol. 23, p. 58. September 1965.
- 16. S.E. Turner and M.K. Gurley, " Evaluation of AMPI-KENO Bench = ark Calculations for High Density Spent Fuel Storage Racks", Nuclear Scionee and Enoineerino, GC 80(2):230-237,. rebruary 1982.
.r e 6 l'
- l. b l
t l-L j.- ' r O A-8
._ ._ - ~ . - . . _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
.?
Table 1 RESULTS OF 27-GROUP (SCALE) AMPI-KENO IV CALCULATIONS OF B&W CRITICAL EXPERIMENTS Experiment Calculated a Westinghouse Number k.ft 6 k.tr
! O.9889 2 0.0049 -0.008 II 1.0040 2 0.0037 -0.012 III 0.9985 2 0.0046 -0.008 II 0.9924 2 0.0046 -0.016 ;
-I 0.9907 2 0.0039 -0.008 II O.9989 2 0.0044 -0.002 III 0.9932 2 0.0046 -0.013 IIII 0.9890 t 0.0054 -0.007 d+. IIV 0.9830 2 0.0038
- ... -0.013
. IV , 0.9852 t 0.0044 -0.016 i
IVI 0.9875 210.0042 --0. 015 - IVII .0'9811 2 0.0041 .-0.015 w , IVIII 0.*784 t 0.0050, -0.015
'III' O.9888 2 0.0033 -0.016
.II. G.9922: 2 0.0048 -0.011 III 0'.9783 2 0.0039 -0.017 Mean- ,0.9894 . t 0.0011(1) -0.0120 t 0.0010 Bias. 0.0106
- 0.0019(2) 0.0120;2 0.0010 Bias.(95%/95%). 0.0106 2.0.0048 0.0120 2 0.0023
'A I
;UE
.t i l
~t2 calculated from individual standard deviations. i Calculated from k.tr values and used as reference.
d A-9 I .
~. -
u
.p
. 1 i f
i i Table 2 RESULTS OF CASHO-2E BENCHMARK (INTERCOMPARISON) CALCULATIONS k.(1) B&W Experiment (13- AMPI-KENO IV(2) CASHO-2E Ek i III 1.1203 2 0.0032 1.1193 0.0010 i IVII 1.1149
- O.0039 1.1129 0.0020 IV 1.1059 2 0.0038 1.1052 0.0007 1
Interpolatedt23 1.1024 2 0.0042 1.1011 0.0013 l
.f IIV, 1.0983 t 0.0041 1.0979 0.0004 "IIII 1.0992 2 0.0034 1.0979 0.0013 - i Hean 1 0.~0038 0.0011 -!
Uncertainty t 0.0006
~ Typical BWR fuel -0.9212 2 0.0027' O 3218: <- 0.0006-
-rack 1
(2). Infinite array of central assemblies of: the 9-assembly V&W. - critical. configuration (Ref. 9). 'l (23
- k. from AMPX-KENO corrected for bias of 0.0106-ok. .
- t3)-. Interpolated froa 7 Fig. 28 of Ref. 9 for soluble boron-
\
concentration:at the. critical condition.' , ; h A - le n
"~
i l l 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction I A primary objective. in the design of the high density spent fuel storage racks for the-TMI-I spent-fuel pool is to ensure adequate : I
. cooling of the-fuel assembly cladding. In the following section, a brief synopsis of the design basis, the method of analysis, and the' numerical results-is provided.
Similar methods of thermal-hydraulic analysis have been used in previousi licensing efforts .on high ' density spent fuel racks for I Fermi 2'(Docket-.50-341), Quad Cities'.1 and 2 (Dockets.50-254 and= l 50-265), Rancho Seco-(Docket 50-312),. Grand Gulf Unit 1 (Docket ; 50-416 ) ~, . Oyster' Creek -(Docket . 50-219),.-Virgil C. Summer'(Docket-50-395 ) , .Diablo ' Canyon ' 1 and 2 (Docket Nos. 50-275 and 50-323),: ! Byron Units 1 and: .2- (Docket 50'-454 r 455 ) , St.- Lucie Unit. One
'% (Docket . 50-335 ) , - Millstone Point I~(50-245), Vogtle Unit..2 (50-
,425), Kuosheng. Units 1L&.2..(Taiwan PoweriCompany), Ulchin ' Unit.- 2 f
(Korea Electric Power Company) , 'and J. A'. FitzPatrick -(New York ! Power' Authority)'. The- analyses to!'be carried! out for the thermal-hydraulic !
.qualificationiof. the rack array may be broken down - into the-following categories: ,
.(1). 1 Pool? ' decay. ' heat evaluation and pool bulk- f temperature variation with time.- l
. i
./
5-1 _ _ _ _ . - .r .--___ . - - . -me - - - < - - - -
u l fl ud (ii) Determination of the maximum pool local temperature at the instant when the bulk temperature reaches-its maximum value. 1 (iii) Evaluation of the maximum fuel cladding temperature l to establish that bulk nucleate boiling at any location resulting in two _ phase conditions '
. environment around the fuel is not possible. 1 (iv) Evaluation of the time-to-boil if all heat ,
rejection paths from the cooler are lost. Compute the effect of a blocked fuel cell opening (v) on the local water and maximum cladding s temperature. The following sections present a synopsis of the methods employed to perform'such analyses and final results. t s
- 5.2 SPENT FUEL COOLING SYSTEM DESCRIPTION '
4- The-principal ~ functions >of the Spent Fuel, Cooling System are the removal of aecay heat from~the spent. fuel stored in the pools.it-h serves and maintain 1.ng the clarity of, and a low activity _ level in"
- the water of the pools.
Cleanup of: pool water is accomplished by ( diverting:partiof the flow,Lmaintained-for removal of decay-heat, through filters and/or demineralizers of the' Liquid Waste Dispos'al l System as. described in Section 11.2.1 of FSAR.- 1 .
- 512.11 System Functions- I
; Ina addition .to its' principal functions: of circulating ' spent fuel .
pool water for decay, heat removal,.the equipment of1 he t Spent Fuel _ i A
' coolingLSystem isfdesigned to. fulfill the followingt functions:- f Q l a.. . Transfer- borated- refueling . water -from the borated water *
; storage 1 tank to thetfuel transfer, canal.
t
- , .O
*x
+ . .
d D'
l 1 j 1
- b. Transfer water from the ' fuel transfer canal to the borated
]A- . water storage tank.
- c. Circulate refueling water through cleanup equipment.
p -
- 1) During transfer from the fuel transfer canal to the borated water storage tank, or ,
t
- 2) During storage in the borated water storage tank.
- d. Circulate fuel- transfer canal water through cleanup equipment. e e.. " Skim" pool water. surfaces for the removal of any floating debris.
t i Draw down. and replace ' the water in the spent fuel shipping f.- cask pit to prevent " dunking"' of the spent fuel building
- crane > hook and-cables.
5 ^. 2 2 2 - System Descriotion Spent ~ fuel 'isJ cooled by pumping spent fuel storage poni watei
- through: coolers and back to the spent fuel storage po.W . Either ~
. (~y ' " of-. the two spent fuel cooling pumps and either of the-two spent i
fuel coolers .'may - be aligned-to cool both-spent fuel. pool A ' and-spent-fuel pool B during: normal refueling operation or,to transfer Y g , refueling water'in either direction.- Both pumps and coolers;will I belusedtoiremovedecayheatfromspentfuelstoredinspentfuel ~
-pool ~ A and ' ' spent- fuell Pool B, ~ if ' required, ' and at .the - time of-entire cors? offload.- >
4 4 h
?
-9
- . - - . .~- .- . -- . . - . . - . . . - .
s
,' l
.i l
r ;
.] The following supplemental-means for cooling the spent fuel pools ]
4 may be used in addition to the spent fuel pool cooling system:
,o 1. The Decay Heat Removal System, f' ,. 2. The Forced Ventilation System (to improve the cooling effects of pool surface evaporation). j
- 3. Reclaimed -water (for pool water makeup as well as for its cooling effects).
.The' borated water = recirculation pump is used to accomplish water circulation from either spent fuel pool, the fuel transfer canal, or the borated water . storage tank for. cleanup or " skimming" !
~
functions. It -is also used' to empty - spent fuel pool A, if required, and to: lower and raise the water level in the spent fuel cask pit as required'for the placement,-loading, and removal of a the spent: fuel shipping cask. l.. e 5.2.3 Performance Reauirements n.
~" O .
i The first -design basis of the' system :is based on .the normal $ crefueling: operation with approximately 80 assemblies being removed-l! . from'the ~ unit each time.- ' The removed fuel assemblies will have J. - been in the reactor for three cycles,.at the timeLof discharge. C The.. second design - basis for the system ' considers that it' is possible to unload the reactor- vessel -for : maintenance or 4 inspection atra time wh'en a maximum of 1640 spent fuel assemblie's , are already residing.in the spent fuel pools. ! J ,,
'i s' -
> I l
l, 5-4
. . ___ _ __l
p .i j l 1 5.2.4 Methods of Ooeration I Spent fuel cooling functions are monitored and controlled from the l Main' Control Room. All other functions of the Spent Fuel Cooling ; I System are accomplished by local manipulation of valves and l control of equipment. However, after a piping lineup for filling or emptying the fuel transfer canal (via a spent fuel cooling pump) has been set up, the transfer may be monitored and controlled from the Main Control Room. 5.2.5 Leakaae Considerations Whenever.m leaking fuel assembly is transferred from the fuel transfer canal to the spent fuel storage pool, a small quantity of , fission products may enter the spent fuel cooling water. A purificatioc loop is - provided within the Liquid Waste Disposal System for removing-these fission products and other contaminants g V from the water. A small quantity of flow from the spent fuel cooling pumps is diverted to a radiation monitor. This provides 1 monitoring of radiation levels in the spent fuel pool water to
' indicate,when cleanup.should be initiated.
The-fuel-handling and storage. area housing the spent fuel storage
- pool- is ventilated on a controlled basis, normally. exhausting .
circulated air to the outside through - the unit vent. The fuel j Handling Building ; includes a modification to physically separate the1 fuel handling area- from .the Auxiliary J Building and the Fuel l Buildingi accessway. In addition, -a separate _ ESF ventilation system is l'n' service to support fuel movement. The Fuel Handling Building and Auxiliary Building exhaust duct ventilation monitors .; are described in Section.11.4.3 of the FSAR. 5-5 h (a
hf Provisions have been made to air-test the valved and flanged ends of each fuel transfer tube for leak-tightness after it has been used. .A valve (on the Fuel Handling Building side) and blind flange (on the Reactor Building side) are used to isolate each fuel transfer tube. 5.3 DECAY HEAT LOAD CALCULATIONS The decay . heat load calculation is performed in accordance with ; the provisions of "USNRC Branch Technical Position ASB9-2,
" Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981. For purposes of this licensing _
l application, it is assumed that the pool contains an inventory.of 1640 assemblies accumulated through scheduled discharges from 1976' L to 2019 (Table 1.1). Further,.since the decay heat load increases
- l. monotonically with reactor exposure time, an upper bound of 1460
~
o full power - operation ' days is assumed for all stored fuel. The l (mj' cumulative decay heat load is computed for the instance of L hypothetical normal discharge #23 in the year 2021. As shown in Table 5.4.1, the ratio of this decay heat load due to the inventory- of previously stored' fuel to'.the average assembly operating. power- is 0.1176. p - This . decay heat load is assumed to remain . invariant for the duration- of the- pool temperature evaluations of- the pool 1-. temperature evaluations performed in the wake of-normal-and full core offloads discussed below. L l I. l l 5-6 C-
o s 5.4 DISCHARGE SCEFMIOS The following discharge scenarios are examined: l Case-1: ,ermal discharge A normal batch of 80 assemblies with 6 years of reactor exposure time at full power is discharged in the pool at the end of a normal two year operating -cycle. There ara 22 previously discharged batches in the pool. The normal discharge occurs at tne rate of 6 assemblies per hour after , 150 hours of: decay in the reactor. One fuel pool cooler is active and running. Normal discharge 1 Case 2: Same as~ Case 1 except two fuel pool coolers are operating. Case 3: This case differs from_the prsceding two cases only. in the manner of deployment of_ fuel pool' coolers. It E is
. assumed that-at the~' commencement of-discharge to the pool two.
coolers are working. -After a certain length of time, one cooler.is.taken out of service. The amount'of time needed
- before the second . cooler is withdrawn from : service -is : set - -
such that the peak pool bulk waterntemperature remains below ' 160'F.
-Case 43 Full; core offlo'ad'(BOC) i The? so-called; Beginning-of-cycle (BOC)=. full, core i
. offload condition corresponds .to ,the Lemergency reactor offload condition wherein a full core offload occurs after-36' days;of reactor operation. Two : coolers ; are operating- inL parallel.
Reactor' decay time and rate.of discharge are the'same in all-L cases. Case 53 Full ~ core offload-(EOC) i
.i The end-of-cycle (EOC), full core _ offload Lis the scheduled i offload- ' scenario. ~After two years' of normal. reactor operation, the' full core-is offloaded to the pool.,
Additional and 5.4.3. data on the five ' cases ' described in Tables 5.4.2 , l 5-7 ,
' I.
t l h u 5.5 ByLK POOL TEMPERATURES A ntun'aer of simplifying assumptions were made which render the analysis conservative. These include: o The heat exchangers are assumed to be fouled to their design maximum. Thus, the temperature effectiveness, p, for the heat exchanger utilized in the analysis is the- ; lowest postulated value calculated from heat exchanger ' thermal hydraulic codes. p ' is assumed constant in the calculation. o As of now,.no heat exchanger. tubes in the coolers have had to be plugged. However, out of a total of 328 tubes in the cooler, 80 are assumed to have been plugged -in this analysis. Thus, the -temperatures reported herein-are. considerably greater than the actual values until 80 7 tubes are-plugged in each cooler. The mathematical? formulation can be explained with reference t.o ' i the simplified" heat- exchanger alignment of Figure 5.5.1. A:-Q.A. validated' heat. exchanger . . thermal rating code is. used to H
-].-
' calculate the temperature effectiveness p of the cooler. . This is.
. done:by simulating the-geometrie details of the-cooler.-in Holtec's o
- proprietary- . heat = ' exchanger rating computer code. Figure 5.5.2 !
shows: 'the . variation - of. p .with the number of tubes 1 plugged. L o
' Referring lto the , spent fuel i pool / cooler - system, the . governing -
differential equation can be written.byLutilizing conservation'of .; I; energy:. , 1. u p dT ' l C = QL - QHX- (5-1) O dr t lQL = Pcons + Q (I) - QEV (T, ta) i . r where: r I C: Thermal capacitance of t.'u.a pool l h 5-8 ; L l N +
QL t - Heat load to the heat exchanger Q(t): Heat generation rate from recently discharged fuel, which is a specified 1' function of time, t. Pcons = Pos Heat generation rate from "old" fuel. QHX: Heat removal rate by the heat exchanger. QEV-(T,ta): Heat loss to the surroundings, which is a function of pool temperature T and ambient temperature ta ' QHX is a- non-linear function of time if we assume the temperature effectiveness p is constant during the , calculation. QHX- can, however, be written in terms' of effectiveness p as follows: QHX = Wt - Ct P (T - ti) - ( 5-2 )_ where: Wt t Coolant. flow rate Ct Coolant specific heat: b
'pt. Temperature effectiveness' T Pool-_ water; temperature, g , tit Coolant = inlet temperature- , ;
We recall that p has been obtained from the heat exchanger: rating e program.Q(t) is'.specified according to the-provisions of'"USNRC. s
- Branch Technical Position =ASB9-2, " Residual Decay Energy for Light Water- Reactors for . Long . Term Cooling", ' Rev. 2, .. July, 1981. ' QgI)-
is a function of; time, number'of assemblies' and operating. time.- , t' During the fuel transfer, the heat' load in the pool will increase ; h , with= respect to rheirate=of fuel transfer and equals to Q(r).-after the fuel ~ transfer.. 1 5-9 1 i i {.. f
_. - -_ . . - . . - . - . - . ~ . . . - . . . - . . - . . . - . -
'l O QEV is a non-linear function of pool temperature and ambient tempere.ture. QEV contains the heat evaporation loss through the pool surface, natural' convection from the pool surface and heat conduction through the pool walls and slab. Experiments show that the heat conduction takes only about 4% of the total heat loss (5.5.1], therefore, can be neglected. The evaporation heat and natura convection heat loss can be expressed as:
QEV = m A As + hc As 6 (5-3) where: m: Mass evaporation rate h: Latent heat of pool water ;
.A, Pool surface area he: Convection heat transfer coefficient at pool .i surface
..h 6 = T-ta . T he, temperature - dif ference between pool water and.
ambient air. g a l- The mass evaporation rate m can be obtained as a nonklinear-function of 6. We, therefore, have ;
- i. m = ho.(6)-(Wp
_ - Was)' (5-4}- where: b Wps: Humidity ratio of saturated moist. air at pool water surface temperature T. 1-Wast Humidity ratio of saturated moist air at ambient. temperature ta
.ho(6): Diffusion coefficient at pool water surface. up is a non-linear function of'6.
'O 5-10
t The non'-linear single order differential equation (5-1) is solved using Holtec's Q.A. validated numerical integration code "ONEPOOL". This equation is solved as an initial value problem by noting that the cooler heat removal rate must equal the heat generation rate from previously discharged assemblies. Hence: , Weool P (Tin - teool) = PCONS where the parameters are as follows: PCONS: Heat generation rate from previously stored assemblies Weool: Coolant thermal flow rate P: Temperature . effectivecess of the fuel pool ; cooler. r- Tint. Coincident pool water temperature (initial
. (>3 value before beginning of discharge) tcool Coolant inlet temperature
'The above. equation yields:-
PCONS Tin > "
. + teool-Weool P The value of Tin computed- from - the. above formula is. the initial value of the pool water ' temperature -(at the start of fuel o discharge)'.
l
-l 1
5-11
- O p
?O Figures 5.5.3 through 5.5.7 provide the bulk pool temperature profiles for the normal discharge, and full core offload scenarios l respectively. Table 5.5.1 gives the peak water temperature,
' coincident time, and coincident heat load to the cooler and coincident heat loss to the ambient for five cases. Figure 5.5.8 shows the variance of the pool maximum temperature for case 1 versus the number of tubes plugged in the cooler. There is one cooler in operation for case 1. The maximum pool bulk temperature, Tb ulk of 160'F in the TMI-1 pool is quite - consistent with the values in other PWR plants licensed in recent years. For example, Tb ulk for Indian Point Unit 2 (Docket 50-247) is'180'F (licensed in April, 1990) and that for Diablo Canyon Units 1 & 2 (Docket Nos. 50-275 and 50-323) is 188'F for normal discharge. Diablo Canyon was licensed in 1986, and relicensed (after ASLB hearings) in 1988. The next step in the analy' sis is to determine the temperature. rise ! profile of the pool water if all forced indirect cooling modes are suddenly-lost. Make-up water is provided with a fire hose. Clearly, the most critical instant of loss of cooling is when pool' water hascreached its maximum value. It is assumed that cooling
/~ .. water is added through a fire hose at the rate of G lb/hr. The ]- - cooling water is at temperature, tcool. The governing enthalpy balance' equation for this condition can be written as F
dT ( C +LG (r - to)) = Pcons + Q (t += tins + G tcool)~- QEV dr , t where-- water is - assumed to '_ have specific heat of unity, and the time coordinate t.is measured from the instant-maximum. pool water temperature is' reached. to is the time coordinate when.the direct addition (fire hose) cooling' water application is begun. rins is the time coordinate measured from the instant of reactor shutdown r to'when maximum pool water temperature is-reached. T is the P- 5-12
-O
._ ._ _ . _ . . ~ . _ _ _ _ _ _ ._ ._ _
t I l l dependent variable (pool water temperature). For conservativeness, QEV is assumed constant af ter pool water temperature reaches l 170*F. I i A Q. A. validated numerical quadrature code is used to integrate the foregoing equation. The pool water heat up rate, time-to-boil,'and subsequent water evaporation-time profile are generated and compiled-for safety evaluation. The ' time-to-boil output results are presented in Table 5.5.2. Figures 5.5.9 through 5.5.13 show the plot of the inventory of i water .in the- pool af ter loss-of-coolant-to-the-pool condition begins. a
'5. 6 Local Pool Water Temoerature
,. I6 this section,: ' a . . summary of the methodology, calculations and-
~
results for local pool' water temperature is presented.
)
i 5. 6. lI Basis-p lIn order.to determine an upper bound.on the maximum fuel cladding-
. temperature, a-series of conservative assumptions are7made. The most important. assumptions are listed below:
0 The fuel-pool'will contain spent fuel with varying time- -i after-shutdown (ts). Since the-heat emission fallstoff rapidly with increasing ts, it is conservative to assume that all fuel - assemblies are _ from the latest Lbatch discharged simultaneously'~in-the shortest possible time and- they all have had the maximum. postulated years -of. ' operating time in the reactor. The heat emission rate- ~ of .each fuel assembly -is assumed : to be - equal and: maximum.
v
, f~/
O As snown'in the pool layout drawings, the modules occupy an irregular floor space in the pool. For the
, hydrothermal analysis, a circle circumscribing the -
actual rack floor space is drawn (Fig. 5.6.1). It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal layout pitch. O The actual downcomer space around the rack medule group varies. The nominal downcomer gap available in the pool is assumed to be the total gap available around the ideali::ed cylindrical. rack; thus, the maximum resistance
.to -downward flow is incorporated into. the analysis ,
4 (Figs. 5.6.2 and 5.6.3) (i.e. minimum - gap between the pool . wall and rack module, including seismic kinematic effect). O No downcomer flow is assumed to exist between the rack modules. 7 O lNo heat transfer is assumed to occur between pool water and the= surroundings (wall, etc.)
- 5. 6. 2 ' Model Descriotion In this manner, - a -conservative : idealized - model for the rack
.-. assemblage is ' obtained. - The water .. flow is axisymmetrie about the vertical 1 axis ob the circular rack assemblage, and thus', the flow; -
is two-dimensiona.' (axisymmetric'. three-dimensional) . Fig.i 5. 6.2 shows a. typical " flow chimneya rendering of the'thermallhydraulics. model'. The governing equation-to characterize the flowifield.in the. pool'caninow.be written. .The resulting integral' equation can be solved for . the J lower E plenum velocity- field (in the-' radial direction) andi axial velocity, (in-cell velocity: field), by-using d the - method ; of : collocation. The hydrodynamic. loss coefficients:
.which enter into the-formulation:of the integral equation arelalso
-taken 'from -well-recognized sources (Ref. 5.6.1) and .wherever.
< discrepancies lin: reported values exist, - the conservative values' l
>5-14
'D.
V, 1
.y* v- et
are consistently used. Reference 5.6.2 gives the details of mathematical analysis used in this solution process. After the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature. The knowledge of the overall flow field enables pinpointing of the storage location with the minimum axial flow (i.e, maximum water outlet temperatures). This is called the most " choked" location. In order to find an upper bound on the temperature in a typical ; cell, it is assumed that it is located at the most choked ! location. Knowing the glob-1 plenum velocity field, the revised axial flow through this choked cell can be calculated by solving the Bernoulli's equation for the flow circuit through this cell. Thus, an absolute upper bound on the water exit temperature and , maximum fuel cladding temperature is obtained. In view of these
. aforementioned assumptions, the temperatures - calculated in this manner' overestimate the temperature rise that will actually occur in-the' pool. Holtec's earlier computer code THERPOOL*, based on the theory of Ref. 5.6.2, automates this calculation. The analysis' procedure embodied in THERPOOL has been accepted by the -Nuclear-iRegulatory- Commission _on several dockets. The ' Code THERPOOL'for local temperature analyses includes the calculation THERPOOL has buen us'ed in qualifying the spent fuel pools for Enrico Fermi Unit 2 (1980),. Quad Cities I and II - (1981), .
Oyster: Creek (1984), V.C. Summer (1984), Rancho Seco (1983), Grand Gulf I (1985), Diablo Canyon I and II (1986), among others. Ii 5-15 oe
. . . -. - - . . . . . - .- -~ .. -. ..- - .-
r t. m f
. - of void generations. The effect of void on the conservation equation, crud layer in the clad, flux trap temperature due to s gamma heating, and the clad stress calculation when a void exists, are all incorporated in THERPOOL. The peaking factors are given in Table 5.6.1.
5.7 Claddina Temeerature t
- The maximum. specific power of a fuel array qA can be given by: :,
qA = q Fxy (1) where: i Fxy =. radial peaking factor q '= average fuel assembly. specific-power LTheL maximum: temperature rise of- pool- water in the most ,. ., . disadvantageously placed fuel assembly is computed for.all loading-- -,
?p Q cases'. Having' determined the maximum local water-temperature in the- pool, it .is: now' possible to determine - the maximum fuel g cladding _ temperature. .A-fuel rod can produce Fg. times the' average-b,l 1 heat emission' rate over alsmall length,;where Fz is the axial rod-L peaking-factor.eThe-axial heat distribution.in a rod is generally 4 a maximum in the l central region, and tapers of f . at its two l
extremities. lx' . g L :Itfcan- be shown that' the power ' distribution ' corresponding 'to the : chopped cosine power emission rate'is'given by L
' t rr ( a - + : x ) - .
'q(x) --qA sin [
5-16 h- \
g where: 11 active fuel length at- chopped length at both extremities in the power curve
, x: axial coordinate with origin at the bottom of the active fuel region The value of a is given by 1z ;
a= 1-2z where: , 1 1 1 2 z= o- - + n Fz- n2 F22 nF 2 n2 1 where:Fz i is~the axial peaking factor. The cladding temaerature Te 'is governed by a third' order ;
> differential equat;.on which has the form of e
i d3 T- d2 T. dT dlx3
+ at -
a2 = f-(x) dx-2 dx 1 where ai, . ' a2 - and f(x) are functions of . x, and fuel assembly-1 geometric -properties. < The solution of (this - dif ferential equation
- with ' appropriate boundary- conditions. provides . the fuel. cladding -
temperature and: local water temperature profile. 1 In - order to introduce some additional conservatism ~in .the analysis, we!. assume that the fuel cladding has a crudedeposit af - i
.005 0F -sq.ft.-hr/ Btu crud resistance, which -' covers the entiro surface.
5-17 i
Table 5.6.2 provides the key input data for local temperature analysis. The results of maximum local pool water temperature are presented in Table 5.7'.l. ! 5.8 Blocked Cell Analysis Calculations are also performed assuming that 50% of the top , opening in the thermally limiting storage cell is blocked due to a horizontally placed (misplaced) fuel assently. The corresponding , maximum local pool water temperature and local' fuel cladding temperature data are also presented in Table 5.7.1. 3 In all-cases, there is no incic ence of localized nucleate boiling of the pool water. 5.9; REFERENCES F% Cr'CTION 5
-5.5.1 Wang,- Yu, " Heat Loss to ' the Amelent From- Spent Fuel-Pools: Correlation of Theory with Experiment", Holtec- ,
Report HI-90477, Rev. O, April 3, 1990. 1 5.6.1 General Electric Corporation, R&D Data. Books, " Heat Transfer and Fluid Flow", 1974 and updates.. 5.6.2 Singh,- K.P. et al., " Method for Computing the Maximum-Water .Tempe' rat'ure- in- a^' Fuel-' Pool -Containing- Spent, Nuclear ~ Fuel", Heat Transfer Engineering,uVol. 7, No.-;1 - 2,1 pp. 72-82_(1986). s i I 5-18 L O- . i
.# _. , , u,. .
..-.#._.. _ - . . - .. _ # ~, ,., -. ..m-a-t
+
r Table 5.4.1 i t FUEL SPECIFIC POWER AND POOL CAPACITY DATA 1 Total water' volume ofLPool 635645 gallons ! Specific Operating Power of l
- a Fuel Assembly: 60.3E+06 Btu /hr. '
Dimensionless' decay power of "old" discharges: .1176 j i O V w i b k i L
. j
- a. j r
?
li 9 b,> < 1 b w I 5-19
.. . . .- . - ._..- - ~ - _ - - -. . ..
1
,.7..
s Table 5.4.2 t DATA FOR SCENARIO,S 1 through 3 ; t} CASE NO. 1 2 3 l-
~
Poolthgrmalcgpacity 4.241 4.241 4.241 C x 10" , Btu / F No. of Coolers 1 2 2 ; No. of. Discharges Considered for the Analysis 1 1 2 Time between Shutdowns,. hrs. -- -- 1680' Cooler Inlet Temp.,.OF 95 95 95. 1 Coolagt Flow Rate- 1.49 49 1.49 . x.10 ; lb/hr V a ( i t f 5-20 A. U
i n r i Table 5.4.3
. DATA-FOR SCENARIOS 1 THROUGH 5 i 1
Time After
' Shutdown when Offload . Expo. .
~c- Case _ Discharge. No. of -Transfer Begins Time- Time ID Assemblies (hrs) No. (hrs) (hrs) u,
^1 Discharge 1 80 150 '13.33 35040*- ;
2: . Discharge 1 80 150 13.33 35040 3 Discharge:1 ~80 ;150 13.33 35040
])
~ Discharge 1 80- 150' 13.33' 35040 '
4' 80' 864 Discharge 2' + ,'150? :29.5 ' 97~ 17520 a- g 80 . '35040 5 DischargeEl' + 150 29.55 97 17520
=a -
- Some assemblies -may ~have, in-core, exposure time of. 6 years.
However, the1effect offthis term is negligible. 2.-n 1
- 5-21 !
m s t
)! , +
Table 5.5.1 7 9e' BULK POOL TEMPERATURE RESULTS WHEN t CONSIDERING HEAT LOSSES TO THE AMBIENT ASSUMING 90 TUBES PLUGGED IN EACH COOLER ! c, Coincident F Heat Load Coincident Maximum Coincident to the Heat-Case Bulk Pool Time After Cogler, No. 10 Btu /hr Logses, Temp., 'F Shutdown, hr. 10 Btu /hr t
+ :(
1 .158.35 220 15.21 1.36 ' I 2 .129.73- 197 16.69 0.35 >J],[ 3 0 -)
.3: "
- 148.89' 473 12.94 0.89 ll '4: 156.05. -205; 29.31- 1.23-
.i' l' 5 '; '154.56 207 28.62 '1.15-l E
5-22 1 -
)
,i
41 Table 5.5.2 TIME-TO-BOIL RESULTS WHEN CONSIDERING HEAT LOSSES TO THE ENVIRONMENT Time -to-Boil, hr Case Number G = 0 GPM 1 .18.96 2 27.56 13' 27.16 I
-. 4 10.2:
5- 10.56-
-n i;
1 ) I
--o' 5-23 l
)
I s V l f} ,.
- l; .
)
Table 5.6.1 Peaking Factor Data A Radial Bundle Peaking Factor 1.7 'i Rod to bundle maximum power ratio' 1.065 Maximum axial power peaking. 1.57 factor Total' peaking; factor-(product of. , all~three above) 2.84 10
.w f
i, b . 1 j. n l L-L l: l' ! 1 . 5-24 , s. i:
1 44 - l 1
' l
[,
.r l m
)
l 1 b Table 5.6.2 l Data for Local Temperature l Type of Fuel Assembly PWR (15x15) u-
, ' Fuel' Cladding Outer Diameter, inches 0.43 g ,
Fuel-Cladding Inside Diameter, inches. 0.377 l
-t Storage Cell- inside-Dimension, inches 9.0 l
[: LActive-fuel length, inches 141.8
.-y q
s
, %_A No.:of Fuel-Rods / Assembly. 208. i
,[ fOperating' Power ser_ Fuel; Assembly. 49.53. , f
'Po'.x'-6, Btu /hr Cell pitch, inches. 9.2 yu
. Cell; height, inches. -
163. , w li '; [ ' - .\. R , [, Plet m. radius,' feat ~20.5_ ; yw
- j. 7.: Bottom hei.ght, ' inches -- 4~.25
.!o
' Min.: gap between. pool. wall-
- and outer-rack periphery, inches 2 1
,o . r i 1
c' 4 ,g
,*s 4 6 \ (. kyc q, '
" O' 25 m ,
s l_ l'
, + , ~ . - .
- 3. ,
O :
':I r
',' Table 5.7 1 Local and Cladding Temperature Output Data for the~ Maximum Pool Water Condition (Case 1)
. +- Maximum Loca Maxin Cladding WaterTemp,gF bz
. Condition Temp,gF ,
L No' Blockage 220.7 251.9 i 50%' Blockage: 231.7 -260.2 L 1 1 Ii. 4 [- I.- l
}
t .e , b j l i f ,.;' f
,1 0- 5-2e 1
i, l i e O > l j i 4 SPENT FUEL W,' T POOL W,C#
- C,. T "
l
- l 4 O y W,T W,9 i
l l
.t' wt[ ,
FIGURE 5.5.1 Pool Bulk Temperature Model for Normal Discharge Scenario 5-27
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Titd ,$FTER REACTOR SHafrOOWii. HR [ t
- .. Figure 5.5.7. . BULK', POOL' TEMPERATURE FOR' CASE 5 [
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10EALIZATION OF RACK ASSEMBLY 5-40 FIGURE 5.6.1
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.',*A',,,* 1'.*, t t , y,- -A4, Q. . 6. .,; t '. . .,s . .*. ., . c . r,; ; r . .;;. * , . . 6j PLENUM CONVECTION CUR RENTS W WE POOL 5-42 FIGURE 5.6.3 0 -. . ._
i l O l 6.0 RACK STRUCTURAL CONSIDERATIONS l The purpose of this section is to demonstrate the structural l adequacy of the TMI Unit i spent fuel rack design under normal and accident loading conditions following the guidelines of the USNRC Standard Review Plant (Ref. 6.0.1). The method of analysis presented uses a time-history integration method similar to that previously used in the licensing reports on high density spent fuel racks for Enrico Fermi Unit 2 (USNRC Docket No. 50-341), Quad Cities 1 and 2 (USNRC Docket Nos. 50-254 and 50-265), Rancho Seco (USNRC Docket No. 50-312), Grand GuJf Unit 1 (USNRC Docket No. 50-416), Oyster Creek (USNRC Docket No. 50-219), V.C. Summer (USNRC Docket No. 50-395), Diablo Canyon Units 1 and 2 (USNRC Docket Nos. 50-275 and 50-323), Vogtle Unit 2 (USNRC Docket No. 50-425) and Millstone Point Unit 1 (USNRC Docket No. 50-245). The results show that the high density spent fuel racks are structurally adequate to resist the postulated stress combinations associated with level A, B, C, and D conditions as defined in Ref. 6.0.2. . 6.1 ANALYSIS OUTLINE (FOR NEW PROPOSED RACK MODULES) The spent fuel storage racks are Seismic Class I equipment. They are required to remain functional during and after a-Safe Shutdown Earthquake (Ref. 6.1.1). These racks are neither anchored to the . l pool- floor nor attachad to the sidewalls. The individual rack modules are not interconnected. Furthermore, a particular rack may be completely loaded with fuel assemblies (which - corresponds to O 6-1
greatest rack inertia), or it may be completely empty. The coefficient of friction, p, between the supports and pool floor is another indeterminate factor. According to Rabinowicz (Ref. l 6.1.2), the results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of p to be 0.503 ' with a standard deviation of 0.125. The upper and lower bounds (based on twice the standard deviation) are thus 0.753 and 0.253, respectively. Analyses are therefore performed for single rack simulations assemblies with values of the coefficient of friction equal to 0.2 (lower limit) and 0.8 (upper limit), respectively. A typical rack is also analyzed using y = .5 to establish that an intermediate friction coefficient does not lead to results which are not bound by the two limiting values of p. In order to predict the limiting conditions of rack module seismic response, the following module geometries are analyzed:
- 1. 12x15 Region 2 rack
- 2. 8x13 Region 1 rack
- 3. 8x13 Region 2 rack The racks are considered fully loaded, partially loaded, or almost O empty, with various coefficients of friction used to identify the worst case response for rack movement and for rack structural l integrity. Af ter completion of reracking, the gaps between the rack modules and those between the racks and walls will be in the .
I manner of Figure 1.1. However, intermediate configurations during reracking may exist when a rack module is far away from the walls, or other modules. It is necessary to demonstrate that even the intermediate configurations meet all seismic criteria. With this objective in mind, in all cases studied, the rack is assumed to be the only rack occupying the pool so as to minimize hydrodynamic effects from adjacent racks. This is conservative since any significant hydrodynamic mass induced by small clearances between l
1 I structures results in reductions in movements and stresses. We show in this report that even with the neglect of the hydrodynamic mass induced by the small gaps between structures, the resulting motion is such that no rack-to-rack or rack-to-wall impacts will occur. It is noted that module types 1 and 2 above belong to the ' grouping of modules which are planned to be installed in the first batch. The third rack type (8x13 module) analyzed herein belongs to the batch of racks which will be installed at a suitable time in the future. The seismic analyses were performed utilizing the time-history method. Pool slab acceleration data in three orthogonal directions was developed and verified to be statistically independent. 1 The objective of the seismic analysis of single racks is to determine the -tructural response (stresses, deformation, rigid body motion, etc.) due to simultaneous application of the three statistically independent, orthogonal seismic excitations. Thu's, recourse to approximate statistical summation techniques such as l the " Square-Root-of-the-Sum-of-the-Squares" method (Ref. 6.1.3) is avoided. For nonlinear analysis, the only practical method is simultaneous application of the seismic loading to a nonlinear , model of the structure. Pool slab acceleration data are developed from specified response ! spectra for two earthquakes: SSE and OBE. Seismic time histories are calculated from the plant response spectra using it equipment. damping. Figures 6.0a, b show the plant horizontal response spectrum for OBE at levels 302' and 329', respectively. In ;- accordance with the GPUN seismic specification, an OBE horizontal ' l l 6-3 i f o
i responce can be generated by linear interpolation at any level i between the given levels. The spent fuel pool floor is at level 305'; however, for additional conservatism in the results, the actual OBE spectrum used to develop the time histories was generated for level 312'. In accordance with the TMI FSAR, SSE horizontal spectrums are twice OBE, and vertical spectrums are 2/3 of horizontal. Figure 6.0c shows the final OBE horizontal spectrum, the SSE horizontal spectrum, and the vertical spectrum. . As noted above, these spectra, at level 312', are used to generate time histories for fuel rack excitation which are considered to act at level 305'. Figures 6.1 - 6.12 show the time-histories and comparison of the corresponding velocity spectra and the design spectra for SSE and OBE conditions. To examine the potential for , l overturning, SSE seismic excitations amplified by a factor of 1.5 are imposed on the rack module which exhibits maximum kinematic response under the foregoing analyses. The soismic analysis of a single rack is performed in three steps, namelyt
. 1. Development of a nonlinear dynamic model consisting of inertial mass elements, spring, gap, and friction elements.
- 2. Generation of the equations of motion and inertial coupling and solution of the equations using the
" component element time integration scheme" (Refs .
6.1.4 and 6.1.5) to determine nodal forces and displacements.
- 3. Computation of the detailed stress field in the rack just above the baseplate and in the support legs using the nodal forces calculated in the previous >
step. These stresses are checked against the design limits given in Section 6.5. l A brief description of the dynamic model follows.
. 6-4
i 6.2 FUEL RACK - DYNAMIC MODEL Since the racks are not anchored to the pool slab or attached to the pool walls or to each other, they can execute a wide variety l of motions. For example, the rack may slide on the pool floor (so-called "slidincj condition"); one or more legs may momentarily lose contact with the liner (" tipping condition"); or the rack may l experience a combination of sliding and tipping conditions. The l structural model should permit simulation of these kinematic l events with inherent built-in conservatisms. Since the modules are designed to preclude the incidence of inter-rack impact, it is also necessary to include the potential for inter-rack impact phenomena in the analysis to demonstrate that such impacts do not occur. Lift off of the support legs and subsequent liner impacts must be modelled using appropriate impact (gap) elements, and Coulomb friction between the rack and the pool liner must be l simulated by appropriate piecewise linear springs. The elasticity l of the rack structure, relative to the base, must also be included in the model even though the rack may be nearly rigid. These special attributes of the rack dynamics require a strong emphasis on the modeling of the linear and nonlinear springs, dampers, and compression only stop elements. The term non-linear spring is the generic term to denote the mathematical element representing the situation where the restoring force exerted by the element is not linearly proportional to the displacement. In the fuel rack simulation the coulumb friction interface between the rack support leg and the liner is a typical example of a non-linear spring. The model outline in the remainder of this section, and the model description in the following section, describe the detailed modeling technique to simulate these effects, with considerable emphasis placed on the nonlinearity of the rack seismic response. 6-5 3 (a 1 6 m .
i 6.2.1 outline of Model for Comeuter Code DYNARACK
- a. The fuel rack structure is a folded metal plate assemblage welded to a baseplate and supported on four legs. An odd-shaped module may have more than four legs. The rack structure itself is a very rigid structure. Dynamic analysis of typical multicall racks has shown that the motion of the structure is captured almost completely by modelling the rack ;
as a twelve degree-of-freedom structure, where the movement , of the rack cross-section at any height is described in terms ' of six degrees-of-freedom of the rack base and six degrees of l freedom defined at the rack top. The rattling fuel is modelled by five lumped masses located at H, .75H, .5H,
.25H, and at the rack base, where H is the rack height as measured from the base.
- b. The seismic motion of a f ue 's rack is characterized by random rattling of fuel asramblies in their individual l storage locations. Assuming a certain statistical coherence (i.e. assuming that all fuel elements move in-phase within a rack) in the vibratic, of the fuel assemblies exaggerates the 1 computed dynamic loading on the rack structure. This )
assumption, however, greatly reduces the required degrees-of-freedom needed to model the fuel assemblies which are represented by five lumped masses located at different levels of the rack. The centroid of each fuel assembly mass Q can be located, relative to the rack structure centroid at that level, so as to simulate a partially loaded rack. ,
- c. The local flexibility of the pedestal is modelled so as to account for floor elasticity, and local rack elasticity just above the pedestal.
d.,The rack base support may slide or lift off the pool floor.
- e. The pool floor has a specified time-history of seismic accelerations along the three orthogonal directions.
- f. Fl'ld coupling between rack and fuel assemblies, and between rack and wall, is simulated by introducing appropriate inertial coupling into the system kinetic energy. Inclusion L of these effects uses the methods of Refs. 6.2.3 and 6.2.4 :
for rack / assembly coupling and for rack / rack coupling (see ' Section 6.2.3 of this report). 6-6 L i '
~ {
_ , _ _ _ _ _ _. _ _ . _ _ __ _ i
_ _ _ _ . . - _- . . _ - ~ _ . - - . - _. -
- g. Potential impacts between rack and fuel assemblies are
-] accounted for by appropriate " compression only" gap between masses involved.
elements
- h. Fluid dam due to viscous effects between rack and assemblies, ping and between rack and adjacent rack, is conservatively neglected.
I l
- 1. The supports are modeled as "comprescion only" elements for the vertical direction and as " rigid links" for transferring horizontal stress. The bottom of a support leg is attached to a frictional spring as described in Section 6.3. The cross-section inertial properties of the support legs are ,
compi.ed ' and used in the final computations to determine- l
.upport leg stresses. ,
l
- j. The effect of sloshing is negligible at the level of the top of the rack and is hence neglected.
- k. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at the top and bottom of the rack in the two horizontal directions.
- 1. Rattling of fuel assemblies inside the storage locaticns causes the " gap" between the fuel assemblies and the cell wall to change from a maximum of twice the nominal gap to a theoretical zero gap. Fluid coupling coefficients are based p J on the nominal gap.
- m. The form drag due to motion of the fuel ' assembly in the storage cell, or that due to movement of a rack in the pool, which har been previously considered in licensing high density racks, have also been neglected in this analysis for added ,:onservatism.
- n. The cross coupling effects due to the movement of fluid.from one. interstitial (inter-rack) space to'the adjacent one is modelled. using potential flow and Kelvin's circulation theorem. This formulation has been reviewed and approved by the Nuclear Regulatory Commission, during the post-licensing multi-rack analysis for Diablo Canyon Unit I and II reracking
-project. The coupling coefficients are based on a consistent modelling of the fluid flow. While updating of the fluid' flow coefficients, based on the current gap, is permitted in the algorithm, the analyses here are conservatively carried i
6-7 o o , l
out using the constant nominal gaps that exist at the start of the event. For conservatism in this analysis, all A simulations are carried out assuming no rack-to-rack D hydrodynamic interaction. That is, it is assumed that any single rack can be placed in the pool at its specified location and withstand a seismic event. The results show , that even with this restriction on the hydrodynamic effect, the racks will not impact each other when they are all placed together in the pool. Figure 6.13 shows a schematic of the model. Twelve degrees of freedom are used to track the motion of the rack structure. Figures 6.14 and 6.15, respectively, show the inter-rack impact springs (to track the potential for impact between racks or between rack and wall) and fuel assembly / storage cell impact springs at a particular level. As shown in Figure 6.13, the model for simulating fuel assembly motion incorporates fiva rattling lumped masses. The five rattling masses are located at the baseplate, at quarter height, at half height, at three quarter height, and at the top of the rack. Two degrees of freedom are used to track the motion of each rattling mass in the horizontal plane. The vertical motion of each rattling. mass-is assumed to be the same as the rack base. Figures 6.16, 6.17 and 6.18 show the modelling scheme for including rack , elasticity and the degrees of freedom associated with rack elasticity. In each plane of bending a shear and a bending spring are us . .! to simulate elastic effects in accordance with Ref. l 6.2.1. Table 6.3 gives spring constants for these bending springs as well as corresponding constants for extensional and torsional rack elasticity. l l t l' l l s 6-8 l O
6.2.2 Model Demerletion h The absolute degrees of freedom associated with each of the mass locations are identified in Figure 6.13 and in Table 6.1. The rattling masses (nodes 1*, 2*, 3*, 4*, 5*) are described by translational degrees-of-freedom q7-q16* Ui(t) is the pool floor slab displacement seismic time-history. Thus, there are twenty-two degrees of freedom in the system. Not shown in Fig. 6.13 are the gap elements used to model the support legs and the impacts with adjacent racks. 6.2.3 Fluid Couclinc , An effect of some significance requiring careful modeling is the
" fluid coupling ef f ect" (Refs. 6.2.2 and 6. 2. 3 ) . If one body of mass (mi) vibrates adjacent to another body (mass m2), and both
-bodies are submerged in a frictionless fluid medium, then Newton's equations of motion for the two bodies have the form I] (mi + M11)=X1+M12 X2 = applied forces on mass mi + 0 (x12)
M21 X1 + (m2 + M22) X2 = applied forces on mass m2 + 0 (X22 ) N N X1 ,- X2 denote absolute accelerations of mass mi and m2e respectively. M11, M12s M21, and M22 are fluid coupling coefficients which t depend on the shape of the two' bodies, their relative disposition,
'l etc.'Fritz.(Ref. 6.2.4) gives data for Mij for various body shapes and arrangements.-The above equation indicates-that the effect of-the fluid is to add-a certain amount of mass to the body (M11 to body 1), and an external force which is proportional to the 6-9
acceleration of the adjacent body (mass m2). Thus, the acceleration of one body affects the force field on another. This O- force is a strong function of the interbody gap, reaching large values for very small gaps. This inertial coupling is called fluid coupling. It has an important effect in rack dynamics. The lateral motion of a fuel assembly inside the storage location will encounter this effect. So will the motion of a rack adjacent to .- another rack if the racks are closely spaced. These effects are included in the equations of motion. For example, the fluid coupling is between nodes 2 and 2* in Figure 6.13. Furthermore, the rack equations contain coupling terms which model the effect of fluid in the gaps between adjacent racks. The coupling terms modeling the effects of fluid flowing between adjacent racks are computed assuming that all adjacent racks are vibrating 180 0 out of phase from the rack being analyzed. Therefore, only one rack is considered surrounded by.a hydrodynamic mass computed as if there were a plane of symmetry located in the middle of the gap region. As noted previously, it is desired to permit any single rack to be h placed in the pool in the absence of any or all of the other racks. Therefore, all analyses in this report assume that the racks are isolated. That is, inter-rack hydrodynamic effects are neglected by assuming a large gap to exist between racks. Finally, fluid virtual mass is included in the vertical direction vibration equations of the rack; virtual inertia is also added to the governing equation corresponding to the rotational degree of freedom, q6(t) and q22(t). 6-10 O t
~,t-
< t i
6.2.4 Damoinc O l In reality, damping (Ref. 6.2.5) of the rack motion arises from material hysteresis (material damping), relative intercomponent motion in structures (structural damping), and fluid viscous effects (fluid damping). In the analysis, a maximum of 1% structural damping is imposed on elements of the rack structure during OBE and SSE simulations. Material and fluid damping due to fluid viscosity are conservatively neglected. The dynamic model has the provision to incorporate form drag effects; however, no form drag has been used for this analysis. 6.2.5 Imoact Any fuel assembly node (e.g., 2*) may impact the corresponding structural mass node 2. To- simulate this impact, four compression-only gap elements around each rattling fuel assembly node are provided (see Figure 6.15). The compressive loads developed in these springs provide the necessary data to evaluate the integrity of the cell wall structure and stored array during O the seismic ev.nt. Figure 6.14 shows the 1ocation of the impact springs used to simulate any potential for inter-rack or rack-to-wall impacts. Section 6.4.2 gives more details on these additionalimpact springs. Since there are five rattling masses, a total of 20 impact springs are. used to model fuel assembly-cell wall impact. 6.3 ASSEMBLY OF THE DYNAMIC MODEL The cartesian coordinate system associated with the rack has the following nomenclature O x = Horizontal coordinate along the short direction of rack rectangular platform O 6-11
I I O y = Horizontal coordinate along the long direction of O the racx rectaneutar platform 0 z = Vertical coordinate upward from the rack base 1 If the simulation model is restricted to two dimensions (one horizontal motion plus vertical motion, for example) for the Durcotip of model clarification oniv, then a descriptive model of the simulated structure which includes gap and friction elements is shown in Figure 6.19. The impacts between fuel assemblies and rack show up in the gap elements, having local stiffness K, I in Figure 6.19. In Table 6.2, gap elements 5 through 8 are for the vibrating mass at the top of the rack. The support leg spring rates Ks are modeled by elements 1 through 4 in Table 6.2. Note that the local compliance . of the concrete floor is included in Ks. To simulate sliding potential, friction elements 2 plus 8 and 4 plus 6 (Table 6.2) are shown in Figure 6.19. The friction of the support / liner interface (3 is modeled by a piecewise linear spring with a suitably large stiffness Kg up to the limiting lateral load, pN, where N is the current compression load at the interface between support and liner. At every time step during the transient analysis, the current value ' of N (either zero for liftoff condition, or a compressive finite value) is computed. Finally, the support rotational friction springs KR reflect any rotational restraint that may be offered by the foundation. This spring rate is l calculated using a modified Bousinesq equation (Ref. 6.2.6) and is included to simulate the resistive moment of the support to counteract rotation of the rack leg in a vertical plane. This i O 6-12
- m
g . rotation sprinej is also nonlinear, with a zero spring constant value assigned af ter a certain ' limiting condition of slab moment loading is reached. The nonlinearity of these springs (friction elements 9, 11, 13, and 15 in Table 6.2) reflects the edging limitation imposed _on the base of the rack support legs and the shifts in the centroid of load application as the rack rotates. If this effect is neglected; any support leg bending, induced by liner / baseplate friction forces is resisted by the leg acting as a beam cantilevered from the rack baseplate. This leads to higher predicted loads at the support leg - baseplate junction than if the moment resisting l capacity due to floor elasticity at the' floor is included in the model. The spring rate Ks modeling the effective compression stiffness of the structure in the vicinity of the support, is computed from the equations' O 1 1 1 1
- - -+ _ +
Ks K1 K2 K3 where; K-i
= spring' rate of the- support leg treated as a tension-compression membsr
= local spring rate of pool slab K2
=
'K3 spring rate of folded' plate cell structure . above support leg y
h 6-13
As described in the preceding section, the rack, along with the base, supports, and stored fuel assemblies, is modeled for the general three-dimensional (3-D) motion simulation by a twenty-two degree of freedom model. To simulate the impact and sliding phenomena expected, up to 64 nonlinear gap elements and 16 nonlinear friction elements are used. Gap and friction elements, with their connectivity and purpose, are presented in Table 6.2. Table 6.3 lists representative values for the modules used in the dynamic simulations. For the 3-D simulation of a single rack, all support elements (described in Table 6.2) are included in the model.* Coupling between the two horizontal seismic motions is provided both by any offset of the fuel assembly group centroid which causes the rotation of the entire rack and/or by the possibility of liftoff G v Since inter-rack or rack-to-wall impact is not to occur - in the subject modules, only 8 gap elements are used around the bottom and top edges of the rack instead of the twenty described in ' Table 6.2. Since their purpose is only to signal if-an impact occurs, the exact number utilized has no bearing on the final reported results. O 6-14
1 i of one or more support legs. The potential exists for the rack to p be supported on one or more support legs during any instant of a complex ~3-D seismic event. All of these potential events may be simulated during a 3-D motion so that a mechanism exists in the model to simulate the real behavior. I 6.4 TIME INTEGRATION OF THE EQUATIONS OF MOTION I l 6.4.1 Time-History Analysis Usino Multi-Decree of Freedom l Rack Model Having assembled the structural model, the dynamic equations of motion corresponding to each degree of freedom are written by using Lagrange's Formulation. The system kinetic energy can be constructed including contributions from the solid utructures and from the trapped and surrounding fluid. A single rack is modelled in detail. The system of equations can be represented-in matrix notation as: (M) {q=> = {Q) + {G}
,q : where the vector {Q} is a function of nodal displacements and V velocities, and-{G} depends on the coupling inertia and the ground acceleration. Premultiplying the above equations by (M]-1 i
renders the resulting equation uncoupled in mass. We have: {q"} = (M)-1 {Q} + [M)-1 {G}
- As noted earlier, in the numerical simulations run to ' verify structural integrity during a seismic event, the rattling fuel i
assemblies are assumed to ' move .tn phase. This will provide-
-maximum impact force level, and induce additional conservatism in the time-history. analysis.
I 6-15 Q_ 2' .- l
W This equation set is mass uncoupled, displacement coupled at each instant in time, and is ideally suited for numerical solution using a central difference scheme. The proprietary, USNRC qualified, computer program "DYNARACK"* is utilized for this purpose. Stresses in various portions of the structure are computed from known element forces at each instant of time and the maximum value of critical stresses over the entire simulation is reported in sundnary form at the end of each run.
'h
- similar This code has been previously utilized in racks for Enrico Fermi Unit 2 (USNRC Docket No. 50-341),
licensing of
-Quad Cities- 1 and 2 (USNRC Docket Nos. 50-254 and 265),-Rancho Seco (USNRC Docket No. 50-312), Oyster Creek- (USNRC Docket No.
50-219), V.C. Summer (USNRC Docket No. 50-395), and Diablo Canyon 1 and 2 (USNRC Docket Nos. 50-275 and 50-323), St. Lucie Unit I _(USNRC Docket No. 50-335), Byron Units I and II (USNRC Docket Nos. 50-454, 50-455),1 Vogtle 2 (USNRC Docket 50-425), and Millstone Unit 1-(USNRC Docket 50-245). M 6-16
\; , O In summary, dynamic analysis of typical multicell racks has shown that the motion of the structure is captured almost completely by the behavior of a twenty-two degree of freedom structure; therefore, in this analysis model, the movement of the rack cross-section at any height is-described in terms of the rack degrees of freedom (q1(t),...q6(t) and gi7-q22(t)). The remaining degrees of freedom are associated with horizontal movements of the fuel assembly masses. In this dynamic model, five rattling masses are used to represent fuel assembly movement in the horizontal , plane. Therefore, the final dynamic model consists of twelve degrees of freedom for the rack plus ten additional mass degrees of freedom for the five rattling masses. The totality of fuel mass is . included in the simulation and is distributed among the five rattling masses. ! 6.4.2 Evaluation of Potential for Inter-Rack Imoact p since racks are usually closely spaced, the simulation includes impact springs to model the . potential for inter-rack impact. To account ' for this potential, yet still retain the simplicity of simulating . only a. single rack, gap elements are located on the rack at the top and at the baseplate level. Figure 6.14 shows-the , location of these - gap elements. _ Where impacts between racks is permitted by specification, twenty gap elements at each level would be used as shown. The rack design specification precludes
' any impacts between racks or between ' rack and walls during any
-single event; therefore~ only sixteen impact springs are retained (8 at
- ap : 'and 8 at baseplates) solely to - demonstrate that the postulated gaps ~do not close completely 'dee to rack motion. In this analysis, additional conservatism is incorporated in every..
analysis? byJ assuming ' that rack-to-rack hydrodynamic terms- are 6-17
- p. > )
I'
-. - neglected; Thus the impact springs lack only the propensity for ? rack-to-wall impact. The resulting deflection analysis shows that rack-to-rack impacts do not occur even if all racks are in the t- pool.
L 6.5 STRUCTURAL ACCEPTANCE CRITERIA There are two sets of criteria to be satisfied by the rack
- modules i
- a. Kinematic Criterion -l This criterion seeks to ensure that the rack is a physically stable structure. The racks are designed to '
preclude -inter-rack impacts. Therefore, pl.ysical stability of the rack .is considered along with the criterion that inter-rack impact or rack-to-wall impacts do not occur.
- b. Stress Limits The- stress limits of the ASME Code, Section -III, Subsection NF, 1986 Edition are used. The following -
loading combinations are applicable-(Ref. 6.5.1)-and are s j. ;
.. . J .
7, consistent with the plant FSAR commitments. Loadino Combination Stress Limit D+L Level A service-limits D + L + To ' D + L'+-To + E D + L + Ta + E- Level B service limits D + L + To + Pg 4 D + L + Ta_+ E' Level D service limits D+L+Fd The functional capability : of-the fuel racks-should be demonstrated.
)
1 6-18 1 4 v - ,
i i l a ;
.l l
h- where: l D = Dead weight-induced stresses (including fuel assembly weight) 1 L = Live Load (0 for the structure, since there are no moving objects in the rack load path). 1
=
Fd Force caused by the accidental drop of the i heaviest load from the maximum possible I height. I Pg = Upward force on the racks caused by postulated i stuck fuel assembly E = Operating Basis Earthquake (0BE) E' = Safe Shutdown Earthquake (SSE) To -= Differential temperature induced loads (normal I or upset condition) l Ta = , Differential temperature induced loads (abnormal design conditions)
.The conditions .Ta and To cause local thermal stresses to be
~
l
-producedi The worst: situation = will be obtained when an isolat d L
storage- location has a - fuel assembly' which is generating- heat at ' l the maximum postulated' rate. The surrounding: storage locations are assumed to contain no fuel. The heated = water makes unobstructed I
, contact . with the 'inside..of the storage walls,-. thereby producing.
the maximum possible temperature difference between . the adjacent . J E f cells'. The secondary stresses thus produced ' are limited - to the Tbody of the rack; that-.is, the support legs ldo not1 experience the j secondary-(thermal). stresses. yi , p l l-
?
r q
..b 6-19' . 4 I,
I Jl
-. - .__________-______k
2 4 6.6 MATERIAL PROPERTIES The data on the physical properties of the rack and support materials, -obtained from the ASME ' Boiler & Pressure Vessel Code, Section III, appendices, are listed in Table 6.4. Since the maximum pool-bulk temperature is less than 200 F, this is used as the reference- design temperature for evaluation of material l properties. 6.7 STRESS LIMITS FOR VARIOUS CONDITIONS The following stress limits are derived from the guidelines of the , ASME Code, Section III, Subsection NF, in conjunction with the material properties data of the preceding section. 6.7.! Normal and Uoset Conditions (Level A or Level B)
- a. Allowable stress in tension on a net section-
=Ft = 0.6 Sy or Ft =-(0.6) (25,000) = 15,000 psi (rack material)
Ft = is equivalent to primary membrane stresses h Ft = (.6) (25,000) = 15,000- psi (upper part-of support feet) i
= (.6) (106,300) = 63,780- psi (lower part of j support feet) '
'b. -On the gross section, allowable stress in shear is:
Fy = .4 S (.4)y(25,000) --10,000 psi (main' rack body): l Ft=- (.4)-(25',000) = 10,000 psi-(upper part of
. sr.pport feet);
= ( .4 ) - (106,300) = 42,520- psi (lower part-of support feet) h 6-20
~
,,.t..
)
)
, c. Allowable stress in compression, Fat k12 2
[1 - /2Ce Sy r . Fa " l 5 ki- kl 3 3 (( ) + (3.( ) /8Ce) .(( ) /8C e )} 3 r r
-where:
(2n E) 2 12
/ :
Ce = ( ) sy l k1/r ' for the main rack - body is based on.the full I
. height and cross section . of the honeycomb region..
i Substituting numbers,-- we 'obtain, for .both support legiand honeycomb regions i T h' Fa.=:15,000 psi (main rack body) Fa " 15,000 psi (upper.part of. support' feet)-
= 63,780 psi.(lower part of support feet) '
a
-d. Maximum. allowable ~ bending stress. at the outermost ;
fiber due to flexure about one plane of symmetry:- ,
-1 F b'= 0.60 S y = 15,000 psi (rack body) i Fb =.15,000 psi-(upper part of-support feet) !
= 63,780. psi (lower part of support. feet)- i o
.e. Combined' flexure and: compression
_i fa _
.Cmx fb x. Cmyf by -!
+ + . <1 -i Fa l DxF bx' .- .;Dy Fby : l s
' i:
h . 6-21 m < t
.ab
-r .s - where: : 0: fa = Direct compressive stress in the 1 section fx b = Maximum flexural stress along x-axis fby = Maximum flexural stress along y- , axis
=
Cmx Cmy = 0.85 fa Dx = 1 - i F'ex , fa Dy=1- 1 where 12 n2 E F'ex,ey =
- e b 2 4
23 ( ki xty )- - IDxty-and the subscripts - x , y. reflect the particular_ ; bending-plane,of_ interest.
- f. Combined flexure and: compression-(or tension):
fa' fx b fby
+ + <'1.0 0.6S y-- Fx b Fby- i i
Tn= r.bove requirement should 'ba met for both the i direct tension'or compression case. t
-h. 6-22 l i l
4 i-
.f z, .-
6.7.2 Level D Service Limits
-( ..
j LF-1370 (ASME Section III, Appendix F), states that the limits for the Level D condition are the minimum of 1.2 (Sy /Ft) or (0.7Su/Ft )- times the corresponding limits for Level A condition. Since 1.2 S y is greater than 0.7 Su for the lower part of the support feet, the factor is 1.54 for the lower section under SSE conditions. The facto.: for the upper portion of the support foot is 2.0. ;
-Instead of. tabulating the results of these six different stresses as dimensioned values, they are presented in a dimensionless form.
These so-called stress - factors are defined as the ratio of the , actual developed stress to its specified limiting . value. With this definition, the. limiting value of each stress factor is 1.0 for the OBE and 2.0 (or 1.54).for the SSE -condition. 'I 6.8 RESULTS FOR THE ANALYSIS OF SPENT FUEL RACKS
.USING A SINGLE RACK MODEL AND-3-D SEISMIC MOTION d AL eomplete synopsis.of the analysis of the modules subject ta the postulated earthquake motions, is presented in a summary Table 6.5 ,
.which gives the bounding values of stress factors Ri-(l'= 1.. 7).
Theistress factors are defined as: j
= Ratio of direct tensile.or compressiveistress on m.-
.R1 not' section to its allowable value - (note support:
feet only support compression).
= Ratio of gross shear on a . net ~ section in the x-R2 direction to its allowable value.
R3
=
Ratio - of . maximum bending stress 'due to bending about the x-axis to its allowable value . for -the section
, s~ .
'v' 6-23 .
p p ?
q i l lO R. = Rat a of -xaum bendine stress due to bendue about the y-axis-to its allowable value R5
=
Combined flexure and compressive factor (as defined in 6.7.le above) ! R6
= Combined flexure and tension (or compression) [
, factor (as defined in 6.7.lf above)
=
R7 Ratio of gross shear on a net section in the y- I direction to-its allowable value. ; As stated before, the allowable value of R1 (i =1,2,3,4,5,6,7) is d l'for the OBE condition and 2 for the SSE (exceot for the lower i section of the succort where the factor is 1.54)
= The- dynamic, analysis _ gives the maximax (maximum in time and- in space)? values of;the stress factors at critical locations in the rack module. Values are also obtained for maximum rack displacements.and for critical impact loads. Table 6.5 presents critical'results for:the stressefactors, and rack to-fuel impact f load.. Table 6.6 presents maximum results for horizontal =
1' displacements; at - the : top . and bottom of the rack in the x andLy. direction. "x" is ' alwtv s the short direction of the rack. . In 4 Table 6.6',; fc each run, both .i.ne maximum value. of the sum of all ! support foot loadings'(4 supports) as well as the maximum value on any" single: foot is: reported. The tableaalso.gives: values: for the
. maximum vertical load and the corresponding net shear force at the liner at' essentially the same time _ ' instant, ~and for thee maximum -
. net shear load and the corresponding vertical force at a support
. foot:at essentially the same time instant. !
-The.results presented in Tables 6_.5, 6.6 represent'the totality of runs - carried out. The critical ' case for structural integrity .
ccalculations l's included. t 6-24
V-The results corresponding to SSE give the highest load factors. Note that the results given for the SSE yield maximum stress factors (Ri) that are below the limiting value for the OBE condition for all sections.The critical load factors reported for the support feet are all for the upper segment of the foot and for SSE simulations are to be compared with the limiting value of 2.0. Results for the lower portion of the support foot are not critical and are not reported in the tables. Analyses show that significant margins of safety exist against local deformation of the fuel storage cell due to rattling impact of-fuel assemblies. Overturning has also been considered. This has been done by assuming a multiplier of 1.5 on the SSE horizontal earthquakes ' (more con's ervative . than the USNRC Standard Review Plan) and checking predicted displacements. The horizontal displacements do ~ not - grow ' 'to such an extent as to imply any possibility for overturning. Run D61 presents - the maximum displacements for the case where' .- the horizontal excitation level is -increased' by 50%. The increases in movement due to increased excitation are seen by comparing the results,with Run D51. 6.9 IMPACT ANALYSES-6.9.1 Imcaet Leadina Between Fuel Assembiv and Cell Wall
~
The local stress.in a cell wall is conservatively estimated from t the peak impact loads obtained from the dynamic simulations.' Plastic analysis is- used to obtain the limiting-
' impact load. The limit load is calculated as 4315 lbs. per' cell O 6-25
i g I n- which is much greater than the loads obtained from any of the 1 i-simulations. L 6.9.2 Imoacts Between Adiacent Racks All of~the dynamic analyses assume, conservatively, that the racks are isolated. However, the displacements are obtained from the dynamic analyses are less than 50% of the rack-to-rack spacing or rack-to-wall spacing if the pool is assumed fully populated. Therefore, we conclude that no impacts between racks or between racks and walls occur during the SSE event. 6.10 WELD STRESSES Critical weld locatioits under seismic loading are at the bottom of the rack at ; the baseplate connection and . at the welds on the support legs. Results from the dynamic analysis using the i simulation codes are surveyed and the maximum loading is used to qualify the welds on'these locations. ; h 6.'10.1- Baseolate to Rack Welds and Cell-to-Cell Welds Section NF permits, .for the SSE ' condition, an allowable weld i stress r= .'42'Su;= 29,820 psi. Based.on-the worst case of all runs reported, the maximum weld stress for the baseplate to rack weldsLis 7541 psi for SSE conditions. This value occurs using a fuel weight of.3000 lbs. per cell. . The weld between baseplate and support leg is checked using limit analysis techniques'. The structural - weld at that location is
. considered safe if the interaction curve' is' such that a derived function of F/Fy and M/My is below a limiting value.
Fy ,' My are the limit load and moment under direct load only and-6-26 otl
i direct moment only. F, M are the absolute values of the actnal peak force and moments applied to the weld section. For the worst case' simulation, this criterion requires that M/M y be less than
.951. The calculated value for the critical design case of normal fuel loading As .053. '
The critical area that must be considered for fuel tube to fuel tube welds is the weld between the fuel tubes. This weld is ' M scontinuous as we proceed along the tube length. Stre' eses in'the fuel tube to fuel . tube welds develop along the len.3th of each fuel tube due to fuel assembly impact with the tube wall. This occurs if fuel assemblies in adjacent tubes are moving L - out of phase with one another~so that impact loads in two adjacent h tubes are in' opposite directions which would tend to. separate the , L ' channel from the tube at the weld. The critical load.that can be ' transferred in this . weld region for the SSE condition is calculated as 5271 lbs. at every fuel tube connection to adjacent-0; tubes. in upper bound to the 1oad reguired to be transferred is V2;x-380 x 2 = 1075 lbs. where we have=used a maximum impact load of 380 lbs. -(from Table
~
~
-1 6.5), assumed t'wo impact locations are . supported. by each ~ weld
' region, and . have increased the load by V2 to . ' account for 3-D
' effects..
I 6 .10 ~. 2 .Heatino of an Isolated cell- 'l l l Weld - stresses due to heating :of . an isolated hot cell are also
' computed.-The' assumption used is that.a single cell is-heated, over its entire length, to a temperature above the value I associated with all surrounding cells. No. thermal. gradient'in the vertical' direction is assumed so that the results are o
6-27 1 I
i O conservative. Using the temperatures associated with this unit, analysis shows that the weld stresses along the entire cell length do not exceed the allowable value for a therinal loading condition. Section 7 reports a value for this thermal stress. 6.11 SPENT FUEL POOL SLAB MODEL ; 6.12 SPENT FUEL POOL SLAB ANALYSIS AND RESULTS 6.13 DEFINITION OF TERMS USED IN SECTION 6.0 S1, S2, S3, S4 Support designations pi Absolute degree-of-freedom number i qi Relative degree-of-freedom number i p- Coefficient of friction Ui Pool floor slab displacement time history in the i-th direction x,y coordinates horizontal direction
'O z coordinate vertical direction K
I. Impact spring between fuel assemblies and cell K Linear component of friction spri'ng f Ks Axi:1 spring at support leg locations N Compression load in a support foot K R: Rotational spring provided by the pool slab j O' (/ 6-28
6 F i 1 Subscript i When used with U or X indicates O air etic = <t - 1 x-airectie=, 1-2 y-direction, i = 3 z-direction) j 6.14 REFERENCES FOR SECTION 6 6.0.1 USNRC Standard Review Plan, NUREG-0800 (1981). 6.0.2 ASME, Boiler & Pressure Vessel Code, Section III, Subsection NF (1983). 6.1.1 USNRC Regulatory Guide 1.29, " Seismic Design ! Classification," Rev. 3, 1978. t 6.1.2 " Friction Coefficients of Water Lubricated Stainless Steels' for a Spent Fuel Rack Facility," Prof. Ernest Rabinowicz, MIT, a report for Boston Edison Company, 1976. 6.1.3 USNRC-Regulatory Guide 1.92, " Combining Modal Responses .
,and Spatial Components in Seismic Response Analysis," 1 Rev. 1, February, 1976.
6.1.4 "OT Position for Review and' Acceptance of Spent Fuel. Storage and _ Handling Applications",: dated April 14,
. 1978, and January 18, 1979 amendment thereto. 3 6.2.2 " Mechanical- Design' of Heat Exchangers and Pressure vessel a components,." Chapter 16, K.P. Singh and A.I.
'Soler, Arcturus Publishers, Inc., 1984.
6.2.3- " Dynamic coupling in a closely. Spaced Two-Body System Vibrating lin Liquid Medium: The Case ~ of Fuel : Racks , "
, K.P. Singh'and A.I.'Soler,-3rd International Conference on Nuclear Power Safety, Keswick', England, May 1982.
6.2.4- R.J..'Fritz, "The Effects c of Liquids on . ' the Dynamic - Motions of Immersed-Solids," Journal of Engineering for~ Industry, Trans. of the ASME, February 1972, pp 167-172..
~6.2.5 USNRC Regulatory. Guide 1.61,." Damping Values for
[; seismic Design of Nuclear Power Plants ," 1973. h- 6-29 i i
cs . .- ,. -} I t 6.3.1 "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering,"
~
. S.
Levy and J.P.D. Wilkinson, McGraw Hill, 1976. 6.3.2 " Dynamics of Structures," R.W. Clough and J. Penzien, McGraw Hill (1975). i m I
~
I 6-30
~
t l '_ 1
-, t I
p , 1 l: Table 6,1 1 DEGREES OF FREEDOM Displacement Rotation Location, Uy Ux_ Uz 6x- By Oz
;(Node) 1 P1 P2 P3 94 95 96 2 P17 P18 P19 q20 921 922 Point 2 is assumed attached to rigid rack-at the' top most point, 2* p7 P8- .
3* pg p10 1
] 4*
-5*'
.P11 P12
_p13: p14 ' 1* P15- P16 wheret l pi- '= qi(t)-+ U l(t) 1:- 1,7,9,11,13,15,17-
= qi(t) + U 2(t) i =_2,8,10,12,14,16~,18.
= qi(t) + U 3(t). 1 = 3,19 1
.Ui (t) are the 3.known: earthquake displacements.-
)
l
, l N
6-31
. - - - -~ -. ... .- - - - - .- ._-
i
~
Table 6.2 NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS I. Nonlinear Serinas (Gao Elements) (64 Total) Number Node Location Descriotion 1 Support S1 Z compression only element Z compression only element 2 Support S2 i 3 Support S3 Z compression only element 4 Support S4 Z compression only element 5 2,2* X rack / fuel assembly impact element 4 6 2,2* X rack / fuel assembly impact elemer.t 7 2,2* Y rack / fuel assembly impact element 8 2,2* Y rack / fuel assembly impact element 9-24' Other. rattling masses for nodes 1*, 3*, 4* and 5*
.G 25- Bottom cross- Inter-rack impact elements section of rack (around' edge) 1 Inter-rack impact elements
. Inter-rack. impact elements
. Inter-rack impact elements
. Inter-rack impact elements
. Inter-rack impact elements
. . Inter-rack impact' elements 44 Inter-rack impact elements 45 Top cross-section Inter-rack impact' elements
.- of rack _ Inter-rack impact elements
. (around edge) Inter-rack impact elements
. Inter-rack-impact elements-
.; Inter-rack impact elements
. . Inter-rack impact elements
. Inter-rack' impact elements 64 Inter-rack impact elements = '
O. ' 6-32
L- .4 +, l i 1 i O It l l' Table 6.2 (continued) 1 L NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS 1 L II. Friction Elements (16 total)
. ;- 1 Number Node Locai;.iSD Descriotion 1 Support S1 X direction friction 2' Support S1 Y direction friction
- 3. Support S2 X direction friction ,
L 4 Support S2 Y' direction friction
-5 Support S3' X direction friction 6_ Support S3 Y direction friction 7' Support S4 X direction friction 8' Support S4 Y direction friction 9 S1 X Slab moment 10 S1 Y.Jlab moment
; .. .11 S2 .X Slab moment
]'- 13 S2 S3 Y Slab moment--
X Slab moment a 14 S3 Y. Slab moment
'15 S4 X Slab moment
-16 L S4 .Y Slab moment'
, c i
1 5
~
e 6-33
- t Y -
10 Table 6.3 TYPICAL INPUT DATA FOR RACK ANALYSES (1b-inch units)
< (Region 1) (Region 2)
Module B Module C Support Foot Spring 1.65 x 106 1.91 x 106 Constant Ks (#/in.) Frictional Spring 2.338 x 109 2.338 x 109 Constant Kg (#/in.) . Rack to Fuel Assembly- .273 x 106 .472 x 106 (Impact Spring Constant (t/in.) Elastic Shear Spring for- 1.54 x 106 (x) 4.224 x 106 (x)
' Rack (9/in.)-- 5.75 x 10 (y) 7.746 x 10 (y)
Elastic Bending Spring 1.580 x 10ll(y) 4,43 x 10 ll(y) for-Rack-(4-in/in.) 4.203 x 10ll(x) 6.89 x 10ll(y) Elastic Extensional Spring i2.45 x 108 4.236-.x=107
~~
_(5/in.'), ,
' Elastic' Torsional fpring 8.298-x 109 1.517 x 109
-(6-in./in.)-
' Foundation: Rotational 5.668 x 107 5.668.x 107 Resistance' Springs KR
.(#-in./in.).
, : Gaps'(in.)'(for hydrodynamic calculations).
(hi, h3 are' ,x faces;4and hi hi 20. 20.
,+y
- h2 , h4 -are faces, h2 20. _ _ 20.-
.respectively) h3 6.125- 4.875 h4 20. 5.125 l
-o 6-34 a
1; ;s .
~
4
/
4 LOT Table 6.4 RACK MATERIAL DATA (200'F) Young's Yield Ultimate Modulus Strength Strength t Material E (psi) Sy (psi) Su (psi) 304 S.S. . 27.9 x 106 25000 71000 Section-III Table' Table Table Reference I-6.0 I-2.2 I-3.2 i SUPPORT'MATERIALLDATA (200 F) v& ' Material 1-hSTM-240, Type:303 - 27.'9 x 106 25,000- 171000
-(upper part.of support. psi- . psi psi' feet).
2 ASTM-~564-630. 27.9 x 106 106,300 140,000- <
- (age hardened at - . psi psi- psi m
, 1100'F). i S
h i
)
6-35
, 'a
, "*r 5
~
o _ a . o, m-
. Table 6.5 STRESS FACTORS AND RACK TO l FUEL IMPACT. IDAD .
. Rack / Fuel.
- Impact : Load .. (1bs)
PerLCell'at' .
- Worst Location Along Height..
. (Critical:
Run . Remarks . Location) Rf " R? 2 R3 Rg R3 R6 R. 7 B02 B rack, 8 x13' , f231.. .036 .023. .093 ..134 .178 .208 .025* Full,-Normal ._ Fuel,.SSE .154- ...040 .089 .056 .211 .221 .043**
- p'.8 u
B03 Same as'B02 212.3 .036. .028. .093 .135 . 167 .194 .021 except'. p= .2 -.154 .040 .073 .C'7 .197 .208 .037 F01- F Rack'8x13 I380. .024 - .'024 .111' .202 .203 .237 .023 i Normal ~ Fuel- ; SSE, Full
.164 .039 .062 .055' .179 .184 .035 y=.8-F02 Same as F01 380. . 024 --
.024 .111 .202 .203 .237 .023 -!
except p=.5 .164 - .044 -.059 .068 .186 .192 .034 j
- i
* ' Upper., values are for. rack cell cross-section just;above baseplate. ;
i
** Lower values are for support foot' female cross'-section just below attachment to baseplate.
- 1
. . . _ . - _ _ - i . . ,. i
. s .. . . , . . ~ . ,.& _ . J, . ...w ., . , _ ~ .__.m. . . , .
w-
..=
..~- ,, . ,
n
~
OJ - O I O :
^
- Table'6.5'(continued)
- STRESS :-FACTORS . AND ' RACK TO - FUEL IMPACT IDAD JRack/ Fuel -__
' Impact Load (lbs)-
Per~ Cell at-Worst-Location Along Height ' (Critical Run Remarks ~ .. Location) :Ri . R2 R3 R-4 R3 R6 R7 F03 Same-as F01 -380. .024 . .024 .110 .202 .203 .237 .022* except-p = .2- :.164 ..
.039; .0f .063 .199 .207 .034**
D51 D rackI12x15 '332. .033 .022 . . ' t, .112 .152 .172 .025 m Full'with-normal fuel, .255 .036 .091' .072 .328 .344 .046 6w SSE,:# = .8
' i' D57.. D rack, 20 283. .009 - .005 . .018 .019 .029 .034 .005 cells loaded (centrally) .041. .-.010 .015- .018 .054 .057 .008 y= . 8, SSE Normal-fuel D60' D rack, 1/2' 1256. .017' .012 .049 .105 .125 .146 .013 ;
Load,'SSE . Normal ~ Fuel- .129 .021 .045 .051 .155 .161 .023 in Positive ' X-quadrant
-g = .8, SSE D11. Same as.D60, 256. '017
. - .011
.049 .104 . 124 .146 .012 Except.~ -
p = .2- .129 .026 .060 . .061 . 172 . 181 .030
- l t7,4"- %.4 -Ai= - * ' '
.s m.,- .v .* We v e =w-- ,4.-qg,i%-.
3 [ f
~
.'g:-
Table.6.5 (continued) STRESS FACTORS AND RACK TO FUEL ' IMPACT IDAD
- . Rack / Fuel _ .
l Impact-Load (lbs)- ! Per-Cell at. Worst ~ Location Along Height-(Critical R3 R6 R7 Location) R2 R3 Rg. Run: Remarks: - R,
.017 .018 .029 .033 .004.
D56 D rack, 20 319. .009 .005. Cel1s loaded .020 .060 .063 .010 centrally, SSE .041 .010 .019
- e f Normal Fuel co - y = .2 .
.105 .057 .120 .152 .012 D70 D ra)ck, SSE 256. .017 .012
!- Normal Fuel
.134 .022 .050 .039 .166' .122 .021**
! :Ioad in Positive I y quadrant, l - y = : .2 D61 hD Rack, Full
-Load, Normal Fuel,-1.5 SSE Not Applicable.
to check stability (compare with
. D51), p = .8 he l
.m
_ _ . _ _ _ _ _ _ b
(
~
- Table 6.5 (continued)'
STRESS FACTORS AND RACK:TO FUEL . IMPACT IDAD. Rack / Fuel _. . Impact. Load.(lbs)
- Per Cell'at-
-Worst' Location
-Along' Height ~
criC cal
.Run Remarks -Location) - Ri .R2_ R3 Rg R3 R6 R7 D71 D rack, Normal 256. .017 . 011 .104 .056 .129 .151- .012*
Fuel-Load in
- .053 .188 - .198 Positive Y- .133 . 029 .066 .031**
Quadrant, SSE e y-= .2 , e
$ D55 D Rack, Normal 326.. .033' .
. ~020 .078 .111 .151 173 .025 Fuel, Full, SSE-p.= .2
.?55 . . 055 .116 .103 .356 .377 .058 .
t 4
- Upper values are for rack cell cross-section just above baseplatd. ,
** Lower values'.are for support foot female cross-section just below attachment to baseplate. l b
i e 6
.e , ,,,-m .a v w ,
>.y .- ..n...,_
--- :.y 'h-
} ., O - O .O; Table ~6.6 ! RACK DISPLACEMENTS AND SUPPORT IDADS -- l -(all'~-loads'are in lbs). FLOOR -IDAD MAXIMUM SHEAR
- l (sum of all. '. MAXIMUM IDAD AND support feet). VERTICAL ' COINCIDENT- ,
in a ack IDAD VERTICAL RUN . (x ' 10{) (1bs) (x 10- 5 ): IDAD.(x lb') DX (in.) DY (in.)** a I
- $ D51 4.065 l.441 .2118 (1.356) .0832 .0421
.1364 D57 .6562 .2324 . 0 3 6'3 (.2187) .0160 .0121
.0215 .0003 .0003 D60 2.055 .7301 .0991'-(,3861) .0367 .0281
.0550 D11 2.055 -.7264 .1214T(.6069) .0379 .0291
.0964 D56 .6562 .2388 .0443 (.2241) .0169 .0137 i
.0443 l
I
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- """"'71 -
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Table ti.6 (continued) i RACK DISPIACEMENTS AND ~ SUPPORT IDADS (all loads are in lbs). FIDOR IDAD MAXIMUM SHFAR*
'(sum of all MAXIMUM LOAD AND support feet) VERTICALy COINCIDENT IDAD VERTICAL 5
RUN in (x a 10{)ack -(lbs) (x 10 ) IAAD (x 10') DX (in.) DY (in.)** cs - D70 2.052 .755 .1004 (.32057) .0427 .0277' [' 4
.062- .0010 .0015 D61 NA NA -- i' NA + .0917 .0619 i ll i *
.0023 .0016
, '. s - . D71 2.052 -
".753 o .1274--(.6371) .0442 .0288
;. .1112 . .0031 .0031 i D55 4.065 1.439 , .26979 (1.349) .0784 .0495
. .21688 .0060 .0068 B02 2.407 .8726 .17877 (.62846) .0977 .0357
.16853 .0023 .0009 9
i B03 2.407 .8738 .1509 (.7555) .0963 .0469
.1001 .0092 .0185 i
l i l s. I I .
=
~
O .. O O: - Table'6.6 (continued) RACK DISPLACEMENTS AND SUPPORT IDADS
-(all loads are in 1bs).
FLOOR IDAD MAXIMUM SHEAR
- MAXIMUM- b LOAD AND (sum of all support feet) VERTICAL- COINCIDENT IDAD VERTICAL RUN in (x a10y)ack 5
. (lbs) (x 10 ) .'i IDAD (x 10 )
3 DX (in.) DY (in. ) ** t
.14937 (.56262) 7 F01 2.156 .9263 .1320 .0503 3 . ,
+ '.19116 .0030 .0012 i
FM 2.156 .9263 . 165714(.05001) '
.1320 .0503
.10138', .0030 .0012 !
i!! F03 2.156 .9250+ .15231 (.83002) .1363 .0523
.09915 .0076 .0068 -
l, i
- The value in parenthesis is the vertical load at the instant when the shear load is maximum.
The second.value in the column is the shear load when the vertical load is maximum (or near maximum). The maximum vertical and shear loads generally do not occur at the save instant. '
** Upper values are top movements; lower values are baseplate movements (not necessarily at the '
same time). l i L i i
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! TECHillCAt. n'flCT10f!S ?!ueber I-p]g pl 017:SI0ft ES-02:
Title Revision no. Seismic Criteria 3 i 1 l l~ EXHIBIT E T-02 , t I
. ,. ,,- _p *
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TMI-1 SESMIC RESPONSE Aux /F.H./ Control Bldgs. Elev. 329 Ft.- O In. ; 4
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..i..... ,,. . . .....,, .. !
$.00 400.00 800.00 1200.00 1600.00 ime Ste as (.O' sec./s :e a) L l FIGURE 6.1 '
. t
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- 6E g! I li II i l
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, ill li l ll } lll i ll ll ld
- 4d i Il l ll l l E I I I OE d
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- q. :
7:: o3 TMI VT SSE AC CELERATION TIME HISTORY ! .00 4do.o'o ' ' ' ' ' ' 8do'.o'o ' ' ' ' ' ' 12bo.do' ' ' ' ' ' 16bo.oo i ~~ime S~e as (.0 ' sec./ step) l FIGURE 6.3 I
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9 m: N- ' o: o-T ! d: f . N: o:
- 'q( ., s %
b i o- 11 *
)
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- b Lij : i
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j o: oE TMI SSE F 1 SPECTRUMC OM 3ARIS ON 15s Dampin g o- . d i 2 .1 10 11 10 rR EQUEN CY (-z.) FIGURE 6.4
.i
O O O o 9 m: N: c: o-I 11 - d: i N-b l [ o: o-
, Y1 l I
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- i o TMI SSE F 2 SPECTRUMC OM 3ARIS ON 1% Equipment D ampin g -* - * - + -*-+-+ ++-.
l Q~ , m 2 l 11 10 10 REQUENCY ( z ) . FIGURE 6.5
O :O O o o I (d : e_
~
0 o I o: - f W: r_ II k j _ is ' O 1 41 li K h oo: - I ll N
"l
.- % P y w w -
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- TMI SSE VT . SPECTRUM COIAPARISO 4 1 s Equipme nt Daraping c:
- 1. o. -
+ 2 l 11 10 10 l
j R EQU EN CY
~
(-z.) FIGURE 6.6
i O JO O L oi o! , , ; . ll 1l liil, i li i l I i11 di :
- i. '. :l
- n
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l il ; ' I
. !l l]! ll %Eo- -
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!(j o
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' E! i l TMI H1 OBE A CELERATION TIME HISTORY h.0d ' ' ' ' ' 'idO.00' ' ' ' ' ' ' ' 53dO.00 ' ' ' ' ' ' ' ' ' bO.dd 12 ' ' ' ' ' 16bO.00 Time S:e as (.O' sec./ste a) . FIGURE 6.7
p O - O O P l O[
' ~
i O l l l- l
] . ll ! II bl'IbilN]bLIllllI h, hk*~
I i O o: O e-5 em ! ! ! . ei l l l . YE l l ! af TMI H2 'OBE A CELERATION TIM HISTORY
.0d'4dO.$0' ' 8 dO.dO ' ' ' ' ' ' ' 12bd.bd ' ' ' ' 16
' bO.00
, ~~ime Ste as- (.0' sec./s :e 3) FIGURE 6.8 j __ . - . --._- _ _/
o
;o; ;
o o. di
'O i1 I il ! ! E! ~
ll O d- (II ' I l l ! O I-: ' O
- ob -
b ! in: TMI VT OBE A ELERATION TIME HISTORY i h.0d'4dO.dO 8dO.dO 12b0.dd ' 16b0.00 ime Steas (.0' sec./s ea) FIGURE 6.9 ,_ _i
d'- 0; O O , o-o_ e y_ , o- ' a.: N-
"5 'd' MQ i :
i
- i. Ai
! II ; Oo: l 1
- 0 0: I \ h I b (0 .$
b \Co ~ dL_.-I l ;
. _ o- ;
v,f : i l i 1 - i l Ld -
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o-1 t 1 - a: o-TMI OB E H1 SPECTRUM COMPARISON 15s Equipment Damping , 4 . . . . . . . ...i 11 10 10
- rREQU ENCY ( z.) .
FIGURE 6.10
~
^
_ . _- __ _ . _ _ _ . . . . ~ . . . _ . .. ,_ . . . _ . . . _ . _ _ . . _ . . . _ . _ . . . .
e # O O O - o o,_ , e4 - s- -
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O-t W O.'_ 7 o_ _ l o- TMi OBE' H2 SPECTRUM COMPARISON 1 s Equipment Damping o-4 . . .i . . . . ..i; 11 10 10 rREQUENCY ( z.) FIGURE 6.11
-O- .O- O o
9, o: o:
. . o. _ I m:
1
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( 9: I o: , i : I oE TMi OBE VT SPECTRUM COMPARISON 1 5 Equipment Damping : o-g . . . . .
..4 R EOU E \ CY ( z.)
t FIGURE 6.12 j
,, - =- --. , -
, ~-c- . - - . _ - - - - - - - - _ _ _ . . - - - _ - - - - - - - - . - - _ _ _ _ _ - -
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- pg CENTER LINE A H/4 H ,
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=
TYP. FRICTION ELEMENT
,;/ .
e , ' .V a
- SCHEMATIC MODEL Fog cyNRACK 6-58 FIGURE 6.13
TYFICAL TOP IMPACT ELEMENT o j -
$W 6 i
%m 5:
/ -
/ / /
< 's l
RACK STRUCTURE O T(P. BOTTOM IMP ACT ELEMENT. f
/ /
'A
'W p
g$w _ M h
/
i h ' E 2 f ir.r . . 1- .
. f~\
.L' {G)-
, RACK.TO RACK If.tPACT SPR!flG3 1 '
. 59 FIGURE 6.14
~
-- .-e,., - - . . . ,
i 1 i
.\
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. 4 .
\
CELL WAL L l 7 I P M ASS i XB A --
,'l uf
/ \
f FUEL ASSEMELY/ CELL ! d IMPACT SFRlNG l f .
/ '
..~ /
/
/
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/
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i
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s
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,1 - l lMPACT SPRING ARR ANGEMENT AT NODE I -
-{
6-60
. - FIGURE 6 .1 '- l ,
6
-O O i 5
l I 17 ' i l ; I ! qi l 20' 7, P 4 L 9 3g 921 l A. Y 92 k 93 1 FIGURE 6.16 1 l i 4 l DEGREES OF FREEDOM MODELLING RACK MOTION
. _ _ ~- - --n- -
e ~ w a _ _ _ -_ __w_ --
- - _ _ __w____
o
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; ji [
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- i. ,
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95 . i i I I I i I . FIGURE 6.17
. RACK DEGREES OF FREEDOM FOR I_Z l' LANE BENDING Qa
l ;: . I' O 0 2 q 8 1 1 8 H O G N I D - N E B E . N . A L - P LE Z Y - R . - O F - M o O D E E 8 R
! 1 F -
6 5 F O E - R S U E E G R I - g} F G E D K - _ C --
- A R
_ 4 q- - 2 A ~ i i C e ~
. ~
a
-0 N: - ,'
=
r , mAm w
,. ,t RIEL ASSY/ CELL IMPACT' SPRING , K ;
o s '" 4 3 ; 0.25H F/2 <
'M3 W "
b 0.25H
, PJCX N v y
\' 4 s y L b 0.25H
- H/2 s'
-.y y O .
v i TYPICAL RATTLING PASS 0.25H i, 6 *
. FRICTIO.4 '
'INTI?JACE mob 9 SUPP0F.T LIG SPRING, K S- I
. SPRING, I -
f N O~ b
, ,,,u, .
' FOUNDATION .
ROTATIONAL CCSLIANCE SPRING, K ' R
=
. . (, ' FIGURE 6.19 2-D kIEW 0F RACK MODEL 6-64
- m - e.--', - -v - - - , - <----n - . -- - -- - __._ _ _ -- - _ --
O l l 7.0 ACCIDENT ANALYSIS and MISCELTANEOUS STRUCTURAL EVALUATIONS l l 7.1 Introduction i l The TMI Safety Analysis Report has presented results of analyses of several types of accidents which could potentially affect the
, spent fuel storage pools. Installation of the proposed high l density racks will enable the storage of increased amounts of l spent fuel in the TMI plant spent fuel pool. Accordingly, accidents involving the spent fuel pool have been reevaluated to ensure that the proposed spent fuel pool modification does not change the present degree of assurance to public health and safety. The following accidents have eeen considered i
O Fuel Pool - Earthquake Loading Loss of Water ) 0 Fuel Storage Building - Earthquake Loading O Refueling Accidents - Dropped Fuel Dropped Gates ., 7.2 Results of Accident Evaluation 7.2.1 Fuel Pool The ef fects 'of - earthquake loadings on the fuel racks and spent fuel pool floor are discussed in Section 6.0 of-this report. The loss of - cooling water in the spent fuel pool is discussed in Section 5 of this report. i f 7-1 s
; O 7.2.2 Fuel Storace Buildina The ability of the Fuel Storage Building to resist earthquakes has not been affected by the spent fuel pool densification. Therefore, the information presently contained in the FSAR is still valid.
4 7.2.3 Refuelina Accidents This section considers three (3) accidents associated with the handling of fuel assemblies, the movement of inter-pool canal gates and movement of miscellaneous equipment over the pool. 7.2.3.1 Drooned Fuel Assembly The . consequences of dropping a new or spent fuel assembly as it is being moved over stored fuel is k discussed below.
- a. Droceed Fuel Assembly Accident I A fuel assembly is dropped from elevation 355'-0" above a storage location and impacts the base of the module. -Local f ailure 'of the baseplate is acceptable; however, the rack design should ensure that gross struct' ural failure does not occur and-the suberiticality of the adjacent fuel assemblies is not-violated. Calculated results show that the fuel assembly will not hit the liner and that there will be no change in the spacing between fuel tubes. It is also shown that the load transmitted to the liner through the support is appropriately .
dif fused .. through the bearing pads located on the liner. Although local deformat;.on of the baseplate occurs, it is demonstrated that the liner .is not damaged by impact. 7-2
i 1 l
- b. Droceed Fuel Assembly Accident II One fuel assembly dropping from elevation 355'-0" l above the rack and hitting the top of the rack. 1 Permanent deformation of the rack is acceptable, but is required to be limited to the top region such that the rack cross-sectional geometry at the level of the top of the active fuel (and below) is I not altered. Analysis dictater. that the naximum I local stress at the elevation sf the top of. the active fuel region is less than material yiel.d point. Thus, the functionality of the rack is not affected. Although local deformation occurs, it is i confined to a region above the active fue) area,
- c. Droceed Fuel Assembly Accident III l
This postulated accir%nt is identical to (a) above I except that the fue?, assembly is assumed to drop in I an inclined manner on top of the rack. Analyses show that the straight drop case. (case b above) i bounds the resultr.
]
l Other heavy. objects carried over the fuel racks are smaller r~) - in weight than the fuel assembly. Therefore, fuel assembly
.q
.l drop scenario governs the rack design. Thi:: analysis limits
.l 'the maximum elevation of the b'ottom of all heavy objects l carried over the pool-to.355'-0". -
Analysis of Drop Case-I
- indicated ; that- ' 3 7 the~ drop were: to- occur on to a cell directly' above ~ a _ support leg' then the _ local slab bearing stress will be exceeded. -Bearing pads are interposed between the rack feet.and'the: pool liner to diffuse the pressure, and F t
" . bring the' pool slab surf ace pressure _ to within the ACI ,
. allowable value. ,
7-3 1 I w - - _ __- - +
N . [
.O 7.3 LOCAL BUCKLING OF FUEL CELL WATT.S This sub-section and the next one presents details on the secondary stresses produced by buckling and by temperature effects.
The allowable local buckling stresses in the fuel cell walls are obtained by using classical plate buckling analysis. The following formula for the critical stress has been used. Tr2 ' E t2
+ Ocr "
12 b2 (1 . p2) where E'= 27.9 x 106 poi, y is Poison's ratio, t = .075", b = 9".
'The factor is suggested in (Ref. 7.3.1) to be 4.0 for a long panel loaded as shown in Figure 7.1.
'l n
t 0 - 7-4 N&r
t i I f. q.] 1 eoe the given data acr < 7005 psi It should be noted that this calculation is based on the applied N stress-being uniform along the entire length of the cell wall. In the actual fuel rack, the compressive stress comes from consideration of overall bending of the rack structures during a yr seismic event and as such is negligible at- the rack top and c maximum at the rack bottom. It.is conservative to apply the above p u , equation to-the rack cell wall if we compare a tcwith the maximum [ compressive stress anywhere in the cell wall. As shown in Section ( 6, '.this -local buckling stress limit is' not violated .anywhere in - ( the body _of the rack modules, since the maximum compressive stress 4 in'the outermost cell'la o = 15000
- R6 (from Table 6.5.with R6 =
, .y .237)~= 3555-psi.- 7 < t .?t i
- h. ." 7.4 ANALYSIS OF WELDED JOINTS IN RACK ;
LQ L, - In-rack ' weldsd joints are exa; %d under the loading conditions l ' arising . from therma . ' eff acts' due t o an isolated hot shell, - in this sub-section.- i A thermal J gradient e. between cells . will - develop ' when an isolated
,atoraga- location contains 'a > fuel assembly' emitting maximum-postulated heat, . while:- che surrounding 11oc~ations L are - empty. We can - obtain a conservative estimate of ! weld stresses along; the L length of- an. L isolated hot . cell by considering- . a. . beam. strip. ;
H n
-' uniformly . heated by 60 F, and ? ' restrained' f rom . growth ' along one-lLlongedge(Figure 17.2)..
M l 4 1! C w. f . , l L m
. . . _ . . _ _ . _ . . . _ _ __ _ . _ . _ _ m __ .- _ _ _
- < < rl ,
i I O"t . Using a ' shear beam theory, and subjecting the strip to a uniform { temperature rise AT = 60*F, we can calculate an estimate of the maximum value of the average shear stress in the strip. 1 The final result for wall maximum shear stress, under conservative l
. restraint assumptions is given as ;
.1
-p I
,, EaAT l
.g Imax "
v..
.931
' {
where E.= 28 x 106 psi, a = 9.5 x 10-6 in/in F and AT = 60 F. Therefore,'we obtain an estimate of maximum weld shear stress in , g .an isolated hot cell, due-to thermal gradient, as-
, , rmax = 17325 psi -j q" . 'Since '- this 'is a secondary thermal stress, we use ' the allowab!a O!
.shearf stress criteria - for faulted conditions as a guide (r <
V .42Su)* . a r
,7.5 ' REFERENCES FOR SECTION 7i p
7.2.'1 " Stanbards ~
~
of ' Tubular Exchanger. Manufacturer'a-
' Association" ,L6th' Edition, Section 12 (1978).
7.2.2 " Fluid Mechanics" , by-M.C. . Potter and J.F.,Foss, RonaldL '
." ; j;c i (1975),-p.'454..
'7/.3.1 " Strength of Materials", S.P. Timoshenko, 3rd Edition,. :
, - Part II,ipp.194-197;-(1956). .
4 ( l / k 5
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b + c. b ).4
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..=- +-t -
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{O: ) 8.0 ANALYSIS OF FUEL POOL STRUCTURE I 8.1 Introduction ;
, The TMI-1 spent fuel pool is a nuclear safety related, seismic W category' I, reinforced concrete structure. Five percent (5%)
- y. structural damping for SSE is prescribed. The object of the
"- analysis la to demonstrate the structural adequacy of the pool I slab structure to withstand the increased loadings due to high
, density fuel storage. The loading on the pool structure can be t broken down into the following discrete items:
a . (a) Dead weight of the pool structure.itself.
'(b) ' Dead weight of rack r~ tules and that of the stored 4- fuel. .
... -i
'l E
(c);' Weight of the water mass in the pool.
-l[ U(d) Time . dependent vertical and horizontal (shear) 1.>rc e S transmitted.by. rack support foot to the' slab during an . e ,
l~ " SSE or-1/21SSE (also termed OBE). 7(e) _Self excitation of the slab.and the. water mass-during an
~
SSE or-1/2-SSE (or OBE).
. .(f)--Pressure. l loading 1on sthe:: pool walls cause'd by.; fluid coupling effects.
- v. .(g). Thermal gradient:across pool slab and walls..
i i s, H Thog TMI -fuel; pool is analyzed.. using: a' finite' element modelling: \ 'l. ; _ j~ l: 'scheme. ' Load combinations mandated by NUREG-0800,; SRP' 3. 8k 41 (Ref.
, 8 .~1'. 1 )1 J a r e - applied: and 'it is ' demonstrated that! structural;
' integrity is maintained when the fuel pool-is assumedVto be fully: j J1'oaded L with .high . density: fuell racks. and; all storage ' locations : are' -[
8 ' I i
- l:-
y . i N' , N, . : -------A-
't.-
y; j
, y- occupied by fuel assemblies. This implies that the structural j t J .
analysis is:1 carried out assuming that there are 1494 cells loaded j
- with fresh or irradiated fuel assemblies in Pool A. A general I purpose -finite element code (Reference 8.1.2) is utilized to perform :the analyses.
' /
l The critical regions examined are the fuel pool slab and the most
~
i critical wall sections adjoining the pool slab. Both moment and 1 shear . capacities = of the' critical regions are checked for' i structural integrity. Gross section integrity under bending and ahear is evaluated. Also evaluated are. local punching and bearing integrity in - the vicinity of a fuel rack pedestal. Structural ,
' capacity' evaluations are carried out in accordance with the !
V _ l' requirements. ' of ' the' American Concrete Institute- (Ref. 8.1.3); as i . l. delineated in - the Safety Analysis report for TMI ' Unit 1 (Ref.
- l J 8.1. 4 )'. ; lit ' Lis aot'ed , however, that factored loads specified in-I L
reference' ( 8 '.1.1 ) . are- more severe- than those used in- the' ! applicabl'e L ACI codes n ( 8.1.3 ) . Therefore, in this analysis, the ! LA ; ' higher load'. factors!of SRP-3.8.4 have been used together with the; q h allowable concrete 1and i reinforcement .' loads as called for in the m , LACI. Codes. This constitut'esuthe most conservativeLapproach to the , l structural qualification.of the. pool structure.
- 8.2 'MODEL b'
s y Ck t JThe; fuel' pool modeliis - constructed using iinformation from ~ as-H ' b'uiltL THI-1Estructural: drawings . . Afdescription of the-portion of: " l, the:. pool.modelled'for; analysis is ne 'ven.
- j. '
, s >
- d f
w >
.[
.f-.
-)
s . u { 8 4 7
- m. .
I I 4 Q.
..__.-___-------__--------------2-------L--
1 l 9 A fuel pool slab is a 5 foot thick reinforced concrete slab 24 ft. wide and 63 feet long. The slab is located at elevation 300'- 305' and its long direction is aligned along the North-South l direction. The West edge of the slab is supported by a 5 foot
; thick vertical reinforced wall which is fixed at the mat level (281') and extends above the slab to a level exceeding 348'. The East edge of the slab is supported by a 5 foot thick wall from mat level (281') to the slab, and by a 9 foot thick wall from level 305' to 348'. The walls extend higher than the 348' level, but we y end our modelling at this height and assume free edges at this level. The north wall is a 5 foot t. hick wall extending from the slab to level 348'. There is no wall below the slab along the north. edge. The south edge of the slab is supported by a 5 foot thick wall extending from level 281' up-to and beyond level 348'. l 3 p It is-clear from-the-above description that the West Wall has the '
t largest length to thickness ratio, and therefore, will present the limiting condition of structural strength, i 1 The pool slab is' assumed to be loaded with -12 high density fuel s racks having a total of 1494 cells. Each cell is assumed to 1 containc a 1550 'lb. weight normal fuel assembly. All fuel - pool walls above the pool slab are assumed toshave-a free. edge at level j
, , 348'.-- Horizontal restraint ~is provided-to the-West wall at two !
vertical lines of ' nodes above the 305' level. This restraint simulates the effect ~ of adjacent wall' . structure : which is not modelled Figure 8.1 shows - a' layout of the entire 3-D finite k element -model. The gridwork in . dif farent regions .shows the totality of elements used.
- j Jih ' 8-3 h g m ;
ti LA-__=_------_---------------.-----_ - - - - - _ -
7 4
)
i r ' h 7' l
?i- The finite element discretization of the pool structure consists D QL- of 1142 shell elements, 1180 nodes, and over 6600 degrees-of-freedom.
L l T The material properties for the shell elements used in the finite j element mod.'l are calculated using standard procedures for '
. reinforced concrete sections and reflect the different levels of reinforcement present in different sections of the pool structure.
' 't ;
-Accordingly, varying material properties are used for different regions of the structure. l 4 ;
g I 8.3 LOADING CONDITIONS J The following . six finite element analyses are carried out .to a qg
- obtain' the- response of the pool structure for 'six different y floading conditions:-
- c .o ' 1. . Dead loading from concrete, reinforcement, and 40 feet
;A- 'of ~ hydrostatic head. The loading ' is applied as a - lg ;
E M. ' vertical gravitational load 1 for the structure and a
, surface pressursion slabs and walls for the hydrostatic .
T surface pressure -at level '305' isi 2496 UM , Thead. ilbs/ft 2,he:
- 2. Dead' loading due:to weightsJof rack plus'fulltfuel. load.
t;
- P . These ' loads; are applied 1 as concentrated ' loads- at floor.
m:. N' slab node points . at the - locationF of . the. 48 fuel rack
- ,, . pedestals (four: support legs per. rack;; total of 12! '
T::, 5 ll racks ) .* D These : loads . are . obtained - f rom the results . of-j,S the rack dynamic analysis' described in Section 6. g g: .- E i
*r
%"[, *= Thel foot- loading ; from Module ,E .which has five support legs,: 3 e ; .:has also: been represented' by four loaded locations by coalescing. .,
~'two.. adjacent leg: loads. +!
~
_"L-lpp + '
-O ;i g
8 4 20,1 > t . . m a
,/.s ,i- ,!
/>.; ' _
,b ) '?
5., 1
- 3. Seismic vertical loading due to racks plus fuel load applied at the 48 pedestal load points. The loading applied is obtained from the dynamic analyses .used to qualify the rack structure and is adjusted to reflect the randomness of motion of the dif ferent racks. This loading represents the results from an SSE event and is appropriately scaled to simulate an OBE event. The peak (instantaneous)' loading from the worst pedestal in each rack is used for additional conservatism.
i 4. Seismic horizontal loading due to . structure. weight, reinforcement, and hydrodynamic head applied against the , Wese wall. The loading'is applied as a .lg horizontal ~
, accele. ration- to the structure' plus a hydrodynamic - pre:sure applied to the West wall to maximize stresses 4
in the wskest- location in the pool. ,
, 5 '.
A. thermal: temperature gradient of.57'F"is applied across
-the: walls and floor slab'to simulate the heating,effect of ? the ' wat.cr ; in the pool. . This gradient is calculated ,
based or. 9ha maximum pool water temperature anticipated 4 X_. o dor r
,ted in Section-5.of'this report.
. 6.- ~A frequency analysis of the pool structure assuming that t- '
.allicoritained: fluid moves vertically with the pool-slab..
fQ, h This frequency- analysis is Lused. to ' determine the ' appropriate.cseismic amplifiers to apply to load' cases'l and:41to: simulate SSE or OBE. events. m # We!- will1 Lrefer ~' to the ab'o vel loading conditions: as . case il L ~t hrough. case'6, respectively;
>OfDall loading conditions > mandated?in1(8.1.'1), the. conditions.thati <
g.V applyltotthis structure.that-are deemed-critical: ares
, ; t J , '
, A.. :1.4D W 1.9E: o l' y 'B '. . .75i(1.4DL+:1.9E + '1~.7,To-l o ,
CM _D'+ E'~+:To: .
"j Ewheresi'
~'
j < m EDi ^r$) dead. load! >
" '~ .
-{
E' = Safe = Shutdown. Earthquake J ' E' = OperatingLBasis-Earthquake. ! 4 1To.
'= SteadyiState thermal:. load 1 e ,
;6 -
, 7 i
a . ,
! +
l .
- l. .
- u. 1
)I W
.c ye, g L ., , , , . y .- .--i '
---------------------------~--
- , .i
, m _ .
- .- ~-
M, - f{
!ut ,
m
' ,. . . r4> -
N
-The appropriate-load = cases are formed from the individual finite
- w M element analyses as follows: -
m D = = case 1 + case 2 E' = SSE amplifier x case 1 + SSE amplifier x case 4 + m4, case 3 '
^
E. = OBE amplifier x case 1~+ OBE amplifier x case 4 + .p p OBE ~ reduction factor x case 3 To- = casez4 $'< 4 I Load combinations are formed-in the appropriate manner'to maximize
-critical stress resultants.
,a '
l llN ' 9' For-analysis of_ fuel pool bending _ moment dit.tribution, the seismic gu Lamplifiers were conservatively based on the peak g-level responses ! f i' f from,the acceleration response spectrum given in (Ref; 8.3.1). The 'T oM ,m use:of this maximum amplification:isia direct consequence of:the- <
%y '
~
.o, i. assumptionLthat.,allLof theEfluid mass responds _in' phase 1with the.
J P "4 o
', lowest vertical response of'the slab. g .N ? h ,
y h)hb a
.a 8.4 RESULTS OF ANALYSES .
,Mff!!y. * '
, l:The['ANSYSj po'stprocessing? module is ' used; to . form the1 appropriate: j 41oad! cases 6:Tand . the i bendingj moment' distribution .in .the critical. ?
tareas a checked 0 againsti the ; allow ableD limits .Z The Eultimate S moments -
~
J Q $ ifor[esch isection: are : computed in accordance with allowablaf stress
]
j his!G, ,% t - - -
.. .. a-j-
f b;y. levels.;given;in (8.1;3'and<8'.l.4). The: allowable'stressclevels:forf "y
;%:e 'THILUnitLl?are' -;
- g. x ,
*s , , ,
uy
- .!- +
s.concreteL __
=. 5000/ psi : - .. . r reinforcement =
~ 40000 ' psi :( for till . barf'or staller)1 7'
Q Syf'.W N"
.7 .. . . . .. t m
2 ' f/t yp l$a ,h n$ 4WeLcheck the%safetyffactor:-(defined to:beithe allowabluivalue.ofJ , Q jg[!$n :
~c . . . . . ..
- bendingimoment:dividedLby/the calculated value.ofLbending(moment)L R ' ;B f. L l l.ab critics 1c regions! of the . pooli structure.. Table--8.'lishowsLthe i 4mmA, - -
, .. y. h ; ' ' t l[I'W
- h. h '
Cg[!j a j tl( - f:p , 's.e{ ' [k. ,-
. 1 ~.. t j
,{
D ' m t96
. rl
'og .u lT; l a : {l
}
s ; m.. ,y , , j/N .[ , [. M <L N blNNNOL
'4 > "
P *
'(
l .?
=
x ' ' l. l.. l .. l' - X results obtained. Note that these are safety factors based on the ! b -factored load conditions as mandated in (8.1.1). l-l The floor slab perimeter is also checked against gross shear u failure under factored load conditions. More realistic values for [ seismic amplifiers ' are obtained for this shear calculation by f
, l using the methods in Ref. 8.3.1 to estimate the vibrating fluid !
mass. The lowest vertical response (frequency) is calculated at 16-18-HZ and the seismic amplifiers are based on this frequency
. range. Local. bearing: strength and punching shear calculations are l performed in .accordance with Ref. 8.3.1.- Table 8.2 presents I .
results of the shear calculations. , 8.5 POOL LINEP.,i 1 The" pool; liner.is~ subject'to in-plane strains due to mcVement of 3 the rack. support m.m .. feet,during the; seismic-event. Calculations are
.made to establish .- , thatEthe
. , m .Flin,. # , , .e,rmm.3 will ' not fall due to cyclic j
,O .f straining - caused ' by - the"rac.R, foot ) movements . It,is found that the < u
- - .s
' cumulative damage f actor is r wel1~ below.;thF ASME . Code -limit of one,- l 1
even if c 1,000LeydlAs"of gc strainind"2.niislasisumeh. . x ;. . n. .. u. . . g t I.O se j j ' h: , if '-l 4 ..-e y , 8 .' 6 CONCLUSIONS w ;gigy:ya ;j3c.;wW,. Q"j'Q _..mu .gyf
, ;i. 4 yrc 5.s. c $ + W.:tseg- ^
Critical #regionC affocted' b~y76 Edin'g.';thef fuislipool' completely with - _ j m F ' "+ [' ' J high' density ^ . racks ! are d. . xamiAAd.[~..f, oT,#.T AtrucI 4 . . . - . . , ,, 1$ integrity 1 under
- . H
; bending; 'and shearing. ; act1ErCw'Iti.is(.,,date
...>>- . . ~ .
M,dthat.. adequate-h
,s "
s'afety) factors exist?assumiwngthat*kil"rackiiT.. ,. m ~ragfully leaded with . Mj m . .
- o. .
. .. ; m. _ ,
, normalN fue'l? and: that the f actored iload L combinations are checked
.# e w, 4 ,i o-m .:)%against thA -appropriate stiructural' dAsign[stiengths. ., . ~ . .
3 N. c.f;i ' '
. , s'd W *,, f p 3 ..u- ) - V. U
- tML,,4, , M, .4. J.ais _ . .; .i .. .i =41.ptu
. u.@ '
.@J
' Thh - '[D".d I--"*'*ip}
M[ * ' "*f W # ' ?/. -
)
W t i e > 'j si 3 1
, i:
2,
,.u ' ' .. .
,8, 7 3
t , i-m
.p ,t .
IIN. l , i; / , diMG q j' l y' <-
' i, ' j; i i
- s. .- z. n a r - - - _ - - _ _ __ = _. : -_ . - _L
f O The horizontal rack reac'tions at the liner pedestal interface (developed doc to friction) can be shown to have a negligible effect on the gross bending moment distributions obtained. These local frictional loads are checked to ascertain their local effect on.the liner. The liner is shown to have a large margin against fatigue failure.
- ,j
.o ll r ;
l i: I J b e>, , f i l: I , li , I 4-h L 8-8 L n - l/ 5 [ 1, .
. .. . _ 1
- 4. ,i t
i 8 '. 7 REFERENCES FOR SECTION 8 8 .1 ' NUREG-0800, SRP for Review of Safety Analysis Reports for Nuclear Power Plants, Section 3.8.4, July 1981. t 6- 8.2 .ANSYS User's Manual, Swanson Analysis Rev. 4.3, 1987.
. 8.3 ACI 318-63, Building Code Requirements for Reinforced
. Concrete, American Concrete Institute, Detroit, Michigan. i
-0.4 -GPU Nuclear: Three Mile Island Nuclear Station Unit 1, Final Saf ety - Analysis ' Report (updated version), Volume 3, section o 5,c7/89. i
% 8.5 . Fritz, . R'.J. ,- The Ef fect of Liquids on the Dynamic Motions of 1, ' Immersed' -Journal of Engineering for Industry,
.g LFebruary, f1972, solids,-
pp 167-170. [ i:
)
, i e [._
IM ' t i
.,'.l /
,I . $
o n,
=
.a .;
L u 4 e , l ,,5
,kh '(
fk q '_' l3 l
) {
l j 4.h.- l.1
't 8-9 ,
U -'\_ p)7 s t .I . . -'
.- . . +
t i 4' q%. c.
-j j
4 r j
'4 , ,
1
.% 'u .
L(~~W
. . , u ;-
m) - n= , s Table 8.1 )
@/ o- SAFETY FACTORS FOR' BENDING OF' POOL STRUCTURE REGIONS ,
)
. i
' yty;- !
J REGION FACTOR OF SAFETY l
.y
-}
, (n
' ,i; L ' 4 Slab bending-at West 1.28 I Wall,_(E-W) l 1/? , -Floorislab bending'in 1.29
'D ,.- Center'(E-W)r
- ql
'1
,Fborfslab bending at 1.77 1 Enst; Wall.(E-W)" .
& ni o
- MM,
'F10or: slab (bending' 1.69 h4 Mt. SouthiWalls(N-S)'
o l
, .Fl'oor slab bending at 1.34 t
, g[ ^ ~ .C nter.:(N-S)i L . $m$ % .t . , ifY' Floor; slab { bending at--
-. North Walli(N-S) ,
'1.40 3,,
mm 4 ; ,1 s . . -
,3..
off
- i
": West ! wal'11' bending .' above' 1. '0 4 * ' :g floor? slabs -M am
}$ , , ., .i ,
West'. wall bending:- .2.28 'i
, bel ~ow floor.si'ab q
,..- g_
. g l
, iVery ; conservative ' 'and lower ' bound ~. value since i considerable. 3
. local, 'section and1 concrete - reinforcement .are neglected in Lthis J
c' calculation. '1!! T ' 4.Ai
,.p i
'M
'Q, . 1 l
, .)
xj- b 5;ki 8-10
..d " D i
4
/ 1 .
6 i j l w . 1 t
., f[ , ,, , , . .,-
~
[. w )/l$b-[. l 2 .
%;ic , ,
s r y'k' tf '..
,t y 5_ ,
'l' y
l lt ll : '
-( i t- l. ,J i -
1...r, . Q i_ ,,tr. .y -'
!6 'U .
s < 4 ( 1 ' t Table 8.2 sj
,.l -'
POOL SLAS SHEAR SAFETY FACTORS-1 e t.- ., i 3.i ' w/ jf ,
.- 1 M FACTOR OF SAFETY ;
ka- s
,'. 5 i ,
E' .d k( Uf#y,
; . Floor slab gross shear- 1.257 .;
1 o Bearing capacity _under: 4.13 g@e pedestal .>
.o 1 5 fn f .I,~ T
' j il r; 4 *c
, f Punching-Shear 3.59- 2
. .~
's'.:.
,g . I j! '!b ' ' f ! <;,g" ,9
' ;<r r , s .
A. \ ' > g , [jk+'e i(g}p,)i
" 0; ,
.t f, '.
. - - g Q. m-)!,;g f, ,
r ; f i ', : ;5
. '.c.\ \ ?,
\, a ,*
- ifft,
't.
.e
*I 'k i . h j'. j -
y, h, OihD :. I
? y q;-;
.p,- 1 r - ' 4 .
n- ' } g In=,r / s
'1A.\ \ <
b
- l 1
't 4 -15_t -} r 4 , , y<
t e t 1 ;4
- 1 I.
, . . ' .,; ] ?
/I.s. ,
-y -ti i,;
$i J i <
) I t '
t t,
.? s y s s
i ... 8-11 - ,., ..W:. ,
. ,'f' l
c i ,i P .
v y s- o i
+, ;
- u.
n 1
-, ) p .-
, 1 i .i , l
, i , , ,
, , . . - _ ___.______4- - - - - _ . _ _ _ _ _ -
/ O O O l yl1 I / s
,L f-N s
s M Y :. . ..
" .' 2: -
;2 .. .
' 1 918
.ff Nr ST=5 .!!
ll /> . 5 E_EMENTS 4Y., . FIGURE 8.1 OVERALL FINITE ELEMENT MODEL OF THI SPENT FUEL POOL
ru , n
.j ;'t t
J r [ , 9.0 RADIOLOGICAL EVALUATION l l v 9 '.1 Fuel'Handlina Accident ; E i 9.1.1. Assumptions and Source Term Calculations -!
+
; An evaluation of the consequences of a fuel handling _j g accident has been made for fuel of 4.6 wt% initial enrichment
. burned to 60,000 MWD /MTU, with the reactor conservatively assumed i f> 3 ms'* ito have been operating at 2568 MW thermal power'(30.97 MWD /KgU~ l,
- u. .
h, ? specific: power) immedioi.ely prior to reactor shutdown. The fuel j
- handling-accident wa's assumed to result in the release of all R 4
gaseous' fission proaucts contained in the fuel-rod gaps of the. j outer row of' rods 'in a _ fuel assently. at- the time' of the accident 1
' , , a j3 ,
(565 rods assumed damaged) . Estimates of the gap inventories were made' using a both the escape rate method and the assumptions Lidentifled in Regulatory Guide'l.25 .
]q i
Most of the c gaseous fission products having a'sig-nificant impact on theHoff-sitet doses 1 are : the short ~ lived ' d nuclides'ofLIodineLand-XenonLwhich reach saturation. inventories; .]
. during in-c' ore . operation'.- - These Einventoriesidepen'ds primarily onithe fue1~ specific' power. :The. inventory.of long-lived Kr-85
~
' 4 L(l'O.73 2 year, half-life), ;however,Jis naarly; proportional to .the- I!
1
~
L 1 assumed fuel = discharge burnup. Kr-85 has only a minor impact'on. off-sit'e . doses, primarily : affecting the? whole-body beta dose.
~ Since the of f-site radiological' consequences'-are dominated by the - 1 y
short lived radionuclides. (which areiat =saturatilon' concentration j
' independent' of. fuel burnup),- thel calculated dosesLwill not differ. 1 >
appreciablyI fromo those of previousEeval6ations: andTare nearly- 1
;. ;independentiof Lenrichment- (or burnup) . .Results of the evaluation ^
Ep d 54 , a n
'" Assumptions used for evaluating the potential radiologi- ,
,' .' cal consequences of a fuel. handling accident in the fuel! handling - '" l and storage facility for. boiling and. pressurized water reactors".' l l . . n
.9-1 1 d
j l 1 L # . d - : .- .- m
_~ .. . . - - . -. _ I l 1 l p confirm that the of f-site doses are essentially the same as those l AJ previously reviewed and accepted. - I The present evaluation uses values for the 2-hour
- atmospheric dispersion f actor (X/Q) and filter efficiencies that ]
have previously been reviewed and accepted. Core inventories j of fission products were estimated with the ORIGEN (QBNL Isotope Qgneration) code based upon a reactor power of 2568 MWt. The estimates of fission product inventories were derived from ORIGEN runs 'at various burnups and cooling times for initial enrichments of 3. 9% and 4.4 %. Thereafter, a small extrapolation, primarily I I affecting;the Kr-85 inventory, was used to determine the fission product. inventories-for 4.6% initial enrichment fuel burned to 60,000' MWD /MtU. I Calculations were made for 72 hours cooling time as
'the. source term for the fuel handling accident and for 120 days i : (source ~ term for a subsequent cask-drop accident evaluation). ,
! Reg LGuide o l.25. specifies that 10% of the Iodine and noble gas
- nuclides (except 30% for Kr-85) are present in the fuel rod gaps.
>Gapfinventories'were estimated.using-both these Reg Gu'de 1.25 assumptions and by the escape rate coefficient; method, based upon the ;B&W coefficients usedT in previous submittals and listed Jbelow:L
<o Noble Jases' (Xe; and Kr)f .l.'O x 104, o Halogens!(I2 a n d .' B rz ) 2.0 x 10'8, and j o Tellurium (I2 precursor) 4.0 x' 10'?.
9 summarizesi the
~
. Table' inventory' calculations for- those-radionuclides, significant 7in the fuel-handling and cask drop
. : accidents. l l
- d. ,_2
'\:
;I
- . . , ._. . , _ . ~__ ..___.. _ ___ .._ _ . . . . _ _ . _ . . . . _ .
I 1
- t. 1 '- '-
.T q
The following equation, from Reg Guide 1.25, was used
; -to calculate-the thyroid dose ~(D) from the inhalation of radio-lodine, F, I, F P ' B R .s ( x/Q )
D=q Rads
. t DF, DF, .
. summed,over all Iodine radionuclides.
- F' =.fraction.of _
fuel rod B= Breathing' rate = Iodine inventory-in gap 3.47 x 10' cubic -G, : space meters'per second f m , 3' '. I, = core IodineLradio- '[
.nuclide inventory at Ri= Dose conversion 1 time'of-theLaccident factor (rads / curie)_ j f
y (curies)-- from Reg. Guide 1.25 @i . N . ?u . Fi = - fractionsof core: ' 'tWf 'r - 1 damaged ~so as.to- . _ '(X/Q) = atmospheric
. ' release' Iodine in-the diffusion factor 4 3 %e , p. .. ,
rod.gapt(56/208x1/177) (6.8 x 10 sec/m ) 7 j;cj' } ' lPl= Core peaking factor' i M s
;(1470)_
DF, = effective. Iodine-p,-
~~U decontamination
@ ,; ' ' DF(=decontamination effective Iodinefactor' ., factor-for pooli water.-(= 100) M for-filters (= 10) me, n; y }g:",0[/j{ ;The gap -inventories listedL in Table: 9-1 are 'the product. - p of I 9 ]g% w (core inventory) and F,'(the fraction existing ;inL the gap) . , 4 m' pp, M e f Ng . .The functionLused to calculate the external whole. body > M.k ; J >
< v. doseifrom beta' (D,');or_ gammar l(D ) radiation in the cloud;uses many c N w. .; .
4 T; Lof'the terms defined above and is given by: r w 'n agn y > $cgAy
-D,'.=q.0.23
( x/Q ) . F. P G j . E, ,- 'and l nm 4 , . a w,t ' x a
. ? ,.y ' -Dr.=: q _ o. 25L (x/Q) . F P[G3 E ' '
x we l~ "!_'Jhf
.. ri y a}[jQue s
{ i;I ' ' @!\W,, , ( Zi!i j& '
'9 - 3 ID[hd '
a $ji y i
! y _' 5 y ,.g ,.-,
k.g g . . -1 . . ,
; e D
l' Gi is the gap inventory of the gaseous radionuclides of Xe and Kr and the functions above are summed over all the noble gases. These functions assume the noble gas decontamination factors in l l water and the. filters are 1.0. The gap inventories of radio-iodines make'a negligible contribution to the whole body doses, l. D, or Dr , because of the large decontamination f actors appropriate to the iodines. l L i 9.1.2 Results 1
! A summary of the assumptions used to evaluate the fuel
$ handling' accident-is given in Table 9-2. The minimum time after -i kg shutdown'when fuel. assemblies would be movad was conservatively assumed- to .be 72 hours as identified in the Technical 1 Specifications.- At 72 hours, the consequences at the site boundary of a' fuel handling accident releasing all of the gaseous > fission product radioactivity in the gaps of 56 rods in e damaged assembly over the entire course of the accident are: 'i Ih Escape Rate Reg Guide Method 1.25 , Inhalation thyroid dose = 1.1 Rads 4.8 Rads Whole body beta dose, D, = 0.40 Rads O.44 Rads Whole body gamma dose, Dr= 0.22' Rads 0.36 Rads i These ~ doses ' are well within the ' limits of 10 CFR Part 100 in. conformance with the acceptance criteria'of SRP:15.7.4~. (Rev.1, i July 1981). o O 9 - 4. 9 ,
g 9.2 Cask Drop Accident V The cask drop accident is the hypothetical drop of a filled shipping cask, very conservatively assumed to involve the rupture of all fuel rods in ten (10) fuel assemblies inside the cask. The accid '- sasumed to occur during transfer to the railcar at 120 days after shutdown. Since the acciden' is assumed to occur at ground level, the atmospheric dispersion factor is smaller than the fuel handling accident and the X/Q previously reviewed and accepted is 1.2 x 10'3 There are no decontamination f actors for water scrubbing and filtering (i.e. , DF, and DF, = 1.0) . On this basis, and using the gap inventories listed in Table - 9.1, the site boundary dose calculated by the methodology of Reg Guide 1.25 would be the following: Escape Rate Reg Guide Method 1.25 Inhalation thyroid dose = 1.7 Rads 8.7 Rads t] b Whole body b? 1 dose, D, = 5.9 Rads 18.4 Rads Whole body gamma dose,.Dr= _ 0.05 Rads 0.02 Rads
- For the' very conservative assumptions used, the thyroid-and whole l' ' body doses ~at 120 days are well within the limits of'10 CFR 100 in conformance with the acceptance criteria of SRP 15.7.4.
9.3 -Solid Radwaste l, The necessity for resin replacement is determined primarily by the requirement for water clarity and the resin is-normally. changed about once'a year. No significant increase in S L the volume ; of solid radioactive wastes is expected with the !+ expanded storag3 capacity. During reracking operations, a .small l, amount of additional resins may be generated by the pool cleanup system on a one-time basis. 9-5 1 l:
9.4 Gaseous Releases wi Gaseous releases from the fuel storage building are combined with other plant exhausts. Normally, the contribution from the fuel storage building is negligible compared to the other releases and no significant increases are expected as a result of the expanded storage capacity. 9.5 Personnel Excosures During normal operations, personnel working in the fuel storage area may be exposed to radiation from the spent fuel pol . Operating experience has shown that the area radiation dose rates, which originate primarily from radionuclides in the pool water, are generally less than 1 mrem /hr but may temporarily P increase to . 2. 5 - 3 mrem /hr during refueling operations. No evidence :has been observed of any crud deposition around the edges of the pool that might cause local areas of high radiation.
.A U
Radiation levels in zones. surrounding the pool are not expected to be significantly. af fected. Existing shielding around the pool (water depth L.id concrete walls) provide more than adequate protection, despite the slightly closer approach to the walls.of the pool. y p Typical concentrations of radionuclides in the poe' water are shown in Table 9.3. .During fuel reload operations, the concentrations will. increase due'to. crud deposits spalling lie 'from, spent fuel assemblies and to activities carried into the ! 7 pool from the primary system.. While these effects may increase b~ the concentrations-(as much as al factor of 10), the pool cleanup system soon reduces the concentrations to the normal operating i.. o range. 9 i__
. . ~ . .-, .. - . . . , _ . . . . - . - - -. . . .
u t
. Operating experience has shown that there have been c [ . negligible concentrations of airborne radioactivity and no increases are expected as a result of the expanded storage ,
capacity. L No increase in radiation exposure to operating personnel is [' expected and therefore neither the current health physics program 4 nor the area monitoring systems need to be modified. I ' s. 9.6 Anticioated Excosure Durina Rerackina l f - l q Total occupational exposure for the reracking operation I is estimated.to.be between 5 and 10 person-rem, as indicated in l l i
- T a b l e . 9 . 4~. While ' individual task ' efforts and exposures may x differ.'from;those'in Table'9.4, the' total'is believed to be a '
A l
' reasonable estimate for planning purposes. At the present time, J the use?ofidivers;is~nottanticipated. The reracking operatilon will utilize detailed procedures prepared with full consideration
", of'ALARA principles. Similar operations have been performed-'in
.a. number of fecilitiesLin/the'past,and thereLis every reason to- l
, a ibelieve that reracking can be' safely and efficiently accomplished
- J l
b,, , c at - Three : MiM . Island, - with~ minimum radiation; exposure Lto: l; m=. ,
,i L!'y f i
, :p_ersonne,l".
~ :z '
w .
- TheiexfstiingJradiation protection _ program at TMI-1'is adequate .for the reracking . operations. . -Where there is 'a: 3
' pot'entdial!; for significanti airborne activity, '
continuous air y p.!* ] ' samplers willi l bel [in: ' operation. . Personnel . wear ' ' protective - r% . . . .. .-a. .
._ .. 1 .
F' clothing, and n Lif?necessary, x respiratory vprotective tiequipment. - t ). >
, ,-Activitiesjare'governedtbyCa2 Radiation Work Permit and-personnel
* "( : monitoringfequipment . may jbe - assigned to . each individual'. As a- ,
kh: l minimum, thisiincludes thermoluminescent dosimeters- and pocket' , jdosimeters .- Additional personnel - monitoring equipment L (i.e , , {[
""4
~
extremityibadgesimay.be utilized'as required. : Work, personnel. a w . L .,: . I, v - . -
.~ -e-+- V -
4Y
l 1
-1 1
. traffic, and the novement of equipment will be- monitored and 1 I
,._O!; contro11ed to minimize contamination and to assure that exposures
' 'are maintained ALARA.
l l In re-racking, the existing storage racks will be 1
. removed, decontar,inated as much as possible by washing and wipe- l idowns, packaged and snipped. to a licensed processing / disposal facility. Ship, n -containers and procedures will conform to
?
;g Federal ~ . DOT regulations and the requirements of any State DOT l
office through which the shipment may pass. r
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- iTabie!941 RADIONUCLIDE lNVENT0 RIES. AND' CONSTANTS' -
^
1TOTAliGAP lNVENTORY, CURIEf , sHu N DECAYf RegiGuide 1.25 -DOSE
,. k' d ,
C -- CONVERSION E _(WEV) E (MEV) l 1 NUCUDE : INVENTORY CONST.
- CURIESV Af1/ HRS 72 hrs ' 120. days 72 hrs 120._ days . Ri - ,
l: & r c : 1-i 31^ 575 E+7 3.593E-3 11.2 E+6 50 ::: 6.0 E+6 250 1.48E+6 0.186 0.389 l
~
Il-132 1.1L E+8 iO28E-1 0 ;O- 0; O: : 5.35E+4
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I-133I 1.74 EA8 3.33dE-2 2.7 E+4 s 0~ ' 1.3 E+6 0" 4.0 E+5 :0.419 0.597 i 1-134: :1.5 E+8 7.920E--I '0 -0 0 0 2.5 E+4. - -
.I-i35 ~ 1.2 E+8 1.048E-1 0' O . 0-. 1.24E+5 - 0.394 1.456
- p. m l
l-Kr-85W .. 1.6 E+6 1.548E-1 0 0 0' O - - l Kr 1.6 E+6 7.374E-6 " 1.5 E+ 6 ' 1.5 E+6 - 4.8 E+5 4.7 E+5 0.251 0.002 l l Kr-87 E c 3.0 E+7 5.436E-1 0 0.. 0 0 - - Kr I4.4 E+7- 2.441E-1 'O. 0 0 0 - - 1 I l~ l. l Xe-131ni 16.5 E+5.- 2;407E-3 : 8.0 E+4; .94- 5.5 E+4 76 - 0.163 i. i. Xe-133W 3.3 E+6 1.289E-2 4.0 E+4 'O- ~ 1.4 E+5 0 - 0.233 p
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Xe-133 ~ 1.4 E+8 5.508E-3 5.9 E+6 11' 9.5 E+6 l18 0.102 0.081 ~ Xe-135: 3.6 E+7 7.632E-2 .Of. 0 0. 0 0.309 0.262 u , f(*) Gap' inventory of 0 means the contribution is negligible. _a. , 3 ss % ~ - - - - . . v- .
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.. g). Table 9.2 DATA AND ASSUMPTIONS FOR THE EVALUATION OF THE FUEL RANDLING ACCIDENT T : 1. . Source Term Assumntions VALUES Core power level, MWT 2568
. Fuel burnup, MWD /MTU 60,000 Analytical method ORIGEN
- 2. Release Assumnt12DA Number of failed fuel 56 rods in 1 of rods 177 assemblies
' By escape rate
, Fraction of core J inventory released to method gap by Rea Guide 1.21 with coefficients:
% of the Iodines - 10 3.0 x 10 - Halogens
%~of the Xenone - 10 1.0 x 10'I4 - Noble Gas
% of Kr-85 - 30 4.0 x 10 - Tellurium
'gl Assumed power peaking 1.70 factor
- Inventory'in gap Table 9.1 available for release m , Pool decontamination factors 2 For Iodines 100 For Noble gases 1 Filter decontamination-factors For Iodines 10 For noble gases 1 Atmospharic Dispersion, 6.8 x 10 sec/m3 (x/Q) Breathing rate 3.47 x 10 m3 /sec h 1 Ill 9 - 10
= I
. = - _ _ _ _ . .
l.
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i Table 9.3 Typical Concentrations of Radionuclides in the Spent-Fue: Pool Water Concentration Nuclide uC, /ml , Ag-110M 4.6 x 10'8 1 Co-58 1. 5 x - 10 5 0, c -eo Cs-134 4 4 x 1o ' 3.2 x 10'. 1 Cs-137 6.4 x 10 1 J l i i i
/.
9 - 11 l
M' U Table 9.4 Preliminary Estimate of Person-Rem Exposures During Raracking ' Number of Estimated Step Personnel Hours Excosure(D Remove empty racks 5 40 0.5 to 1.0 Wash and Decon racks 3 10 0.08 to 0.2 Clean and Vacuum Pool 3 25 0.3 to O.6 Partial installation 5 20 0.25 to 0.5 of new rack modules Move-fuel to new racks 2 150 0.8 to 1.5 Remove remaining racks 5 120 1.5 to 3.0 ' Wash and Decon racks 3 30 0.2 to 0.4 ()' Install remaining new rack modules 5 35 0.4 to 0.8 Prepare old racks for 4 80 1.0 to 2.O(2) shipment Total Exposure, person-rem 5 to 10 d'
. Assumes minimum dose rate of 2 1/2 mR/hr (expected) to a maximum of 5 mR/hr, except for pool vacuuming operations which assumes.4 to 8 mR/nr.
(2) Maximum expected exposure, although details of preparation and packaging of old racks for shipment have not yet been deter-mined. 1 9 - 12 ,
9 10.0 IN-SERVICE SURVEILLANCE PROGRAM 10.1 Purcose i This'section describes the programmatic commitmonts made by GPU l Nuclear for in-service surveillance of the neutron absorption ! material to comply with the provisions of Section IV(8) of the OT I l Position Paper (Ref. 10.1.1). l All material used within a storage system for spent nuclear fuel 1 are qualified to a level of performance predicated upon calculated I worst case environmental conditions and are based on accelerated I i testing of the materials to levels of service life corresponding ! to that environment. Because such environmental compatibility testing in the laboratory conditions is accelerated, it is prudent ! that each of the system components be monitored to some extent l O throughout the service ' life to assure that the actual in-service f performance remains within acceptable parameters as defined by the i accelerated testing. For many_ of the materials, monitoring throughout the service lifa is relatively easy, however, the poison material is encased in a stainless steel jacket precluding a direct visual or physical examination during the in-service condition. j i
\
A poison surveillance program is presented in this section which l allows access to representative poison samples without disrupting 'I the integrity of the storage system. This program includes not
]
only the capability to evaluate the material in.a normal use mode, I but also to forecast changes that might occur within the storage 3 1 10-1 i A V L g t; L -
4 h system at a time significantly prior to the normal use mode occurrence- of such changes. A program to make a non-intrusive assessment of representative " poison" panels in the rack modules is also presented. . Recent and ongoing concerns regarding the integrity of the poison L material heightened the need for a comprehensive poison L surveillance program which ensures that the suberiticality requirements of the stored fuel array are maintained to be in keeping with the USNRC guidelines, and any material degradation l- trends are pinpointed in sufficient time to enable remedial action l by the owner of the nuclear power plant. Towards this end, a two track program of poison surveillance has
- j. been adopted. The primary vehicle for forecasting future trends ;
is to Jtilize the " coupons" which are removed at periodic ! intervals, and subject to various tests to acquire inferential data on the current and future performance. This approach, currently in common use, will be referred to as " coupon surveillance". l- The other vehicle for poison surveillance is direct testing of installed poison panels in the fuel racks. This approach will be referred to as " direct surveillance". A brief discussion of both I approaches-is presented below. 1' Coucon Surveillance: This procedure consists of preparing poison coupons carefully encased in a stainless steel metal jacket, and suspending them from a " coupon tree". Two coupon trees are prepared. One tree is suspended in a storage location selected in such a manner that L 10-2
i f-' ~ it_is surrounded by a random batch of spent fuel discharged from-. the pool. This batch of coupons is designated for "long term surveillance". The batch of coupons in the second " tree": is intendt.d for accelerated ' exposure. This tree is placed in the center of a group of freshly discharged fuel assemblies each time a new batch
'is discharged to the pool. The group of assemblies surrounding l the accelerated expos'ure tree should be those which have - the highest values of radial peaking factor. The object, of course, [
is to subject this " tree" to the maximum radiation exposure in the fuel pool in the minimum amount of time. f l Direct Surveillance In contrast to " coupon surveillance", direct surveillance involves' testing the poison panels themselves. As a result, cracks in the poison panel, or even depletion of boron carbide can be f--i identified.- However, since the testing is essentially a remotely ; d - ex'e cuted ' operation, the physical state of the poison (viz its ! state of embrittlement, warpage, etc.)'cannot be ascertained. ( L i The. principal direct testing. technique is the so-called Blackness i Testing Method. It is.a rapid, albeit imprecise, method. It is { well . suited for an expeditious assay of all storage locations to l
' detect the incidence 'of B-10 deficiencies. If the test yields 1 suspect locations ', then those shall be examined by' the much I slower, but. markedly more accurate, procedure of neutron. l l radiography. This two step process was successfully used to l
conclusively identify the cracks in the Quad Cities: rack module-poison panels.
, /
10-3 1
m i l' l l l f]; 10.2 COUPON SURVEILLANCE 10.2.1 Descriotion of Test Coucons ] i The poison used in the surveillance program must be representative of the material used within the storage system. It must be of the l same composition, produced by the same method, and certified to the same criteria as the production lot poison. The sample coupon , must be the same thickness as the poison used within the storage system and shall meet the referenced Holtec drawing dimensional requirements. Each poison specimen must be encased in a stainless ,, E ' steel jacket of an alloy identical to that used .in the storage I system, formed so as to encase the poison material and fix it in a position and with tolerances similar to that for the storage racks. The jacket would be similar to that for the storage racks. The jacket would be closed by quick disconnect clamps or screws I with-lock nuts in such a manner as to retain its form throughout the use period and also allow rapid and easy opening without p contributing mechanical damage to the poison specimen contained therein. Consistent with the USNRC OT Position Paper requirements of a , statistically acceptable sample size, a total of sixteen jacketed poison specimens shall be used. Eight specimens will be suspended from each-tree. Tree No. 1 will be used for long term testing. The second tree containing eight coupons will be used for accelerated testing. The specimen location must be adjacent to a designated storage cell with design ability to allow for removal of the strip, providing access to a particular speci:nen. Coupons are arranged in'each tree in groups of two at roughly 30" spacings with the11owest coupon _ pair located at approximately 3' from the ) rack baseplate. At each elevation where the coupons are located, l p 1 10-4 ev-r- 2_+
i
'l] the two coupons are independently disconnectable. The pair of coupons face opposing walls in the storage cell at each coupon elevation.
Recognizing that the reactivity of the fuel is maximum at its mid-height, the coupon located nearest to the mid-height shall be removed for the first surveillance test. The remaining coupons ahall be relocated as necessary, such that the next coupons to be rer.cVed will receive maximum cumulative exposure. 1 1 10.2.2 Benchmark Data i
. The following benchmark tests shall be performed on test coupons j derived from the same production run as the actual poison panels.
,(1) Length, width, thickness and weight measurements (ii) Wet chemistry (iii) Neutron attenuation measurement More information on these tests follows in a subsequent section. 1
(
-10.2.3 Coucon' Reference Data Prior to encasing the coupons, each coupon shall be carefully calibrated. Their width, thickness, length and weight shall be carefully measured and recorded. The_ wet chemistry will be
. performed on a strip taken from the same Boral' plates from which the coupons are made'to provide a benchmark B-10 loading data.
10-5 g
i l. 10.2.4 Lona Term surveillance (} o l-For long term surveillance it is necessary to define the incremental exposure period (TLONG) for the coupons. This period is defined as the time elapsed between the removal of successive coupons. This period has been set at 5 years for the TMI Unit I L pools. At the time of the first off-load of spent fuel, a specimen tree is located in a storage cell designated as a long-term testing location and surrounded by freshly discharged spent fuel assemblies from the peak power region of the reactor core. The location of the specimen strip must be controlled and positioned I in the center of the first off-load from the core. One coupon is removed at scheduled intervals, and is examined for loss of its L physical-and neutronic properties. , 10.2.5 Accelerated surveillance O At the time of the first off-load of spent fuel, the accelerated speciren strip is surrounded by storage cells containing fuel assemblies from the peak power region of the reactor core. At the time: of the second off-load of tae fucl assemblies, the accelerated specimen strip is withdrawn from the fuel pool and one coupon is taken for evaluation. The specimen strip is replaced in the ' fuel pool .in a new location, where it is again surrounded by
, peak power region fuel assemblies. The storage cell that was vacated may now be used to store a fuel assembly. This arrangement is repeated at each ' of f-load of fuel. By evaluation of the specimens, an accelerated monitor of environmental effects on the poison will be obtained.
10-6 b m
)
l 0
75 The period between successive removal of accelerated program (_) coupons is referred to as Tacc. Initially Tace shall correspond to the outage interval. In later years, the plant staff may revise it based on data obtained from the coupons and other means, such as direct surveillance. 10.2.6 Insoection Tests to be oerformed The following tests will be performed to infer the integrity of the poison material (for both accelerated and long term coupons). (1) Physical tests: l Dimensional stability l a) b; Weight stability , c) Visual inspection t (2) Chemical tests: a) Wet chemistry test to determine the areal ; density of Boron carbide. ' n k- In the event of indication of poison material degradation by the above inspections, then, at the discretion of GPU Nuclear, further evaluation will be made using neutronic tests. (3) Neutronic tests: a) Neutron Radiograph of the Poison Material b) Neutron Attenuation measurements of the poison material. 10-7
l l j l
.?
-10.2.7 Methods (a) Dimensional stability is measured by either vernier calipers or micrometer.
(b) Weight Stability: 'The sample- is - weighed on a l 1 suitable grams balance and record dry weight in grams and milligrams. (c) Final examination: The surface integrity of the sample is examined by. a nuclear physicist or technician qualified for visual examination under a < nuclear Q.A. program. l (d) Wet chemistry: It is a process wherein the l aluminum in Boral is chemically dissolved . in an ! acid solution leaving Boron carbide precipitate which can be dried and weighed to determine the B 4C content in the coupon. l (e) Neutron Radiography (Optional) t Neutron radiography consists of passing a thermalized beam of neutrons through the sample (i.e., surveillance coupon) and allowing the beam l
-to impinge on a sheet of photographic film- '
,o positioned immediately behind the sample. Upon V developing the film, the. resulting image (called the radiograph) is reversed, i.e., where neutrons are absorbed by the sample, there will be less darkening of the film than in areas where. neutrons were not absorbed. To enhance the film response to thermal neutrons, a thin backing plate- of
- gadolinium -is positioned in intimate contact with the filut cmulsion.
)
Low energy electrons, arising from neutron reactions with the' gadolinium,. produce a more intense darkening of the film than would result from directly -impinging neutrons alone. Exposure times required for acceptable image contrast depends upon the intensity of the thermal neutron beam and the absorption. properties of the sample.
-With Boral samples, neutron absorption is high and good contrast is expected for areas. of mission or -
significantly lower - absorption. Experience has i e 10-8 j ( . ( . {[ L M
1
~
l A shown that Kodak single-sided-emulsion, SR5 X-ray 1 i)j film produces- good contrast neutron radiographs I vith minimum graying -from background gamma. I Iadiation in the neutron beam. Neutron radiography 1 saall be optionally performed on the coupon using a l qtalified' procedure and the radiograph will be evaluated by a qualified nuclear scientist. (f) Neutron Transmission Test (Optional) Neutron transmission measurements of the Boral sample coupons require a collimated source of thermalized neutrons, a fixture to maintain geometry, and an appropriate neutron detector. All measurements are made relative to a standard sample and counting is performed for a sufficient period I of time to assure the desired statistical confidence limits. A beam -of neutrons, well collimated and. l thermalized, is passed through the sample (in a perpendicular direction), and the number- of neutrons emerging is determined by counting with an i appropriate neutron detector (e.g., BF 3, LiI . (tl) l or equivalent). By comparing the counting rate for I a-surveillance sample in the neutron beam with the l p corresponding rate when the standard sample is in V the beam, the relative transmission can .be { determined. 10.3 ACCEPTANCE CRITERIA A plant procedure, entitled HPP-90310-5, "In-Use Surveillance Program-for Boron Neutron Absorber Material" has been develope-1 to execute the commitments made in this licensing submittal.
. Equipment requirements, step-by-step instructions for executing inspections. and acceptance criteria are described in- that. '
procedure for use by plant personnel. 3 1 10-9 ] O ; i i I 1
.A
[. , i g. 10.4 DIRECT POISON PANEL TESTING-Benchmark Data
~1 0.4.1 A minimum of 10% of all poison panels shall be examined by Blackness testing upon initial installation of the racks in the pool. The strip chart recordings shall be saved, with careful Q.A. validated notations to pinpoint the relevant poison panel locations. These strip charts will serve as benchmarks in the event that Blackness testing on irradiated cells is carried out at some point in time in the future.
l 10.4.2 Blackness Testino on Irradiated Panels GPU Nuclear, at . Its option, may conduct Blackness testing on irradiated panels at some point in time in the future. The fuel assemblies will be removed, as far away as possible, from the l cells previously benchmarked by Blackness testing. Then these-l cells will be Blackness - tested. At the same time, 10% of other ! E h locations, where fuel .has been previously s tore.d , shall be randomly Blackness tested. A comparison of the strip chart recording of irradiated panel with the reference unirradiated i chart will indicate the onset of any material degradation. In the event that any degradation is noniced, a neutron radiograph L of the suspected location may be obta.!ned. In the event of a poaitive identification of defect, the- responsible utility , personncl will determine what.further course of action is required and USNRC will be notified. 10-10
~
-i f
n 10.5 ACCEPTANCE CRITERIA POR BLACKNESS TESTING V The strip chart produced by the neutron logging device used in Blackness testing must be interpreted by a nuclear physicist j skilled in this very specialized measurement operation. In ( general, the Blackness testing cannot detect cracks in the Boral panels which are under 3/8" in width, or depletion in B-10 concentration under 10%. A qualified physicist can, however, discern the presence of anomalies which are due to over 3/8" wide cracks or over 10% depletion of B-10. Strip chart recordings showing anomalies which extend more than 3/4 inch in length will be considered to certainly point to a potential defect, l-
10.6 REFERENCES
FOR SECTION 10 l- l 10.1.1 OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", by Brian K. Grimes, USNRC, April 14, 1978, and Revision dated January 18, 1979. f3 l \_) L l l p . 10-11 *
'(
l 8
11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 11.1 The specific need to increase the limited existing storage capacity of the Spent Fuel Pool at Three Mile
-U Island Unit -1 is based on the continually-increasing inventory in the pool,- the prudent requirement to g maintain full-core off-load capability, and a lack of viable economic alternatives.
The inventory increase can be predicted by the following fuel assembly discharge schedule: Number of Assemblies Cumulative Iggg Discharced Inventerv
-1990 80 440 1991 76 516 1993 68 584*'
1995 68 652-1997 68 720** i 2001 68 788 2003 68 924 l 2005 68 992 2007 68 1000 _ 2009 ** 177 1237
=
Loss of full-core off-load capability with current racks. End of current licensed plant life. Last discharge allowable with current full poel capacity . of.749 assemblies. Includes control rods and burnable poison rod assemblies stored in Spent Fuel Assembly locations. Additionally, there are no existing or planned contractual-arrangements for TMI-1 fuel storage or fuel reprocessing. 11-1 z M
l l l
, l x
s 11.2 The proposed construction contemplates the re-racking of both TMI-1 spent fuel pools (A and B) using free-standing,-high density, poisoned spent fuel racks. The engineering, design, and licensing will be completed for full re-racking of both pool A and B. However, actual. construction will proceed no* with new rack construction i and installation for a portion 'of pool A only. This partial installation will provide sufficient TMI-1 pool I storage capacity to maintain a full core off-load capability to the end of current plant licensed life. E Additional new racks may be constructed and installed as dictated by any future changes to the planned-operating j i lifetime of THI-1. Should all the designed high ' density, poison racks be installed, TMI-1 would have sufficient capacity to store all the spent fuel L discharged if.TMI-1 were to operate through 2021. ! L-ll l The total capital cost is estimated to be approximately
$4.2 million as detailed below.
q Engineering & Design *
$ 800K Rack Fabrication S 3,000K:
. Rack Installation -$ 1,500K Total $ 5,400K l-h j 11.3 Many alternatives were considered . prior to proceeding-9 with re-racking, which is not'the;only technical option available to ' increase on-site storage capacity.
.Re-L racking does however, enjoy a cost advantage over other-technologies, as shown:
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;G 11-2 li
_.m__.__.i._______________.______._______________.__________ -
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. .y Capital Costs ,
Tvoe of Storace ($/KcU)* i l- - Re'-Rack S10-25 Puel Consolidation $25-55 Dry Cask Storage $88 Storage Vault $79 ' New Pool $115 From EPRI NF-3580, May 1984 There are no ceceptable alternatives to increasing the i on-site spent fuel storage capacity of TMI-1. . First, there are no commercial independent spent fuel storage facilities operating in the U.S. Second, the adoption of the Nuclear Waste Policy Act' (NWPA) created a de f acto throw-away nuclear fuel cycle. Since the cost of D- spent fuel reprocessing is not offset by the ~ salvage - V value of the residual uranium, reprocessing represents l an added cost for the nuclear fuel cycle which already , includes the NWPA Nuclear Ware Fund fees. In any' event, there are no domestic reprocessing facilities. Third, j since the other GPU Nuclear operating power plant has-a: spent fuel storage capacity problem as well, shipment.of spent fuel from 'MI-1 T to the other system nuclear-power 1 plant is not an. effective solution. Fourth,.at $500,000 per day replacement power cost, shutting down the TMI-1 ? nuclear power plant. is many-times more expensive than- i !j simply re-racking the existing spent fuel pools. l
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11.4 The expansion of the TMI-1 Spent Fuel Pool "A" capacity ) is expected to require the following primary resources: l W Stainless Steel - 80 Tons. Boral Neutron Absorber - 22 Tons, of which 10 Tens is Boron Carbide powder and 12 tons is aluminum. ' The requirements for stainless steel and aluminum represent' a smdll fraction of total world output of l , these metals - (less than .0001%). Although the fraction of world production of Boron Carbide required for the ! l- fabrication is somewhat higher than that of stainless steel or. aluminum, it is unlikely that the commitment ~of .
\
Boron Carbide to this project will affect other l alternatives. Experience has shown that the production
.of Boron Carbide is highly ' variable and depends upon 1
.need, and can easily be expanded to accommodate domestic needs, l
11.5. Due to the additional heat-load arising : from increased l Spent Fuel Pool inventory, the anticipated maximum bulk pool temperature increases from a previously-licensed' 140*F to 160'F, as' detailed in the- calculations ! described in Section 5.0 of this- report. This temperature is arrived at ~ ' by assuming . fully-degraded Spent Fuel Pool Coolers (fully-fouled, maximum allowable number of tubes plugged), which is extremely conservative given the actual ' condition 'of the 'TMI-1 coolers. The resultant total. heat-load (worst case).is 29 million BTU /HR, which is only .5% of the total. plant heat loss to the environment and well' within the-capability of' plant cooling systems. I i O ; 1 11-4 l l
I I I L b The increased bulk pool water temperature will result in 3 an increase in the pool water evaporation rate.- This ! has been calculated as sufficient to increase the relative humidity of the Fuel Handling Building atmosphere by 5%. This increase is within the capacity of both normal and the ESF Ventilation System. The net result of the increased heat loss and ' water vapor emission to the environment is negligible. 1 O , 9 i LO 11-5 I}}