ML20203B072

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Rev 6 to HI-92828, Licensing Rept for Spent Fuel Storage Capacity Expansion
ML20203B072
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 11/16/1998
From:
HOLTEC INTERNATIONAL
To:
Shared Package
ML20203B069 List:
References
HI-92828, HI-92828-R06, HI-92828-R6, NUDOCS 9902100215
Download: ML20203B072 (11)
v • d • e


Text

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i l H O' L T E C 1NTERNATIONAL i

! LICENSING REPORT l FOR i

SPENT FUEL STORAGE CAPACITY EXPANSION FORT CALHOUN STATION DOCKET 50-285-Holtec Report HI-92828 Omaha Public Power District Omaha, Nebraska e38*88Eci$$8$$385 P PDR I

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H O -L T E C INTERNATIONAL REVIEW AND CERTIFICATION LOG  :

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DOCUMENT NAME: LICENSING REPORT FOR SPENT FUEL STORAGE '

CAPACITY EXPANSION H 28 HOLTEC DOCUMENT LD. #

HOLTEC PROJECT NUMBER 20330 maha Public Power District CUSTOMER / CLIENT l

REVISION BLOCK ISSUE QUALITY PROJECT NO. AUTHOR REVIEWER ASSURANCE MANAGER

& DATE & DATE & DATE & DATE ORIGINAL wA% 9/Hf9t b

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Must be Project M mager or his Designee.

NOTE: Signatures and printed names are required in the review block.

This document conforms to the requirements of the design specification and the applicable sections of the governing codes.

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SUMMARY

OF REVISIONS LOG Report HI-92828 Preliminary Issue: May 14, 1992.

Revision 0: Issued August 21, 1992. l Revision'1: Issued October 5, 1992.  ;

contains the following number of pages:

Title Page 1 Review and Certification Log 1

  • Table of Contents 4 List of Tables 4 List of Figures 5  ;

Section 1 9  !

Section 2 21 Section 3 18 Section 4 29 Section 5 34 Section 6 . 129 Note: Section 6 includes a page 35a ,

Section 7 7 Section 8 15 Section 9 15 Section 10 6 Section 11 6 l

Revision 2: Issued October 8, 1992. '

Revision 2 incorporates OPPD's comments.

Revision 2 contains the same number of pages as Revision 1 with the exceptions of Section 9 (now 14 pages) and Section 11 (now 7 pages).

Revision 3 incorporates OPPD's comments, and revises the l following pages: 4-4, 4-11, 4-22, 5-1, 8-8, 9-2, 9-5, 9-9, j and 9-12.

i Revision 3 contains the same number of pages as Revision 2.

Revision 4 incorporates OPPD's comments and revises the following number of pages: 1, ii, v, vii, viii, and 7.3.

Section 4 is revised-to add calculations for considering CEA.

Contains same number of pages except Section 4 now contains 34 pages.

i Revision 5 incorporates OPPD's comments and revises the i following pages: 1-4, 1-7,2-5, 2-7, 2-8, 3-3, 3-5, 3-6, 4-1,

! 4-4, 4-5, 4-12, A-2, A-12, 5-1, 5-2, 5-4, 5-9, 5-15, 6-2, 6-3, i

6-4, 6-5, 6-6, 6-7, 6-9, 6-11, 6-15, 6-19, 6-23, 6-26, 6-30, 6-35a, 6-46, 7-2, 8-6, 9-3, 9-4, 9-5, 9-11, 10-3, and 10-6.

SUMMARY

OF REVISIONS LOG (continued)

Report HI-92828 Revision 6 incorporates revised time-to-boil calculation results in Chapter 5 of the report. The following items are modified:

Section 5.8 on page 5-12 Table 5.8.3 on page 5-21 Figure 5.8.5 on page 5-33 Figure 5.8.6 on page 5-34 1

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  • le TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 MODULE LAYOUT FOR INCREASED STORAGE 2.1 Background 2-1 l 2.2 Multi-Region Storage 2-2 2.3 Material Considerations 2-3 2.3.1 Introduction 2-3 2.3.2 Structural Materials 2-3 2.3.3 Poison Material 2-4 j

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2.3.4 Compatibility with Coolalit 2-6 2.4 Existing Rack Modules and 2-6 I Proposed Reracking Operation '

2.5 Heavy Load Considerations for the 2-6 Proposed Reracking Operation 3.0 RACK FABRICATION AND APPLICABLE CODES 3.1 Fabrication Objective 3-1 3.2 Rack Module for Region I 3-1 3.3 Rack Module for Region II 3-3 3.4 Codes, Standards, and Practices 3-5 for the Fort Calhoun Station Spent Fuel Racks 4.0 CRITICALITY SAFETY CONSIDERATIONS 4.1 Design Basis 4-1 4.2 Summary of Criticality Analyses 4-4 4.2.1 Normal Operating Conditions 4-4 4.2.2 Abnormal and Accident Conditions 4-6 4.3 Reference Fuel Storage Cells 4-7 4.3.1 Reference Fuel Assembly 4-7 4.3.2 Region 1 Fuel Storage Cells 4-7 4.3.3 Region 2 Fuel Storage Cells 4-8 4.4 Analytical Methodology 4-9 4.4.1 Reference Design Calculations 4-9 4.4.2 Fuel Burnup Calculations and 4-10 Uncertainties 4.4.3 Effect of Axial Burnup 4-11 Distribution 4.4.4 Long-term Changes in Reactivity 4-12 4.5 Region I Criticality Analyses and 4-13 Tolerances 4.5.1 Nominal Design 4-13 4.5.2 Uncertainties due to Tolerances 4-13 i 4.5.2.1 Boron Loading Tolerances 4-13

! 4.5.2.2 Boral Width Tolerance 4-14 l 4.5.2.3 Tolerances in Cell 4-14 Lattice Spacing i

'l s TABLE OF CONTENTS (continued) 4.5.2.4 Stainless Steel Thickness 4-14 Tolerances 4.5.2.5 Fuel Enrichment and 4-15 Density Tolerances 4.6 Region 2 Criticality Analyses 4-16 4.6.1 Nominal Design Case 4-16 4.6.2 Boundary Cells (Region 3) 4-17 4.6.3 Uncertainties due to Tolerances 4-18 4.6.3.1 Boron Loading Tolerances 4-18 4.6.3.2 Boral-Width Tolerance 4-18 4.6.3.3 Tolerance in Cell Lattice 4-18 Spacing 4.6.3.4 Stainless Steel Thickness 4-19 Tolerance 4.6.3.5 Fuel Enrichment and 4-19 Density Tolerances 4.7 Fuel Currently in Storage 4-19 4.8 Abnormal and Accident Conditions 4-20 4.8.1 Temperature and Water Density 4-20 Effects 4.8.2 Eccentric _ Fuel Positioning 4-20 4.8.3 Dropped Fuel Assembly 4-21 4.8.4 Lateral Rack Movement 4-21 4.8.5 Abnormal Location of a Fuel- 4-21 Assembly 4.9 References for Section 4 4-23 Appendix A to section 4 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 Introduction 5-1 5.2 . Spent Fuel Cooling System and Cleanup 5-2

System Description

5.3 Decay Heat Load calculations 5-3 5.4 Discharge Scenarios 5-3 5.5 -Bulk Pool Temperatures 5-4 5.6 Local Pool' Water Temperature 5-8 5.6.1 Basis 5-7 5.6.2 Model Description 5-8 5.7 Cladding Temperature 5-10 5.8 Results 5-12 5.9.; References for Section 5 5-13 6.0 SEISMIC / STRUCTURAL CONSIDERATIONS 6.1 Introduction 6-1 6.2 Analysis Outline 6-1 6.3 Artificial Time-Histories 6-6 ii

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l TABLE OF CONTENTS (continued) 6.4 Rack Modeling for Dynamic Simulations 6-10  !

6.4.1 General Remarks 6-10 6.4.2 The 3-D 22 DOF Model for l Single. Rack Module 6-12 6.4.2.1 Assumptions 6-12 6.4.2.2 Model Details 6-13 6.4.2.3 Fluid Coupling Details 6-14 6.4.2.4 Stiffness Element Details 6-15 6.4.3 Whole Pool Multi-Rack (WPMR) 6-17 l Model j 6.4.3.2 General Remarks 6-17  :

6.4.3.2 Whole Pool Fluid Coupling 6 17 l 6.4.3.3 Coefficients of Friction 6 -18 I 6.4.3.4 Modeling Details (-18 6.5 Acceptance Criteria, Stress Limits and U-19 Material Properties 6.5.1 Acceptance Criteria 6-19 6.5.2 Stress Limits for Various Conditions 6-21 6.5.2.1 Normal and Upset Conditions 6-21 (Level A or Level B) 6.5.2.2 Level D Service Limits 6-23 6.5.2.3 Dimensionless Stress Factors 6-23 6.5.3 Material Properties 6-24  ;

6.6 Governing Equations of Motion 6-24  !

6.7 Results of 3-D Nonlinear Analyses of 6-26 Single Racks 6.7.1 Impact Analyses 6-28 6.7.2 Weld Stresses 6-29 6.8 Results from Whole Pool Multi-Rack Analyses 6-30 6.9 Bearing Pad Analyses 6-32 6.10 References for Section 6 6-33 7.0 ACCIDENT ANALYSIS AND MISCELLANEOUS STRUCTURAL EVALUATIONS 7.1 Introduction 7-1 7.2 Refueling Accidents 7-1 7.2.1 Dropped Fuel Assembly 7-1 7.2.2 Gate Drop onto the Top of a Rack 7-2 7.3 Local Buckling of Fuel Cell Walls 7-3 7.4 Analysis of Welded Joints in Rack 7-4 due to Isolated Hot Cell 7.5 References for Section 7 7-5 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Introduction 8-1 l 8.2 Description of Spent Fuel Pool Structure 8-1 t

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  • s TABLE OF CONTENTS (continued) 8.3 Definition of Loads 8-2 8.3.1 Static Loading 8-2 8.3.2 Dynamic Loading 8-2 8.3.3 Thermal Loading 8-3 8.3.4 Loads from Adjacent Structure 8-3 8.4 Analysis Procedures 8-3 8.4.1 Finite Element Analysis Model 8-3 8.4.2 Analysis Methodology 8-4 8.4.3 Load Combinations 8-5 8.5 Results of Analyses 8-6 8.6 Pool Liner 8-8 8.7 Conclusions 8-8 8.8 References for Section 8 8-8 9.0 RADIOLOGICAL EVALUATION 9.1 Radiological Consequences of Accidents 9-1 9.1.1 Fuel Handling Accident 9-1 9.1.2 Cask Drop Accident 9-3 9.1.3 Spent Fuel Pool Gate Drop 9-4 Accident 9.1.4 Seismic Event in the Spent 9-4 Fuel Pool 9.2 Solid Radwaste 9-5 9.3 Gaseous Releases 9-6 9.4 Personnel Exposures 9-7 9.5 Anticipated Exposure During Reracking 9-8 9.6 Conclusions 9-10 9.7 References for Section 9 9-11 10.O BORAL SURVEILLANCE PROGRAM 10.1 Purpose 10-1 10.2 Coupon Surveillance Program 10-2 10.2.1 Coupon Description 10-2 10.2.2 Surveillance Coupon Testing Schedule 10-3 10.2.3 Measurement Program 10-4 10.2.4 Surveillance Coupon Acceptance 10-4 Criteria 10.3 References for Section 10 10-5 11.0 ENVIRONMENTAL COST-BENEFIT ASSESSMENT 11.1 Introduction 11-1 11.2 Project Cost Assessment 11-3 11.3 Environmental Assessment 11-5 11.4' Conclusions 11-6 l

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5.8 ,

Results It is shown in'the calculations that the heat load of the "old -

fuel assemblies- (background heat load) in the pool (27 refueling cycles of 1159 assemblies) is 3.60 x 10' Btu /hr during the final discharges specified in Section 5.4. Table 5.8.1 gives the general input .for the bulk pool temperature analyses. The maximum bulk. ~ pool temperature results - are : presented in Table 5.8.2. The time varying bulk pool: temperatures and heat load in the pool.are plotted vs. time-after-shutdown in Figures 5.8.1 to 5.8.4. It is-shown from the analyses that the maximum bulk pool temperature resulted from the limiting full core offload is 135*F at 122 hours0.00141 days <br />0.0339 hours <br />2.017196e-4 weeks <br />4.6421e-5 months <br /> after reactor shutdown. The coincident heat load to the cooling system is 20.72 x 10' Btu /hr (excluding 0.25 x 10 8 Btu /hr evaporation. hqat losses). The maximum calculated temperature is well below the ' temperature guidelines for both normal and abnormal conditions specified in the NUREG-0800 Standard Review Plan, Section 9.1.3. The ' temperature limit of 140*F prescribed in' the FCS USAR is also maintained.

The loss-of-cooling events have been considered for the specified discharge scenarios. The loss of all forced cooling is assumed to occur at the instants of peak pool temperatures.

Tr.ble 5.8.3 summarizes the results of the time-to-boil and maximum evaporation rate. The calculated minimum time from the loss- of pool cooling until the pool boils is 7.22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> and the g~

maximum boil-off rate is 45.6 gpm. The water elevations in the pool are plotted vs. time after loss-of-cooling in Figures 5.8.5 and 5.8.6 assuming no makeup water is added to the pool.

Table 5.8.4 gives the results of the maximum local water temperature and maximum local fuel cladding temperature for the limiting discharge scenario (Case 1) . Calculations are performed assuming non-bloci. age and 50% blockage, respectively. The blockage is assumed to occur in the thermally limiting storage cell'by a 5-12

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Table 5.8.3 [

TIME-TO-BOIL RESULTS Time-to-Boil (Hours) Maximum (Without Make-Up Evaporation Case No. Water) Rate (GPM) 1 7.22 45.6 RG 2 9.17 38.5 I

5-21

s HOLTEC INTERNATIONAL FCS SFP LOSS OF COOLING - CASE 1 (LAST REFUELING FULL CORE OFFLOAD) .

45 40 ,

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30 40 50 60 70 80 0 10 20 TIME AFTER LOSS OF COOUNG, HRS FIGURE 5.8.5 WATER ELEVATION AFTER LOSS-OF-COOLING - CASE 1 In

'9 HOLTEC INTERNATIONAL FCS SFP LOSS OF COOLING - CASE 2 (EMERGENCY FULL CORE OFFLOAD) .

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0 10 20 30 40 50 60 70 80 90 TIME AFTER LOSS OF COOUNG, HRS  !

FIGURE 5.8.6 WATER ELEVATION AFTER LOSS-OF-COOLING - CASE 2 M

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