ML18016A853
| ML18016A853 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/18/1999 |
| From: | Gupta V, Pellet S HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML18016A852 | List: |
| References | |
| HI-971760(NP), HI-971760(NP)-R03, HI-971760(NP)-R3, NUDOCS 9903220071 | |
| Download: ML18016A853 (353) | |
Text
ENCLOSURE 2 TO SERIAL: HNP-99-032 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 SPENT FUEL STORAGE RE-DESIGNATION OF PROPRIETARY INFORMATION LICE<NSING REPORT FOR EXPANDING STORAGE CAPACITY IN HARRIS SPENT FUEL POOLS 'C'ND 'D', RE<VISION 3 (NON-PROPRIE<TARY VERSION) 990$ ~007g 9cypegg PDR ADQCK 05000400 P PDR
0 555%% Holtec Center, 555 Lincoln Drive West, Marlton, N) 08053 HOLTEC I N T E R N AT 0 I N A L Telephone (609) 797-0900 Fax (609) 797-0909 LICENSING REPORT for EXPAt'6)ING STORAGE CAPACITY HARRIS SPENT FUEL POOLS C AND D HOLTEC INTERNATIONAL 555 LINCOLNDRIVE WEST MAIKTON,NJ OS053 HOLTEC PROJECT NO. 70324 HOLTEC REPORT HI-971760 REPORT CATEGORY: A REPORT CLASS: SAFETY RELATED CLIENT CONTRACT NO. XTA7000024
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REVIEW AND CERTIFICATION LOG FOR MULTIPLEAUTHORS Sheet 1of2 REPORT NUMBER: PROJECT NUMBER:
REVISION 0 REVISION I REVISION 2 REVISION 3 Document Portion Author Reviewer Author Reviewer Author Reviewer Author Reviewer P.
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-~~m5y 5-z t9.gf QA APPROVAL I/I P~OV.s / P> P/I p~l S/sik v, GvP 4'L PROJECT MANAGER'I l/l3/fr This document conforms to thc requirements of the Design Specification and the applicable sections of the governing codes.
Note: Signatures and printed names arc rcquircd in thc review block.
A revision of this document will bc ordered by thc Project Manager and carried out ifany of its contents is materially affecte during evolution of this project.
The determination as to thc need for revision will bc made by the Project Manager with input from others, as dccmcd necessary by him.
Must be Project Manager or his dcsigncc.
Distribution: C: Client M: Designated Manufacturer F: Florida Office THE REVISION CONTROL OF THIS DOCUMENT IS BY A "
SUMMARY
OF REVISIONS LOG" PLACED BEFORE THE TEXT OF THE REPORT. Form: RCL.O2
REVIEW AND CERTIFICATION LOG FOR MULTIPLEAUTHORS Sheet 2 of 2 REPORT NUMBER: PROJECT NUMBER:
REVISION 0 REVISION I REVISION 2 REVISION 3 Document Portion Author Reviewer Author Reviewer Author Reviewer Author Reviewer Chapter 9 H -z,z-9 k5. 'f I 4'5 s fg Sigh I~ 8 a/p nf/~
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t %s-s 58 Sg z.fb'f c s iS QA APPROVAL t ~~a-%8's .yes <<l~+ ~ i'm L3 PROJECT MANAGERt This document conforms to the requirements of thc Design Specification and thc applicable sections of thc governing codes.
Note: Signatures and printed names are required in the review block.
A revision of this document will bc ordered by the project Manager and carried out ifany of its contents is materially affected during evolution of this project.
Thc determination as to thc need for revision will bc made by thc Project Manager with input from others, as dccmcd ncccssary by him.
Must be Project Manager or his designee.
Distribution: C: Client M: Dcsignatcd Manufacturer F: Florida Office THE REVISION CONTROL OF THIS DOCUMENT IS BY A
SUMMARY
OF REVISIONS LOG PLACED BEFORE THE TEXT OF THE REPORT. Form: RCL.02
SUMIVfARYOF REVISIONS Revision 3 contains the following pages:
COVER PAGE 1 page REVIEW AND CERTIFICATION LOG 2 pages QA AND ADMINISTRATIVEINFORMATIONLOG 1 page
SUMMARY
OF REVISIONS 1 page TABLE OF CONTENTS 9 pages
1.0 INTRODUCTION
10 pages 2.0 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION 28 pages 3.0 MATERIAL,HEAVYLOAD, AND CONSTRUCTION CONSIDERATIONS 16 pages 4.0 CRITICALITYSAFETY EVALUATION 29 pages APPENDIX 4A 25 pages 5.0 THERMAL-HYDRAULICCONSIDERATIONS 26 pages 6.0 STRUCIVRAL/SEISMIC CONSIDERATIONS 75 pages 7.0 FUEL HANDLINGAND CONSTRUCTION ACCIDENTS 25 pages 8.0 FUEL POOL STRUCIVRE INTEGRITY CONSIDERATIONS 13 pages 9.0 RADIOLOGICALEVALUATION 4 pages 10.0 INSTALLATION 9 pages 11.0 ENVIRONMENTALCOST/BENEFIT ASSESSMENT 9 pages TOTAL 283 pages Revision 1 contains changes to incorporate comments from CP&L letter 10003481-009 (see Holtec letter 70324SP11.
Revision 2 contains changes to incorporate additional minor CP&L comments. Section 10 "Boral Surveillance Program" has been deleted with following sections renumbered accordingly.
Revision 3 changes the proprietary designation of some information based on discussions with the NRC. This revision does not modify any technical information or textual content.
Holtec International Rl Holtec Report HI-971760
C TABLE OF CONTENTS
1.0 INTRODUCTION
1-1 1.1 Rafhomcas 1-6 2.0 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION...... . 2-1 2.1 Intraduchon . 2-1 2.2 2-2 2.3 2-4 2.4 2-11 2.5 2-11 2.6 2-12 2.6.1 2-12 2.6.2 2-13 2.6.3 2-15 3.0 MATERIAL,HEAVYLOAD, AND CONSTRUCTION CONSIDERATIONS 3-1 3.1 Introduction 3-1 3.2 3-1 3.3 3-1 3.4 3-3 3.5 3-3 3.6 3-8 4.0 CRITICALITYSAFETY EVALUATION 4-1 4.1 4-1 4.2 4-3 4.2.1 4-3 4.2.1.1 PWR Fuel Results 4-3 4.2.1.2 BWR Fuel Results 44 4.3 4-5 4.3.1 4-5 4.3.2 4-6 4.4 47 4.4.1 4-7 4.4.2 4-8 4.4.2.1 PWR Fuel Burnup Calculations 4-8 4.4.2.2 BWR Fuel Burnup Calculations and Comparison to Vendor Calculations.... 4-8 4.4.3 4-9 4.4.3.1 PWR Fuel Axial Burnup Distribution . 4-9 4.4.3.2 BWR Fuel Axial Burnup Distribution . . 4-10 4 4,4 . 4-10 4.5 4-11 4.5.1 4-11 4.5.2 4-11 4.5.2.1 Boron Loading Tolerances 4-12 Holtec International Holtec Report HI-971760
TABLE OF CONTENTS 4.5.2.2 Boral Width Tolerance .. . 4-12 4.5.2.3 Tolerance in Cell Lattice Spacing and Cell Box Inner Dimension .. . ~ . 4-12 4.5.2.4 Stainless Steel Thickness Tolerance . 4-12 4.5.2.5 Fuel Enrichment and Density Tolerances . . 4-12 4.6 4-13 4.6.1 4-13 4.6.2 . 4-14 4.6.2.1 Boron Loading Variation . 4-14 4.6.2.2 Boral Width Tolerance Variation .... 4-14 4.6.2.3 Tolerance in Cell Lattice Pitch and Inner Box Dimension . ~ ..... 4-14 4.6.2.4 Stainless Steel Thickness Tolerances ~ 4-15 4.6.2.5 Fuel Enrichment and Density Variation 4-15 4.6.2.6 Zirconium Flow Channel . 4-15 4.7 . 4-15 4.7.1 . 4-16 4.7.2 . 4-16 4.7.3 4-17 4.7.4 4-17 4.7.5 . 4-18 4.8 .... 4-19 Appendix 4A "Benchmark Calculations"..... ~... ..
~ ~ . Total of 25 Pages including 6 figures 4A.1 4A-1 4A.2 4A-3 l0 4A-4 4A.3 4A.4 4A-5 4A.4.1 4A-5 4A.4.2 4A-5 4A.4.3 4A-5 4A.S 4A-6 4A.6 4A-7 5.0 THERMAL-HYDRAULICCONSIDERATIONS 5-1 5.1 Intrmiur.lion ~ ~ 5-1 5.2 5-2 5.3 5-3 5.4 5-4 5.4.1 5-8 5.5 5-8 5.6 5-11 6.0 STRUCTURAL/SEISMIC CONSIDERATIONS . 6-1 6.1 Inti:~ebon 6-1 6.2 6-1 6.2.1 6-2 6.3 6-5 Holtec International Holtec Report HI-971760
TABLE OF CONTENTS 6.3.1 . 6-5 6.4 6-6 6.5 . 6-7 6.5.1 6-7 6.5.1.1 . 6-7 6.5.1.2 . 6-9 6.5.2 . 6-10 6.5.2.1 6-11 6.5.3 6-12 6.5.4 6-13 6.5.5 . 6-14 6.6 . 6-15 6.6.1 . 6-15 6.6.2 6-16 6.6.3 6-18 6.6.4 6-20 6.7 6-21 6.8 . 6-23 6.8.1 . 6-23 6.8.2 . 6-25 6.8.3 6-26 6.8.4 6-27 6.8.4.1 6-27 6.8.4.2 6-29 6.8.'4.3 6-29 6.9 6-30 6.9.1 6-30 6.9.1.1 6-31 6.9.2 6-32 6.9.3 6-33 6.9.4 6-34 6.9.5 6-37 6.9.6 6-39 6.9.7 6-40 6.10 6-41 6.11 ..... 6-42 6.12 RaQzmcas ..... 6-43 7.0 FUEL HANDLINGAND CONSTRUCTION ACCIDENTS...'........... 7-1 7.1 Inimducfion 7-1 7.2 7-1 7.3 7-3 7.4 7-4 7.5 7-5 7.5.1 7-5 Holtec International ln Holtec Report HI-971760
TABLE OF CONTENTS 7.5.2 7-5 7.6 7-6 7.7 7-7 7.8 7-8 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS ... .... 8-1 8.1 Introduction 8-1 8.2 8-1 8.3 8-2 8.3.1 8-2 8.3.2 8-2 8.3.3 8-3 8.4 8-3 8.4.1 8-3 8.4.2 8-4 8.4.3 8-6 8.5 8-7 8.6 8.7 P~~
Candusions 8-8
. 8-9 8.8 8-10 9.0 RADIOLOGICALEVALUATION 9-1 9.1 ~ ~ ~ 0 9-1 9.2 9-1 9.3 0 ~ 9-1 94 9-2 10.0 INSTALLATION.. 0 10-1 10.1 Infxoduofion 10-1 1.2 ~ ~ ~ 10-4 10.3 10-5 10.4 10-6 10.4.1 10-6 10.4.2 2uxification....... 10-6 10.5 10-6 10.6 10-8 10.6.1 10-8 10.6.2 10-8 10.6.3 10-8 10.7 ~ ~ 0 10-9 11.0 ENVIRONMENTALCOST/BENEFIT ASSESSMENT ...... 11-1 11.1 ~ ~ 11-1 11.2 0 11-1 11.3 11-1 Holtec International 1V Holtec Report HI-971760
TABLE OF CONTENTS 11.3.1 .. 11-6 11.4 .. 11-6 11.5 .. 11-7 11.6 .. 11-8 11.7 .. 11-9 Holtec International Holtec Report HI-971760
TABLE OF CONTENTS T Ader Key Haris Plant Information 1-7 2.1.1 Geometric and Physical Data for Pool C Rack Modules 2-17 and 2-18 2.1.2 Geometric and Physical Data for Pool D Rack Modules . 2-19 2.5.1 Module Data for Harris Spent Fuel Racks ...... 2-20 3.3.1 Boral Experience List - PWRs .. 3-9 and 3-10 3.3.2 Boral Experience List - BWRs 3-11 and 3-12 3.3.3 1100 Alloy Aluminum Physical Characteristics 3-13 3.3.4 Chemical Composition - Aluminum (1100 Alloy) 3-14 3.3.5 Chemical Composition and Physical Properties of Boron Carbide 3-15 3.5.1 Heavy Load Handling Compliance Matrix (NUREG-0612) 3-16 4.2.1 Summary of Criticality Safety Calculations for PWR Fuel Racks 4-20 4.2.2 Summary of Criticality Safety Calculations for BWR Fuel Racks 4-21 4.3.1 PWR Fuel Characteristics 4-22 4.3.2 BWR Fuel Characteristics 4-23 4.4.1 Reactivity Allowance for Uncertainty in Burnup Calculations and the Effect of Axial Burnup Distributions for PWR Fuel . . 4-24 4.5.1 Comparison of MCNP-4A and CASMO-3 Calculations . 4-25 4.7.1 Reactivity Effects of Temperature and Void .. ~ ~.......... ~ . ... ~ . 4-26 4A.1 Summary of Criticality Benchmark Calculations ... ~... ..
~ ~ 4A-9 thru 4A-13 4A.2 Comparison of MCNP4a and Keno5a Calculated Reactivities for Various Enrichments . . 4A-14 4A.3 MCNP4a Calculated Reactivities for Critical Experiments with Neutron Absorbers 4A-15 4A.4 Comparison of MCNP4a and KENOSa Calculated Reactivities for Various "B Loadings .. 4A-16 4A.S Calculations for Critical Experiments with Thick Lead and Steel Reflectors . 4A-17 4A.6 Calculations for Critical Experiments with Various Soluble Boron Concentrations 4A-18 4A.7 Calculations for Critical Experiments with MOX Fuel .. 4A-19 5.1.1 Partial Listing of Rerack Applications Using Similar Methods of Thermal-Hydraulic Analysis ............. 5-12 and 5-13 5.2.1 Decay Periods for a Bounding Pools C and D Storage Configuration ..... ~ . . . . 5-14
~ ~
5.2.2 Fuel Assemblies Input Data for Decay Heat Evaluation ....... ~ . 5-15 5.2.3 Bounding Decay Heat Input from Stored Fuel in Pools C and D 5-16 5.4.1 Bounding Fuel Assemblies Hydraulic Flow Resistance Parameters 5-17 5.5.1 Pools C and D Dimensional Data ... 5-18 Holtec International vi Holtec Report HI-971760
TABLE OF CONTENTS 5.5.2 Bulk and Local Temperature Summary . 5-19 6.2.1 Partial Listing of Fuel Rack Applications Using DYNARACK ......... 6-45 and 6-46 6.3.1 Rack Material Data (200'F) (ASME - Section II, Part D) ~ 6-4/
6.4.1 Time-History Statistical Correlation Results .. . 6-48 6.5.1 Degrees-of-Freedom ..................................... ~...... 6-49 6.5.2 (MR216) Numbering System for Gap Elements and Friction Elements in the Pool D Campaign I Model.................. .. ~ 6-50 and 6-51 6.9.1 Comparison of Bounding Calculated Loads/Stresses vs. Code Allowables at Impact and Weld Locations 6-52 7.1 Impact Event Data . 7-9 7.2 Material Definition 7-10 8.5.1 Bending and Shear Strength Evaluation 8-11 9.4.1 Preliminary Estimate of Person-REM Dose During Reracking .. 9-4 Holtec International vii Holtec Report HI-971760
TABLE OF CONTENTS 1.1 Harris Fuel Handling Building Plan Layout 1-8 1.2 Storage Configuration for Pool C . 1-9 1.3 Storage Configuration for Pool D . 1-10 2.1.1 Pictorial View of Typical Harris Rack Structure .. 2-21 2.6.1 Seam Welded Precision Formed Channels . 2-22 2.6.2 Three PWR Cells in Elevation View 2-23 2.6.3 Three BWR Cells in Elevation View 2-24 2.6.4 Composite Box Assembly 2-25 2.6.5 Typical Array of Storage Cells 2-26 2.6.6 Support Pedestal for Holtec PWR Rack 2-27 2.6.7 Support Pedestal for Holtec BWR Rack 2-28 4.2.1 Burnup Versus Enrichment for PWR Fuel 4-27 4.3.1 a Two Dimensional Representation of the Calculational Model Used for the PWR Storage Rack Analysis . 4-28 4.3.2 a Two Dimensional Representation of the Calculational Model Used for the BWR Storage Rack Analysis 4-29 4A.1 MCNP Calculated k-eff Values for Various Values of the Spectral Index ..~ ~... 4A-20 4A.2 KENOSa Calculated k-eff Values for Various Values of the Spectral Index ..... 4A-21 4A.3 MCNP Calculated k-eff Values at Various U-235 Enrichments 4A-22 4A.4 KENOSa Calculated k-eff Values at Various U-235 Enrichments ............. 4A-23 4A.5 Comparison of MCNP and KENOSa Calculations for Various Fuel Enrichments . 4A-24 4A.5 Comparison of MCNP and KENO5a Calculations for Various Boron-10 Areal Densities .. 4A-25 5.3.1 C and D Pools Minimum Total Cooling System Requirements Curve at 137' Bulk Pool Temperature 5-20 5.4.1 Harris C and D Pools Physical Configuration . 5-21 5.5.1 Plan View of the Harris Pools C and D and CFD Model 5-22 5.5.2 Perspective View of the Harris Pools C and D and CFD Model . 5-23 5.5.3 Peak Local Water Temperature in the Rack Cells 5-24 5.5.4 Pools Interconnecting Channel Flow Velocity Vectors Elevation View Plot . ~ ..... 5-25 5.5.5 Pool Cooling Inlet/Outlet Piping Region Flow Velocity Vectors Plot ...... .... 5-26 6.3.1 Phased Storage Configuration for Pool C . .. 6-53 6.3.2 Phased Storage Configuration for Pool D .. . 6-54 6.4.1 SFP Time History Accelerogram (X-Direction, OBE) . 6-55 6.4.2 SFP Time History Accelerogram (Y-Direction, OBE) . 6-56 6.4.3 SFP Time History Accelerogram (Z-Direction, OBE) . 6-57 Holtec International viii Holtec Report HI-971760
TABLE OF CONTENTS 6.4.4 SFP Time History Accelerogram (X-Direction, SSE) 6-58 6.4.5 SFP Time History Accelerogram (Y-Direction, SSE) 6-59 6.4.6 SFP Time History Accelerogram (Z-Direction, SSE) 6-60 6.5.1 Schematic of the Dynamic Model for DYNARACK . 6-61 6.5.2 Fuel-to-Rack Impact Springs at Level of Rattling Mass . 6-62 6.5.3 Two Dimensional View of the Spring-Mass Simulation . 6-63 6.5.4 Rack Degrees-of-Freedom and Bending Springs . 6-64 6.5.5 Rack-to-Rack Impact Springs 6-65 6.5.6 Bottom Rack Impact Springs - Campaign I - Pool C . 6-66 6.5.7 Top Rack Impact Springs - Campaign I - Pool C 6-67 6.5.8 Bottom Rack Impact Springs - Campaigns II and III- Pool C 6-68 6.5.9 Top Rack Impact Springs - Campaigns II and III - Pool C 6-69 6.5.10 Bottom Rack Impact Springs - Campaign I - Pool D 6-70 6.5.11 Top Rack Impact Springs - Campaign I - Pool D .. ~ ~ ~ 0 6-71 6.5.12 Bottom Rack Impact Springs - Campaign II - Pool D 6-72 6.5.13 Top Rack Impact Springs - Campaign II - Pool D 6-73 6.8.1 Vertical Pedestal Time History Loading Plot . 6-74 6.9.1 Rack Fatigue Finite Element Model ~ ~ . ~ . 6-75 7.2.1 Shallow Drop on a Peripheral Cell 7-11 7.2.2 Plan View of Impactor and Impact Zone (Shallow Drop Event) . 7-12 7.2.3 Deep Drop on a Support Leg Location . 7-13 7.2.4 Deep Drop on a Center Cell Location 7-14 7.3.1 Heaviest Rack Drop 7-15 7.5.1 Shallow Drop: Finite Element Model Detail - Impacted Region 7-16 7.5.2 Maximum Cell Deformation for Shallow Drop on Exterior Cell . 7-17 7.5.3 Shallow Drop: Maximum Deformation - Impacted Region Plan 7-18 7.5.4 Plan View of Deep Drop Scenarios ~.....
~ ~ 7-19 7.5.5 Maximum Baseplate Deformation from Deep Drop Scenario 7-20 7.6.1 Gate Drop Finite-Element Model 7-21 7.6.2 Gate Drop Finite-Element Model, Detail of Impacted Region .. 7-22 7.6.3 Gate Drop Finite-Element Model, Detail of Impacted region (Plan) 7-23 7.6.4 Gate Drop Maximum Deformation .. 7-24 7.6.5 Gate Drop Maximum Deformation, Impacted Region Plan 7-25 8.2.1 Pool Structure Dimensions .. 8-12 8.4.1 Fuel Handling Building Finite Element Model . 8-13 Holtec International 1X Holtec Report HI-971760
1.0 INTRODUCTION
The Harris Nuclear Plant (HNP) is a single unit pressurized water reactor installation located in the extreme southwest corner of Wake County, North Carolina, and the southeast corner of Chatham County, North Carolina. The HNP installation is owned by the Carolina Power sc Light Company (CP8cL) and the North Carolina Eastern Municipal Power Agency (NCEMPA), located in Raleigh, North Carolina. CPS'as the overall responsibility to ensure that plant operations are performed without undue risk to the health and safety of the public. Table 1.1 contains key overview data for HNP's PWR Unit.
HNP was originally named Shearon Harris Nuclear Power Plant (SHNPP) and was initially designed as a four unit nuclear reactor site, of which only Unit 1 was completed. The Fuel Handling Building (FHB), however, was constructed to service all four Units as originally envisioned. During initial licensing, the possibility of transhipment from other Units was recognized and consequently the Spent Fuel Pools were licensed to store both PWR and BWR fuel. Transhipped fuel from the Robinson and Brunswick plants is already in stored in pools A and B.
The FHB is a long narrow structure intended to be sandwiched between the nuclear plants, in order to service all four Units. Each end of the building contains two large pools, with the South end pools (A and B) originally intended to service Units 1 and 4 and the North end pools (C and D) designed to service Units 2 and 3. The layout of the FHB and pools in relationship with Unit 1 is shown in Figure 1.1. The two pools in each end of the building were originally designated as the "New Fuel Pool"for the smaller of the two pools and the "Spent Fuel Pool" for the larger pool. These four pools have since been re-designated as pools A, B, C, and D, where pools A and D represent the smaller pools. All four pools are interconnected through "gated" passages and are capable of storing spent fuel.
Holtec International Holtec Report Hl-971760
Pools A and B, located at the South end of the building, have already been racked and are nearly full. Pool A contains six Region 1 type (6 x 10 cell) PWR racks and three (11 x 11 cell)
BWR racks for a total storage capacity of 723 assemblies. Pool A has been, and will continue to be, used to store fresh (unburned) fuel, recently discharged Harris fuel and transshipped fuel. Pool B contains six (7 x 10 cell), five (6 x 10 cell), and one (6 x 8 cell) PWR Region 1 style racks. Pool B also currently contains seventeen (11 x 11 cell) BWR racks, twelve of which have been supplied by Holtec International. Pool B is licensed to store one more (11 x 11 cell) Holtec BWR rack which would increase the total pool storage capacity to 2946 assemblies. The combined pool A and B licensed storage capacity is 3669 assemblies.
Projected operation of the Harris Unit and transhipments from the Robinson and Brunswick Units will continue to demand incremental increases in spent fu'el storage capacity. The Carolina Power 8~, Light Company, HNP's principal owner and operator, has entered into a contract with Holtec International of Marlton, N.J. to design maximum density spent fuel storage racks for pools C and D. Under the proposed capacity expansion, fuel storage racks will be installed in cainpaign phases on an as needed basis. This process is consistent with the incremental capacity expansions already performed in pool B.
Pools C and D are unused and are located in the north end of the Harris Fuel Handling Building. Pool C will provide storage for both PWR and BWR fuel. This pool has nominal dimensions of 27 feet wide by 50 feet long and at maximum storage density can accommodate 927 PWR and 2763 BWR assemblies. Pool D will contain only PWR fuel and with nominal dimensions of 20 feet wide by 32 feet long can accommodate 1025 maximum density storage cells. Proposed storage configurations for pools C and D are provided in Figures 1.2 and 1.3, respectively.
The configuration shown in Figure 1.2 represents the mixture of PWR and BWR storage which will accommodate future storage needs based on the best information currently available. To provide the greatest flexibility in mixture of fuel types, the storage racks were sized to allow Holtec International 1-2 Holtec Report HI-971760
erchangeability. The dimensions of the 9x9 PWR storage rack are nearly identical to those of the 13x13 BWR rack. Therefore, configurations other than those shown in Figure 1.2 are possible by replacing one rack type by the other. The complete geometric fungibility between the 9x9 PWR and 13x13 BWR rack modules affords CP8rL the latitude to alter the mix between PWR and BWR storage as the precise need for the two types of spent nuclear fuel storage become known. Interchanging of PWR and BWR modules would be performed after appropriate safety evaluations supported by reanalysis of the criticality, thermal-hydraulic, and structural analyses are successfully conducted to support such a substitution under Subpart 50.59.
The new Holtec racks are free-standing and self-supporting. The principal construction materials for the new racks are SA240-Type 304L stainless steel sheet and plate stock, and SA564-630 (precipitation hardened stainless steel) for the adjustable support spindles. The only non-stainless material utilized in the rack is the neutron absorber material which is a boron carbide and aluminum-composite sandwich available under the patented product name Boral .
The new Holtec racks are designed to the stress limits of, and analyzed in accordance with, Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel (B8cPV)
Code. The material procurement, analysis, and fabrication of the rack modules conform to 10CFR50 Appendix B requirements.
The rack design and analysis methodologies employed in the Harris storage capacity expansion are a direct evolution of previous rerack license applications. This Licensing Report documents the design and analyses performed to demonstrate that the new Holtec racks meet 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 the 1979 Addendum thereto [1.0. 1].
Holtec International 1-3 Holtec Report HI-971760
Sections 2 and 3 of this report provide an abstract of the design and material information on the new racks.
The criticality safety analysis requires that the neutron multiplication factor for the stored fuel array be bounded by the USNRC k, limit of 0.95 under assumptions of 95% probability and 95% confidence. The criticality safety analysis provided in Section 4 sets the requirements on the Boral panel length and the areal B-10 density for the new high density racks.
Thermal-hydraulic consideration 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 prescribed for the spent fuel pool to satisfy the pool structural strength, operational, and regulatory requirements. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 5.
Demonstrations of seismic and structural adequacy are presented in Section 6.0. The analysis sho'ws that the primary stresses in the rack module structure will remain below the allowable stresses of the ASME BEcPV Code (Subsection NF) [1.0.2]. The structural qualification also includes analytical demonstration that the subcriticality of the stored fuel will be maintained under all postulated accident scenarios in the Harris Final Safety Analysis Report (FSAR).
The structural consequences of these postulated accidents are evaluated and presented in Section 7 of this report.
Section 8 contains the structural analysis to demonstrate the adequacy of the spent fuel pool reinforced concrete structure. A synopsis of the geometry of the Harris reinforced concrete structure is also presented in Section 8.
The radiological considerations are documented in Section 9.0. Sections 10, and 11 discuss the salient considerations in the installation of the new racks, and a cost/benefit and Holtec International 1-4 Holtec Report HI-971760
environmental assessment to establish the prudence of CPEcL's decision to exercise the wet storage expansion option, respectively.
All comp'ter programs utilized to perform the analyses documented in this licensing report are benchmarked and verified. These programs have been utilized by Holtec International in numerous rerack applications over the past decade.
The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety in respect to all considerations of safety specified in the OT Position Paper, namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanical integrity.
Holtec International 1-5 Holtec Report HI-971760
[1.0.1] VSNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978, and Addendum dated January 18, 1979.
[1.0.2] ASME Boiler A, Pressure Vessel Code,Section III, Subsection NF, and Appendices (1995).
Holtec International 1-6 Holtec Report HI-971760
Table 1.1 KEY HARRIS PLANT INFORMATION ITEM DATA Docket Number 50-400 Capacity, MWe 940 Applied to NRC 9-4-71 Construction Permit 1-27-78 Commercial Operation 1986 Present Capacity Pool A 723 Pool B 2946 TOTAL 3669 Holtec International 1-7 Holtec Report HI-971760
OF UNIT 4 CONTAINMENT FUEL TRANSFER (NOT CONSTRUCTED)
CANALS CD
~NORTH CD C4 POOL B 2i 1 41 OF UNIT 1 CONTAINMENT OF UNIT 3 CONTAINMENT FUEl TRANSFER (NOT CONSTRUCTED) E(IUIPMENT CANALS HATCH CASK LOADINill POOL C POOL POOL D I
45 COLUMN LINE
( NOT CONSTRUCTED )
IDENTIFIERS FIGURE l.i; HARRIS FUEL HANDLING BUILDING PLAN LAYOUT
600" 2.50" SOItp SUPPORT LEG 2.50" OO 4.06" 4.06" OO OO I
OO 32 4I 4.06" 4.06"
.625" MIN.(TYP) 2.50" 2.50" TOTAL CELL COUNT:
927 CELLS PWR 2783 CELLS - BK FIGORE 1.8; STOME CONFIGURATION FOR POOL C H[-971780
384"
.625"(TYP) 5.50" 6.10" 00 0 0 0 10 240" 0 0 0 0 0 6.10" SUMP LEG TOThL CELL COUNT: SUPPORT 1025 CELLS PWR FIGURE 1.3; STORAGE CONFIGURATION FOR POOL D HI-971760
2.0 OVERVIEW OF THE PROPOSED CAPACITY EXPANSION 2.1 In!radIICIian In its currently proposed fully implemented configuration, Pool C will contain eleven PWR racks and nineteen BWR racks. Pool D will contain twelve PWR racks. All storage racks arrays will consist of free-standing modules, made from Type 304L austenitic stainless steel containing prismatic storage cells interconnected through longitudinal welds. A panel of Boral cermet containing a high areal loading of the B-10 isotope provides appropriate neutron attenuation between adjacent storage cells. Figure 2.1.1 provides a schematic of the typical Region 2 storage module proposed for Harris. Data on the cross sectional dimensions, gross weight and cell count for each rack module in pools C and D are presented in Tables 2.1.1 and 2.1.2, respectively.
In the parlance of wet storage technology, the Harris modules are of the so-called non-flux-trap genre and are referred to as Region 2 style racks. Region 2 PWR racks have enrichment/burnup limitations placed on them and storing PWR spent nuclear fuel will be subject to burnup compliance restriction. The BWR storage racks do not require any such limitations since the criticality analyses are, performed for the maximum reactivity over burnup.
Each new rack module is supported by four legs which are remotely 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 undulations in the fuel pool and cask pit floor flatness.
A bearing pad interposed between the rack pedestals and the pool liner serves to diffuse the dead load of the loaded racks into the reinforced concrete structure of the pool slab.
Holtec International 2-1 Holtec Report HI-971760
The overall design of the Harris racks is similar to those presently in service in the spent fuel pools at many other nuclear plants, among them Zion Nuclear Station of the Commonwealth Edison Company, Donald C. Cook of American Electric Power, and Connecticut Yankee of Northeast Utilities. Altogether, over 50 thousand storage cells of the Harris design have been provided by Holtec International to various nuclear plants around the world.
2.2 The key design criteria for the new Harris spent fuel racks are set forth in the classical USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", April 14, 1978 as modified by amendment dated January 18, 1979.
The individual sections of this report expound on the specific design bases derived from the above-mentione'd "OT Position Paper". Nevertheless, a brief summary of the design bases for the Harris racks are summarized in the following:
- a. Disposition: All new rack modules are required to be free-standing.
dd
- Allf -
did d I d k l lly pl (against tipping or overturning) if a seismic event (which is 150% of the postulated OBE or 110% of the postulated SSE) is imposed on any module.
.Allpl y l A k dl <<ll!'d limits postulated in Section III subsection NF of the 1995 ASME Boiler and Pressure Vessel Code.
- The spatial average bulk pool temperature is required to remain under 137'F t in the wake of a normal refueling.
In addition to the limitations on the bulk pool temperature, the local water temperature in the Harris pools must remain'ubcooled (i.e., below the boiling temperature coincident with local hydraulic pressure conditions).
The 137 F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of re-evaluating systems and components to allow for an increase the allowable bulk pool temperature.
Holtec International 2-2 Holtec Report HI-971760
ggl 2 II bbl<<gk'yld fuel of 5 w/o enrichment and 40 MWD/MTUburnup while maintaining the reactivity s 0.95.
- The reracking of Harris must not lead to violation of the off-site dose limits, or adversely affect the area dose environment as set forth in the Harris FSAR. The radiological implications of the installation of the
- g. ~TI new racks also need to be ascertained and deemed to be acceptable.
bgy fg II d << <<ifyg load combinations set forth in NUREG-0800, SRP 3.8.4 must be demonstrated.
- In addition to satisfying the primary stress criteria of Subsection NF, the alternating local stresses in the rack structure are evaluated to ensure that the "cumulative damage factor" due to at least ten SSE events does not exceed 1.0.
- Tl>> gly fb II d yll I-pl \ d gd Ig seismic event must be demonstrated. A material fatigue evaluation is performed in accordance with ASME B&PV Code. The alternating local stresses in the liner are evaluated to ensure that the "cumulative damage factor" due to at least ten SSE events does not exceed 1.0.
- Tb b gpd b ffll lyllk gb 2 p on the liner continues to satisfy the ACI limits during and after a design basis seismic event.
- k. :I~ 2 fp I d&p ( lldl Ig of a fuel assembly, for instance), it is necessary to demonstrate that the subcriticality of the rack structure is not compromised.
Tb fld I I dl d b Id for executing the reracking must be demonstrated to be within the "state of proven art".
The foregoing design bases are further articulated in Sections 4 through 9 of this licensing report.
Holtec International 2-3 Holtec Report Hl-971760
The following codes, standards and practices are used as applicable for the design, construction, and assembly of the Harris fuel storage racks. Additional specific references related to detailed analyses are given in each section.
(1) AISC Manual of Steel Construction, 1970 Edition and later.
(2) ANSI N210-1976, "Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design).
(3) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code Section III, 1986 Edition; ASME Section V, 1986 edition; ASME Section VIII, 1986 Edition; ASME Section IX, 1986 Edition; and ASME Section XI, 1986 Edition.
(4) ASNT-TC-1A June, 1984 American Society for Nondestructive Testing (Recommended Practice for Personnel Qualifications).
(5) American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI318-63),and (ACI318-71).
(6) Code Requirements for Nuclear Safety Related Concrete Structures, ACI349-85/ACI349R-85, and ACI349.1R-80.
(7) ASME NQA-1, Quality Assurance Program Requirements for Nuclear Facilities (8) ASME NQA-2-1989, Quality Assurance Requirements for Nuclear Facility Applications.
(9) ANSI Y14.5M, Dimensioning and Tolerancing for Engineering Drawings and Related Documentation Practices.
(10) ACI Detailing Manual - 1980.
Holtec International 2-4 Holtec Report Hl-971760
b.
(1) E165 - Standard Methods for Liquid Penetrant Inspection.
(2) A240 - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure Vessels.
(3)- A262- Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.
(4) A276 - Standard Specification for Stainless and Heat-Resisting Steel Bars and Shapes.
(5) A479 - Steel Bars for Boilers A Pressure Vessels.
(6) ASTM A564, Standard Specification for Hot-Rolled and Cold-Finished Age-Hardening Stainless and Heat-Resisting Steel Bars and Shapes.
(7) C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder.
(8) A380 - Recommended Practice for Descaling, Cleaning and Marking Stainless Steel Parts and Equipment, (9) C992 - Standard Specification for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.
(10) ASTM E3, Preparation of Metallographic Specimens.
(11) ASTM E190, Guided Bend Test for Ductility of Welds.
(12) American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section II-Parts A and C, 1995 Edition.
(13) NCA3800 - Metallic Material Manufacturer's and Material Supplier's Quality System Program.,
.ABMEB II dd V IE d,d I IE-W ld g and Brazing Qualifications, 1995 Edition.
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d.
(1) ANSI 45.2.1 - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants.
(2) ANSI N45.2.2- Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase).
(3) ANSI - N45.2.6 - Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).
(4) ANSI-N45.2.8, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Plants.
(5) ANSI - N45.2.11, Quality Assurance Requirements for the Design of Nuclear Power Plants.
(6) ANSI-N45.2.12, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.
(7) ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123).
(8) ANSI N45.2.15 Hoisting, Rigging, and Transporting of Items For Nuclear Power Plants.
(9) ANSI N45.2.23 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1. 146).
(10) ASME Boiler and Pressure Vessel,Section V, Nondestructive Examination, 1995 Edition.
(11) ANSI - N16.9-75 Validation of Calculation Methods for Nuclea'r Criticality Safety.
e.
(1) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and the modifications to this document of January 18, 1979.
Holtec International 2-6 Holtec Report HI-971760
(2) NUREG 0612, "Control of Heavy Loads at Nuclear Power Plants",
USNRC, Washington, D.C., July, 1980.
(1) ANSI/ANS 8.1 (N16.1) - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.
(2) ANSI/ANS 8.17, Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors.
(3) N45.2 - Quality Assurance Program Requirements for Nuclear Facilities
- 1971.
(4) N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974.
(5) N45.2.10 - Quality Assurance Terms and Definitions -1973.
(6) ANSI/ANS 57.2 (N210) - Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
(7) N14.6 - American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for Nuclear Materials.
(8) ANSI/ASME N626-3, Qualification and Duties of Personnel Engaged in ASME Boiler and Pressure Vessel Code Section III, Div. 1, Certifying Activities.
(9) ANSI Y14.5M, Dimensioning and Tolerancing for Engineering Drawings and Related Documentation Practices.
(1) 10CFR20 - Standards for Protection Against Radiation.
(2) 10CFR21 - Reporting of Defects and Non-compliance.
(3) 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants.
Holtec International 2-7 Holtec Report HI-971760
(4) 10CFR50 Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
(5) 10CFR61 - Licensing Requirements for Land Disposal of Radioactive Material.
(6) 10CFR71 - Packaging and Transportation of Radioactive Material.
(1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 2 Proposed).
(2) RG 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors.
(3) RG 1.28 - (ANSI N45.2) - Quality Assurance Program Requirements .
(4) RG 1.29 - Seismic Design Classification (Rev. 3).
(5) RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Material.
(6) RG 1.38 - (ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants.
(7) RG 1.44 - Control of the Use of Sensitized Stainless Steel.
(8) RG 1.58 - (ANSI N45.2.6) Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel.
(9) RG 1.60 - Design Response Spectra for Seismic Design of Nuclear Power Plants.
(10) RG 1.61 - Damping Values for Seismic Design of Nuclear Power Plants, Rev. 0, 1973.
(11) RG 1.64 - (ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear Power Plants.
(12) RG 1.71 - Welder Qualifications for Areas of Limited Accessibility.
Holtec International Holtec Report Hl-971760
(13) RG 1.74 - (ANSI N45.2.10) Quality Assurance Terms and Definitions.
(14) RG 1.85 - Materials Code Case Acceptability - ASME Section 3, Div. 1.
(15) RG 1.88 - (ANSI N45.2.9) Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records.
(16) RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis.
(17) RG 1.122 - Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components.
(18) RG 1. 123 - (ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants.
(19) RG 1.124 - Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, Revision 1, 1978.
(20) RG 3.4 - Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities.
(21) RG 3.41 - Validation of Calculational Methods for Nuclear Criticality Safety, Revision 1, 1977.
(22) RG 8.8 - Information Relative to Ensuring that Occupational Radiation Exposure at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA).
(23) DG-8006, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants".
(24) IE Information Notice 83 Fuel Binding Caused by Fuel Rack Deformation.
(25) RG 8.38 - Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June, 1993.
(1) CPB 9.1 Criticality in Fuel Storage Facilities.
Holtec International 2-9 Holtec Report HI-971760
(2) ASB 9 Residual Decay Energy for Light-Water Reactors for Long-Term Cooling.
(1) SRP 3.2.1 - Seismic Classification.
(2) SRP 3.2.2 - System Quality Group Classification.
(3) SRP 3.7.1 - Seismic Design Parameters.
'1 (4) SRP 3.7.2 - Seismic System Analysis.
(5) SRP 3.7.3 - Seismic Subsystem Analysis.
(6) SRP 3.8.4 - Other Seismic Category I Structures (including Appendix D), Technical Position on Spent Fuel Rack.
(7) SRP 3.8.5 - Foundations for Seismic Category I Structures, Revision 1, 1981 ~
(8) SRP 9.1.2 - Spent Fuel Storage, Revision 3, 1981.
(9) SRP 9.1.3 - Spent Fuel Pool Cooling and Cleanup System.
(10) SRP 9.1.4 - Light Load Handling System.
(11) SRP 9.1.5 - Heavy Load Handling System.
(12) SRP 15.7.4 - Radiological Consequences of Fuel Handling Accidents.
(1) AWS D1.1 - Structural Welding Code, Steel.
(2) AWS D1.3 - Structure Welding Code - Sheet Steel.
(3) AWS D9.1 - Welding of Sheet Metal.
(4) AWS A2.4 - Standard Symbols for Welding, Brazing and Nondestructive Examination.
(5) AWS A3.0 - Standard Welding Terms and Definitions.
Holtec International 2-10 Holtec Report HI-971760
(6) AWS A5.12 - Tungsten Arc-welding Electrodes.
(7) AWS QC1 - Standards and Guide for Qualification and Certification of Welding Inspectors.
2.4 The governing quality assurance requirements for design of the Harris spent fuel racks are enunciated in 10CFR50 Appendix B. The quality assurance program for design of the Harris racks is described in Holtec's Nuclear Quality Assurance Manual, which has been reviewed and approved by the Carolina Power & Light Company. This program is designed to provide a flexible but highly controlled system for the design, analysis and licensing of customized components in accordance with various codes, specifications, and regulatory requirements.
The manufacturing of the racks will be performed in accordance with the requirements setforth in 10CFR50 Appendix B.
The Harris rack modules are designed as cellular structures such that each fuel assembly has a prismatic square opening with conformal lateral support and a flat horizontal bearing surface.
The basic characteristics of the Harris spent fuel racks are summarized in Table 2.5.1. The design of the PWR and BWR storage racks are very similar. The major differences are in the cell inside dimension and pitch, the baseplate flow holes, the support legs, and the poison width and length.
A central objective in the design of the new rack modules is to maximize their structural rigidity while minimizing their inertial mass. Accordingly, the Harris modules have been designed to simulate multi-flange beam structures. The multiple flanges are formed from the numerous cell walls in the rack cross-sectional array. These cells are connected through intermittent welds. The weld lengths, location, and size were chosen during the original Holtec International 2-11 Holtec Report HI-971760
4 design of this rack style/series to ensure adequate strength and to adjust the natural frequency of the rack modules to avoid resonance. In general, this effort has resulted in excellent detuning characteristics with respect to the applicable seismic events.
2.6 This subsection presents an item-by-item description of the anatomy of the Harris rack modules in the context of the fabrication methodology. The object of this section is to provide a self-contained description of rack module construction for the Harris fuel pool to enable an independent appraisal of the adequacy of design.
The requirements in manufacturing the high density storage racks for Harris may be stated in four interrelated points:
- 1. The rack modules are fabricated in such a manner that there is no weld splatter on the storage cell surfaces which would come in contact with the fuel assembly.
- 2. The storage locations are constructed so that redundant fiow paths for the coolant are available.
- 3. The fabrication 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 stipulated seismic input motion.
Holtec International 2-12 Holtec Report HI-971760
In addition to the composite box assembly, the baseplate and the support legs constitute the principal components of the Harris fuel rack modules. The following description provides details of all of the major rack components.
- The rack module manufacturing begins with fabrication of the "box". The boxes 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. The minimum weld penetration is 80% of the box metal gage. This process results in a square cross section box, as shown in Figure 2.6.1. The clear inside dimension of the PWR box cell is 8.80".
A sheathing is attached to each side of the box with the poison material installed in the sheathing cavity. The design objective calls for attaching Boral tightly on the box surface. This is accomplished by die forming the internal and external boral sheathings to provide end flares with smooth edges, as shown in Figure 2.6.4. The flanges of the sheathing are welded to the box using skip welds and spot welds. The sheathings serve to locate and position the poison sheet accurately, and to preclude its movement under seismic conditions. The PWR Boral dimensions are 145" long and 7.5" wide.
The square cross section box with Boral panels affixed to its external surfaces is referred to as the "composite box assembly". Each composite box has at least two one inch diameter lateral holes punched near its bottom edge to provide auxiliary flow holes. For those cells located over support legs, four flow holes are required to compensate for the loss of the baseplate flow holes described below.
The composite boxes are arranged in a checkerboard array and welded edge-to-edge to form an assemblage of storage cell locations, as shown in Figure 2.6.5.
Filler panels and corner angles are welded to the edges of boxes at the outside boundary of the rack to complete the formation of the peripheral cells. The inter-box welding and pitch adjustment are accomplished by small longitudinal connectors. The connectors are sized and placed to ensure that the 8.8" inside cell clear dimension on developed boxes is maintained after inclusion of any reductions from the sheathing. This assemblage of box assemblies results in a honeycomb structure with axial, flexural and torsional rigidity depending on the extent of intercell welding provided. It can be seen from Figure 2.6.5 that all Holtec International 2-13 Holtec Report HI-971760
four corners of each interior box are connected to the contiguous boxes resulting in a well-defined path for "shear fiow".
Basaplatl:: A 3/4 inch thick baseplate provides a continuous horizontal surface for supporting the fuel assemblies. The baseplate has a 5 inch diameter hole in each cell location, except at lift locations. For the liftlocations the flow holes are modified to provide a smaller hole to match the dimensions of the BWR rack flow holes (3.8125" diameter) and slotted ears to allow insertion and engagement of the lifting rig. Matching the BWR flow hole dimensions allows for a single tool to be used to liftboth rack styles. The cross sectional area of the modified lift location flow holes is only slightly smaller than the 5" diameter holes, because of the added area provided by the slotted ears. The location of all baseplate holes coincide with the cell centerlines. The baseplate is attached to the base of the cell assemblage by fillet welds and extends horizontally, approximately 1/4" beyond, the periphery of the rack.
- As mentioned in the preceding section, Boral is used as the neutron absorber material, Each storage cell side is equipped with one integral Boral sheet (poison material), except for the outer walls of some of the peripheral rack cells. Only one Boral sheet is required between adjacent cells containing fuel. Therefore, outer rack walls which face each other do not both require Boral, consequently one of the two racks may be fabricated without poison along one outer wall.
lv. Shearing: As described earlier, the sheathing serves as the locator and retainer of the poison material.
~:All pp I g 6 pj Pl pp 6 I Ilg The inch diameter top (female threaded) portion is made of austenitic steel 2.6.6.
material. The bottom (male threaded) part is made of 17:4 Ph series stainless steel to avoid galling problems. Each support leg is equipped with a readily accessible socket to enable remote leveling of the rack after its placement in the pool. The support legs are located at the centerlines of cells to ensure accessibility of the levelling tool through the 5 inch diameter flow hole in the baseplate.
The assembly of the rack modules is carried out by welding the composite boxes in a vertical fixture with the precision fabricated baseplate serving as the bottom positioner.
Holtec International 2-14 Holtec Report HI-971760
- The fabrication of the boxes for the BWR racks is similar to that of the PWR racks. The inside cell clear dimension is reduced to 6.06" to properly support the smaller BWR assembly. Each box has two inch diameter lateral holes punched near its bottom edge to provide auxiliary flow holes.
The smaller BWR assembly dimensions require the Boral and sheathing to be smaller than the PWR version. The BWR Boral dimensions are 150" long and 5" wide for inside cell walls and 3.5" wide for rack periphery walls, if available, otherwise 5" wide will also be used at these locations. Sheathing is similar to the PWR version, except smaller to accomodate the smaller Boral dimensions.
The composite box assemblage is prepared in identical fashion to the PWR racks and will also appear much the same as the typical array shown in Figure 2.6.5.
Baseplate: The 3/4" baseplate is fabricated to run continuously beneath the entire array of cell assemblage. Similar to the PWR racks, flow holes are placed on the centerlines of each rack cell. The flow'holes in the BWR racks are prepared to accommodate and support the bottom BWR assembly nozzle by providing a 3.8125 inch opening which is tapered larger at the top to conform to the configuration of the BWR lower fitting. At lift locations ears are included to allow for insertion and engagement of the lift rig: The baseplate is attached to the base of the cell assemblage by fillet welds and extends horizontally approximately 1/4" beyond the periphery of the rack cells.
- As mentioned in the preceding section, Boral is used as the neutron absorber material. Each storage cell side is equipped with one integral Boral sheet (poison material), except for the outer walls of some of the peripheral rack cells. Only one Boral sheet is required between adjacent cells containing fuel. Therefore, outer rack walls which face each other do not both require Boral and consequently one of the two racks may be fabricated without poison along one outer wall. Similarly, for cells facing the pool walls, Boral shielding is not required.
~:
iv. Sheathing: As described earlier, the sheathing serves as the locator and retainer of the poison material.
V. All pp hg h dj bl yp I i Rg 2,5.7.
The inch diameter top (female threaded) portion is made of austenitic steel Holtec International 2-15 Holtec Report HI-971760
material. The bottom (male threaded) part is made of 17:4 Ph series stainless steel to avoid galling problems. gussets plates are welded to each of the support legs. Each support leg is equipped with a readily accessible socket to enable remote leveling of the rack after its placement in the pool. The support legs are located at the centerlines of cells to ensure accessibility of the levelling tool through the 3.8125 inch diameter flow hole.
One advantage of the BWR assembly configuration is that it provides open areas at its base to allow coolant flow directly up and through the assembly. The rack design exploits this advantage by allowing flow through the baseplate holes, even at locations where support legs would normally interfere. This flow is accomplished by providing diameter flow holes in the support legs.
The assembly of the rack modules is carried out by welding the composite boxes in a vertical fixture with the precision fabricated baseplate serving as the bottom positioner.
An elevation view of three PWR and BWR storage cells is shown in Figures 2.6.2 and 2.6.3, respectively.
Holtec International Holtec Report HI-971760
Table 2.1.1 GEOMETRIC AND PHYSICAL DATA FOR POOL C RACK MODULES Rack Number of Number of Dimension (inches) Shipping Submerged I.D. Type Cells Cells Per Weight (lbs) Weight (lbs) tt N-S Module N-S Direction E-W Direction Al PWR 99 99.5 81.5 14,770 12,850 A2 PWR 99 99.5 81.5 15,620 13,590 B1 PWR 81 81.5 81.5 12,250 10,660 B2 PWR 81.5 81.5 12,940 11,260 B3 PWR 81 81.5 81.5 12,250 10,660 B4 PWR 81 81.5 81.5 12,600 10,960 B5 PWR 81 81.5 81.5 12,250 10,660 B6= PWR 81 81.5 81.5 12,600 10,960 B7 PWR 81 81.5 81.5 12,250 10,660 B8 PWR 81 81.5 81.5 12,250 10,660 B9 PWR 81 81.5 81.5 11,910 10,360 C1 BWR 13 50.5 81.5 9,710 8,450 C2 BWR 13 50.5 81.5 9,710 8,450 Dl BWR 88 50.5 69.0 8,460 7,360 D2 BWR 88 50.5 69.0 8,460 7,360 All dimensions are rounded off to the nearest 0.5 inch, and all weights are rounded off to the nearest 10 lbs.
See Figure 1.2 for pool configuration.
Holtec International 2-17 Holtec Report HI-971760
Table 2.1.1 (Cont'd.)
GEOMETRIC AND PHYSICAL DATA FOR POOL C RACK MODULES'ack Number of Number of Dimension (inches) Shipping Submerged I.D. Cells Cells Per Weight (lbs) Weight (lbs)
N-S E-W Module N-S Direction E-W Direction El BWR 13 13 169 81.5 81.5 15,370 13,370 E2 BWR 13 13 169 81.5 81.5 15,700 13,660 BWR 13 13 169 81.5 81.5 15,700 13,660 BWR 13 13 169 81.5 81.5 15,700 13,660 BWR 13 13 - 169 81.5 81.5 15,700 13,660 BWR 13 13 169 81.5 81.5 15,370 13,370 E7 BWR 13 13 169 81.5 81.5 15,700 13,660 E8 BWR 13 13 169 81.5 81.5 15,700 13,660 E9 BWR 13 13 169 81.5 81.5 15,370 13,370 Fl BWR 13 143 81.5 69.0 13,380 11,640 BWR 13 143 81.5 69.0 13,380 11,640 BWR 13 143 81.5 69.0 13,380 11,640 BWR 13 143 81.5 69.0 13,380 11,640 F5 BWR 13 143. 81.5 69.0 13,380 11,640 F6 BWR 13 143 81.5 69.0 13,100 11,400 All dimensions are rounded off to the nearest 0.5 inch, and all weights are rounded off to the nearest 10 lbs.
See Figure 1.2 for pool configuration.
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Table 2.1.2 GEOMETRIC AND PHYSICAL DATA FOR POOL D RACK MODULESt Rack Number of Number of Dimension (inches) Shipping Submerged I.D. Type Cells Cells Per Weight (lbs) Weight (lbs) tt N-S E-W Module N-S Direction E-W Direction A1 PWR 10 8 80 90.5 72.5 12,080 10,510 A2 PWR 10 8 80 90.5 72.5 12,460 10.840 A3 PWR 10 8 80 90.5 72.5 12,080 10,510 A4 PWR 10 8 80 90.5 72.5 12,460 10.840 A5 PWR 10 8 80 90.5 72.5 11,770 10,240 A6 PWR 10 8 80 90.5 72.5 12,150 10,570 B1 PWR 10 9 90 90.5 81.5 13,900 12,090 B2 PWR 10 9 90 90.5 81.5 13,900 12,090 B3 PWR 10 9 90 90.5 81.5 13,550 11,790 C1 PWR 11 8 88 99.5 72.5 13,200 11,480 PWR 11 88 99.5 72.5 13,620 11,850 D PWR 11 9 99 99.5 81.5 15,190 13,220 All dimensions are rounded off to the nearest 0.5 inch, and all weights are rounded off to the nearest 10 lbs.
See Figure 1.3 for pool configuration.
Holtec International 2-19 Holtec Report HI-971760
Table 2.5.1 MODULE DATA FOR HARRIS SPENT FUEL RACKS Parameter PWR BWR Storage cell inside dimension (nominal) 8.80 in. 6.06 in.
Cell pitch (nominal) 9.00 in. 6.25 in.
Storage cell height (above the baseplate) 169 in. 169 in.
Baseplate hole size (away from pedestal) 5.0 in. 3.8125 in.
Baseplate thickness 0.75 in. 0.75 in.
Support leg height (nominal) 5.5 in. 5.5 in.
Support leg type Remotely adjustable legs Remotely adjustable legs Number of support pedestals Remote lifting and handling provisions Yes Yes Poison material Boral Boral Poison length 145 in 150 in.
Poison width 7.5 ill 5.0 in. (interior) /
3.5 in. (exterior)
'arrower 3.5" exterior Boral will be used subject to availability, otherwise the 5.0" width will be used.
Holtec International 2-20 Holtec Report HI-971760
FIGURE P..i.i; PICTORIAL VIEW OF TYPICAL EEARRIS RACK STRUCTURE III - 971760
I ALIXIL ARY FLOW HOLE
( TYPICAL )
WELO SEAN FIGURE 2.6.1; SEAM WELDED PRECISION FORMED CHANNELS HI-971760 2-22
OEVELOPEO CELL PITCH FUEL POISON PANEL ASSEMBLY I I I { I I CELL ACTIVE HEIGHT POISON LENGTH I I I FLOW HOLE SHEATHING (TYP)
BASEPLATE
~~ BASEPLATE HOLE FIGURE 2.6 2; THREE PWR CELLS IN ELEVATION VIEW HI-971760 2-23
OEVELOPEO CELL FUEL ASSEMBLY PITCH I I POISON PANEL I I I
CELL ACTIVE HEIGHT POISON LENGTH I
I I FLQM HOLE SHEATHING (TYP)
BASEPLATE BASEPLATE HOLE FIGURE 2.6.3; THREE HWR CELLS IN ELEVATION VIE%
HI-971760 2-24
C)
K)
CD I
L3 UJ I
I LU C3 C3 LL
/S ~ I I) l I l l' ~
f a i ~ ~ i ~ ~
5
~ g ~
sl
CELL BASEPLATE LEVELING TOOL SOCKET I I 55'05 I I I I I I I I DRAIN HOLE ELEVATION C1 RACK CORNER PLAN FIGURE 2.6.6; SUPPORT PEDESTAL FOR PIER RACK HI-971760
LEVELING TOOL SOCKET BASEPLATE CELL 5.5'40.5'QSSET DRAIN HOLE I~LEV '~IO T
RACK CORNER PLAN FIGURE 2.6.7; SUPPORT PEDESTAL FOR HOLTEC BWR RACKS HI-971760
3.0 MATERIAL,HEAVYLOAD, AND CONSTRUCTION CONSIDERATIONS 3 1 Intmdulion Safe storage of nuclear fuel in the Harris pools requires that the materials utilized in the rack fabrication be of proven durability and be compatible with the pool water environment.
Likewise, all activities during the rack installations must comply with the provisions of NUIKG-0612 to eliminate the potential of construction accidents. This section provides a synopsis of the considerations with regard to long-term service life and short-term construction safety.
3.2 The following structural materials are utilized in the fabrication of the new spent fuel racks:
- a. ASME SA240-304L for all sheet metal stock
- b. Internally threaded support legs: ASME SA240-304L
- c. Externally threaded support spindle: ASME SA564-630 precipitation hardened, stainless steel (heat treated to 1100'F)
- d. Weld material - per the following ASME specification: SFA 5.9 ER308L 3.3 The racks employ Boral',a patented product of AAR Manufacturing, as the neutron absorber material. Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum. Boron carbide is a compound having a high boron content in a physically stable and chemically inert form. The 1100 alloy aluminum is a lightweight metal with high tensile strength which is protected Rom 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 nuclear reactor or a spent fuel pool. Boral has been shown [3.3.1] to be superior to alternative materials previously used as neutron absorbers in storage racks.
Holtec International Holtec Report HI-971760
Boral has been the most widely used neutron absorbing material in fuel rack applications over the past 20 years. Its use in the spent fuel pools as the neutron absorbing material can be attributed to its proven performance (over 150 pool years of experience) and the following unique characteristics:
The content and placement of boron carbide provides a very high removal cross-section for thermal neutrons.
ii. Boron carbide, in the form of fine particles, is homogeneously dispersed throughout the central layer of the Boral panels.
iii. The boron carbide and aluminum materials in Boral do not degrade as a result of long-term exposure to 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 willbe manufactured by AAR Manufacturing under the control and surveillance of a Quality Assurarice/Quality Control Program that conforms to the requirements of 10CFR50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants". As indicated in Tables 3.3.1 and 3.3.2, Boral has been licensed by the USNRC for use in numerous PWR and BWR spent fuel storage racks and has been extensively used in international nuclear installations.
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 vessels and storage tanks, chemical equipment, reflectors, and sheet metal work.
It has high resistance to corrosion in industrial and marine atmospheres. Aluminum has atomic number of 13, atomic weight of 26.98, specific gravity of 2.69 and valence of 3. The physical, mechanical and chemical properties of the 1100 alloy aluminum are listed in Tables 3.3.3 and 3.3.4.
Holtee International 3-2 Holtec Report HI-971760
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.
Boron Carbide: The boron carbide contained in Boral is a fine granulated powder that conforms to ASTM C-750-80 nuclear grade Type III. The material conforms to the chemical composition and properties listed in Table 3.3.5.
References [3.3.2], [3.3.3], and [3.3.4] provide further discussion as to the suitability of these materials for use in spent fuel storage module applications.
3.4 All materials used in the construction of the Holtec racks have an established history of in-pool usage. Their physical, chemical and radiological compatibility with the pool environment is an established fact.
Austenitic stainless steel (304L) is perhaps the most widely used stainless alloy in nuclear power plants, since it provides both high strength and non-corrosive properties.
3.5 The Fuel Handling Building auxiliary crane willbe used for installation of the new storage racks in pools C and D. The Spent Fuel Cask Handling Crane (CHC) cannot be used for rack installation, since travel limitations prohibit its movement over the spent fuel pools. Storage capacity will be increased starting in the south end of pool C and proceeding north. This installation pattern will enable the storage racks to be manipulated without lifts over spent fuel.
Holtec International 3-3 Holtec Report HI-971760
0 The auxiliary crane is a single failure proof crane and is currently rated for 10 tons.
hoist willbe attached to the auxiliary crane hook to prevent submergence A 20 ton of the auxiliary crane hook. The auxiliary crane was used for installation of storage racks in pool B. Rigging and procedures for pools C and D rack installation willbe similiar to those used previously.
The maximum liftweight during rack installation is determined by the following table.
Item Weight (Ibs)
Rack 15,700 (maximum)
LiftRig 1,200 Rigging 500 20 ton hoist 1,420 Total Lift 18,820 The rack sizes were limited to ensure that the crane and lifting components remain single failure proof and it may be seen that the maximum liftof 18,820 lbs is below the auxiliary crane rating of 20,000 lbs. As a result, the auxiliary crane, which can travel over both pools C and D, is qualified to accept the anticipated load during the rack installation project.
A remotely engagable lift rig, meeting NUREG-0612 [3.5.1] stress criteria, will be used to lift the new modules. The rig is designed for handling both PWR and BWR racks. The new rack lift rig consists of independently loaded liftrods in a liftconfiguration which ensures that failure of one traction rod will not result in uncontrolled lowering of the load being carried by the rig (which complies with the duality feature called for in Section 5.1.6(3a) of NUREG 0612).
The rigs have the following attributes:
Holtec International 3-4 Holtec Report HI-971760
- a. The traction rod is designed to prevent loss of its engagement with the rig in the locked position. Moreover, the locked configuration can be directly verified fi'om above the pool water without the aid of an underwater camera.
- b. The stress analysis of the rigs willbe carried out using a finite element code, and the primary stress limits in ANSI 14.6-1978 [3.5.2] willbe shown to be met by detailed analysis.
- c. The rigs willbe load tested with 300% of the maximum weight to be lifted. The test weight willbe maintained in the air for 10 minutes. All critical weld joints will be liquid penetrant examined to establish the soundness of all critical joints.
Pursuant to the defense-in-depth approach of NUREG-0612, the following additional measures of safety will be undertaken for the racking operation.
The crane used in the project willbe given a preventive maintenance checkup and inspection per the Harris maintenance procedures before the beginning of the racking operation.
ii. Safe load paths willbe developed for moving the new racks in the Fuel Handling Building. The racks will not be carried directly over any fuel located in the pool.
iii. The rack upending and laying down will be carried out in an area which precludes any adverse interaction with safety related equipment.
iv. All crew members involved in the use of the lifting and upending equipment will be given training similar to that utilized in previous rack installation operations.
The rack installation activities willrequire Harris PNSC approval and willbe conducted in accordance with written procedures which willbe reviewed and approved by Carolina Power 0 Light.
The proposed heavy loads compliance will be in accordance with the objectives of the CPRL, NRC-approved submittal to NUREG-0612. The guidelines of NUREG-0612 call for measures to "provide an adequate defense-in-depth for handling of heavy loads near spent fuel...". The NUREG-0612 guidelines cite four major causes of load handling accidents, namely Holtec International 3-5 Holtec Report HI-971760
operator errors rigging failure lack of adequate inspection inadequate procedures The Harris racking program ensures maximum emphasis on mitigating the potential load drop accidents by implementing measures to eliminate shortcomings in all aspects of the operation including the four aforementioned areas. A summary of the measures specifically planned to deal with the major causes is provided below.
Operator errors: As mentioned above, CP&L plans to provide comprehensive training to the installation crew. All training shall be in compliance with ANSI B30.2 [3.5.3].
Rigging failure: The liAing device designed for handling and installation of the new racks at Harris has redundancies in the liftlegs and lifteyes such that there are four independent load members. Failure of any one load bearing member would not lead to uncontrolled lowering of the load. The rig complies with all provisions of ANSI 14.6 [3.5.2], including compliance with the primary stress criteria, load testing at 300% of maximum liftload, and dye examination of critical welds.
The Harris rig design is similar to the rigs used in the initial racking or the rerack of numerous other plants, such as Hope Creek,'Millstone Unit 1, Indian Point Unit Two, Ulchin II, Laguna Verde, J.A. FitzPatrick and Three Mile Island Unit 1.
Lack of adequate inspection: The designer of the racks will develop a set of QC hold points which willrequire inspections and approvals prior to proceeding. Additional hold points will be established for activities during the installatin process. These inspection points have been proven to significantly reduce any requirement for rework or instances of erroneous installation in numerous prior rerack projects.
Inadequate procedures: CP&L is developing various operating procedures to address operations pertaining to the rack installation effort, including, but not limited to, mobilization, rack handling, upending, lifting, installation, verticality, alignment, dummy gage testing, site safety, and ALARAcompliance. Many of the procedures willbe the same or revisions to those developed and currently in use for rack installations in pool B.
The series of operating procedures planned for Harris rack installations are the successors of the procedures successfully implemented in previous projects.
Holtec International 3-6 Holtec Report HI-971760
0' Table 3.5.1 provides a synopsis of the requirements delineated in NUREG-0612, and their intended compliance.
Holtec International 3-7 Holtec Report HI-971760
[3.3.1] "Nuclear Engineering International," July 1997 issue, pp 20-23.
[3.3.2] "Spent Fuel Storage Module Corrosion Report," Brooks Ec Perkins Report 554, June 1, 1977.
[3.3.3] "Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools," Brooks k, Perkins Report 578, July 7, 1978.
[3.3.4] "Boral Neutron Absorbing/Shielding Material - Product Performance Report," Brooks 0 Perkins Report 624, July 20, 1982.
[3.5.1] NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," July 1980.
[3.5.2] ANSI N14.6-1978, Standard for Special Lifting Devices for Shipping Containers Weighing 10000 Pounds or more for Nuclear Materials," American National Standard Institute, Inc., 1978.
[3.5.3] ANSVASME B30.2, "Overhead and Gantry Cranes, (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist)," American Society of Mechanical Engineers, 1983.
[3.5.4] ANSVASME B30.20, "Below-the-Hook LiftingDevices," American Society of Mechanical Engineers, 1993.
[3.5.5] CMMASpecification 70, "Electrical Overhead Travelling Cranes," Crane Manufacturers Association of America, Inc., 1983.
[3.5.6] ANSVASME B30.20, "Below-the-Hook LiftingDevices," American Society of Mechanical Engineers, 1993.
Holtec International 3-8 Holtec Report HI-971760
Table 3.3.1 BORAL EXPERIENCE LIST - PWRs Plant Utility Docket No. Mfg. Year Maine Yankee Maine Yankee Atomic Power 50-309 1977 Donald C. Cook Indiana & Michigan Electric 50-315/316 1979 Sequoyah 1,2 Tennessee Valley Authority 50-327/328 1979 Salem 1,2 Public Service Electric & Gas 50-272/311 1980 Zion 1,2 Commonwealth Edison Co. 50-295/304 1980 Bellefonte 1, 2 Tennessee Valley Authority 50-438/439 1981 Yankee Rowe Yankee Atomic Power 50-29 1964/1983 Indian Point 3 NY Power Authority 50-286 1987 Byron 1,2 Commonwealth Edison Co. 50<54/455 1988 Braidwood 1,2 Commonwealth Edison Co. 50456/457 1988 Yankee Rowe Yankee Atomic Power 50-29 1988 Three Mile Island I GPU Nuclear 50-289 1990 Sequoyah (rerack) Tennessee Valley Authority 50-327 1992 Donald C. Cook (rerack) American Electric Power 50-315/316 1992 Beaver Valley Unit 1 Duquesne Light Company 50-334 1993 Fort Calhoun Omaha Public Power District 50-285 1993 Zion 1 & 2 (rerack) Commonwealth Edison Co. 50-295/304 1993 Salem Units 1 & 2 (rerack) Public Gas and Electric Company 50-272/311 1995 Haddam Neck Connecticut Yankee Atomic Power 50-213 1996 Company Gosgen Kernkraftwerk Gosgen-Daniken AG 1984 (Switzerland)
Koeberg 1,2 ESCOM (South Africa) 1985 Beznau 1,2 Nordostschweizerische Kraftwerke 1985 AG (Switzerland)
Holtec International 3-9 Holtec Report HI-971760
Table 3.3.1 (Cont'd.)
BORAL EXPERIENCE LIST - PWRs Plant Utility Docket No. Mfg. Year 12 various Plants Electricite de France (France) 1986 Ulchin Unit 1 Korea Electric Power Company 1995 (Korea)
Ulchin Unit 2 Korea Electric Power Company 1996 (Korea)
Kori-4 Korea Electric Power Company 1996 (Korea)
Yonggwang 1,2 Korea Electric Power Company 1996 (Korea)
Sizewell B Nuclear Electric, pic (United 1997 Kingdom)
Angra 1 Furnas Centrais-Electricas SA 1997 (Brazil)
Holtec International 3-10 Holtec Report Hl-971760
Table 3.3.2 BORAL EXPERIENCE LIST - BWRs Plant Utility Docket No. Mfg. Year Cooper Nebraska Public Power 50-298 1979 J.A. FitzPatrick NY Power Authority 50-333 1978 Duane Arnold Iowa Electric Light & Power 50-331 1979 Browns Ferry 1,2,3 Tennessee Valley Authority 50- 1980 259/260/296 Brunswick 1,2 Carolina Power & Light 50-324/325 1981 Clinton Illinois Power 50-461/462 1981 Dresden 2,3 Commonwealth Edison Company 50-237/249 1981 E.I. Hatch 1,2 Georgia Power 50-321/366 1981 Hope Creek Public Service Electric & Gas 50-354/355 1985 Humboldt Bay Pacific Gas & Electric Company 50-133 1985 LaCrosse Dairyland Power 50-409 1976 Limerick 1,2 Philadelphia Electric Company 50-352/353 1980 Monticello Northern States Power 50-263 1978 Peachbottom 2,3 Philadelphia Electric 50-277/278 1980 Perry 1,2 Cleveland Electric Illuminating 50-440/441 1979 Pilgrim Boston Edison Company 50-293 1978 Susquehanna 1,2 Pennsylvania Power & Light 50-387,388 1979 Vermont Yankee Vermont Yankee Atomic Power 50-271 1978/1986 Hope Creek Public Service Electric & Gas 50-354/355 1989
& Light
'991 Shearon Harris Pool B Carolina Power 50-401 Duane Arnold Iowa Electric Light & Power 50-331 1993 Pilgrim Boston Edison Company 50-293 1993 LaSalle 1 Commonwealth Edison Company 50-373 1992 Millstone Unit 1 Northeast Utilities 50-245 1989 James A. FitzPatrick NY Power Authority 50-333 1990 Hope Creek Public Service Electric & Gas Company 50-354 1991 Holtec International 3-11 Holtee Report HI-971760
Table 3.3.2 (Cont'd.)
BORAL EXPERIENCE LIST - BWRs Plant Utility Docket No. Mfg. Year Duane Arnold Energy Iowa Electric Power Company 50-331 1994 Center Limerick Units 1,2 PECO Energy 50-352/50- 1994 353 Shearon Harris Pool Carolina Power 8c Light Company 50-400 1996
'B'ine Mile Point Unit 1 Niagara Mohawk Power Corporation 50-220 1997 Chinshan 1,2 Taiwan Power Company (Taiwan) 1986 Kuosheng 1,2 Taiwan Power Company (Taiwan) 1991 Laguna Verde 1,2 Comision Federal de Electricidad 1991 (Mexico)
Holtec International 3-12 Holtec Rcport HI-971760
Table 3.3.3 1100 ALLOYALUMINUMPHYSICAL CHARACTERISTICS Density 0.098 lb/in3 Melting Range 1190'F - 1215'F Thermal Conductivity (77'F) 128 BTU/hr/ft /F/ft Coefficient of Thermal Expansion 13. 1 x 10~ in/in-'F (68'F - 212'F)
Specific Heat (221'F) 0.22 BTU/lb/'F Modulus of Elasticity 10 x 10'psi Tensile Strength (75'F) 13,000 psi (annealed) 18,000 psi (as rolled)
Yield Strength (75'F) 5,000 psi (annealed 17,000 psi (as rolled)
Elongation (75'F) 3M5% (annealed) 9-20% (as rolled)
Hardness (Brinell) 23 (annealed) 32 (as rolled)
Annealing Temperature 650'F Holtec International 3-13 Holtec Report HI-971760
Table 3.3.4 CHEMICALCOMPOSITION - ALUMINUM (1100 ALLOY) 99.00% min. Aluminum 1.00% max. Silicone and Iron 0.05-0.20% max. Copper 0.05% max. Manganese 0.10% max. Zinc 0.15% max. Other Holtec International 3-14 Holtec Report HI-971760
Table 3.3.5 CHEMICALCOMPOSITION AND PHYSICAL PROPERTIES OF BORON CARBIDE CHEMICALCOMPOSITION (WEIGHT PERCENT)
Total boron 70.0 min.
B'sotopic content in natural boron 18.0 Boric oxide 3.0 max.
Iron 2.0 max.
Total boron plus total carbon 94.0 min.
PHYSICAL PROPERTIES Chemical formula B4C Boron content (weight percent) 78.28%
Carbon content (weight percent) 21.72%
Crystal structure rhombohedral Density 0.0907 lb/in'elting Point 4442'F Boiling Point 6332'F Holtec International 3-15 Holtec Report HI-971760
Table 3.5.1 HEAVY LOAD HANDLINGCOMPLIANCE MATRIX(NUREG-0612)
Criterion Compliance
- 1. Are safe load paths defined for the Yes movement of heavy loads to minimize the potential of impact, if dropped, on irradiated fuel?
Will procedures be developed to cover: Yes identification of required equipment, inspection and acceptance criteria required before movement of load, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?
- 3. Will crane operators be trained and Yes qualified?
- 4. Will special lifting devices meet the Yes guidelines of ANSI 14.6-1978?
I
- 5. Will non-custom lifting devices be Yes installed and used in accordance with ANSI B30.20, latest edition?
- 6. Will the cranes be inspected and tested Yes prior to use in rack installation?
- 7. Does the crane meet the intent of ANSI Yes B30.2-1976 and CMMA-70?
Holtec International 3-16 Holtec Report HI-971760
4.0 CRITICALITYSAFETY EVALUATION 4,1 II The high density spent fuel PWR and BWR storage racks for Harris Pools C and D are designed in accordance with the applicable codes listed below. The rack design and fuel storage configuration acceptance criteria is to show that the effective neutron multiplication factor, k,n, is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and Qooded with un-borated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerances. All uncertainties are statistically combined, with unce'itainties applied conservatively to calculate the final k,tr which must be shown to be less than 0.95 with a 95% probability at a 95% confidence level [4.1.1]. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under credible abnormal and accident conditions, the reactivity will not exceed the limiting design basis value.
4
. Applicable codes, standards, and regulations or pertinent sections thereof, include the following:
~ General Design Criteria 62, Prevention of Criticality in Fuel Storage and Handling.
~ USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3
- July 1981.
~ USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.
~ USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981.
~ ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.
~ ANSVANS-57.2-1983, Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
Holtec International 4-1 Report HI-971760
~ ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication factor, k,tn including uncertainties, shall be less than or equal to 0.95. The infinite multiplication factor, k; is calculated for an infinite array, neglecting neutron losses due to leakage Rom the actual storage rack, and therefore is a higher and more conservative value. In the present criticality safety evaluation of the Harris storage racks, the design basis criterion was assumed to be a k;, of less than 0.95, which is more conservative than the limit specified in the regulatory guidelines.
To ensure that the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made:
~ Moderator is un-borated water at a temperature (4'C) that results in the highest reactivity.
~ In all cases (except for the assessment of peripheral effects and certain abnormaVaccident
. conditions where neutron leakage is inherent), the infinite multiplication factor, k;was used rather than the effective multiplication factor, k,n (i.e., neutron loss &om radial and axial leakage neglected).
~ Neutron absorption in minor structural members is neglected, i.e., spacer grids are analytically replaced by water.
~ The racks were assumed to be fully loaded with the most reactive fuel authorized to be stored in the facility. In the analysis, no credit was taken for any control rods or burnable poison (IFBA rods for the PWR fuel or gadolinia for BWR fuel), or soluble boron in the pool water which may be present.
~ In-core depletion calculations assume conservative operating conditions, highest fuel and moderator temperature, an allowance for the soluble boron concentrations during in-core PWR operations and an allowance for voids during in-core BWR operations.
The PWR spent fuel storage racks are designed to accommodate any and all of the fuel assemblies listed in Table 4.3.1 with a maximum enrichment of 5 wt% "'U. To assure the acceptability of the racks for storage of any and all of the above assembly types, the most Holtec International 4-2 Report HI-971760
~ '
reactive fuel assembly type was identified and used as the design basis fuel assembly. The Westinghouse 15x15 assembly was determined to have the highest reactivity at zero burnup and as a function of burnup for an initial 5 wt% "'U enrichment and therefore was used as the design basis PWR fuel assembly.
The BWR spent fuel storage racks are designed to accommodate any and all of the fuel assemblies listed in Table 4.3.2 with a maximum planar average enrichments of 4.6 wt.% 'U.
Each fuel assembly type was analyzed independently to determine its acceptability in the rack. It is noted that individual fuel rods can have enrichments that are less than or greater than the maximum planar average enrichment.
4.2 Summa of Criticalit Anal ses 4.2.1 Normal 0 eratin Conditions 4.2.1.1 PWR Fuel Results The design basis PWR fuel assembly is a 15 x 15 Westinghouse fuel assembly containing UO, at a maximum initial enrichment of 5.0 wt% "U. All fuel assembly types listed in Table 4.3.1 were also evaluated and the Westinghouse 15x15 assembly was shown to exhibit the highest reactivity for the high density PWR storage racks at Harris.
The NRC guidelines specify that the limiting k,tr of 0.95 under normal storage conditions should be evaluated in the absence of soluble boron. Consequences of abnormal and accident conditions have also been evaluated assuming no soluble boron, where "abnormal" refers to conditions (such as higher water temperatures) which may reasonably be expected to occur during the lifetime of the plant and "accident" refers to conditions which are not expected to occur but nevertheless must be protected against.
The criticality analyses of the spent fuel storage pool are summarized in Table 4.2.1 for the design basis storage conditions. The maximum k;, is 0.9450 (95% probability at the 95% con-Holtec International 4-3 Report HI-971760
fidence level) for the enrichment-burnup combinations shown in Figure 4.2.1. The calculated maximum reactivity includes burnup-dependent allowances for uncertainty in depletion calculations and for the axial distribution in burnup. Reactivity allowances for manufacturing tolerances and calculational uncertainties are also included. As cooling time increases in long-term storage, decay of Pu-241 and growth of Am-241 results in a significant decrease in reac-tivity, which willprovide a continuously increasing subcriticality margin for the next 100 years.
The racks can safely accommodate fuel of various initial enrichments and discharge fuel burnups, provided the combination falls within the acceptable domain above the curve in Figure 4.2.1. For convenience, the minimum (limiting) burnup data for unrestricted storage can be described as a linear function of the initial enrichment (E, in weight percent "'U) which conservatively encompasses the limiting burnup data. The equation for this curve is shown in Figure 4.2.1 and provided below.
For Unrestricted Storage of the following PWR fuel assemblies Westinghouse 17x17 Std Westinghouse 17x17 V5 Westinghouse 15x15 Siemens 17x17 Siemens 15x15 the enrichment must be less than or equal to 5 wt% "'U and the burnup must satisfy the minimum burnup requirements Minimum Burnup in MWD/MTU= 12114*E-19123 The burnup criteria willbe implemented by appropriate administrative procedures to ensure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2, prior to fuel transfer into Spent Fuel Pools C or D.
4.2.1.2 BWR Fuel Results AllBWR fuel assembly types being considered were explicitly analyzed to determine the acceptability for storage in Spent Fuel Pool C. The maximum planar average enrichment was Holtec International 44 Rcport HI-971760
assumed for all rods in the assembly and no credit was taken for gadolinia which might be present.
The criticality safety was evaluated at the burnup corresponding to a k;, of 1.32 in the Standard Cold Core Geometry (SCCG). SCCG is defined as an infinite array of fuel assemblies on a 6-inch lattice spacing at 20'C, without any control absorber or voids.
4 The maximum k;, in the BWR storage rack was determined to be 0.9443 (95% probability at the 95% confidence level) including all known calculational and manufacturing uncertainties. In addition, a conservative allowance of 0.01 bk for possible differences between fuel vendor calculations and those reported here was included. This allowance also encompasses any uncertainty in the burnup calculations.
The basic calculations supporting the criticality safety of the Harris fuel storage racks for the design basis fuel are summarized in Table 4.2.2. For the design basis fuel, the fuel storage rack satisfies the USNRC criterion of a maximum k,rr less than or equal to 0.95.
The acceptance criteria for storage of spent BWR fuel in Harris Pool C can be summarized in the following manner.
For Unrestricted Storage of the following BWR fuel assemblies GE 3, GE 4, GE 5, GE 6, GE 7, GE 8, GE 9, GE 10, GE 13 the maximum planar average enrichment must be less than or equal to 4.6 wt.% "'U and the k;, in standard cold core geometry must be less than or equal to 1.32 4.3 In ut Parameters 4.3.1 Reference PWR Fuel Assembl and tora e ell The design basis PWR fuel assembly is a 15x15 array of fuel rods with 21 rods replaced by 21 control rod guide tubes. Table 4.3.1 summarizes the PWR fuel assembly design specifications Holtec International 4-5 Report HI-971760
for all fuel assemblies analyzed. Figure 4.3.1 shows the calculational model of the PWR spent fuel storage cell containing a 15x15 assembly.
The design basis for the Region 2 type storage cells is fuel of 5.0 wt.% 'V maximum initial enrichment burned to 41,447 MWD/MTU. The storage cells, consisting of an egg-crate structure, are composed of stainless steel walls with a single fixed neutron absorber panel, Boral,'n a 0.107 inch channel. These cells are located on a lattice spacing of 9.017 + ~~:..;-: inch. The 0.075 + ~6<-':."g* thick steel walls define a storage cell which has a 8.8 + ~l- ',t'nch nominal inside, dimension. The Boral absorber has a thickness density of 0.098 + ~ inch and a nominal B-10 areal of 0.0302 g/cm'minimum of ~a~ g/cm'). The width of the Boral absorber panel is 7.5 + ~>'-'."". inches. Boral panels are not needed or used on the exterior walls of modules facing non-fueled regions, i.era the pool walls. However, at least one boral panel is used between storage racks.
4.3.2 Reference BWR Fuel Assembl and Stora e Cell The design basis BWR fuel assembly, used for uncertainty calculations, is a standard 8x8 array of BWR fuel rods containing UO, clad in Zircaloy (60 fuel rods with 4 water rods). Design parameters for all BWR fuel assemblies analyzed are summarized in Table 4.3.2. Figure 4.3.2 shows the calculational model of a BWR storage rack cell containing an 8x8 assembly.
The BWR storage cells, consisting of an egg-crate structure, are composed of stainless steel walls with a single fixed neutron absorber panel, Boral, in a 0.08 inch channel. These cells are located on a lattice spacing of 6.25 a ~i-;.",:: inch. The 0.075 + E"i*.:rQ thick steel watts define a storage cell which has a 6.06 + ~<':.-'i; inch nominal inside dimension. The Boral absorber has a thickness of 0.075 + QQ inch and a nominal B-10 areal density of 0.0162 g/cm'minimum of Q~:-'-";"Qt g/cm'). The width of the Boral absorber panel is 5.0 + ~i"=) inches. Boral panels are not needed or used on the exterior walls of modules facing non-fueled regions, i.e., the pool walls. However, at least one boral panel is used between storage racks. The boral panel used on the outside of the Holtec International '4-6 Report HI-971760
BWR racks is 3.5 inch in width. The minimum B-10 loading on these panels is identical to the loading on the internal panels.
4.4 Anal ical Methodolo 4.4.1 Reference Desi Calculations In the fuel rack analyses, the primary criticality analyses of the high density spent fuel storage racks were performed with a two-dimensional multigroup transport theory technique, using the CASMO-3 computer code [4.4.1 - 4.4.4]. Since CASMO-3 can not be directly compared to critical experiments, a calculational bias is not available for CASMO-3. Therefore, independent verification calculations were made with a Monte Carlo technique utilizing the MCNP-4A computer code [4.4.5]. Benchmark calculations, presented in Appendix A, indicate a bias of 0.0009+ 0.0011 for MCNP-4A, evaluated at the 95% probability, 95% confidence level. The MCNP-4A bias and uncertainty were included in the MCNP-4A to CASMO-3 comparison as discussed in Section 4.5.
CASMO-3 was also used for burnup calculations and for evaluating small reactivity increments associated with manufacturing tolerances. In the geometric model used in the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions (zero neutron current) were used in the axial direction and at the Boral and steel plates between storage cells. These boundary conditions have the effect of creating an infinite array of storage cells in all directions.
MCNP-4A was used to determine reactivity effects, to calculate the reactivity for fuel misloading outside the racks and to determine the effect of having PWR 'and BWR racks adjacent to each other. MCNP-4A Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of the MCNP-4A calculated reactivity, a minimum of 600,000 neutron histories in 200 generations of 3000 neutrons each, are accumulated in each calculation.
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4.4.2 Burnu Calculations and Uncertainties CASMO-3 was used'for burnup calculations during core operations. CASMO-3 has been extensively verified [4.4.4, 4.4.6] against Monte Carlo calculations, reactor operations, and heavy-element concentrations in irradiated fuel. In addition, Johansson [4.4.7] has obtained very good agreement in calculations of close-packed, high-plutonium-content, experimental configurations.
4.4.2.1 PWR Fuel Burnup Calculations Since critical experiment data with spent fuel is not available for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations. Assuming the uncertainty in depletion calculations is less than 5/e of the total reactivity decrement, a burnup dependent uncertainty in reactivity for burnup calculations may be assigned. Table 4.4.1 summarizes results of the burnup analyses to determine the allowances for uncertainties in burnup calculations. The reactivity allowances for uncertainties in burnup are listed for three different burnup ranges: less than 30,000 MWD/MTU, between 30,000 and 40,0000 MWD/MTU,and between 40,000 and 45,000 MWD/MTU. The appropriate uncertainty was used for each burnup range in determining the acceptable burnup versus enrichment combinations depicted in Figure 4.2.1. The allowance for uncertainty in bumup calculations is believed to be a conservative estimate, particularly in view of the substan-tial reactivity decrease with aged fuel as discussed in Section'4.4.4.
4.4.2.2 BWR Fuel Burnup Calculations and Comparison to Vendor Calculations CASMO-3 was used to perform depletion calculations and to calculate the k;, in the SCCG. As discussed, there are no depleted fuel critical experiments with which to benchmark CASMO-3's d epletion calculations. Therefore a reactivity allowance for uncertainty in depletion is needed.
Instead of using 5% of the reactivity decrement, as in the case of the PWR assemblies, a flat Holtec International 4-8 Report HI-971760
reactivity allowance of 0.01 hk is used. This value is not statistically combined with the other uncertainties but rather added directly to the calculated k;The allowance is used to also encompass any potential differences between the SCCG calculations performed here and'he vendor calculations.
4.4.3 Effect of Axial Burnu Distribution Initially, fuel loaded into the reactor willburn with a slightly skewed cosine power distribution.
As burnup progresses, the burnup distribution willtend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. 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 assemblies 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.
V Generic analytic results of the axial burnup effect have been provided by Turner [4.4.8] based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup, becoming positive at burnups greater than about 30,000 MWD/MTU. The trends observed [4.4.8] suggest the possi-bility of a small positive reactivity effect above 30,000 MWD/MTUincreasing to slightly over 1% hk at 40,000 MWD/MTU.
4.4.3.1 PWR Fuel Axial Burnup Distribution Calculations for the Harris storage racks with PWR fuel of three different average burnups were made using an axial burnup distribution representative of spent PWR fuel'. At lower burnups, the The axial burnup distribution measured on spent fuel from the Surry plant was used as representative of PWR fuel.
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reactivity increment is smaller as indicated in Table 4.4. 1, being negative at 30,000 MWD/MTU and at lower burnups. No credit is taken for this negative reactivity effect at the lower burnups other than the suggestion of additional conservatism. Furthermore, the reactivity significantly decreases with time in storage (Section 4.4.4 below) providing a continuously increasing margin below the 0.95 limit.
The appropriate reactivity allowance for the effect of axial burnup distribution was used for each burnup range in determining the acceptable burnup versus enrichment values in Figure 4.2.1.
4.4.3.2 BWR Fuel Axial Burnup Distribution The burnup at which k;, in the SCCG reaches 1.32 is approximately 12,000 MWD/MTU.As discussed above and in [4.4.8] the effect of using the explicit axial burnup distribution as opposed to an average burnup distribution results in a negative effect on reactivity. Therefore, no reactivity allowance for axial burnup distribution is applied to the BWR fuel analysis.
4:4.4 Lon Term Reactivit Chan es At reactor shutdown, the reactivity of the fuel initially decreases due to the growth of Xe-135.
Subsequently, the Xenon decays and the reactivity increases to a maximum at several hundred hours when the Xenon is gone. Over the next 30 years, the reactivity continuously decreases due primarily to Pu-241 decay and Americium growth. At lower burnup, the reactivity decrease will be less pronounced since less Pu-241 would have been produced. No credit is taken for this long-term decrease in reactivity other than to indicate additional and increasing conservatism in the design criticality analysis.
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0 4.5 PWR Stora e Rack Criticali Anal ses and Tolerance Variations 4.5.1 Nominal Desi Case The principal method of analysis for the racks was the CASMO-3 code, using the restart option in CASMO-3 to analytically transfer fuel of a specified burnup into the storage rack configuration at a reference temperature of 4'C (39'F). Calculations were made for fuel of several different initial enrichments and, at each enrichment, a limiting k;,value was established which includes reactivity allowances for manufacturing tolerances, the uncertainty in the burnup analyses and for the effect of the axial burnup distribution on reactivity. The restart CASMO-3 calculations (cold, no-Xenon, rack geometry) were then interpolated to define the burnup value yielding the limiting k;value for each enrichment. A line was fitted to these converged burnup values and this line defines the boundary of the acceptable domain shown in Figure 4.2.1.
An independent MCNP-4A calculation was performed to verify the acceptability of the reference criticality analyses. Fuel of 5.0 wt% initial enrichment was analyzed by MCNP-4A and by CASMO-3. The results of this comparison are presented in Table 4.5.1. In comparing the MCNP values to the CASMO-3 values, the MCNP-4A calculational bias and calculational statistics were included. In addition, the MCNP-4A model correctly included the effect of axial neutron leakage which the CASMO-3 calculations conservatively neglect. Since the MCNP-4A model is at 20 'C and the CASMO-3 model is at 4 'C, a temperature correction had to be applied to the MCNP-4A result. The MCNP-4A result confirms that the reference CASMO-3 calculations are conservative.
4.5.2 Uncertainties Due to PWR Rack Manufacturin Tolerances Allreactivity allowances for manufacturing tolerances are summarized below and listed in Table 4.2.1. Since the tolerances are statistically independent, the allowances are statistically combined into a single reactivity allowance which was used in the final calculations (see Table 4.2.1).
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4.5.2.1 Boron Loading Tolerances The Boral absorber panels used in the storage cells are 0.098 inch thick with a nominal B-10 areal density of 0.0302 g/cm'. The manufacturing limit of B-10 loading assures that at any point the minimum B-10 areal density willnot be less than ~@-,"-."g g/cm . Differential CASMO-3 calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of
+ 0.0041 LUc.
4.5.2.2 Boral Width Tolerance The reference storage cell design uses a Boral absorber width of 7.50 + ~Q inch. CASMO-3 differential calculations show that this tolerance results in a reactivity uncertainty of +0.0009 4k.
4.5.2.3 Tolerance in Cell Lattice Spacing and Cell Box Inner Dimension Since the Region 2 style racks do not utilize a water gap between storage cells, the manufacturing tolerance on inner box dimension is identical to the tolerance on the storage cell lattice spacing. The inner box dimension is 8.8 d: ~! inches. This corresponds to an uncertainty in reactivity of + 0.0017 hk as determined with CASMO-3.
4.5.2.4 Stainless Steel Thickness Tolerance The nominal thickness of the stainless steel box wall is 0.075 inch with a tolerance of + ~t', '.
inch, resulting in an uncertainty in reactivity of + 0.0005 dk as calculated with CASMO-3.
4.5.2.5 Fuel Enrichment and Density Tolerances The maximum fuel enrichment was specified as 5.0 wt.% '"U+ 0.0/- 0.05. This uncertainty results in a negative reactivity effect which is not credited in this analysis. The UO, density was t'>>'"--- g.-"-',-'-",.-"':~: ..'. CASMO-3,calculations show that the reactivity allowance for this tolerance is 2 0.0014 LHc Holtee International 4-12 Report HI-971760
4.6 BWR Stora e Rack riticalit Anal ses and Tolerance Variations 4.6.1 Nominal Desi Case The two-dimensional CASMO-3 code was used as the principal method of analysis for the Harris spent fuel pool BWR racks. CASMO-3 was used to perform depletion calculations on the fuel assembly and using the restart option in CASMO-3 the fuel of a specified burnup was analytically transferred into the storage rack at a reference temperature of 4'C (39'F). The same fuel of a specified burnup was also analytically transferred into the standard cold core geometry (SCCG) configuration which is an infinite lattice with 6 inch spacing at a temperature of 20'C without any burnable absorber or control blades and no voids. AllXenon which was present during the depletion calculations was removed during the restarts in the rack and SCCG. The reactivity effects of the natural uranium blanket normally located at the ends of the assemblies were conservatively neglected since an infinite fuel length was used.
IU All fuel assemblies specified were analyzed at the maximum enrichment specified. The maximum k;, in the SCCG was specified as 1.32. Using the CASMO-3 results, the burnup corresponding to a k;in the SCCG of 1.32 was determined and the corresponding k,, in the rack was determined. The reactivity adjustments were added to the rack k; to determine the maximum value and this was compared against the 0.95 k,tr limit. Based on this analysis, all specified fuel assemblies are acceptable for storage as stated in Section 4.2.1.2. Table 4.2.2 provides the final results of the BWR fuel assembly calculations.
An independent MCNP-4A calculation was used to verify the acceptability of the reference criticality analyses. Fuel of 4.6 wt% initial enrichment was analyzed by MCNP-4A and by CASMO-3. The results of this comparison are presented in Table 4.5.1. In comparing the MCNP values to the CASMO values, the MCNP-4A calculational bias and calculational statistics were included. In addition, the MCNP-4A model correctly included the effect of axial neutron leakage which the CASMO-3 calculations conservatively neglect. Since the MCNP-4A model is at 20 'C Holtec International 4-13 Report HI-971760
the CASMO-3 model is at 4 'C,'
i'nd temperature correction had to be applied to the MCNP-4A result. The MCNP-4A result confirm that the reference CASMO-3 calculations are conservative.
4.6.2 Uncertainties Due to Manufacturin Tolerances The reactivity effects associated with manufacturing tolerances are discussed below and shown in Table 4.2.2. Since the tolerances are statistically independent, the allowances are statistically combined into a single reactivity allowance which was used in the final calculations (see Table 4.2.2).
4.6.2.1 Boron Loading Variation 1
The Boral absorber panels used in the storage cells are nominally 0.075 inch thick, with a B-10 areal density of 0.0162 g/cm'. The manufacturing tolerance limit in B-10 content, including both thickness and composition tolerances assures that the minimum boron-10 areal density will not be less than ~Q g/cm'. Differential CASMO-3 calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of + 0.0053 hk.
4.6.2.2 Boral Width Tolerance Variation The reference storage cell design uses a Boral panel width of 5.00 inches. The tolerance on the Boral width is + Q,"~ inch. Calculations using CASMO-3 showed that this tolerance corresponds to a 2 0.0018 hk uncertainty.
4.6.2.3 Tolerance in Cell Lattice Pitch and Inner Box Dimension Since the Region 2 style racks do not utilize a water gap between storage cells, the manufacturing tolerance on inner box dimension is identical to the tolerance on the storage cell lattice spacing. The inner box dimension is 6.06 2 ~~, "*"
inches. This corresponds to an uncertainty in reactivity of + 0.0037 hk as determined with CASMO-3.
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0 4.6.2.4
~ ~ Stainless Steel Thickness Tolerances The nominal thickness of the stainless steel box is 0.075 d: gg': inches. The maximum positive reactivity effect of the expected stainless steel thickness tolerances was calculated to be
+ 0.0005 LQc using CASMO-3.
4.6.2.5 Fuel Enrichment and Density Variation The maximum planar average fuel enrichment was specified for each fuel assembly analyzed.
Therefore, there is no reactivity allowance for variations in enrichment since the absolute maximum was used for all calculations.
The UO, density was specified for each fuel assembly analyzed. The maximum tolerance on the density is d: ~8""'" g/cc. CASMO-3 calculations show that the reactivity allowance for this tolerance is 2 0.0023 d,k.
4.6.2.6 Zirconium Flow Channel Elimination of the zirconium flow channel results in a small (approximately 0.0024 d,k) decrease in reactivity. More significant is a positive reactivity effect resulting from potential bulging of the zirconium channel, which moves the channel wall outward toward the Boral absorber. It is conservatively assumed that the maximum bulging that could occur would result in the channel touching the cell walls. Since this would not occur over the entire length of the channel, the model assumed that the entire channel was enlarged so that the mid-point of the channel wall was placed equidistant between the nominal channel outer dimension and the cell wall. This results in an incremental reactivity of+ 0.0045 bk as determined with MCNP-4A.
4.7 Abnormal and Accident Conditions Strict administrative controls on the fuel transfer to Pools C and D willpreclude fuel which is outside the range of the previously stated acceptance criteria from being brought into the spent Holtec International 4-15 Report HI-971760
fuel pool. Therefore, the only potential abnormal and accident conditions that exist are the misplacement of a fuel assembly outside the rack or the dropping of a fuel assembly on top of the rack. It is also possible to inadvertently place a BWR spent fuel assembly in the PWR rack.
4.7.1 Tem erature and Water Densi Effects The spent fuel pool temperature coefficient of reactivity is negative. Using the minimum temperature of 4'C, therefore, assures that the true reactivity will always be lower than the calculated value regardless of the temperature. Temperature effects on reactivity have been calculated and the results are shown in Table 4.7.1. Introducing voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown in the table. Boiling at the submerged depth of the racks would occur at approximately 122'C.
4.7.2 Dro ed Fuel Assembl 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 12 inches (which is considered infinite), including an estimated allowance for deformation under seismic or accident conditions. At this separation distance, the effect on reactivity is insignificant.
It is also possible to vertically drop an assembly into a location occupied by another assembly.
Such a vertical impact would at most cause a small compression of the stored assembly, reducing the water-to-fuel ratio and thereby reducing reactivity. In addition the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no neutron interaction between the two assemblies.
Dropping an assembly into an unoccupied cell could result in a localized deformation of the baseplate of the rack. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in the active fuel height no longer being covered by the boral. The immediate eight surrounding fuel cells could also be affected.
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However, the amount of deformation for these cells would be considerably less. The amount of localized deformation would not exceed three inches for a PWR assembly and would therefore be considerably less for the lighter BWR assembly. The criticality effect of this drop accident has been conservatively analyzed and it has been shown that this localized event (nine storage cells at most) has a negligible impact on reactivity.
4.7.3 Lateral Rack Movement Lateral motion of the rack modules under seismic conditions could potentially alter the spacing between rack modules. Region 2 storage cells do not use a flux-trap and the reactivity is therefore insensitive to the spacing between modules. The spacing between modules is suffi-:
ciently large to preclude adverse interaction even with the maximum seismically-induced reduction in spacing.
4.7.4 Abnormal Location of a PWR or BWR Fuel Assembl Strict administrative controls willprevent an unacceptable assembly, as determined by the acceptance criteria stated in Section 4.2, from being transferred to Hams Pools C and D.
Therefore, the only potential mislocation of a fuel assembly is the mislocation of a fuel assembly of equal or lower reactivity to the design basis outside a PWR or BWR rack.
Since the racks will have a Boral panel on the outside face (when the outside face is not against a wall) the reactivity effect of a misloaded fuel assembly outside the rack is negligible because of the neutron leakage that occurs from the rack itself. Therefore, the conservative infinite lattice calculations that were performed have k;, values that are higher than any potential mislocation accidents.
Another mislocation event could occur with a BWR assembly. This would be the inadvertent placement of a BWR assembly in the PWR racks. Since, the BWR assembly is significantly smaller than a PWR assembly, the reactivity effect of placing a BWR assembly in the PWR rack is negligible. The reverse scenario of misplacing a PWR Holtec International 4-17 Report HI-971760
dj assembly in the BWR rack is impossible because of the size of the PWR assembly.
4.7.5 Eccentric Fuel Positionin The fuel assembly is assumed to be normally located in the center of the storage rack cell and in the case of the BWR rack there are bottom fittings and spacers that mechanically restrict lateral movement of the fuel assemblies. Nevertheless, MCNP-4A calculations were 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 positioning results in a decrease in reactivity (by about 0.0051 for the PWR assemblies and 0.0091 for the BWR assemblies). The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells.
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4.8 References
[4.1.1] M. G. Natrella, Ex erimental Statistics National Bureau of Standards Handbook 91, August 1963.
[4.4.1] A. Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly Burnup Program," AE-RF-76-4158, Studsvik report (proprietary).
[4.4.2] A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604, 1977.
[4.4.3] M. Edenius, A. Ahlin, and B. H. Forssen, "CASMO-3 A Fuel Assembly Burnup Program, Users Manual", Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986.
[4.4.4] M. Edenius and A. Ahlin, "CASMO-3: New Features, Benchmarking, and Advanced Applications," Nuclear Science and Engineering, 100, 342-351, (1988).
[4.4.5] J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).
[4.4.6] E. E. Pilat, "Methods for the Analysis of Boiling Water Reactors (Lattice Physics),"
YAEC-1232, Yankee Atomic Electric Co., December 1980.
[4.4.7] E. Johansson, "Reactor Physics Calculations on Close-Packed Pressurized Water Reactor Lattices," Nuclear Technology, Vol. 68, pp. 263-268, February 1985.
[4.4.8] S. E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.
Holtec International 4-19 Report HI-971760
Table 4.2.1 Summary of Criticality Safety Calculations for the PWR Fuel Racks Fuel Assembly Westinghouse 15xl5 Enrichment 5%
Temperature 4'C Burnup &om Calculation (MWD/MTU) 41,352 Burnup &om Curve (MWD/MTU) 41,447 CASMO-3 k;, 0.9126 Uncertainties UO, density 0.0014 Inner box dimension 0.0017 Box wall thickness 0.0005 Boral width 0.0009 B-10 loading 0.0041 Burnup 0.0160 Total Uncertainty at 95%/95% 0.0167 Effect of Axial Burnup Distribution 0.0157 Maximum k, 0.9450 Re ulato Limit 0.9500 Notes:
- 1. Only the most reactive assembly is shown.
- 2. The total uncertainty is a statistical combination of the manufacturing uncertainties.
Holtec International 4-20 Report HI-971760
Table 4.2.2 Summary of Criticality Safety Calculations for the BWR Fuel Racks Fuel Assembly GE3 GE4 GE7 GE8 GE9 GE10 GE13 Temperature 4'C 4'C 4'C 4'C 4'C 4'C 4'C SCCG k;, 1.32 1.32 1.32 1.32 1.32 1.32 1.32 Enrichment 4.6 4.6 4.6 4.6 4.6 4.6 4.6 CASMO-3 k; 0.9163 0.9140 0.9192 0.9214 0.9207 0.9201 0.9227 Uncertainties UO, density 0.0023 Inner box 0.0037 dimension Box wall 0.0005 thickness Boral width 0.0018 B-10 loading 0.0053 Total uncertainty 0.0071 0.0071 0.0071 0.0071 0.0071 0.0071 0.0071 at 95%/95%
Channel bulging 0.0045 0.0045 0.0045 0.0045 0.0045 0.0045 0.0045 Uncertainty for 0.0100 0.0100 0.0100 0.0100 0.0100 0.0100 0.0100 burnup and vendor comparison k,,
'aximum 0.9379 0.9356 0.9408 0.9430 0.9423 0.9417 0.9443 Re ulato Limit 0.9500 0.9500 0.9500 0.9500 0.9500 0.9500 0.9500 Notes:
- 1. The total uncertainty is a statistical combination of the manufacturing uncertainties.
- 2. The GE 13 assembly has part length rods. Two CASMO-3 calculations were performed: one with all rods present and the other with only the full length rods present. The most reactive configuration was the second and the k; from this configuration is presented.
- 3. The GE 5 and GE 6 are identical to the GE 7 for the fuel parameters analyzed and therefore the GE 5 and GE 6 have a maximum k;equivalent to the GE 7.
"4. The enrichment is the planar average enrichment.
Holtec International 4-21 Report HI-971760
Table 4.3.1 PWR Fuel Characteristics Fuel Assembly Westinghouse Westinghouse Westinghouse Siemens Siemens 17x17 Std 17x17 VS 15x15 17x17 15x15 NOTE: All Dimensions in inches Clad O.D. 0.374 0.360 0.422 0.376 0.424 Clad I.D. 0.329 0.315 0.373 0.328 0.364 Clad Material Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Pellet Diameter 0.3225 0.3088 0.3659 0.3215 0.357 Stack Density 10.41 10.41 10.41 K~3 Maximum 5.0 5.0 5.0 5.0 5.0 Enrichment Active Fuel 144 144 144 144 144 Length .
Number Fuel 264 264 204 26$ 204 Rods Fuel Rod Pitch 0.496 0.496 0.563 0.496 0.563 Number of 24/1 24/1 21 21 Thimbles Thimble O.D. 0.482/0.484 0.474/0.476 0.546 0.480 0.544 Thimble I.D. 0.450/0.448 0.442/0.440 0.512 0.448 0.511 The highlighted data in the table above is the property of Westinghouse or Siemens and is proprietary information provided in confidence. Access to this information shall be limited to those individuals having a need for such access and shall not be disclosed or transmitted to any organization without the written permission of Westinghouse or Siemens, respectively.
Holtec International 4-22 Report HI-971760
Table 4.3.2
'WR Fuel Characteristics Fuel Assembly GE 3 GE 4 GE 7 GE 8 GE9 GE10 GE13 NOTE: All dimensions in inches Clad O.D 0.563 0.493 0.483 0.483 0.483 0.483 0.440 Clad I.D. 0.489 0.425 0.419 0.419 0.419 0.419 Clad Material Z1-2 Zl'-2 Zl"2 Zl'-2 Zr-2 Zl"2 Zr"2
'ellet 0.477 0.416 0.410 0.411 0.411 0.411 Diameter Stack Densit 10.31 10.40 10.54 10.58 10.54 10.54 Maximum 4.6 4.6 4.6 4.6 4.6 Enrichment SCCG k; <1.32 <1.32 <1.32 <1.32 <1.32 <1.32 51.32 Active Fuel 144 146 150 150 150 150 Len th Fuel Rod 7x7 8x8 8x8 8x8 8x8 8x8 9x9 Arra Number Fuel 49 63 60 60 60 74 Rods Fuel Rod 0.738 0.640 0.640 0.640 0.640 0.640 0.566 Pitch Number of 0 4 Water Rods Water Rod 0.493 0.591 0.591/ 1.34 1.34 0.980 O.D. 0.483 Water Rod 0.425 0.531 0.531/ 1.26 1.26 I.D. 0.431 Channel I.D 5.278 5.278 5.278 5.278 5.278 5.278 5.278 Channel 0.080 0.080 0.080 0.080 0.080 0.070 0.070 Thickness Notes:
- 1. The GE 13 assembly has 8 part length rods.
- 3. The enrichment is the maximum planar average enrichment.
The highlighted data in the table above is the property of GE and is proprietary information provided in confidence. Access to this information shall be limited to those individuals having a need for such access and shall not be disclosed or transmitted to any organization without the written permission of GE.
Holtec International 4-23 Report HI-971760
0 Table 4.4.1 Reactivity Allowance for Uncertainty in Burnup Calculations and the Effect of Axial Burnup Distributions for PWR Fuel hk Calculated Burnup Applicable Burnup Uncertainty in Burnup Effect of Axial (MWD/MTU) Range (MWD/MTU) Burnup Distribution 45,000 40,000-45,000 0.0160 0.0157 40,000 30,000-40,000 0.0143 0.0090 30,000 < 30,000 0.0110 Negative Notes:
- 1. The uncertainty in burnup was calculated by taking 5% of the reactivity decrement from zero burnup to the calculated burnup using CASMO-3.
- 2. The effect of the axial burnup distribution was calculated using MCNP-4A by comparing results from two cases: the first had a uniform axial burnup and the second had a distributed axial burnup distribution represented by 10 axial zones.
- 3. The effect of the axial burnup distribution is negative at and below 30,000 MWD/MTU, therefore, conservatively no reactivity adjustment was made.
Holtee International 4-24 Report HI-971760
Table 4.5.1 Comparison of MCNP-4A and CASMO-3 Calculations PWR Rack BWR Rack Fuel Assembly W 15x15 GE8 Enrichment 5.0 4.6 Temperature 40C 4'C MCNP-4A k,tr 1.2004 0.9993 Uncertainties Calculational Statistics 0.0020 Bias Uncertainty 0.0011 Total Uncertainty at 95%/95% 0.0023 0.0023 Temperature correction 0.0020 0.0020 from 20'C to 4'C, Bias 0.0009 0.0009 MCNP-4A Maximum k, 1.2056 1.0045 CASMO-3 k,r 1.2076 1.0126 Notes:
- 1. The MCNP-4A calculation correctly includes the effect of axial neutron leakage.
Holtec International 4-25 Report HI-971760
Table 4.7.1 Reactivity Effects of Temperature and Void Incremental Reactivity Effect - dk (relative to reference)
Temperature PWR Rack BWR Rack 4oC (39oF) reference reference 20'C (68'F) -0.0020 -0.0020 60'C (140'F) -0.0093 -0.0093 120'C (248'F) -0.0247 -0.0240 120'C with 10% void -0.0492 -0.0446 Holtec International 4-26 Report HI-971760
0 45000 40000 Acceptable Burnup Dom ain 35000 Burnup=12114*Enrichment-19123 30000 F
25000 20000 Unacceptab le Burnup Do main 15000 10000 5000 2.5 3 3.5 4 4.5 Enrichment (avt% U235)
Figure 4.2.1: Burnup Versus Enrichment for PWR Fuel Holtec International 4-27 Report HI-971760
0 0
Boral~ Box Wall wQ QwQ QQ w Q W GGG w QQQ wQQ QQQ w w Q W Q GGG Q w W W
w QQ QQQ w w QQQ wQ QwQ W G W = guide tubes Figure 4.3.1: This is a two dimensional representation of the calculational model used for the PWR storage rack analysis showing a Westinghouse 15x15 fuel design. This figure was drawn with the two dimensional plotter in MCNP-4A.
Report HI-971760
Boral Box Wall Channel OO OO OO OO OO O Qw Qw O Qw Qw OO OO OO W = water rod Figure 4.3.2: This is a two dimensi'onal representation of the calculational model used for the BWR storage rack analysis showing a GE 8 fuel design. This figure was drawn with the two dimensional plotter in MCNP-4A.
Holtec International 4-29 Report HI-971760
APPENDIX 4A'ENCHMARKCAlCULATIONS 4A.1 Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A.1] is a continuous energy Monte Carlo code and KEN05a [4A.2]
uses group-dependent cross sections. For the KENO5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II[4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errorst (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.
In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the "B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.
Table 4A.1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed 'in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.
One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KENO5a computes and prints the "energy of the average lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KENOSa, the number of fissions in each group may be collected and the EALF determined (post-processing).
Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.
Report HI-971760 Appendix 4A, Page 1
Figures 4A.1 and 4A.2 show the calculated k,<< for the benchmark critical experiments as a
~ ~
function of the EALF for MCNP4a and KENOSa, respectively (UO, fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental error in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B%W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the BED criticals encompass a greater range of critical parameters than the PNL criticals.
Linear regression analysis of the data in Figures 4A.1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KENOSa). The total bias (systematic error, or mean of the deviation from a g<<of exactly 1 .000) for the two methods of analysis are shown in the table below.
Calculational Bias of MCNP4a and KENOSa MCNP4a 0.0009 %0.0011 KENOSa 0.0030 %0.0012 0 The bias and standard error of the bias were derived directly from the calculated k,<<values in Table 4A.1 using the following equationstt, with the standard error multiplied by the one-sided K-factor for 95% probability at the 95% confidence level from NBS Handbook 91 [4A.18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2).
k= Pk, (4A.1)
A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.
These equations may be found in any standard text on statistics, for example, reference
[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENOSa.
Holtec International Report HI-971760 Appendix 4A, Page 2
(4A.2)
(4A.3) where k; are the calculated reactivities of n critical experiments; o< is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95% probability at the 95% confidence level (NBS Handbook 91 [4A.18]).
Formula 4.A.3 is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( 1- k ), is the actual bias which is added to the MCNP4a and KENO5a results.
The second term, Ko-, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95% probability at the 95% confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively.
The bias values are used to evaluate the maximum k,<<values for the rack designs.
KENOSa has a slightly larger systematic error than MCNP4a, but both result in greater precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section sets in KENO5a (SCALE) calculations.
4A.2 The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and 4A.4 show the calculated k,<<values (Table 4A.1) as a function of the fuel enrichment reported for the critical experiments. Lin'ear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENOSa). Thus, there are no corrections to the bias for the various enrichments.
Holtec International Report HI-971760 Appendix 4A, Page 3
As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENOSa for various enrichments.
The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.S, confirm no significant difference in the calculated values of k,~ for the two independent codes as evidenced by the 45'lope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.
4A.3 Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment),
the reactivity worth of the absorbers in the PNL tests is very low and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.
Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A.1) and shows the reactivity worth (b,k) of the absorber.t No trends with reactivity worth of the absorber are evident, although based on the
,calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.
To further confirm the absence of a significant trend with ' concentration in the absorber, a cross-comparison was made with MCNP4a and KENOSa (as suggested in Reg.
Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.
These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45'ine, within an expected 95% probability limit).
The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (hk) change in reactivity due to the absorber.
Holtec International Report HI-971760 Appendix 4A, Page 4
4A.4 4A.4.1 PNL has performed a number of critical experiments with thick steel and lead reflectors.
Analysis of these critical experiments are listed in Table 4A.S (subset of data in Table 4A.1). There appears to be a small tendency toward overprediction of k,~ at the lower spacing, although'there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.
4A.4.2 The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table 4A.1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.
4A.4.3 Various soluble boron concentrations were used in the BOW series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENOSa) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments () 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.
Parallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design.
Holtec International Report HI-971760 Appendix 4A, Page 5
4A5 hlQKEhel The number of critical experiments with PuO, bearing fuel (MOX) is more limited than for UO, fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a k,~ of 1.00, indicating that when Pu is present, both MCNP4a and KENOSa overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KENOSa calculated reactivities are below 1.00, suggesting that a small trend may exist with KENOSa. It is also possible that the overprediction in k,<< for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated k,~ over a wide range of the spectral index (energy of the average lethargy causing fission).
Holtec International Report HI-971760 Appendix 4A, Page 6
4A.6
[4A.1] J.F. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).
[4A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing'Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.
[4A.3] M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460)
Oak Ridge National Laboratory, September 1994.
[4A.4] W.C. Jordan et al., "Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Systems, Inc., Oak Ridge National Laboratory, December 1986.
[4A.5] O.W. Hermann et al., "Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.
[4A.6] R.J. Larsen and M.L. Marx,
, Prentice-Hall, 1986.
[4A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.
[4A.8] G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW-1645-4, Babcock R Wilcox Company, November 1991.
[4A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984.
Holtec International Report HI-971760 Appendix 4A, Page 7
[4A.10] J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods," Trans.
Am. Nucl. Soc. 33: 362-364 (1979).
[4A.11] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 'U Enriched UO, Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.
[4A.12] S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o "'U Enriched UO, Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.
[4A.13] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o "'U Enriched UO~ Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.
[4A.14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.
[4A.15] B.M. Durst et al., Critical Experiments with 4.31 wt % "'U Enriched UO~ Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.
[4A.16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.
[4A.17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.
fAA.IflM.GIB<<II,~,N Standards, Handbook 91, August 1963 ~
I IB I Holtec International Report HI-971760 Appendix 4A, Page 8
Table 4A.1 Summary of Criticality Benchmark Calculations Reference Identification Enrich. MCNP4a KENO5a MCNP4a KENO5a B &W-1484 (4A.7) Core I. 2 46 0.9964 J 0.0010 0.989SJ 0.0006 0.1759 0.1753 B &W-1484 (4A.7) Core H 2.46 1.0008 J 0.0011 1.0015 J 0.0005 0.2553 0.2446 B &W-1484 (4A.7) Core HI 2.46 1.0010 2 0.0012 1.0005 2 0.0005 0.1999 0.1939 B &W-1484 (4A.7) Core IX 2.46 0.9956 2 0.0012 0.9901 2 0.0006 0.1422 0.1426 B &W-1484 (4A.7) Core X 2.46 0.9980 2 0.0014 0.9922 2 0.0006 0.1513 0.1499 B &W-1484 (4A.7) Core XI 246 0.9978 2 0.0012 1.0005 2 0.0005 0.2031 0.1947 B &W-1484 (4A.7) Core XH 2.46 0.9988 J 0.0011 0.9978 J 0.0006 0.1718 0.1662 B &W-1484 (4A.7) Core XIH 2.46 '.0020 2 0.0010 0.9952 2 0.0006 0.1988 0.1965 B &W-1484 (4A.7) Core XIV 2.46 0.9953 J 0.0011 0.9928 J 0.0006 0.2022 0.1986 10 B&W-1484 (4A.7) Core XV" 2A6 0.9910 2 0.0011 0.9909 2 0.0006 0.2092 0.2014 B&W-1484 (4A.7) Core XVI" 2.46 0.9935 2 0.0010 0.9889 2 0.0006 0.1757 0.1713 12 B &W-1484 (4A.7) Core XVH 2;46 0.9962 2 0.0012 0.9942 2 0.0005 0.2083 0.2021 13 B &W-1484 (4A.7) Core XVIH 2.46 1.0036 2 0.0012 0.9931 2 0.0006 0.1705 0.1708 Holtec International Report HI-971760 Appendix 4A, Page 9
Table 4A.1 Summary of Criticality Benchmark Calculations Reference IdentiTication Enrich. MCNP4a KENO5a MCNP4a KENO5a 14 B &W-1484 (4A.7) Core XIX 2.46 0.9961 J 0.0012 0.9971 + 0.0005 0.2103 0.2011 15 B &W-1484 (4A.7) Core XX 2.46 1.0008 J 0.0011 0.9932 J 0.0006 0.1724 0.1701 16 B &W-1484 (4A.7) Core XXI 2.46 0.9994 J 0.0010 0.9918 g 0.0006 0.1544 0.1536 17 B &W-1645 (4A.8) S-type Fuel, w/886 ppm B 2.46 0.9970 + 0.0010 0.9924 + 0.0006 1.4475 1.4680 18 B &W-1645 (4A.8) S-type Fuel, w/746 ppm B 2.46 0.9990 i 0.0010 0.9913 2 0.0006 1.5463 1.5660 19 B &W-1645 (4A.8) SO-type Fuel, w/1156 ppm B 2.46 0.9972 + 0.0009 0.9949 J 0.0005 0.4241 0.4331 20 B &W-1810 (4A.9) Case 1 1337 ppm B 2.46 1.0023 J 0.0010 NC 0.1531 NC 21 B &W-1810 (4A.9) Case 12 1899 ppm B 2.46/4.02 1.0060 J 0.0009 NC 0.4493 NC 22 French (4A.IO) Water Moderator 0 gap 4.75 0.9966 R 0.0013 NC 0.2172 NC 23 French (4A.10) Water Moderator 2.5 cm gap .4.75 0.9952 2 0.0012 NC 0.1778 NC French (4A.10) Water Moderator 5 cm gap 4.75 0.9943 J 0.0010 NC 0.1677 NC 25 French (4A.10) Water Moderator 10 cm gap 4.75 0.9979 J 0.0010 NC 0.1736 NC 26 PNI 3602 (4A.11) Steel Reflector, 0 separation 2.35 NC 1.0004 J 0.0006 NC 0.1018 Holtec International Report HI-971760 Appendix 4A, Page 10
Table 4A.1 Summary of Criticality Benchmark Calculations Reference Identification Enrich. MCNP4a KENO5a MCNP4a KENO5a 27 PNL-3602 (4A.11) Steel Reflector, 1.321 cm sepn. 2.35 0.9980 2 0.0009 0.9992 2 0.0006 0.1000 0.0909-28 PNL-3602 (4A.11) Steel Reflector, 2.616 cm sepn 2.35 0.9968 2 0.0009 0.9964 2 0.0006 0.0981 0.0975 29 PNL-3602 (4A.11) Steel Reflector, 3.912 cm sepn. 2.35 0.9974 J 0.0010 0.9980 2 0.0006 0.0976 0.0970 30 PNL-3602 (4A.11) Steel Reflector, infinite sepn. 2.35 0.9962 J 0.0008 0.9939 2 0.0006 0.0973 0.0968 31 PNL-3602 (4A.11) Steel Reflector, 0 cm sepn. 4.306 NC 1.0003 J 0.0007 NC 0.3282 32 PNL-3602 (4A.11) Steel Reflector, 1.321 cm sepn. 4.306 0.9997 2 0.0010 1.0012 + 0.0007 0.3016 0.3039 33 PNL-3602 (4A.11) Steel Reflector, 2.616 cm sepn. 4.306 0.9994 2 0.0012 0.9974 + 0.0007 0.2911 0.2927 34 PNL-3602 (4A.11) Steel. Reflector, 5 405 cm sepn. 4.306 0.9969 2 0.0011 0.9951 R 0.0007 0,2828 0.2860 35 PNL-3602 (4A.11) Steel Reflector, Infinite sepn. " 4.306 0.9910 2 0.0020 0.9947 g 0.0007 0.2851 0.2864 36 PNL-3602 (4A.11) Steel Reflector, with Boral Sheets 4.306 0.9941 g 0.0011 0.9970 J 0.0007 0.3135 0.3150 37 PNL-3926 (4A.12) Lead Reflector, 0 cm sepn. 4.306 NC 1.0003 2 0.0007 NC 0,3159 38 PNL-3926 (4A.12) Lead Reflector, 0.55 cm sepn. 4.306 1.0025 J 0.0011 0.9997 + 0.0007 0.3030 0.3044 39 PNI 3926 (4A.12) Lead Reflector, 1.956 cm sepn. 4.306 1.0000 + 0.0012 0.9985 2 0.0007 0.2883 0.2930 Holtec International Report HI-971760 Appendix 4A, Page 11
Table 4A.1 Summary of Criticality Benchmark Calculations Reference IdentiTication Enrich. MCNP4a KENO5a MCNP4a KENO5a 40 PNL-3926 (4A.12) Lead Reflector, 5.405 cm sepn. 4.306 0.9971 2 0,0012 0.9946 2 0.0007 0.2831 0.2854 41 PNL-2615 (4A.13) Experiment 004/032 - no absorber 4.306 0.9925 2 0.0012 0.9950 2 0.0007 0;1155 0.1159 42 PNL-2615 (4A.I3) Experiment 030 - Zr plates 4.306 0.9971 ~ 0.0007 NC 0.1154 43 PNL-2615 (4A.13) Experiment 013 - Steel plates 4.306 0.9965 2 0.0007 NC 0.1164 PNL-2615 (4A.13) Experiment 014 - Steel plates 4.306 NC 0.9972 2 0.0007 NC 0.1164 45 PNL-2615 (4A.13) Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 J 0.0010 0.9981 2 0.0007 0.1172 0.1162 46 PNL-2615 (4A.13) Exp. 012 1.62% Boron-Steel plates 4.306 0.9996 2 0.0012 0.9982 g 0.0007 0.1161 0.1173 47 PNL-2615 (4A.13) Exp. 031 - Boral plates 4.306 0.9994 2 0.0012 0.9969 2 0.0007 0.1165 0.1171 48 PNL-7167 (4A.14) Experiment 214R - with flux trap 4.306 0.9991 J 0.0011 0.9956 2 0.0007 0.3722 0.3812 49 PNL-7167 (4A.14) Experiment 214V3 - with flux trap 4.306 0.9969 2 0.0011 0.9963 2 0.0007 0.3742 0.3826 50 PNIA267 (4A. 15) Case173 - OppmB 4.306 0.9974 2 0.0012 NC 0.2893 51 PNL-4267 (4A.15) Case 177 - 2550 ppm B 4.306 1.0057 2 0.0010 NC 0.5509 NC 52 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 J 0.0011 1.0046 2 0.0006 0.9171 0.8868 Holtec International Report HI-971760 Appendix 4A, Page 12
Table 4A.1 Summary of Criticality Benchmark Calculations Reference Identification Enrich. MCNP4a KENO5a MCNP4a KENO5a 53 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0058 + 0.0012 1.0036 + 0.0006 0.2968 0.2944 54 PNL-5803 (4A.16) MOX Fuel- Type 3.2 Exp. 13 20% Pu 1.0083 + 0.0011 0.9989 + 0.0006 0.1665 0.1706 55 PNL-5803 (4A.16) MOX Fuel - Type 3.2 Exp. 32 20% Pu 1.0079 + 0.0011 0.9966 g 0.0006 0.1139 0.1165 56 WCAP-3385 (4A.17) Saxton Case 52 Pu02 0.52" pitch 6.6% Pu 0.9996 + 0.0011 1.0005 f 0.0006 0.8665 0.8417 57 WCAP-3385 (4A.17) Saxton Case 52 U 0.52" pitch 5.74 1.0000 J 0.0010 0.9956 J 0.0007 0.4476 0.4580 58 WCAP-3385 (4A.17) Saxton Case 56 Pu02 0.56" pitch 6.6% Pu 1.0036 + 0.0011 1.0047 + 0.0006 0.5289 0.5197 59 WCAP-3385 (4A.17) Saxton Case 56 borated Pu02 6.6% Pu 1.0008 + 0.0010 NC 0.6389 NC 60 WCAP-3385 (4A.17) Saxton Case 56 U 0.56" pitch 5.74 0.9994 J 0.0011 0.9967 J 0.0007 0.2923 0.2954 61 WCAP-3385 (4A.17) Saxton Case 79 Pu02 0.79" pitch 6.6% Pu 1.0063 + 0.0011 1.0133 + 0.0006 0.1520 0.1555 62 WCAP-3385 (4A.17) Saxton Case 79 U 0.79" pitch 5.74 1.0039 + 0.0011 1.0008 + 0.0006 0.1036 0.1047 Notes: NC stands for not calculated.
EALF is the energy of the average lethargy causing fission.
These experimental results appear to be statistical outliers () 30) suggesting the possibility of unusually large experimental error. Although they could justifiably be excluded, for conservatism, they were retained in determining the calculational basis.
Holtec International Report HI-971760 Appendix 4A, Page 13
Table 4A.2 COMPARISON OF MCNP4a AND KENO5a CALCULATEDREACTIVITIESt FOR VARIOUS ENRICHMENTS Calculated k,ff J 1o Enrichment MCNP4a KENO5a 3.0 0.8465 + 0.0011 0.8478 2 0.0004 3.5 0.8820 + 0.0011 0.8841 + 0.0004 3.75 0.9019 2 0.0011 0.8987 2 0.0004 4.0 0.9132 + 0.0010 0.9140 2 0.0004 4.2 0.9276 J 0.0011 0.9237 2 0.0004 4.5 0.9400 + 0.0011 0.9388 + 0.0004 Based on the GE 8x8R fuel assembly.
Holtec International Report HI-971760 Appendix 4A, Page 14
Table 4A.3 MCNP4a CALCULATEDREACTIVITIES FOR CRITICALEXPERIMENTS WITH NEUTRON ABSORBERS hk MCNP4a Worth of Calculated EALF Ref. Experiment Absorber 'eV) 4A.13 PNL-2615 Boral Sheet 0.0139 0.9994%0.0012 0.1165 4A.7 B8cW-1484 Core XX 0.0165 1.0008%0.0011 0.1724 4A.13 PNL-2615 1.62% Boron-steel 0.'0165 0.999620.0012 0.1161 4A.7 BAW-1484 Core XIX 0.0202 0.9961 %0.0012 0.2103 4A.7 BEcW-1484 Core XXI 0.0243 0.999420.0010 0.1544 4A.7 BOW-1484 Core XVII 0.0519 0.996220.0012 0.2083 4A.11 PNL-3602 Boral Sheet 0.0708 0.9941 20.0011 0.3135 4A.7 BOW-1484 Core XV 0.0786 0.9910%0.0011 0.2092 4A.7 BOW-1484 Core XVI 0.0845 0.993520.0010 0.1757 4A.7 BOW-1484 Core XIV 0.1575 0.9953 20.0011 0.2022 4A.7 BOW-1484 Core XIII 0.1738 1.002020.0011 0.1988 4A.14 PNL-7167 Expt 214R flux trap 0.1931 0.9991 %0.0011 0.3722 BALF is the energy of the average lethargy causing fission.
Holtec International Report HI-971760 Appendix 4A, Page 15
Table 4A.4 COMPARISON OF MCNP4a AND KENO5a CALCULATEDREACTIVITIESt FOR VARIOUS ' LOADINGS Calculated k,~ 1o
', g/cm'.005 MCNP4a KENO5a 1.0381 + 0.0012 1.0340 + 0.0004 0.010 0.9960 + 0.0010 0.9941 + 0.0004 0.015 0.9727 2 0.0009 0.9713 + 0.0004 0.020 0.9541 J 0.0012 0.9560 2 0.0004 0.025 0.9433 J 0.0011 0.9428 2 0.0004 0.03 0.9325 + 0.0011 0.9338 2 0.0004 0.035 0.9234 + 0.0011 0.9251 J 0.0004 0.04 0.9173 J 0.0011 0.9179 2 0.0004 Based on a 4.5% enriched GE 8x8R fuel assembly.
Holtec International Report HI-971760 Appendix 4A, Page 16
Table 4A.5 CALCULATIONSFOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORS Separation, Ref. Case E, wt% cm MCNP4a k,~ KENO5a k, 4A.11 Steel 2.35 1.321 0.998020.0009 0.999220.0006 Reflector 2.35 2.616 0.9968 J0.0009 0.9964 %0.0006 2.35 3.912 0.9974 J0.0010 0.9980 %0.0006 2.35 0.9962 J0.0008 0.9939 %0.0006 4A.11 Steel 4.306 1.321 0.9997 J0.0010 1.0012 J0.0007 Reflector 4.306 2.616 0.9994 %0.0012 0.9974 +0.0007 4.306 3.405 0.9969 +0.0011 0.9951 J0.0007 4.306 0.9910 J0.0020 0.9947 +0.0007 4A.12 Lead 4.306 0.55 1.0025 J0.0011 0.9997 J0.0007 Reflector 4.306 1.956 1.0000 J0.0012 0.9985 +0.0007 4.306 5.405 0.9971 J0.0012 0.9946 J0.0007 Arranged in order of increasing reflector-fuel spacing.
Holtec International Report HI-971760 Appendix 4A, Page 17
Table 4A.6 CALCULATIONSFOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated k,~
Boron Concentration, Reference Experiment ppm MCNP4a KENO5a 4A.15 PNL-4267 0.9974 J 0.0012 4A.8 B&W-1645 886 0.9970 + 0.0010 0.9924 + 0.0006 4A.9 B&W-1810 1337 1.0023 J 0.0010 4A.9 B&W-1810 1899 1.0060 J 0.0009 4A.15 PNL-4267 2550 1.0057 J 0.0010 Holtec International Report HI-971760 Appendix 4A, Page 18
Table 4A.7 CALCULATIONSFOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENO5a Reference Caset EALFtt EALF" PNI 5803 MOX Fuel - Exp. No. 21 1.0041 +0.0011 0.9171 1.0046 J0.0006 0.8868
[4A.16]
MOX Fuel - Exp. No. 43 1.005820.0012 0.2968 1.003620.0006 0.2944 MOX Fuel - Exp. No. 13 1.0083 20.0011 0.1665 0.9989 20.0006 0.1706 MOX Fuel - Exp. No. 32 1.0079 J0.0011 0.1139 0.9966~0.0006 0.1165 WCAP- Saxton 0.52" pitch 0.9996 J0.0011 0.8665 1.0005 J0.0006 0.8417 3385-54
[4A.17J Saxton 0.56" pitch 1.0036 J0.0011 0.5289 1.0047 J0.0006 0.5197 Saxton 0.56" pitch borated 1.0008J0.0010 0.6389 NC NC Saxton 0.79" pitch 1.0063 J0.0011 0.1520 1.0133 J0.0006 0.1555 Note: NC stands for not calculated Arranged in order of increasing lattice spacing.
EALF is the energy of the average lethargy causing fission.
Holtec International Report HI-971760 Appendix 4A, Page 19
s.o 5.1 This section provides a summary of the methods, models, analyses and numerical results to demonstrate the compliance of Harris Spent Fuel Pools C and D with the provisions of Section IIIof the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications", (April 14, 1978) for a bounding configuration. Similar methods of thermal-hydraulic analysis have been used in other rerack licensing projects (see Table 5.1.1).
The thermal-hydraulic qualification analyses for the rack array may be broken down into the following categories:
(i) Evaluation of the long-term decay heat load, which is the cumulative spent fuel decay heat generation from all fuel assemblies stored in the C and D pools.
(ii) Evaluation of the steady-state bulk pool temperatures when forced cooling is available. The bulk pool temperatures are required to be maintained s 137'F t
under normal conditions with fuel pool cooling in operation.
(iii) Determination of the maximum pool local temperature at steady bulk pool temperatures.
(iv) Evaluation of the potential for flow bypass from pool inlet to outlet in the absence of a sparger line to the spent fuel pools racks.
(v) Evaluation of the "time-to-boil" ifall forced heat rejection paths from the pool are lost.
The 137 F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of re-evaluating systems and components to allow for an increase in the allowable bulk pool temperature.
Holtec International 5-1 Holtec Report HI-971760
This section presents a synopsis of the analysis methods employed, and final results. The decay heat load calculation is conservatively 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.
The Pool C and D fuel rack configurations for proposed expansion are depicted in Figures 1.2 and 1.3. A total of 1,952 PWR cells and 2,763 BWR cells will be available in a bounding configuration to maximize fuel storage capacity.
To determine the limiting decay heat in the Harris spent fuel pools, a projected bounding decay period for fuel scenario is considered as shown in Table 5.2.1. The in-core irradiation time and limiting assembly specific power inputs are provided in Table 5.2.2. The C and D spent fuel pools (SFPs) are designated to store old fuel which has been cooled for at least 5 years. The fuel is envisaged to be transhipped from Brunswick and Robinson plants or shuffled &om Harris'ools A and/or B.
Since the decay heat load from the old assemblies varies very slowly as a function of time, the long-term decay heat in the bounding configuration is assumed to be constant. Based on the discharge scenario and fuel assemblies characteristics listed in Tables 5.2.1 and 5.2.2, the combined Pools C and D decay heat rates are determined and summarized in Table 5.2.3.
Holtec International 5-2 Holtec Report Hl-971760
5~
The decay heat load to the two pools (C and D) willbe removed by several passive and active heat rejection mechanisms, as listed below:
(a) Heat loss by pool surface evaporation (b) Radiation heat loss Rom pool surface to fuel handling building extremities (c) Natural convection heat transfer to fuel building air (d) Heat loss through pool concrete walls (e) Forced convection cooling of SFP surface by HVAC system forced air ventilation (0 Forced cooling by SFP water circulation through a heat exchanger In the interest of conservatism, no credit is applied to removing heat by any of the mechanisms listed above from (a) to (e). Consequently, all of the decay heat generated in the C and D pools is considered to be removed by the forced fiow of SFP cooling water circulating through a heat exchanger, which transfers heat to the CCW system. In a forced SFP cooling scenario, hot water from the pool is circulated by a pump through an exchanger cooled by the CCW system. The cooled SFP water is then directed back to the C and D pools. The decay heat load in the C and D pools is from old fuel discharges, which is relatively constant (i.e., steady heat load). Therefore, at equilibrium conditions, the total decay heat load to the pool is equal to the heat removed by the cooling system and a constant bulk temperature is maintained in the C and D pools.
The heat removal capacity of the SFP cooling system is principally characterized by two parameters, namely the water circulation flow rate and the fuel pool inlet water temperature. The bulk pool temperature of pools C and D is required to be maintained at or below 137'F t. The minimum SFP water flow rate required to comply with this bulk pool temperature criterion is thus a function of the fuel pool inlet water temperature. This requirement is graphically illustrated in Figure 5.3.1. A SFP cooling system design point, which is on the curve, satisfies the minimum cooling requirements. A design point above this curve exceeds the SFP cooling The 137'F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of re-evaluating systems and components to allow for an increase in the allowable bulk pool temperature.
Holtec International 5-3 Holtec Report HI-971760
0 requirements. Therefore, Figure 5.3.1 establishes the thermal-hydraulic design basis for SFP cooling system capacity and the final cooling system design shall comply with these flow vs.
inlet temperature parameters.
5.4 In this section, we present the methodology for calculating the local temperatures when forced cooling is available to the Pool C only. The results from evaluations performed with forced cooling in pool C only are conservative, since the pool cooling system willbe connected to both pools and cooling water will be discharged to both pools. Therefore, these evaluations predict conservative local temperatures, especially in pool D.
Truncation of sparger lines has become a standard pool modification procedure in rerack campaigns in recent years. Over a dozen SFPs reracked in the past several years have removed sparger lines to enable a high density storage layout and thus maximize pool capacity. Absence of a sparger in the Harris C and D pools removes the mechanistic feed of cold water into the bottom plenum of the fuel racks. It is not apparent &om heuristic reasoning alone that the cooled water delivered to the pool would not bypass the hot fuel racks and the stored spent fuel in the two pools and exit through the outlet piping. To demonstrate adequate cooling of fuel in the two areas, it is therefore necessary to rigorously quantify the velocity field in the pool created by the interaction of buoyancy driven flows and water ingress/egress. A CFD analysis for this demonstration is required. The objective of this study is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled conditions in the pool is met for the bounding fuel storage configuration. An outline of the CFD approach is described in the following.
Figure 5.4.1 depicts the fuel Pools C and D physical configuration in plan view. The two pools are connected by a transfer canal. Pool piping connections for introducing cooling water and discharge of heated water are shown for both pools. Currently, SFP cooling system design work Holtec International Holtec Report Hl-971760
is in progress to provide a forced cooling system which willprovide suction and discharge to both pools. Thermal-hydraulic adequacy of the two pools shall be conservatively demonstrated by assuming that forced cooling is available to only Pool C. Adequate c'ooling of Pool D is enabled by a buoyancy-driven flow of relatively cooler bulk Pool C water to Pool D through the interconnecting transfer canal. Decay heat inputs to both pools are based on a bounding fuel storage configuration and spent fuel cooling times. The buoyancy-induced cooling of Pool D is demonstrated by performing a rigorous Computational Fluid Dynamics (CFD) analysis of the temperature and flow fields in the two 'pools. The CFD methodology is discussed in the next subsection. An additional assumption about the location of cooling inlet and outlet piping is included in the analysis to result in an extremely conservative thermal-hydraulic portrayal of the two interconnected pools. The pool cooling inlet and outlet piping connections are assumed to be located on the southeast end of the pool. Thus, forced cooling of the pool is in a diagonally opposite (i.e., farthest) corner &om the northwest location of the connection from Pool C to the transfer canal. The forced cooling ingress and egress locations are in close proximity to each other and at the same elevation. The potential for flow bypass from inlet to outlet is conservative, since the modeled locations are closer than the actual relative positions.
There are several significant geometric and thermal-hydraulic features of the Harris SFPs which need to be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there are two regions to be considered. One region is the bulk pool region where the classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat generating fuel assemblies located in the spent fuel racks located near the bottom of the SFP. In this region, water flow is directed vertically upwards due to buoyancy forces through relatively small flow channels formed by stored fuel assembly rod arrays in each rack cell. This situation shall be modeled as a porous solid region in which fluid flow is governed by the classical Darcy's Law:
BP p v,
cpiv(v, /2 8 x, -
K(i)
Holtec International 5-5 Holtec Report Hl-971760
where Gp/5Xi is the pressure gradient, K(i), Vi and C are the corresponding permeability, velocity and inertial resistance parameters and N is the fluid viscosity. Bounding permeability and inertial resistance parameters for the rack cells loaded with PWR or BWR fuel is determined based on &iction factor correlations for laminar flow conditions typically encountered due to low buoyancy induced velocities and small size of the flow channels. A large number of fuel assembly types have been analyzed for hydraulic flow resistance [5.4.1] determination. Table 5.4.1 provides flow resistance parameters which bound all PWR and BWR fuel assembly types which were analyzed in this study.
The pool geometry requires an adequate portrayal of large scale and small scale features, spatially distributed heat sources in the spent fuel racks and water inlet/outlet configuration.
Relatively cooler bulk pool water normally flows down through the narrow fuel rack outline to pool wall liner clearance known as the downcomer. Near the bottom of the racks, the flow turns from a vertical to horizontal direction into the bottom plenum supplying cooling water to the rack cells. Heated water issuing out of the top of the racks mixes with the bulk pool water. An adequate modeling of these features on the CFD program involves meshing the large scale bulk pool region and small scale downcomer and bottom plenum regions with sufficient number of computational cells to capture the bulk and local features of the flow field.
The CFD analysis is performed on the industry standard FLUENT [5.4.2] fluid flow and heat transfer modeling program. The FLUENT code enables buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying "Reynolds's Stresses" to the mean bulk flow quantities with the following turbulence modeling options:
(i) K-e Model (ti) RNG x-e Model (iii) Reynolds Stress Model Holtec International 5-6 Holtec Report HI-971760
The K-e model is considered appropriate for the fuel pool CFD analysis. Rigorous modeling of fluid flow problems requires a numerical solution to the classical Navier-Stokes equations of fluid motion [5.4.3].
The governing equations (in modified form for turbulent flows with buoyancy effects included) can be written as:
Bp,u, Bp,u,u, BQ BQ Bt Bx Bx Bx Bx,
-p P(T-T)g Bp Bx I o l Bp,(tt'p',
J
)
/
where ut are the three time-averaged velocity components. p (up uJ ) are time-averaged Reynolds
/
stresses derived from the turbulence induced fluctuating velocity components u,, p, is the fluid density at temperature TP is the coefficient of thermal expansion, p is the fluid viscosity, g; are the components of gravitational acceleration and x; are the Cartesian coordinate directions. The Reynolds stress tensor is expressed in terms of the mean flow quantities by defining a turbulent viscosity p, and a turbulent velocity scale K/ as shown below [5.4.4]:
P(utul ) =2/3Px5< -P, Bx> Bx, The procedure to obtain the turbulent viscosity and velocity length scales involves a solution of two additional transport equations for kinetic energy (K) and rate of energy dissipation (e). This methodology is known as the K-e model for turbulent flows as described by Launder and Spalding [5.4.5].
Holtec International 5-7 Holtec Report HI-971760
Ifall heat exchanger assisted forced pool'cooling becomes unavailable, then the pool water will begin to rise in temperature and eventually willreach the normal bulk boiling temperature at 212'F. The time to reach the boiling point willbe the shortest when the loss of forced cooling occurs at the point in time when the bulk pool temperature is at its maximum calculated value for a bounding fuel storage configuration. The calculation is conservatively performed for a bounding decay heat load to the pool, no credit for evaporation cooling and no credit for thermal inertia of racks. The amount of water holdup above the racks in the two pools is in excess of 48,000 ft'2.9 x 10'bs) of water. The maximum rate of temperature rise of bulk pools water at a bounding 15.63 million Btu/hr decay heat input (Table 5.2.3) is therefore less than 5.4'F/hr with no water makeup. Ifthe initial temperature is conservatively assumed to be at a uniform maximum bulk average limit of 140'F ', then the time to reach normal boiling point of the bulk pool is in excess of 13 hours. This is a relatively long time period for operator action to start makeup water and re-initiate forced cooling to the pool.
0 A summary of pools dimensional data used to generate a Computational Fluid Dynamics (CFD) model of the two interconnected C and D pools is provided in Table 5.5.1. The CFD model provides a determination of the difference between the peak local and bulk pool temperatures.
The local temperature corresponding to the maximum bulk pool temperature can then be determined by adding this local temperature rise to the bulk temperature limit. In the CFD model, a minimum bounding downcomer gap between racks outline to pool liner is applied as 1
noted in Table 5.5.1. In this manner, the downcomer water flow path hydraulic resistance is maximized. Consequently, the local rack cell temperature predictions shall be conservatively maximized. The background constant decay heat input to the pool is modeled as a uniform volumetric heat source term in the active fuel region of the Pools C and D racks. The total heat The assumption of an initial temperature of 140'F is conservative, since the bulk pool temperature is currently limited to 137'F.
Holtec International
/ 5-8 Holtec Report HI-971760
generating volume is calculated to be 657 m'. Thus, &om the total decay heat input (Table 5.2.3),
the volumetric heat source term is determined to be 6,956 W/m'.
A plan view of the three-dimensional CFD model is presented in Figure 5.5.1. In this view, the two pools with an interconnected transfer canal is depicted. The water inlet/outlet connections are shown modeled in the top left end corner of the Pool C. The racks outline, modeled as a porous media, is depicted in blue color. A perspective view of the CFD model is presented in Figure 5.5.2. The bottom of the transfer canal, as shown in this figure, is at the same elevation as the top of the racks. The average background decay heat is applied to the model as a volumetric heat source term in the active fuel region of the fuel racks. The CFD model of the C and D pools is solved to obtain converged temperature and velocity profiles. The results obtained from the analysis are discussed next.
Peak local water temperature in the rack cells is shown as a contour plot in cross sectional plan view as shown in Figure 5.5.3. The plan view elevation is within the region of the racks above the active fuel region, but below the top of the racks.
An exchange of cold and hot water streams from the Pool D to Pool C is determined by the CFD solution with only pool C cooled by a forced cooling system. This exchange of cold and hot water between the two bulk pools is illustrated as a flow velocity vectors plot (Figure 5.5.4) in the pools'nterconnecting channel. The peak local temperature is 6.8'F above the water temperature at the cooling system discharge from pool. Consequently, the peak local temperature corresponding to the maximum bulk pool temperature limit is obtained by adding this local temperature rise. Table 5.5.2 provides the bulk and local temperature summaries. The peak 143.8'F local temperature is below the local water boiling temperature by a large margin.
Figure 5.5.5 provides a flow velocity vectors plot in the pool cooling inlet/outlet piping region.
The pool inlet piping is modeled to be 12 inches below the pool water level and the pool outlet piping suction is adjacent to the inlet piping discharge at the same elevation. From the velocity vectors plot, it is apparent that no bypass of incoming water to outlet is indicated for an Holtec International 5-9 Holtec Report HI-971760
extremely conservative configuration. In the actual pool piping arrangement for Pool C, the water inlet and outlet connections are widely separated. Consequently, it is concluded that any water bypass &om inlet to outlet is not possible.
Holtec International 5-10 Holtec Report HI-971760
5.~
[5.4.1] Holtec Report HI-951325, "HI-STAR 100 System Thermal Design Package".
[5.4.2] "QA Documentation and Validation of the FLUENT Version 4.3 CFD Analysis Program", Holtec Report HI-961444.
[5.4.3] Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press, 1967.
[5.4.4] Hinze, J.O., "Turbulence", McGraw HillPublishing Co., New York, NY, 1975.
[5.4.5] Launder, B.E., and Spalding, D.B., "Lectures in Mathematical Models of Turbulence", Academic Press, London, 1972.
Holtec International 5-11 Holtec Report HI-971760
Table 5.1.1 PARTIALLISTING OF RERACK APPLICATIONS USING SIMILARMETHODS OF THERMAL-HYDRAULICANALYSIS PLANT DOCKET NO.
Enrico Fermi Unit 2 USNRC 50-341 Quad Cities 1 and 2 USNRC 50-254, 50-265 Rancho Seco USNRC 50-312 Grand Gulf Unit 1 USNRC 50-416 Oyster Creek USNRC 50-219 Pilgrim USNRC 50-293 V.C. Summer USNRC 50-395 Diablo Canyon Units 1 and 2 USNRC 50-275, 50-455 Byron Units 1 and 2 USNRC 50-454, 50-455 Braidwood Units 1 and 2 USNRC 50-456, 50-457 Vogtle Unit 2 USNRC 50<25 St. Lucie Unit 1 USNRC 50-425 Millstone Point Unit 1 USNRC 50-245 D.C. Cook Units 1 and 2 USNRC 50-315, 50-316 Indian Point Unit 2 USNRC 50-247 Three Mile Island Unit 1 USNRC 50-289 J.A. FitzPatrick USNRC 50-333 Shearon Harris USNRC 50-400 Hope Creek USNRC 50-354 Kuosheng Units 1 and 2 Taiwan Power Company Chin Shan Units 1 and 2 Taiwan Power Company Holtec International 5-12 Holtec Report HI-971760
N > . \'
V
'jI XI I
Table 5.1.1 (continued)
PARTIALLISTING OF RERACK APPLICATIONS USING SIMILARMETHODS OF THERMAL-HYDRAULICANALYSIS PLANT DOCKET NO.
Ulchin Unit 2 Korea Electric Power Corporation Laguna Verde Units 1 and 2 Comision Federal de Electricidad Zion Station Units 1 and 2 USNRC 50-295, 50-304 Sequoyah USNRC 50-327, 50-328 La Salle Unit One USNRC 50-373 Duane Arnold USNRC 50-331 Fort Calhoun USNRC 50-285 Nine Mile Point Unit One USNRC 50-220 Beaver Valley Unit One USNRC 50-334 Limerick Unit 2 USNRC 50-353 Ulchin Unit 1 Korea Electric Power Corporation Holtec International 5-13 Holtec Report Hl-971760
Table 5.2.1 DECAY PERIODS FOR A BOUNDING POOLS C AND D STORAGE CONFIGURATION PWR Fuel Assemblies BWR Fuel Assemblies Number of Assys Decay Period Number of Assys Decay Period 172 5 years 456 5 years 172 7 years 456 7 years 172 9 years 456 9 years 172 11 years 456 11 years 172 13 years 456 13 years:
172 15 years 483 15 years 172 17 years 172 19 years 172 21 years 172 23 years 232 25 years Holtec International 5-14 Holtec Report Hl-971760
Table 5.2.2 FUEL ASSEMBLIES INPUT DATAFOR DECAY HEAT EVALUATION Xtem Value PWR Assembly Irradiation Time 1,915 EFPDt PWR Assembly Specific Power 19.11 Mwt BWR Assembly Irradiation Time 2,028 EFPD BWR Assembly Specific Power 4.66 Mwt Effective Full Power Days Holtec International 5-15 Holtec Report HI-971760
Table 5.2.3 BOUNDING DECAY HEAT INPUT FROM STORED FUEL IN POOLS C AND D Decay Heat Load Fuel Assemblies (MillionBtu/hr)
BWR Fuel Assemblies 4.47 PWR Fuel Assemblies 11.16 Total 15.63 (4.57 MW)
Holtec International 5-16 Holtec Report HI-971760
Table 5.4.1 BOUNDING FUEL ASSEMBLIES HYDRAULICFLOW RESISTANCE PARAMETERS Parameter Value Permeability 105 Inertial Resistance Factor m'oltec International 5-17 Holtec Report HI-971760
Table 5.5.1 POOLS C AND D DMENSIONALDATA Parameter Value Pool C: Length 597.88" Width 320.60" Pool D: Length 383.36" Width 237 79" Water Depth 38.5 ft Pools-to-Transfer Canal Channel Width 24I I II Bottom Plenum 6 Pool C Downcomers North Wall 1.44"'.44" South Wall East Wall 2.36" West Wall 2.36" Pool D Downcomers North Wall 5.15" South Wall 5 pll East Wall 5.0" West Wall 5 pll A minimum uniform downcomer gap equal to 1.44" applied to both pools for CFD analysis.
Holtec International 5-18 Holtec Report HI-971760
Table 5.5.2 BULKAND LOCAL TEMPERATURES
SUMMARY
Item 'emperature Local temperature rise above bulk 6.8
('F)
Bulk pool maximum temperature limit 137.0 Peak Local Temperature 143.8 Local temperature values are conservatively computed based on neglecting forced cooling to pool D, as discussed at the beginning of Section 5.4 Holtec International 5-19 Holtec Report HI-971760
3000 2500 E
t2 2000 ACCEPTABLE REGION
~ 1500 E
UN-ACCEPTAB E REGION 1000 500 100 105 110 115 120 125 Spent Fuel Pool Incoming Water Temperature fDeg F]
FIGURE 5.3.1: C AND D POOLS MINIMUMTOTAL COOLING SYSTEM REQUIREMENTS CURVE AT 137 Deg. F BULK POOL TEMPERATURE H1-971760 5-20
MIT TB SCALE R2P.6
~ El. 279'-6 I
I I
S g 200.0 174.0 IRANSFER CANAL g
224 I98.0 I 2I6.0 I
I I
l92.0 278'-6
, EL 240.0 g EL 278'-6 278'-3 EL I
EL 278'-3 I (p I
L 279'-6
,'(B)
~ 69.0 EL L 36.0 PLAN VIEV TRANSFER CANAL VATER LEVEL C(o.ltfi SVSTBI PIPES 0TIERS I 3SFI6-ISA-2Q (l6') 9 7SF4-218-2&3 3SFI6-2S8-2I3 ( l6') OB 7SF3-Im-Za3 CAMIAL 04 3SFI2-6SB-M (12 )
(9 (l2')
' lJJ Cs 3SFI2W-2lk3 3SF12-I74SA-213 ( 12')
Q 3SFI2-17ISB-3k3 ( l2')
3SFI2-176SB-213 (l2')
RAM 3SFI2-I79SA-2(3 ( 12')
80TTN EL 246'-0 ELEVATIS VIEV FRQI SIIITH TB MRTH FIGURE 5.4.1: HARRIS C AND D POOLS PHYSICAL CONFIGURATION 70324%197175% 4 I 5-21
Jan 16 1998 x Grid ( 76 X 27 X 34 ) Fluent 4.32 t Fluent Inc.
FIGURE 5.5.1: PLAN VIEW OF THE HARRIS POOLS C AND D CFD MODEL HI-971760
uclcIcz&ucl&HUsn&EIu&&clu uuuuuuuuuucacauaxuamaaaea I
i uuuuazuuunmuunnuumunw o~~~cx~zx~~~~~~
t auuuuuuuuuuuuumueauuQI I CL&lnCKXCLMMZX&CLCXCL I
II fj cguuauuuuuuunuuuuuuuQ
~~ree44ai~~k85~~~ aaoncxExcL~~~~~cx~
jl I~
l~~~~lsmt Nl
~~ CaS ~~~~ ~~ ~ SRCl ~~ ~ ~ ~~ ~RS~
8%%
~ anaoaanaanann annaannaaonnKl aannnnauaanaauuauaa nnnaonnnannnn
-ononnnnaaannn o
o tuaanaunuanaaaaaannu tanuunnnuuaaununnnnn nannnnnnnnnno oaoonoaanannn
> t t uaaaaaaaunuaaannu auuaunauuanuunuuaua- annnoanannano aoaonnnoaannn iunuauanuuaaaanaananau uunuaauauaanuuauuaua iu t mannnuuuaannaaanunaaa Jan 15 1998 Grid ( 76 X 27 X 34 ) Fluent 4.32 Fluent Inc.
HGURE 5.5.2: PERSPECTIVE VIEW OF THE HARMS POOLS C AND D CFD MODEL Hi 97$ 76p 5-23
3.33E+02 3.33E+02 3.32E+02 3.32E+02 3.32E+02 3.31E+02 3.31E+02 3.31E+02 3.30E+02 3.30E+02 3.29E+02 3.29E+02 3.29E+02 3.28E+02 3.28E+02 3.28E+02 3.27E+02 3.27E+02 3.26E+02 3.26E+02 3.26E+02 3.25E+02 3.25E+02 3.25E+02 3.24E+02 3.24E+02 3.24E+02 3.23E+02 3.23E+02 3.22E+02 3.22E+02 Jan 16 1998 Temperature (K) Fluent 4.32 Z
Lmax = 3.331E+02 Lmin = 3.220E+02 Fluent Inc.
FIGURE 59.3: PEAK LOCALWATER TEMPERATURE IN THE RACK CELLS HI-971760 5-24
3.31E+02 3.31E+02 3.31E+02 3.30E+02 3.30E+02 3.30E+02 3.30E+02 3.30E+02 i
3.30E+02 3.29E+02 3.29E+02 3.29E+02 3.29E+02 3.29E+02 I-3.29E+02 I 3.28E+02 I.
3.28E+02 I 3.28E+02 I J 3.28E+02 I I
/
3.28E+02 r 3.27E+02 3.27E+02 3.27E+02 3.27E+02 3.27E+02 rr 3.27E+02 3.26E+02 3.26E+02 3.26E+02 3.26E+02 Jan 16 1998 Velocity Vectors Colored by Temperature (K) Fluent 4.32 Tmax = 3.310E+02 Tmin = 3.259E+02 Fluent lnc.
FIGURE 5.5.4: POOLS INTERCONNECTING CHANNELFLOW VELOCITYVECTORS ELEVATIONVIEWPLOT Hl-971760 5-2S
1.42E+00 1.37E+00 1.33E+00 1.28E+00 1.23E+00 1.18E+00 1.13E+00 1.08E+00 1.03E+00 9.82E<1 9.33E-01 8.83E-01 8.34E-01 7.85E-01 7.36E-01 e'rr 6.87E-01 6.38E-01 5.89E%1 5.40E-01 4.91E%1 4.42E-01 3.93E-01 3.44E-01 2.95E-01 2.46E41 1.96E-01 ~
~
iI 1.47E-01 I r~
9.83E-02 4.92E-02 r .
1 1.61E-04 rr'elocity Jan 16 1998 Vectors (M/S) Fluent 4.32 Lmax = 1.423E+00 Lmin = 1.609E-04 Fluent Inc.
FIGURE 5.55: POOL COOLING INLET/OUTLETPIPING REGION FLOW VELOCITYVECTORS PLOT Hl-971760-5-26
6.1 This section considers the structural adequacy of the new maximum density spent fuel racks under all loadings postulated for normal, seismic, and accident conditions at Harris. The existing spent fuel storage racks are also examined for stability during the installation process.
The analyzed storage rack configurations with the new racks in place are shown in Figures 1.2 and 1.3.
The analyses undertaken to confirm the structural integrity of the racks are performed in compliance with the USNRC Standard Review Plan [6.1.1] and the OT Position Paper [6.1.2].
For each of the analyses, an abstract of the methodology, modeling assumptions, key results, and summary of parametric evaluations are presented. Delineation of the relevant criteria are discussed in the text associated with each analysis.
0 The response of a free-standing rack module to seismic inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting in impacts and friction effects. Some of the unique attributes of the rack dynamic behavior include a large fraction of the total structural mass in a confined rattling motion, friction support of rack pedestals against lateral motion, and large fluid coupling effects due to deep submergence and independent motion of closely spaced a'djacent structures.
Linear methods, such as modal analysis and response spectrum techniques, cannot accurately simulate the structural response of such a highly nonlinear structure to seismic excitation. An accurate simulation is obtained only by direct integration of the nonlinear equations of motion with the three pool slab acceleration time-histories applied as the forcing functions acting simultaneously.
Holtec International 6-1 Holtec Report HI-971760
Whole Pool Multi-Rack (WPMR) analysis is the vehicle utilized in this project to simulate the i
dynamic behavior of the complex storage rack structures. The following sections provide the basis for this selection and discussion on the development of the methodology.
6.2.1 Reliable assessment of the stress field and kinematic behavior of the rack modules calls for a conservative dynamic model incorporating all key attributes of the actual structure. This means that the model must feature the ability to execute the concurrent motion forms compatible with the free-standing installation of the modules.
The model must possess the capability to effect momentum transfers which occur due to rattling of fuel assemblies inside storage cells and the capability to simulate lift-offand subsequent impact of support pedestals with the pool liner (or bearing pad). The contribution of the water mass in the interstitial spaces around the rack modules and within the storage cells must be modeled in an accurate manner since erring in quantification of fluid coupling on either side of the actual value is no guarantee of conservatism.
The Coulomb friction coefficient at the pedestal-to-pool liner (or bearing pad) interface may lie in a rather wide range and a conservative value of friction cannot be prescribed a priori. In fact, a perusal of results of rack dynamic analyses in numerous dockets (Table 6.2. 1) indicates that an upper bound value of the coefficient of friction often maximizes the computed rack as well as the equivalent elastostatic stresses. 'isplacements In short, there are a large number of parameters with potential influence on the rack kinematics. The comprehensive structural evaluation must deal with all of these without sacrificing conservatism.
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The three-dimensional single rack dynamic model introduced by Holtec International in the Enrico Fermi Unit 2 rack project (ca. 1980) and used in some 50 rerack projects since that time (Table 6.2. 1) addresses most of the above mentioned array of parameters. The details of this methodology are also published in the permanent literature [6.2.1]. Despite the versatility of the 3-D seismic model, the accuracy of the single rack simulations has been suspect due to one key element; namely, hydrodynamic participation of water around the racks. During dynamic rack motion, hydraulic energy is either drawn from or added to the moving rack, modifying its submerged motion in a significant manner. Therefore, the dynamics of one rack affects the motion of all others in the pool.
A dynamic simulation which treats only one rack, or a small grouping of racks, is intrinsically inadequate to predict the motion of rack modules with any quantifiable level of accuracy.
Three-dimensional Whole Pool Multi-Rack analyses carried out on several previous plants demonstrate that single rack simulations under predict rack displacement during seismic responses [6.2.2].
Briefly, the 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the array of variables as follows:
Parametric runs are made with upper bound and lower bound values of the coefficient of friction. The limiting values are based on experimental data which have been found to be bounded by the values 0.2 and 0.8. Simulations are also performed with the array of pedestals having randomly chosen coefficients of friction in a Gaussian distribution with a mean of 0.5 and lower and upper limits of 0.2 and 0.8, respectively. In the fuel rack simulations, the Coulomb friction interface between rack support pedestal and liner is simulated by piecewise linear (friction) elements. These elements function only when the pedestal is physically in contact with the pool liner.
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Rack elasticity, relative to the rack base, is included in the model by introducing linear springs to represent the elastic bending action, twisting, and extensions.
Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface. These interface gaps are modeled using nonlinear spring elements. The term "nonlinear spring" is a generic term used to denote the mathematical representation of the condition where a restoring force is not linearly proportional to displacement.
The fuel assemblies are conservatively assumed to rattle in unison which obviously exaggerates the contribution of impact against the cell wall.
Holtec International extended Fritz's classical two-body fluid coupling model to multiple bodies and utilized it to perform the first two-dimensional multi-rack analysis (Diablo Canyon, ca. 1987). Subsequently, laboratory experiments were conducted to validate the multi-rack fluid coupling theory. This technology was incorporated in the computer code DYNARACK(a.k.a. MR216) [6.2.4] which handles simultaneous simulation of all racks in the pool as a Whole Pool Multi-Rack 3-D analysis. This development was first utilized in Chinshan, Oyster Creek, in earlier projects at the Harris plant [6.2.1, 6.2.3] and, subsequently, in numerous other rerack projects. The WPMR analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses, and also serve to improve predictions of rack kinematics.
For closely spaced racks, demonstration of kinematic compliance is verified by including all modules in one comprehensive simulation using a WPMR model. In WPMR analysis, all rack modules are modeled simultaneously and the coupling effect due to this multi-body motion is included in the analysis. Due to the superiority of this technique in predicting the dynamic behavior of closely spaced submerged storage racks, the Whole Pool Multi-Rack analysis methodology is used for this project.
Holtec International Holtec Report HI-971760
~3 The implementation of the storage capacity increase in pools C and D will be performed on an as needed basis through incremental phases (campaigns). Figures 6.3.1 and 6.3.2 identify the fully implemented configuration and also designates which racks will be included in each of the campaigns. The new high density storage racks are analyzed for the anticipated configurations at the completion of each of the installation campaigns. Evaluated configurations of the two pools are also handled separately, since the'pools are physically separated by the surrounding concrete walls. The analyzed configurations considered are described as follows:
Incremental Incremental Number of I 14 1680 II 10 1260 III 6 750 D 500 525 The materials utilized in fabrication of the rack components are identified in Table 6.3.1.
The cartesian coordinate system utilized within the rack dynamic model has the following nomenclature:
x = Horizontal axis along plant North y = Horizontal axis along plant West z = Vertical axis upward from the rack base 6.3.1 For the dynamic rack simulations, the dry PWR fuel weight is taken to be 1600 lbs and the dry BWR fuel weight is taken to be 680 lbs.
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'1 The synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) are generated in accordance with the provisions of SRP 3.7.1 [6.4.1]. In order to prepare an acceptable set of acceleration time-histories, Holtec International's proprietary code GENEQ
[6.4.2] is utilized.
A preferred criterion for the synthetic time-histories in SRP 3.7.1 calls for both the response spectrum and the power spectral density corresponding to the generated acceleration time-history to envelope their target (design basis) counterparts with only finite enveloping infractions. The time-histories for the pools have been generated to satisfy this preferred (and more rigorous) criterion. The seismic files also satisfy the requirements of statistical independence mandated by SRP 3.7.1.
Figures 6.4.1 through 6.4.3 and 6.4.4 through 6.4.6 provide plots of the time-history accelerograms which were generated over a 20 second duration for OBE and SSE events, respectively.
Results of the correlation function of the three time-histories are given in Table 6.4.1 ~
Absolute values of the correlation coefficients are shown to be less than 0.15, indicating that the desired statistical independence of the three data sets has been met.
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Recognizing that the analysis work effort must deal with both stress and displacement criteria, the sequence of model development and analysis steps that are undertaken are summarized in the following:
Prepare 3-D dynamic models suitable for a time-history analysis of the new maximum density racks. These models include the assemblage of all rack modules in each pool. Include all fluid coupling interactions and mechanical coupling appropriate to performing an accurate non-linear simulation. This 3-D simulation is referred to as a Whole Pool Multi-Rack model.
- b. Perform 3-D dynamic analyses on various physical conditions (such as coefficient of friction and extent of cells containing fuel assemblies). Archive appropriate displacement and load outputs from the dynamic model for post-processing.
- c. Perform stress analysis of high stress areas for the limiting case of all the rack dynamic analyses. Demonstrate compliance with ASME Code Section III,
- Subsection NF limits on stress and displacement.
The dynamic modeling of the rack structure is prepared with special consideration of all nonlinearities and parametric variations. Particulars of modeling details and assumptions for the Whole Pool Multi-Rack analysis of racks are given in the following:
The fuel rack structure motion is captured by modeling the rack as a 12 degree-of-freedom structure. Movement of the rack cross-section at any height is described by six degrees-of-freedom of the rack base and six degrees-of-freedom at the rack top. In this manner, the response of the module, relative to the baseplate, is captured in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom and simulate rack stiffness.
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Rattling fuel assemblies within the rack are modeled by five lumped masses located at H, .75H, .5H, .25H, and at the rack base (H is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the baseplate level. The centroid of each fuel assembly mass can be located off-center, relative to the rack structure centroid at that level, to simulate a partially loaded rack.
Seismic motion of a fuel rack is characterized by random rattling of fuel assemblies in their individual storage locations. All fuel assemblies are assumed to move in-phase within a rack. This exaggerates computed dynamic loading on the rack structure and, therefore, yields conservative results.
Fluid coupling between rack and fuel assemblies, and between rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy.
Inclusion of these effects uses the methods of [6.5.2, 6.5.3] for rack/assembly coupling and for rack-to-rack coupling.
Fluid damping and form drag are conservatively neglected.
Sloshing is found to be negligible at the top of the rack and is, therefore, neglected in the analysis of the rack.
Potential impacts between the cell walls of the new racks and the contained fuel assemblies are accounted for by appropriate compression-only gap elements between masses involved. 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 two horizontal directions. Bottom gap elements are located at the baseplate elevation. The initial gaps reflect the presence of baseplate extensions, and the rack stiffnesses are chosen to simulate local structural detail.
Pedestals are modeled by gap elements in the vertical direction and as "rigid links" for transferring horizontal stress. Each pedestal support is linked to the pool liner (or bearing pad) by two friction springs. The spring rate for the friction springs includes any lateral elasticity of the stub pedestals. Local pedestal vertical spring stiffness accounts for floor elasticity and for local rack elasticity just above the pedestal.
Rattling of fuel assemblies inside the storage locations causes the gap between fuel assemblies and cell wall to change from a maximum of twice the nominal gap to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap in order to provide a conservative measure of fluid resistance to gap closure.
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The model for the rack is considered supported, at the base level, on four pedestals modeled as non-linear compression only gap spring elements and eight piecewise linear friction spring elements; these elements are properly located with respect to the centerline of the rack beam, and allow for arbitrary rocking and sliding motions.
6.5.1.2 Figure 6.5.1 shows a schematic of the dynamic model of a single rack. The schematic depicts many of the characteristics of the model including all of the degrees-of-freedom and some of the spring restraint elements.
Table 6.5.1 provides a complete listing of each of the 22 degrees-of-freedom for a rack model.
Six translational and six rotational degrees-of-freedom (three of each type on each end) describe the motion of the rack structure. Rattling fuel mass motions (shown at nodes 1', 2, 3', 4', and 5'n Figure 6.5.1) are described by ten horizontal translational degrees-of-freedom (two at each of the five fuel masses). The vertical fuel mass motion is assumed ( and modeled) to be the same as that of the rack baseplate.
Figure 6.5.2 depicts the fuel to rack impact springs (used to develop potential impact loads between the fuel assembly mass and rack cell inner walls) in a schematic isometric. Only one of the five fuel masses is shown in this figure. Four compression only springs, acting in the horizontal direction, are provided at each fuel mass.
Figure 6.5.3 provides a 2-D schematic elevation of the storage rack model, discussed in more detail in Section 6.5.3. This view shows the vertical location of the five storage masses and some of the support pedestal spring members.'igure 6.5.4 shows the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending plane a shear and bending spring simulate elastic effects [6.5.4].
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Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model.
Figure 6.5.5 depicts a single rack module with its surrounding impact springs (used to develop potential impact loads between racks or between rack and wall). Figures 6.5.6 through 6.5.13 show the rack numbering schemes used for the WPMR analyses of both pools. These figures also provide the numbering scheme for all of the rack periphery compression only gap elements.
6.5.2 In its simplest form, the so-called "fluid coupling effect" [6.5.2, 6.5.3] can be explained by considering the proximate motion of two bodies under water. Ifone body (mass m,) vibrates adjacent to a second body (mass mg, and'both bodies are submerged in frictionless fluid, then Newton's equations of motion for the two bodies are:
~~ ~~
(m, + M>>) X, + M;~ X, = applied forces on mass m, + 0 (X,2)
M>> X, + (m, + Mg X, = applied forces on mass m, + 0 (X, )
Xand X, denote absolute accelerations of masses m, and mrespectively, and the notation O(X') denotes nonlinear terms.
Mit Mtg Mgt and M~ are fluid coupling coefficients which depend on body shape, relative disposition, etc. Fritz [6.5.3] gives data for M,, for various body shapes and arrangements.
The fluid adds mass to the body (M>> to mass m,), and an inertial force proportional to acceleration of the adjacent body (mass mg. Thus, acceleration of one body affects the force field on another. This force field is a function of inter-body gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a storage location encounters this effect.
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For example, fluid coupling behavior will be experienced between nodes 2 and 2* in Figure 6.5.1. The rack analysis also contains inertial fluid coupling terms which model the effect of fluid in the gaps between adjacent racks.
Terms modeling the effects of fluid flowing between adjacent racks in a single rack analysis suffer from the inaccuracies described earlier. These terms are usually computed assuming that all racks adjacent to the rack being analyzed are vibrating in-phase or 180'ut of phase.
The WPMR analyses do not require any assumptions with regard to phase.
Rack-to-rack gap elements have initial gaps set to 100% of the physical gap between the racks or between outermost racks and the adjacent pool walls.
6.5.2.1 During the seismic event, all racks in the pool are subject to the input excitation simultaneously; The motion of each free-standing module would be autonomous and independent of others as long as they did not impact each other and no water were present in the pool. While the scenario of inter-rack impact is not a common occurrence and depends on rack spacing, the effect of water the so-called fluid coupling effect is a universal factor.
As noted in Ref. [6.5.2, 6.5.4], the fluid forces can reach rather large values in closely spaced rack geometries. It is, therefore, essential that the contribution of the fluid forces be included in a comprehensive manner. This is possible only if all racks in the pool are allowed to execute 3-D motion in the mathematical model. For this reason, single rack or even multi-rack models involving only a portion of the racks in the pool, are inherently inaccurate. The Whole Pool Multi-Rack model removes this intrinsic limitation of the rack dynamic models by simulating the 3-D motion of all modules simultaneously. The fluid coupling effect, therefore, encompasses interaction between every set of racks in the pool, i.e., the motion of one rack produces fluid forces on all other racks and on the pool walls. Stated more formally, both near-field and far-field fluid coupling effects are included in the analysis.
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The derivation of the fluid coupling matrix [6.5.5] relies on the classical inviscid fluid mechanics principles, namely the principle of continuity and Kelvin's recirculation theorem.
While the derivation of the fluid coupling matrix is based on no artificial construct, it has been nevertheless verified by an extensive set of shake table experiments [6.5.5].
6.5.3 Table 6.5.2 lists all spring elements used in the 3-D, 22-DOF, rack model for Campaign I of pool D. This set of elements is chosen since it represents the smallest of the models and provides a sufficient example to describe spring element numbering of Campaign II of pool D and the larger pool C models, which are similar. Three element types are used in the rack models. Type 1, are linear elastic elements used to represent the beam-like behavior of the integrated rack cell matrix. Type 2 elements are the piece-wise linear friction springs used to develop the appropriate forces between the rack pedestals and the supporting bearing pads.
Type 3 elements are non-linear gap elements which model gap closures and subsequent impact loadings (i.e.', between fuel assemblies and the storage cell inner walls, and rack outer
~ ~
periphery spaces.
A detailed numbering scheme for the rack-to-rack and rack-to-wall gap elements for each of the pool models is provided in Figures 6.5.6 through 6.5.13.
Ifthe simulation model is restricted to two dimensions (one horizontal motion plus one vertical motion, for example), for the purposes of model clarification only, then Figure 6.5.3 describes the configuration. This simpler model is used to elaborate on the various stiffness modeling elements.
Type 3 gap elements modeling impacts between fuel assemblies and racks have local stiffness K; in Figure 6.5.3. In Table 6.5.2, for example, type 3 gap elements 5 through 8 act on the rattling fuel mass at the rack top. Support pedestal spring rates K, are modeled by type 3 gap Holtec International Holtec Report HI-971760
elements 1 through 4, as listed in Table 6.5.2. Local compliance of the concrete floor is included in Ks. The type 2 friction elements listed in Table 6.5.2 are shown in Figure 6.5.3 as K,. The spring elements depicted in Figure 6.5.4 represent type 1 elements.
Friction at support/liner interface is modeled by the piecewise linear friction springs with large stiffness K, up to the limiting lateral load IrN, where N is the current 'uitably compression load at the interface between support and liner. At every time-step during transient analysis, the current value of N (either zero ifthe pedestal has lifted off the liner, or a compressive finite value) is computed.
The gap element Ks, modeling the effective compression stiffness of the structure in the vicinity of the support, includes stiffness of the pedestal, local stiffness of the underlying pool slab, and local stiffness of the rack cellular structure above the pedestal.
The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D motions and include all stiffness elements listed in Table 6.5.2.
6.5.4 R
To eliminate the last significant element of uncertainty in rack dynamic analyses, multiple simulations are performed to adjust the friction coefficient ascribed to the support pedestal/pool bearing pad interface. These friction coefficients are chosen consistent with the two bounding extremes from Rabinowicz's data [6.5.1]. Simulations are also performed by imposing intermediate value friction coefficients developed by a random number generator with Gaussian normal distribution characteristics. The assigned values are then held constant during the Holtec International 6-13 Holtec Report HI-971760
entire simulation in order to obtain reproducible results.t Thus, in this manner, the WPMR analysis results are brought closer to the realistic structural conditions.
The coefficient of friction (p) between the pedestal supports and the pool floor is indeterminate. According to Rabinowicz [6.5. 1], results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of p to be 0.503 with standard deviation of 0.125. Upper and lower bounds (based on twice standard deviation) are 0.753 and 0.253, respectively. Analyses are therefore performed for coefficient of friction values of 0.2 (lower limit) and for 0.8 (upper limit), and for random friction values clustered about a mean of 0.5. The bounding values of p = 0.2 and 0.8 have been found to envelope the upper limit of module response in previous rerack projects.
6.5.5 Using the structural model discussed in the foregoing, equations of motion corresponding to each degree-of-freedom are obtained using Lagrange's Formulation [6.5.4]. The system kinetic energy includes contributions from solid structures and from trapped and surrounding fluid.
The final system of equations obtained have the matrix form:
f0' d g c&
= lG]
where:
[M] total mass matrix (including structural and fluid mass contributions). The size of this matrix will be 22n x22n for a WPMR analysis (n = number of racks in the model).
It is noted that MR216 has the capability to change the coefficient of friction at any pedestal at each instant of contact based on a random reading of the computer clock cycle. However, exercising this option would yield results that could not be reproduced. Therefore, the random choice of coefficients is made only once per run.
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q - the nodal displacement vector relative to the pool slab displacement (the term with q indicates the second derivative with respect to time, i.e., acceleration)
[G] - a vector dependent on the given ground acceleration
[Q] ~ - a vector dependent on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom
'l The above column vectors have length 22n. The equations can be rewritten as follows:
d g =
IW + lW 'Gl dE This equation set is mass uncoupled, displacement coupled at each instant in time. The numerical solution uses a central difference scheme built into the proprietary computer program MR216 [6.2.4].
There are two sets of criteria to be satisfied by the rack modules:
Per Reference [6.1.1], in order to,be qualified as a physically stable structure it is necessary to demonstrate that an isolated rack in water would not overturn when an event of magnitude:
~ 1.5 times the upset seismic loading condition is applied.
~ 1.1 times the faulted seismic loading condition is applied.
b.
Stress limits must not be exceeded under the postulated load combinations provided herein.
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6.6.2 The stress limits presented below apply to the rack structure and are derived from the ASME Code, Section III, Subsection NF [6.6. 1]. Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code, Section II, Part D
[6.6.2], and are listed in Table 6.3.1.
- a. Allowable stress in tension on a net section is:,
Ft = 0.6 Sy Where, S= yield stress at temperature, and F, is equivalent to primary membrane stress.
- b. Allowable stress in shear on a net section is:
F= .4S
- c. Allowable stress in compression on a net section F =S .47- k$
444 r kl/r for the main rack body is based on the full height and cross section of the honeycomb region and does not exceed 120 for all sections.
unsupported length of component k = length coefficient which gives influence of boundary conditions. The following values are appropriate for the described end conditions:
1 (simple support both ends) 2 (cantilever beam)
'/i (clamped at both ends) radius of gyration of component Holtec International 6-16 Holtec Report HI-971760
- d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is:
F, = 0.60 Sy (equivalent to primary bending)
Combined bending and compression on a net section satisfies:
fa + mx fbx + my fby where:
f, = Direct compressive stress in the section fb Maximum bending stress along x-axis fb, = Maximum bending stress along y-axis C = 0.85 Cy = 0.85 D 1 - (f,/F',)
Dy 1 - (f,/F',y) ex,ey (m E)/(2.15 (kl/r)y)
E = Young's Modulus and subscripts x,y reflect the particular bending plane.
Combined flexure and compression (or tension) on a net section:
- f. fb, fby (I 0 0.6S Fbx Fb The above requirements a'e to be met for both direct tension or compression.
F= 0.3 S where S is the weld material ultimate strength at temperature. For fillet weld legs in contact with base metal, the shear stress on the gross section is limited to 0.4Sy, where Sy is the base material yield strength at temperature.
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Section F-1334 (ASME Section III, Appendix F) [6.6.2], states that the limits for the Level D condition are the minimum of 1.2 (S/F,) or (0.7S/F,) times the corresponding limits for the Level A condition. S is ultimate tensile stress at the specified rack design temperature. Examination of material properties for 304L stainless demonstrates that 1.2 times the yield strength is less than the 0.7 times the ultimate strength.
Exceptions to the above general multiplier are the following:
a) Stresses in shear shall not exceed the lesser of 0.72Sor 0.42S. In the case of the Austenitic Stainless material used here, 0.72Sgoverns.
b) Axial Compression Loads shall be limited to 2/3 of the calculated buckling load.
c) Combined Axial Compression and Bending - The equations for Level A conditions shall apply except that:
F, = 0.667 x Buckling Load/ Gross Section Area, and the terms F',and F',may be increased by the factor 1.65.
d) For welds, the Level D allowable maximum weld stress is not specified in, Appendix F of the ASME Code. An appropriate limit for weld throat stress is conservatively set here as:
F= (0.3 S) x factor where:
factor = (Level D shear stress limit)/(Level A shear stress limit) 6.6.3 For convenience, the stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value.
The limiting value of each stress factor is 1.0, based on the allowable strengths for each level, for Levels A, B, and D (where 1.2S( .7S).
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The stress factors reported are:
R, = Ratio of direct tensile or compressive stress on a net section to its allowable value (note pedestals only resist compression)
R~ = Ratio of gross shear on a net section in the x-direction to its allowable value
/
R3 = Ratio of maximum x-axis bending stress to its allowable value for the section Ratio of maximum y-axis bending stress to its allowable value for the section R, = Combined flexure and compressive factor (as defined in the foregoing)
R, = Combined flexure and tension (or compression) factor (as defined in the foregoing)
R, = Ratio of gross shear on a net section in the y-direction to its allowable value Holtec International 6-19 Holtec Report HI-971760
6.6.4 The applicable loads and their combinations which must be considered in the seismic analysis of rack modules are excerpted from Refs. [6.1.2] and [6.6.3].
The load combinations considered are identified below:
Loading Combination Service Level D+L Level A D+ L+T, D+L+T +E D+L+T,+E Level B D+L+T +P Level D D+L+T,+E'+L+T,+Fd The functional capability of the fuel racks must be demonstrated.
D Dead weight-induced loads (including fuel assembly weight)
L Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path)
P, Upward force on the racks caused by postulated stuck fuel assembly Fd Impact force from accidental drop of the heaviest load from the maximum possible height.
E Operating Basis Earthquake (OBE)
E' Safe Shutdown Earthquake (SSE) 0 Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition)
Ta Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions)
T, and T, produce local thermal stresses. The worst thermal stress field in a fuel rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing maximum possible Holtec International 6-20 Holtec Report Hl-971760
temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary (thermal) stresses.
6.7 Whole Pool Multi-Rack (WPMR) simulations have been performed to investigate the structural integrity of each rack array. Pools C and D had separate runs performed for the SSE seismic event considering pools filled and partially filled with racks. The partially filled pools represent interim configurations subsequent to the installation campaigns identified for each pool in Figures 6.3.1 and 6.3.2. The configurations were considered with friction coefficients of 0.8, 0.2, and a guassian distribution with a mean of 0.5 (i.e., random coefficient of friction (COF) with upper and lower limits of 0.8 and 0.2). The SSE simulations were performed and conservatively compared against the allowables for OBE events. This process eliminated the need for performing OBE simulations to significantly reduce the number of runs needed. Due to the mild SSE earthquake postulated for Harris, this conservative evaluation technique yielded satisfactory design margins.
The overturning check simulations were performed to determine the behavior of the highest aspect (width/length) ratio racks under both the OBE and SSE events. The overturning check simulations considered a single rack (i.e., no dynamic fluid coupling to walls or other racks) half full with fuel all loaded along the long side of the rack.
'he rack numbering schemes used to identify the racks in each simulation model are introduced in Figures 6.5.6 through 6.5.13. The circled rack numbers in the figures correspond to the rack numbers shown in the following tables.
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The following table presents a complete listing of the simulations discussed herein.
Consideration of the parameters described above resulted in the following r'uns:
Pool C (Campaign I) 0.8 SSE Pool C (Campaign I) 0.2 SSE Pool C (Campaign I) Random SSE Pool C (Campaign Il) 0.8 SSE Pool C (Campaign II) 0.2 SSE Pool C (Campaign II) Random SSE 7 Pool C (Campaign III - Full) 0.8 SSE 8 Pool C (Campaign III - Full) 0.2 SSE 9 Pool C (Campaign III - Full) Random SSE 10 Pool D (Campaign I) 0.8 SSE Pool D (Campaign I) 0.2 SSE 12 Pool D (Campaign I) Random SSE 13 Pool D (Campaign II - Full) 0.8 SSE 14 Pool D (Campaign II - Full) 0.2 SSE 15 Pool D (Campaign II- Full) Random SSE 16 Single Holtec Rack 0.8 OBE x 1.5 Overturning Check 17 Single HoltecRack 0.8 SSE x 1.1 Overturning Check Holtec International 6-22 Holtec Report HI-971760
6.8 The results from the MR216 runs may be seen in the raw data output files. The MR216 output files archive all of the loads and displacements at key locations within each of the rack modules at every time step throughout the entire'time history duration. However, due to the huge quantity of output data, a post-processor is used to scan for worst case conditions and develop the stress factors discussed in subsection 6.6.3.
Further reduction in this bulk of information is provided in this section by extracting the worst case values from the parameters of interest; namely displacements, support pedestal forces, impact loads, and stress factors. This section also summarizes other analyses performed to develop and evaluate structural member stresses which are not determined by the post processor.
6.8.1 A tabulated summary of the maximum displacement for each simulation is provided below with.
the location/direction terms defined as follows:
uxt = displacement of top corner of rack, relative to the slab, in the East-West direction for pool C racks and in the North-South direction for pool D rack modules.
uyt = displacement of top corner of rack, relative to the slab, in the North-South direction for pool C racks and in the East-West direction for pool D rack modules.
Simulations 16 and 17 were performed to evaluate the potential for overturning of a single Holtec rack isolated in the pool without any fluid coupling to adjacent racks or walls. This simulation was performed to account for the unlikely possibility of a seismic event occurring during the installation process.
Holtec International 6-23 Holtec Report HI-971760
The following maximum rack displacements (in inches) are obtained for each of the runs:
Pool Event Run COF Maximum Location/ Rack Displacement Direction (inches)
Pool C Campaign I SSE 0.8 1.132 16 SSE 2 0.2 0.631 uyt SSE 3 Random 0.878 uxt 16 Pool C Campaign II SSE 4 0.8 1.494 uxt 28 SSE 0.2 0.917 uxt SSE 6 Random 0.878 uxt 16 Pool C Campaign III SSE 7 0.8 0.617 uyt 29 SSE 8 0.2 0.740 uyt SSE 9 Random 0.684 uyt Pool D Campaign I SSE 10 0.8 0.520 Uxt SSE 11 0.2 0.390 uyt SSE 12 Random 0.521 Uxt Pool D Campaign II SSE 13 0.8 0.575 uyt SSE 14 0.2 0.595 uyt SSE 15 Random 0.576 uyt Tipover: Single Holtec Rack OBE 16 0.8 0.347 Uyt PWR Tipover: Single Holtec Rack SSE 17 0.8 1.054 uyt PWR The largest displacement of 1.494 occurs in run 4 for rack 28 in the X direction. Since this displacement maintains the centroid of the rack well within the boundaries represented by the support pedestals, there is no possibility of rack overturning (tipover).
Holtec International 6-24 Holtec Report HI-971760
6.8.2 Pedestal number 1 for each rack is located in the +X, -Y corner of each rack. Numbering increases counterclockwise around the periphery of the rack. The following bounding vertical pedestal forces (in kips) are obtained for each run:
Pool Event Run COF Maximum Rack Ped.
Pedestal Load (kips)
Pool C Campaign I SSE 08 122 SSE 0.2 115 SSE Random 123 Pool C Campaign II SSE 0.8 153 SSE 0.2 121 SSE Random 134 Pool C Campaign III SSE 0.8 113 SSE 0.2 110 SSE Random 122 26 Pool D Campaign I SSE 10 0.8 118 SSE 0.2 112 SSE 12 Random 114 Pool D Campaign II SSE 13 0.8 135 SSE 14 0.2 116 SSE 15 Random 130 As may be seen, the highest pedestal load is 153,000 lbs and occurs in run 4 for pedestal 2 of rack 5. Figure 6.8.1 provides a plot of the vertical force of this pedestal transmitted to the bearing pad over the entire duration of the SFP, 0.8 COF, SSE, campaign II simulation.
Holtec International 6-25 Holtec Report HI-971760
6.8.3 The maximum (x or y direction) shear load (in kips) bounding all pedestals for each simulation are reported below and are obtained by inspection of the complete tabular data.
Pool Event Run COF Maximum Rack Friction Load (kips)
Pool C Campaign I SSE 0.8 46 SSE 2 0.2 22.3 SSE 3 Random 41.7 13 Pool C Campaign Il SSE 4 0.8 44.2 13 SSE 5 0.2 22.2 SSE 6 Random 40.9
~ Pool C Campaign III SSE 7 0.8 43.4 SSE 8 0.2 19.7 SSE Random 45.8 26 Pool D Campaign I SSE 10 0.8 45.6 SSE 11 0.2 19.7 SSE 12 Random 34.4 Pool D Campaign Il SSE 13 0.8 .42.3 SSE 14 0.2 22.3 SSE 15 Random 42.4 Holtec International 6-26 Holtec Report Hl-971760
6.8.4 A freestanding rack,'y definition, is a structure subject to potential impacts during a seismic event. Impacts arise from rattling of the fuel assemblies in the storage rack locations and, in some instances, from localized impacts between the racks, or between a peripheral rack and the pool wall. The following sections discuss the bounding values of these impact loads.
6.8.4.1 t
As is often the case with close rack spacing, some rack to rack impacts occur. The following instantaneous maximum impact forces and locations are identified for each of the simulations performed. Listings are only given for those simulations within which an impact occurred.
The element numbering is identified in Figures 6.5.6 through 6.5.13.
Run Impact Load Element Location Run Impact Load Element Location (kips) (kips) 3.0 494 Top 11.3 814 Bottom 8.1 503 Top 8.1 817 Top 8.1 Top 8.1 818 , Top 8.1 583 Top 4.9 831 Bottom 8.1 584 Top 8.4 937 Bottom 8.1 493 Top 5.6 945 Bottom 8.1 494 Top 6.4 991 Bottom 6.7 493 Top 6.5 992 Bottom 8.1 494 Top 8.1 736 Top 8.1 503 Top 1.9 746 Top 8.1 Top 8.1 759 Top 3.0 539 Top 7.9 760 Top 2.1 540 Top 8.1 781 Top 8.1 583 Top 8.1 782 Top 8.1 584 Top 8.1 789 Top 8.1 599 To 8.1 790 To Holtec International 6-27 Holtec Report HI-971760
Run Impact Load Element Location Run Impact Load Element Location (kips) (kips) 8.1 Top 4.9 799 Top 5.3 736 Top 8.1 817 Top 8.1 759 Top 8.1 818 Top 8.1 760 Top 1.2 827 Top 8.1 781 Top 8.1 '28 Top 8.1 782 Top 8.1 835 Top 8.1 799 Top 8.1 836 Top 8.1 800 Top 1.9 914 Top 8.1 817 Top 1.8 946 Bottom 818 Top 8.1 949 Top 8.1 827 Top 8.1 950 Top 8.1 828 Top 8.1 979 Top 8.1 835 Top 8.1 980 Top 8.1 836 Top 2.6 982 Bottom 8.1 907 Top 8.1 986 Top 8.1 908 Top 12.9 992 Bottom 8.1 913 Top 8.1 913 Top 8.1 914 Top 8.1 914 Top 8.1 979 Top 5.3 949 Top 8.1 980 Top 0.8 950 Top 6.7 736 Top 8.0 991 Bottom 16.7 743 Bottom 11.3 992 Bottom 7.7 Bottom 8.1 913 Top 10.7 756 Bottom 8.1 914 Top 4.5 777 Bottom 8.1 949 Top 22.1 778 Bottom 7.8 950 Top 16.1 813 Bottom Holtec International 6-28 Holtec Report HI-971760
6.8.4.2 Storage racks do not impact the pool walls under any simulation.
6.8.4.3 A review of the results from each simulation allows determination of the maximum instantaneous impact load between fuel assembly and fuel cell wall at any modeled impact site.
The maximum values obtained are reported in the following table.
Pool Event Run COF Maximum Fuel Rack Impact Load ebs)
Pool C Campaign I SSE 0.8 532 SSE 2 0.2 562 SSE 3 Random 605 Pool C Campaign II SSE 4 0.8 531 25 SSE 5 0.2 548 SSE 6 Random 535 22 Pool C Campaign III SSE 7 0.8 525 17 SSE 8 0.2 527 17 SSE 9 Random 515 17 Pool D Campaign I SSE 10 0.8 473 SSE 11 0.2 591 SSE 12 Random 473 Pool D Campaign II SSE 13 0.8 472 12 SSE 14 0.2 462 12 SSE 15 Random 472 12 The maximum fuel to cell wall impact load is 605 pounds. Based on fuel manufacturer's data, loads of this magnitude will not damage the fuel assembly.
Holtec International 6-29 Holtec Report HI-97 1760
69 6.9.1 The vertical and shear forces at the bottom casting-pedestal interface are available as a function of time. The maximum values for the stress factors defined in Section 6.6.3 can be determined for every pedestal in the array of racks by scanning this data to select the limiting loads and performing calculations to determine member stresses. These two tasks are performed by a post-processor. With this information available, the structural integrity of the pedestal can be assessed and reported. The net section maximum (in time) bending moments and shear forces can also be determined at the bottom casting-rack cellular structure interface for each spent fuel rack in the pool. This allows the evaluation of the maximum stress in the limiting rack cell (box).
v The tables presented in this section provide limiting stress factor results for male and female pedestals, and for the entire spent fuel rack cellular cross section just above the bottom casting.
These locations are the most heavily loaded net sections in the structure so that satisfaction of the stress factor criteria at these locations ensures that the overall structural criteria set forth in Section 6.6.1 are met.
The tables below develop stress factors for all of the SSE (Level D) simulations based on the associated SSE allowables. However, as stated above the intent is to evaluate the stresses developed from the SSE loadings with the allowables associated with OBE (Level B). Since the OBE allowables are '/a of the SSE allowables, this comparison may be conservatively performed by reducing the acceptable stress ratio to 0.5. This is very conservative, since the actual OBE loads which should be compared against the OBE allowable would be much lower than the SSE loads herein.
Holtec International 6-30 Holtec Report Hl-971760
6.9.1.1 The rack cell dimensionless stress factors for each of the simulations are as follows:
Pool Event Run COF Maximum R6 Rack Stress Factor Pool C Campaign I "SSE 0.8 0.494 SSE 0.2 0.289 SSE Random 0.384 Pool C Campaign II SSE 0.8 0.454 SSE 0.2 0.221 13 SSE Random 0.452 Pool C Campaign III SSE 0.8 0.409 SSE 0.2 0.266 SSE Random 0.432 24 Pool D Campaign I SSE 10 0.8 0.230 SSE 0.2 0.224 SSE 12 Random 0.230 Pool D Campaign II SSE 13 0.8 0.224 SSE 14 0.2 0.227 SSE 15 Random 0.232 The values for all other defined stress factors are also archived. As may be seen, all of the stress factors are well below 1.0. Therefore, the stresses developed during SSE conditions remain below the allowable SSE range and the rack modules are satisfactory to withstand the loadings. Note that stress factors for these SSE simulations are calculated based on SSE allowable strengths. However, since none of the stress factors exceed 0.5, the rack structures also adequately withstand the OBE conditions.
Holtec International 6-31 Holtec Report Hl-971760
6.9.2 The average shear stress in the thread engagement region is given below for the limiting pedestal in each simulation.
Pool Event Run COF Maximum Thread Rack Shear Stress (psi)
Pool C Campaign I SSE 0.8 4,682 SSE 0.2 4,382 SSE Random 4,607 Pool C Campaign II SSE 0.8 5,731 SSE 0.2 4,532 SSE Random 5,019 Pool C Campaign III SSE 0.8 4,232 SSE 0.2 4,120 SSE Random 4,570 26 Pool D Campaign I SSE 10 0.8 3,003 SSE 0.2 2,850 SSE 12 Random 2,901 Pool D Campaign II SSE 13 0.8 3,435 SSE 14 0.2 2,952 SSE 15 Random 3,307 The ultimate strength of the female part of the pedestal is 66,200 psi. The yield stress for the female pedestal material is 21,300 psi, as shown in Table 6.3.1. The male pedestal material has much greater strength and is therefore not a controlling factor in the design. The allowable shear stress for Level B conditions is 0.4 times the yield stress which gives 8,520 psi. The allowable shear stress for Level D conditions is the lesser of: 0.72 S= 15,336 psi or 0.42 S
= 27,804 psi . Therefore, the former criteria controls.
Holtec International 6-32 Holtec Report HI-971760
The largest thread shear stress computed by the post-processor is 5,731 psi. Since this value is below the allowable stresses for both OBE and DBE conditions, the thread shear stresses are within the acceptable range.
6.9.3 Impact loads at the pedestal base (discussed in subsection 6;8.2) produce stresses in the pedestal for which explicit stress limits are prescribed in the Code. The post-processor reports the stress factors in the pedestals which are developed, in part, from these impact stresses.
The reported pedestal stress factors are included in the discussion above in Section 6.9.1.1 along with the rack cell stress factors. However, the post-processor does not develop stress factors for the localized areas of the cellular and baseplate regions of the racks which experience fuel to cell wall, rack to rack, and rack to wall impact loads. These impact loads produce stresses which attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses.
Even though limits on secondary stresses are not prescribed in the Code for Class 3 NF structures, evaluations were made to ensure that the localized impacts do not lead to plastic deformations in the storage cells which affect the subcriticality of the stored fuel array.
Local cell wall integrity is conservatively estimated from peak impact loads. Plastic analysis is used to obtain the limiting impact load which would lead to gross permanent deformation. Table 6.9.1 indicates that the limiting impact load (of 3,238 lbf, including a safety factor of 2.0) is much greater than the highest calculated impact load value (of 605 lbf, see subsection 6.8.4.3) obtained from any of the rack analyses. Therefore, fuel impacts do not represent a significant concern with respect to fuel rack cell deformation.
Holtec International 6-33 > Holtec Report HI-971760
b.
As may be seen from subsection 6.8.4.1, the bottom (baseplate) of the storage racks will impact each other at a few locations during seismic events. Since the loading is presented edge-on to the 3/4" baseplate membrane, the distributed stresses after local deformation will be negligible. The impact loading will be distributed over a large area (a significant portion of the entire baseplate length of about 50.4 (minimum) inches by its 3/4 inch thickness). The resulting compressive stress from the highest impact load of 26,200 Ibs distributed over 37 sq. inches is only 708 psi, which is negligible.
Therefore, any deformation will not effect the configuration of the stored fuel.
Additional impacts will be experienced at the tops of some storage racks. These impacts will result in local yielding of the rack cell walls whenever the load exceeds 8,100 lbs. However, localized damage from all of these impacts occurs above the fuel active region. The fuel configuration and poison areas remain unaffected. Therefore, these impacts are acceptable.
6.9.4 Deeply submerged high density spent fuel storage racks arrayed in close proximity to each other in a free-standing configuration behave primarily as a nonlinear cantilevered structure when subjected to 3-D seismic excitations. In addition to the pulsations in the vertical load at each pedestal, lateral friction forces at the pedestal/bearing pad-liner interface, which help prevent or mitigate lateral sliding of the rack, also exert a time-varying moment in the baseplate region of the rack. The friction-induced lateral forces act simultaneously in x and y directions with the requirement that their vectorial sum does not exceed pV, where p is the limiting interface coefficient of friction and V is the concomitant vertical thrust on the liner (at the given time instant). As the vertical thrust at a pedestal location changes, so does the Holtec International 6-34 Holtec Report HI-971760
0' maximum friction force, F, that the interface can exert. In other words, the lateral force at the pedestal/liner interface, F, is given by P s p N (w) where N (vertical thrust) is the time-varying function of ~. F does not always equal pN; rather, pN is the maximum value it can attain at any time; the actual value, of course, is determined by the dynamic equilibrium of the rack structure.
In summary, the horizontal friction force at the pedestal/liner interface is a function of time; its magnitude and direction of action varies during the earthquake event.
The time-varying lateral (horizontal) and vertical forces on the extremities of the support pedestals produce stresses at the root of the pedestals in the manner of an end-loaded cantilever. The stress field in the cellular region of the rack is quite complex, with its maximum values located in the region closest to the pedestal. The maximum magnitude of the stresses depends on the severity of the pedestal end loads and on the geometry of the pedestal/rack baseplate region.
Alternating stresses in metals produce metal fatigue ifthe amplitude of the stress cycles is sufficiently large. In high density racks designed for sites with moderate to high postulated seismic action, the stress intensity amplitudes frequently reach values above the material endurance limit, leading to expenditure of the fatigue "usage" reserve in the material.
Because the locations of maximum stress (viz., the pedestal/rack baseplate junction) and the close placement of racks, a post-earthquake inspection of the high stressed regions in the racks is not feasible. Therefore, the racks must be engineered to withstand multiple earthquakes without reliance of nondestructive inspections for post-earthquake integrity assessment. The fatigue life evaluation of racks is an integral aspect of a sound design.
Holtec International 6-35 Holtec Report HI-971760
i The time-history method of analysis, deployed in this report, provides the means.to obtain a complete cycle history of the stress intensities in the highly stressed regions of the rack.
Having determined the amplitude of the stress intensity cycles and their number, the cumulative damage factor, V, can be determined using the classical Miner's rule where n, is the number of stress intensity cycles of amplitude a,, and N, is the permissible number of cycles corresponding to ot from the ASME fatigue curve for the material of construction. U must be less than or equal to 1.0.
To evaluate the cumulative damage factor, a finite element model of a portion of the spent fuel rack in the vicinity of a support pedestal is constructed in sufficient detail to provide an accurate assessment of stress intensities. Figure 6.9.1 shows the essentials of the finite element model. The finite element solutions for unit pedestal loads in three orthogonal directions are combined to establish the maximum value of stress intensity as a function of the three unit pedestal loads. Using the archived results of the spent fuel rack dynamic analyses (pedestal load histories versus time), enables a time-history of stress intensity to be established at the most limiting location. This permits establishing a set of alternating stress intensity ranges versus cycles for several seismic events. Following ASME Code guidelines for, computing U, it is found that U =0.464 due to the combined effect of 21 SSE events. This cumulative damage factor is below the ASME Code limit of 1.0 and therefore, fatigue failure I
is not expected. Selection of 21 SSE events represents a conservative evaluation compared to other previous fatigue assessments which were based on the damage resulting from 10 SSE events, as discussed in the Harris FSAR.
Holtec International 6-36 Holtec Report HI-971760
6.9.6 Weld locations subjected to significant seismic loading are at the bottom of the rack at the baseplate-to-cell connection, at the top of the pedestal support at the baseplate connection, and at cell-to-cell connections. Bounding values of resultant loads are used to qualify the connections.
Reference [6.6.1] (ASME Code Section III, Subsection NF) permits, for Level A or B conditions, an allowable weld stress v = .3 S= 19860 psi. As stated in subsection 3.4.2 the allowable may be increased for Level D by the ratio (15336/8520) = 1.8, giving an allowable of 35,748 psi.
Weld dimensionless stress factors are produced through the use of a simple conversion (ratio) factor applied to the corresponding stress factor in the adjacent rack material. A 2.15 factor for PWR racks is based on the differences in material thickness and length versus weld throat dimension and length:
' 2.15165 Ratio 0.0625 4 0.7071 0 7 Similarly, a 1.49 factor for BWR racks is developed as follows:
Ra 1 48736 0.0625 + 0.7071 + 7 The highest predicted weld stress for DBE is calculated from the highest R6 value (see subsection 6.9.1.1) as follows:
Holtec International 6-37 Holtec Report HI-971760
RG ~ [(0.6) F] + Ratio =
0.494 [(0.6) 21,300]~2.144 = 13,574 psi this value is less than the OBE allowable weld stress value, which is 19,860.
Therefore, all weld stresses between the baseplate and cell wall base are acceptable.
The weld between the baseplate and support pedestal are evaluated by development of a finite element model of the bearing pad/base plate interface and appropriate application of the maximum pedestal loads. The maximum weld stress was determined to be 10,194 psi, which is much less than the OBE allowable weld stress value of 19,860 psi.
The results are also shown in Table 6.9.1.
Cell-to-cell connections are made using a series of connecting welds along the cell height. Stresses in storage cell to cell welds develop due to fuel assembly impacts with the cell wall. These weld stresses are conservatively calculated by assuming that fuel assemblies in adjacent cells are moving out of phase with one another so that impact loads in two adjacent cells are in opposite directions; this tends to separate the two cells from each other at the weld.
Table 6.9.1 gives results for the maximum allowable load that can be transferred by these welds based on the available weld area. An upper bound on the load required to be transferred is also given in Table 6.9.1 and is much lower than the allowable load.
This upper bound value is very conservatively obtained by applying the bounding rack-to-fuel impact load from any simulation in two orthogonal directions simultaneously, Holtec International 6-38 Holtec Report Hl-971760
and multiplying the result by 2 to account for the simultaneous impact of two assemblies. An equilibrium analysis at the connection then yields the upper bound load to be transferred. It is seen from the results in Table 6.9.1 that the calculated load is well below'he allowable.
6.9.6 To protect the pool slab from high localized dynamic loadings, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact on these bearing pads during a seismic event and pedestal loading is transferred to the liner. Bearing pad dimensions are set to ensure that the average pressure on the slab surface due to a static load plus a dynamic impact load does not exceed the American Concrete Institute, ACI-349 [6.9. 1] limit on bearing pressures. Section 10.17 of [6.9.2] gives the design bearing strength as fb = $ (.85 f,') e E
where $ = .7 and f,'s the specified concrete strength for the spent fuel pool. e = 1, except when the supporting surface is wider on all sides than the loaded area. In that case, e =
(A,/At)', but not more than 2. A, is the actual loaded area, and A~ is an area greater than A, and is defined in [6.9.2]. Using a value of e ) 1 includes credit for the confining effect of the surrounding concrete. It is noted that this criteria is in conformance with the ultimate strength primary design methodology of the American Concrete Institute in use since 1971. For Harris, the concrete compressive strength is f,' 4,000 psi. The allowable bearing pressure is conservatively computed by taking m=1 to account for lack of total concrete confinement in the leak chase region and a stress reduction factor of /=0.7. Thus, the maximum allowable concrete bearing pressure is 2,380 psi.
The maximum vertical pedestal load is 153,000 lbs (SSE event). The bearing pad selected is 1.5" thick, austenitic stainless steel plate stock. The average pressure at the pad to liner Holtec International 6-39 Holtec Report Hl-971760
interface is computed and compared against the above-mentioned limit. Calculations show that the average pressure at the slab/liner interface is 2,168 psi which is below the allowable value of 2,380 psi, providing a factor of safety of 1 ~ 1.
Therefore, the bearing pad design devised for the Harris pools C and D is deemed appropriate for the prescribed loadings.
6.9.7 The Level A condition is not a governing condition for spent fuel racks since the general level of loading is far less than Level B loading. To illustrate this, the heaviest (fully loaded) spent fuel rack (which is an 11X9 PWR rack) is considered under the dead weight load. It is shown below that the maximum pedestal load is low and that further stress evaluations are unnecessary.
LEVEL A MAXIMUMPEDESTAL LOAD Dry Weight of Largest PWR Holtec Rack 15,700 lbf t Dry Weight of 99 PWR Fuel Assemblies 158,400 lbf Total Dry Weight 174 100 lbf tt Total Buoyant Weight (0.87 x Total Dry Weight) = 151,467 lbf Load per Pedestal 37,867 lbf The stress allowables for the normal condition is the same as for the upset condition. An upset condition pedestal load may be conservatively (bounded on the low side) determined for the Conservative weight corresponding to the heaviest rack, which is a BWR storage rack. The heaviest PWR rack nominal weight is 15,620 lb.
This weight exceeds the weight of the heaviest fully loaded BWR rack, which is
[15,700 lb + (13x13) x 680 lb] = 130,620 lb.
Holtec International 640 Holtec Report HI-971760
purpose of comparing with the load above by dividing the DBE pedestal load by a factor of 2.0. This would result in an OBE pedestal load of 153,000-:2=76,500, which is still much greater than the calculated Level A load. Since this load (and the corresponding stress throughout the rack members) is much greater than the 37,867 lb load calculated above, the Upset (OBE) condition controls over normal (Gravity) condition. Therefore, no further evaluation is necessary for Level A.
6.10 The maximum hydrodynamic pressures (in psi) that develop between the fuel racks and the spent fuel pool walls will occur at those conditions and locations of greatest relative displacements. The greatest displacement was shown in Section 6.8.1 to be 1.494 inches, which occurs in rack 28 under simulation number 4. The maximum hydrodynamic pressure during this simulation was determined to be 19 psi. This hydrodynamic pressure was considered in the evaluation of the Fuel Handling Building and Pool structure.
Holtec International 641 Holtec Report HI-971760
Time history simulations, including all non-linear impact and interface friction effects, have been applied to evaluate the structural margins in the Holtec spent fuel racks.
~ The totality of simulations provide an extensive set of results for loads, stresses, and displacements, which taken together, demonstrate that the spent fuel racks meet the input specification and the governing Code requirements.
Evaluation of structural margins have been performed for the array of racks in each pool with all racks loaded with fuel. The requirements of the specification and the governing Code documents are met for Level A, Level B, and Level D conditions.
Based on all results presented in tabular form above the spent fuel racks are demonstrated to be acceptable for the service intended.
E Holtec International 6-42 Holtec Report HI-971760
0
[61 1] USNRC NUREG-0800, Standard Review Plan, June 1987.
[6.1.2] (USNRC Office of Technology) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, and January 18, 1979 amendment thereto.
[6.2.1] Soler, A.I. and Singh, K.P., "Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984).
[6.2.2] Soler, A.I. and Singh, K.P., "Some Results from Simultaneous Seismic Simulations of All Racks in a Fuel Pool", INNM Spent Fuel Management Seminar X, January, 1993.
[6.2.3] Singh, K.P. and Soler, A.I., "Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (March 1991) ~
[6.2.4] Holtec Proprietary Report HI-961465 - WPMR Analysis User Manual for Pre&Post Processors & Solver, August, 1997.
[6.4.1] USNRC Standard Review Plan, NUREG-0800 (Section 3.7.1, Rev. 2, 1989).
[6.4.2] Holtec Proprietary Report HI-89364 - Verification and User's Manual for Computer Code GENEQ, January, 1990.
Rabinowicz, E., "Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company, 1976.
[6.5.2] Singh, K.P. and Soler, A.I., "Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium: The Case of Fuel Racks," 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.
[6.5.3] Fritz, R.J., "The Effects of Liquids on the Dynamic Motions of il Immersed Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172.
Holtec International 6P3 Holtec Report Hl-971760
[6.5.4] Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering,"
McGraw Hill, 1976.
[6.5.5] Paul, B., "Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment", (Proprietary), NUSCO/Holtec Report HI-88243.
[6.6.1] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF, 1995 Edition.
[6.6.2] ASME Boiler & Pressure Vessel Code,Section II, Part D, 1995 Edition.
[6.6.3] USNRC Standard Review Plan, NUREG-0800 (Section 3.8.4, Rev. 2, 1989).
[6.9.1] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1985.
[6.9.2] ACI 318-95, Building Code requirements for Structural Concrete,"
American Concrete Institute, Detroit, Michigan, 1995.
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. Table 6.2.1 1
PARTIA'L'-;L'ISTING OF.'FUEL RACK APPL'ICATIONS USING DYNARACK PLANT DOCKET NUMBER(s)
Enrico Fermi Unit 2 USNRC 50-341 1980 Quad Cities 1 &2 USNRC 50-254, 50-265 1981 Rancho Seco USNRC 50-312 1982 Grand Gulf Unit 1 USNRC 50-416 1984 0 ster Creek USNRC 50-219 1984 Pil rim USNRC 50-293 1985 V.C. Summer USNRC 50-395 1984 Diablo Can on Units 1 &2 USNRC 50-275, 50-323 1986 B ronUnits1 &2 USNRC 50-454, 50-455 1987 Braidwood Units 1 &2 USNRC 50-456, 50-457 1987 Vo tie Unit 2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 Millstone Point Unit 1 USNRC 50-245 1989 Chinshan Taiwan Power 1988 D.C. Cook Units 1 &2 USNRC 50-315, 50-316 1992 Indian Point Unit 2 USNRC 50-247 1990 Three Mile Island Unit 1 USNRC 50-289 1991 James A. FitzPatrick USNRC 50-333 1990 Shearon Harris USNRC 50-400 1991 Ho e Creek USNRC 50-354 1990 Holtec International 6<5 Holtec Report HI-971760
Table 6.2.1 PARTIAL LISTING OF FUEL RACK APPLICATIONS'USING DYNARACK PLANT DOCKET NUMBER(s)
Kuoshen Units 1 &2 Taiwan Power Com an 1990 Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units 1 &2 Comision Federal de 1991 Electricidad Zion Station Units 1 &2 USNRC 50-295, 50-304 1992 Se uo ah USNRC 50-327, 50-328 1992 LaSalle Unit 1 USNRC 50-373 1992 Duane Arnold Ener Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit 1 USNRC 50-220 1993 Beaver Valle Unit 1 USNRC 50-334 1992 Salem Units 1 &2 USNRC 50-272, 50-311 1993 Limerick USNRC 50-352, 50-353 1994 Ulchin Unit 1 KINS 1995 Yon wan Units 1 &2 KINS 1996 Kori-4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 An raUnit1 Brazil 1996 Sizewell B United Kin dom 1996 Holtec International 646 Holtec Rcport Hl-971760
Table 6.3.1 RACK MATERIALDATA (200'F)
(ASME - Section II, Part D)
Young's Modulus Strength Ultimate Strength Material 10'ield Sr (psi)
SU (psi)
( si)
SA240; 304L S.S. 27.6 x 21,300 66,200 SUPPORT MATERIALDATA (200'F)
SA240, Type 304L (upper 27.6 x 10~ 66,200 part of support feet) 10'1,300 SA-564-630 (lower part of 28.5 x 106,300 140,000 support feet; age hardened at 1100'F)
Holtec International 6-47 Holtec Report Hl-971760
Table 6.4.1 TIME-HISTORY STATISTICALCORRELATION RESULTS OBE Datal to Data2 0.0295 Datal to Data3 0.0392 Data2 to Data3 0.0169 DBE Datal to Data2 0.0183 Datal to Data3 0.0588 Data2 to Data3 0.0299 Holtec International 648 Holtec Report HI-971760
Table 6.5.1 Degrees-of-freedom U e e e, P1 Pz PB q4 Pr PB PB q1O qia Node 1 is assumed to be attached to the rack at the bottom most point.
2'is Node 2 is assumed to be attached to the rack at the top most point.
Refer to Fi ure 6.5.1 for node identification.
3'15 P14 P1B 4 Pa P1B 5 Pie Pro Pn where the relative displacement variables q, are defined as:
pi = qi(t) + Ux(t) i = 1,7,13,15,17,19,21 q(t) + Uy(t) i 2 8 14 16 18 20 22 q,(t)+ U,(t) i = 3,9 q,(t) i = 4,5,6,10,11,12 p, denotes absolute displacement (or rotation) with respect to inertial space q, denotes relative displacement (or rotation) with respect to the floor slab
- denotes fuel mass nodes U(t) are the three known earthquake displacements Holtec International 6-49 Holtec Report HI-971760
Table 6.5.2 (MR216) NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS IN THE POOL D, CAMPAIGN I MODEL h
I. Nonlinear Springs (Type 3 Gap Elements - 212 Total)
Rack Number No. Description Support S1, Z compression-only element Support S2, Z compression-only element Support S3, Z compression-only element Support S4, Z compression-only element X rack/fuel assembly impact element between nodes 2 and 2' rack/fuel assembly impact element between nodes 2 and 2' rack/fuel assembly impact element between nodes 2 and 2' rack/fuel assembly impact element between nodes 2 and 9-24 elements corresponding to the remaining rattling fuel masses 2'mpact at nodes 1', 3', 4'nd 5'similar to elements 5 thru 8) 25-28 Z compression-only elements for Supports S1,S2, S3, and S4 29-48 Impact elements corresponding to the rattling masses at nodes 1', 2',
3', 4'nd 49-144 3-6 elements similar to those above 5'mpact 145-212 varies rack/rack and rack/wall elements at top and bottom of racks (see Figures 6.5.6 and 6.5.7)
Holtec International 6-50 Holtec Report HI-971760
Table 6.5.2 (MR216) NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS IN THE POOL D, CAMPAIGN I MODEL II. Linear Springs (Type 1 Elements - 36 Total)
Rack Number No. Description Rack beam bending element (x-z plane)
Rack shear deformation element (x-z plane)
Rack beam bending element (y-z plane)
Rack shear deformation element (y-z plane)
Rack beam axial deformation element Rack beam torsional deformation element 7-12 Similar to elements 1 thru 6 13-36 3-6 Similar to elements 1 thru 6 III. Piece-wise Linear Friction Springs (Type 2 Elements - 48 Total)
Rack Number No. Description Pedestal 1, X direction Pedestal 1, Y direction 1 ~
Pedestal 2, X direction Pedestal 2, Y direction 5 Pedestal 3, X direction Pedestal 3, Y direction Pedestal 4, X direction Pedestal 4, Y direction 9-16 Similar to elements 1 thru 8 17-48 3-6 Similar to elements 1 thru 8 Holtec International 6-51 Holtec Report Hl-971760
Table 6.9.1 COMPARISON OF BOUNDING CALCULATED LOADS/STRESSES VS.
CODE ALLOWABLES AT IMPACT AND WELD LOCATIONS Item/Location DBE Calculated OBE AHowable Fuel assembly/cell wall impact, lbf. 605
- 3,238 **
Rack/baseplate weld, psi 13,574 19,860 Female pedestal/baseplate weld, psi 10,194 19,860 Cell/cell welds, lbf. 1 711 gpss 3,195 See Section 6.8.4.3.
Based on the limit load for a cell wall. The allowable load on the fuel assembly itself may be less than this value but is greater than 605 lbs.
- Based on the fuel assembly to cell wall, impact load simultaneously applied in two orthogonal directions.
Holtec International 6-52 Holtec Report HI-971760
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HI-971760 Figure 6.42
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Figure 6.4.3 HI-971760
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HI-971760 Figure 6.4.6
ql2 IMPACT RATTLING SPRINGS FVEL MASS (TYP. )
TOP OF RACK P8 qll P7 PI3 10 RACK GEOMETRIC CENTER LINE PI5 BEAM SPRING (6 TOTAL)
IB PI 7 q6 PEOESTAL 3 PEDESTAL 4 P20 PI9 PZ q5
/
PEOESTAL l /
PI P2 PEOESTAL Z P I X
~ IMPACT SPRINGS FOVNOA TION SPRINGS FIGURE 6.5.1; SCHEMATIC OF THE DYNAMIC REPoRT u Hr-97i760 NODEL FOR DYNARACK
CELL WALL AT LEVEL 1 FUEL MASS FUEL ASSEMBLY/CELL Ye IMPACT SPRING
/
/ /
/ /
/
//
Xg FIGURE 6.5.2; FUEL-TO-RACK IMPACT SPRINGS AT LEVEL OF RATTLING MASS 6+2
FUEL ASSY./CELL IMPACT SPRING, Ki TYPICAL FUEL RATTLING MASS RACK C.G.
FRICTION INTERFACE SPRING, Kf SUPPORT LEG SPRING. Ks FIGURE 6.5.3; TWO DIMENSIONAL VIE% OF THE SPRING-MASS SIMULATION 6%3
H/2 bio
, Ksv KBx
'I4 H/2 RACK DEGREES-OF-FREEDOM FOR Y-Z PLANE BENDING WITH SHEAR AND BENDING SPRING H/2 Ksx KBv H/2 FIGURE 6 5 4.- RACK DEGREES OF FREEDOM FOR X-Z PLANE BENDING WITH SHEAR AND BENDING SPRING REPORT ¹ HI-971760
RACK P EL TYPICAL TOP IMPACT ELEMENT RACK STRUCTURE TYPICAL BOTTOM IMPACT ELEMENT RACK 8 EPLATE VEL FIGURE 6 5 5; RACK-TO-RACK IMPACT SPRINGS EPORT 0 HI-971760 6-65
44 49 Os Qs Q7 Qs 5I3 5?I 5ll 81 10 11 W
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1S C2 IG 02 MQIjEL OI3 Qi4 17 C1 18 D1 CIjIIRIllNATE AXES HARRIS SPENT FUEL POOL C FIGURE 6.5.6; RACK IMPACT SPRING NUMBERING SCIIEME ( BOTTOM )
CAMPAIGN I I -971760 6-66
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47I 5I2 05 Qo 07 QB 4Q 5OI 5I7 O'0 Oll 5%
l5 C2 l6 02 Rl MODEL 13 l4 l7 Cl l8 0l COORDINATE AXES AARRIS SPENT FUEL POOL C FIGURE 6 5.7; RACK IMPACT SPRING NUMBERING SCAEME ( TOPI CAMPAIGN I I -9't1760 6%7
7II 7I5 Qi 05 04 743 79 05 Q6 Qs n4 m
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$5 OI4 OI5 OI6 l8 l9 85 9I7 Ozi 22 23 24 9IB Rl 28 IQ inn MIIOEL $4 29 Cl 30 Dl CEKINATE MES HARRIS SPENT FUEL POOL C FIGURE 8.5.8; RACK IMPACT SPRING NUMBERING SCIIEME ( BOTTOM)
CAMPAIGNS If ANB III I -971760
779 748 KI K4 06 06 0~ Qs 78l ell Qio Qll Olg ea 8IZ 8la 8IZ l3 l4 Qn Olo 17 I8 Olg lS 69 9I9 Ogi O. Ogg 9II 9Q 8I9 27 C2 28 02 NQQEL Qgg QZ 29 Cl 30 Dt CQQRQINATE 9V 99 Se 90 99I IXI AXES HARRIS SPENT FUEL POOL C FIGURE 6.5.9; RACK IMPACT SPRING NUMBERING SCHEME ( TOP )
CAMPAIGNS II AND III I -971760 6-69
~ NORTH 145 146 161 162 149 156 168 QI Aj l50 155 ,l07 153 154 165 166 173 180 188 Q~ Q4 A2 A4 187 177 1I8 185 186 193 200 208 Q~ Qb Bl l99 207 HOOEL 197 198 205 206 COOROINATE AXES HARRIS SPENT FUEL POOL D FIGURE 6.5.10; RACK IMPACT SPRING NUMBERING SCHEME ( BOTTOM )
CAMPAIGN I Hl 971760 6-70
147 148 163 164 151 160 172 Ql Al 152 171 157 158 169 170 175 192 Q~ Q4 A2 A4 183 181 182 189 190
'12 195 Q> Qb l06 Z03 Zll MOOEi 201 202 209 210 LXKIPAtE AXH HARRIS SPENT FURL POOL D FIGURE 6.5.11; RACK IMPACT SPRING NUMBERING SCHEME ( TOP)
CAMPAIGN I Hl 971760
289 290 305 306 317 318 329 330 293 300 312 324 336 Qi O~ 04 Al 294 299 3II 323 333 297 298 309 310 321 322 333 334 356 372 O~ Qo Ot OB A2 L2 355 363 37I 345 346 353 354 361 362 369 377 392 408 Qs Qio (y Bi B2 378 3 391 399 407 HOOEL 381 382 389 390 397 398 405 406 COORDINATE AXH HARRIS SPENT FUEL POOL D FIGURE 6.5.IP.; RACK IMPACT SPRING NUIERING SCHEME ( BOTTOM)
CAMPAIGN II Hl 971760 6-72
291 292 307 308 319 320 331 332 295 316 328 Qi Q4 296 Al 3O3 A'n 327 A5 339 301 302 313 314 325 326 337 338 352 360 368 376 Q5 Qs Q7 A4 L2 344 35I 3)g 367 375 349 350 357 358 365 366 373 374 379 388 396 412 Qq QiO Q<l QI2 3BO 387 395 403 4II 3OO<L 385 386 393 394 401 402 409 410 COOROINATE AXH HARRIS SPENT FUEL POOL D FIGURE 6.5.13; RACK IIIPACT SPRING NUMBERING SCHEME ( TOP )
CAMPAIGN II H1 971760
Harr t.s Poo I C Run Ver t t.ca I Pedes ta SI T t. me H t. s to r y Load t. n g Rack 5, Pedes ta I 2 158888 N
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- ch Iarfgu( imalysfs - i>>tR Bach Figure 6.9.1; Rack Fatigue An.alysis Model 6-75
vo 7.1 The USNRC OT position paper [7.1] specifies that the design of the rack must ensure the functional integrity of the spent fuel racks under all credible drop events in the spent fuel pool. This section contains synopses of the analyses carried out to demonstrate the regulatory compliance of the proposed racks under postulated fuel assembly drop scenarios germane to HNP pools C and D.
In addition to the postulated fuel assembly free-fall scenarios, a gate drop accident event was also considered. In this case, the ability of the pool structure to avert primary structural damage (leading to rapid loss of water) needs to be demonstrated.
vz Two categories of fuel assembly accidental drop events are considered. In the so-called "shallow drop" event, a fuel assembly, along with the portion of handling tool which is severable in the case of a single element failure, is assumed to drop vertically and hit the top of the rack. The "depth" of damage to the affected cell walls must be demonstrated to remain limited to the portion of the cell above the top of the "active fuel region", which is essentially the elevation of the top of the Boral neutron absorber. To meet this criterion, the plastic deformation of the rack cell wall should not extend more than 21.3 inches (downwards) from the top of a PWR rack. The distance separating the top of the rack from the Boral in the BWR racks in pools C and D is 13.75 inches. Therefore, to be conservative the smaller BWR dimension of 13.75 inches is selected as the maximum depth of damage of an object falling onto the tops of storage racks.
Holtec International 7-1 Holtec Report HI-971760
By observation, the drop of a PWR assembly onto a PWR rack is more limiting than any other combination of the two fuel types (PWR vs. BWR) with the two rack (PWR vs.
BWR) types. This is obvious because of two reasons. The PWR assembly drop is a more severe case than the BWR assembly case, since the effect of the weight differences (approximately 1600 vs. 680 lbs, respectively) far exceeds the effect of the differences in the impact cross-section zone (about 8.4 vs. 5.5 inches, respectively). The PWR storage rack cell controls as an impact zone over the BWR cell because it is larger (8.4 vs. 6.06 inches, respectively) resulting in less capacity to withstand top of cell or baseplate impacts.
(The nominal cell wall thicknesses of the two rack types is identical).
In order to utilize an upper bound of kinetic energy at impact, the impactor is assumed to weigh 2,100 lbs and the free-fall height is assumed to be 36 inches. The impactor weight corresponds to the heaviest fuel (plus handling tool) which will be handled in pools C and It is readily apparent from the design of the rack modules described in Section 3, that the impact resistance of a rack at its periphery is less than its interior. Accordingly, the potential shallow drop scenario is postulated to occur at the periphery in the manner shown in Figure 7.2.1.
Finally, the fuel assembly assemblage is assumed to hit the rack in a manner to inflict maximum damage. The impact zone is chosen to minimize the cross sectional area which experiences the deformation. Placement of the impact at the 'corner would reduce the impact zone area, but actually increases the cross-sectional area experiencing deform'ation.
Impact at the corner would involve the crushing of two cell walls under the dynamic impact. Therefore, impact on only one cell wall is chosen to simulate the worst case accident. Figure 7.2.2 depicts the impacted rack in plan view.
Holtec International 7-2 Holtec Report HI-971760
The second class of "fuel drop event" postulates that the impactor falls through an empty storage cell impacting the rack baseplate. This so-called "deep drop" scenario threatens the structural integrity of the "baseplate". Ifthe baseplate is pierced, then the fuel assembly might damage the pool liner (which at 3/16" is rather thin) and create an abnormal condition of the enriched zone of fuel assembly outside the "poisoned" space of the fuel rack. To preclude damage to the pool liner, and to avoid the potential of an abnormal fuel storage configuration in the aftermath of a deep drop event, it is required that the baseplate remain unpierced and that the maximum lowering of the fuel assembly support surface is less than the distance from the bottom of the baseplate to the liner.
The deep drop event can be classified into two scenarios, namely, drop through cell located above a support leg (Figure 7.2.3), and drop in an interior cell away from the support pedestal (Figure 7.2.4).
In the former deep drop scenario (Figure 7.2.3), the baseplate is buttressed by the support pedestal and presents a hardened impact surface, resulting in a high impact load. The principal design objective is to ensure that the support pedestal does not pierce the lined, reinforced concrete pool slab.
The baseplate is not quite as stiff at cell locations away from the support pedestal (Figure Il 7.2.4). Baseplate severing and large deflection of the baseplate (such that the liner would be impacted) would constitute an unacceptable result.
7.3 The drop of a rack above spent fuel stored within in-place rack modules is precluded, since racks will not be lifted above spent fuel. The drop of a rack module during installation is also extremely remote, due to the defense-in-depth approach discussed in Sections 3.5 and Holtec International 7-3 Holtec Report HI-971760
11.1. Despite the unlikelihood of this possibility, a rack dropping to the pool floor has been considered. To evaluate the consequences of an accidental, uncontrolled lowering of the heaviest rack module, a 13x13 BWR module conservatively considered with a submerged weight of 16140 lb (actual maximum nominal dry weight is only 15700 lb),
from a height of 480 inches above the pool liner is considered (Figure 7.3.1). The objective of the analysis is to ensure that a rapid loss of pool water will not occur, leading to loss of shielding to the stored nuclear fuel.
7.4 In the first step of the solution process, the velocity of the dropped object (impactor) is computed for the condition of underwater free fall. Table 7,1 contains the results for the three drop events.
In the second step of the solution, an elasto-plastic finite element model of, the impacted region on Holtec's computer Code PLASTIPACT (Los Alamos Laboratory's DYNA3D implemented on Holtec's QA system) is prepared. PLASTIPACT simulates the transient collision event with full consideration of plastic, large deformation, wave propagation, and elastic/plastic buckling modes. For conservatism, the impactor in all cases is assumed to be rigid. The physical properties of material types undergoing deformation in the postulated impact events are summarized in Table 7.2.
Holtec International Holtec Report HI-971760
0 7.5.1 Figure 7.5.1 shows the finite element model utilized in the shallow drop impact analysis.
Dynamic analyses show that the top of the impacted region undergoes severe localized deformation. Figure 7.5.2 shows an isometric view of the post-impact geometry of the rack for the shallow drop scenario. The maximum depth of plastic deformation is limited to 11 inches, which is below the design limit of 13.75 inches. Figure 7.5.3 shows the plan view of the post-collision geometry. Approximately 10% of the cell opening in the impacted cell is blocked.
7.5.2 The deep drop scenario depicted in Figure 7.5.4(b), wherein the impact region is located above the support pedestal, is found to produce a negligible deformation on the baseplate.
The vertical force in the support pedestal remains below the loads generated during seismic events (see Section 6). Therefore, it is concluded that the pool liner will not be damaged.
The deep drop condition through an interior cell depicted in Figure 7.5.4(a) does produce some deformation of the baseplate and localized severing of the baseplate/cell wall welds (Figure 7.5.5). However, the fuel assembly support surface is lowered by a maximum of 2.89 inches, which is less than the minimum distance of 6 inches from the bottom of the baseplate to the liner. Therefore, the deformed baseplate will not strike the liner during this drop event and the pool liner will not be damaged. As stated in Subsection 4.7.2, criticality evaluations performed for this baseplate deformation have shown that the storage configuration remains acceptable.
Holtec International 7-5 Holtec Report HI-971760
v~
Since the primary structural integrity of the slab is unimpaired subsequent to a rack drop to the pool floor liner, catastophic loss of pool water would not occur. Therefore, catastrophic failure of the pool structure or rapid loss of pool water will not occur.
No other credible in-pool drops have been identified. An object potentially carried over the pools is one of the 4,000 pound gates which isolate the pools. These gates are long rectangular metallic structures with a base area of 8 inches by 41 inches. During handling the gate is lifted using a single failure proof crane and double rigging. The rigging complies with the safety margin requirements of NUREG-0612. An accidental drop of the gate is not a credible event, because of the above mentioned defense-in-depth approach to the lifting of this heavy load. Additionally the gates are not located within the pools, but are installed inside of slots within adjacent transfer canals. Nevertheless, analyses were carried out for this accident scenario. A gate drop during handling from 40 feet above the pool liner was evaluated and it has been determined that a primary failure of the water retaining concrete structure will not occur. A gate drop during handling from 15 inches above the top of a PWR rack loaded with fuel was also evaluated. A schematic of the 3D finite element model is depicted in Figures 7.6.1 and 7.6.2. The gate is conservatively considered to strike only three rack storage cell walls, as shown in Figure 7.6.3. This impact zone is conservative, since the dimensions of the gate would span at least four cell walls. The gate is shown to penetrate the rack to a depth of less than 5 inches, as shown in Figures 7.6.4 and 7.6.5. Since this penetration remains above the tops of the stored fuel assemblies, no fuel damage occurs.
Holtec International 7-6 Holtec Report HI-971760
7.7 CIOSIIrC The fuel assembly and gate drop accident events evaluated for the HNP fuel pools were analyzed and found to produce localized damage well within the design limits for the racks. A construction accident event wherein the heaviest rack falls from a 40 foot height onto the pool floor was also considered. Analyses show that the pool structure will not suffer any primary structural damage. A similar conclusion is reached with regard to a gate drop event.
Holtec International 7-7 Holtec Report HI-971760
7.8 Rcfco:aces
[7. 1] "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978.
Holtec International 7-8 Holtec Report Hl-971760
TABLE 7.1 IMPACT EVENT DATA Impactor Drop Impact Weight Impactor Height Velocity Case (lbs) (inches) (inch/sec)
- 1. Shallow drop event 2,100 Fuel Assembly 36 152
- 2. Deep drop event 2,100 Fuel Assembly 205 353
- 3. Construction event 16,140 Rack Module 480 304 Holtec International 7-9 Holtec Report HI-971760
TABLE 7.2 MATERIALDEFINITION Elastic Stress Strain Density Modulus (psi) First Yield Failure Elastic Failure Material (psi) (psi)
Name Type Stainless SA240-304L 490 2.760e+07 2.130e+04 6.620e+04 7.717e-04 3.800e-01 steel Stainless SA240-304 490 2.760e+07 2.500e+04 7.100e+04 7.717e-04 3.800e-01 steel Stainless SA564-630 490 2.760e+07 1.063e+05 1.400e+05 3.851e-02 3.800e-01 steel Concrete 4000 psi 150 3.605e+06 4.000e+03 2.022e+04 1.110e-03 5.500e-02 Holtec International 7-10 Holtec Report HI-971760
FUEL ASSEMBLY 36)l IMPACT REGION
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Figure 7.53; Shallow Drop: Finite Element Model Detail Impacted Region H1-971760
HARRIS - SPENT FUEL RACK UPPER EX STEP 19 TIME = 6.649951GE-002 I
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HARRIS - SPENT FUEL RAC-STEP I9 TIME = 6.6499516E-002 Y
A Z X MAX. DISPL .64042E+00 AT NODE 3500 SCALE FACTOR = 1.0000E+00 Figure 75.3; Shallow Drop: Maximum Cell Deformation Impacted Region Plan HI-971760 7-18
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Figure 755; Maximum Baseplate Deformation from Deep Drop Scenario 7-20 Hl-971760
0 Figure 7.6.1; Gate Drop Finite-Element Model 7-21 Hl-971760
Figure 7.6.2; Gate Drop Finite-Element Model, Detail of Impacted Region 7-22 HI-97176D
8'4 Figure 7.6.3; Gate Drop Finite-Element Model Detail of impacted Region (Plan) 7-23 HI-971760
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8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Intmduction The Harris Spent Fuel Pools (SFPs) C and D are safety related, seismic category I, reinforced concrete structures. Spent fuel is to be placed within storage racks located in both of these areas and they willbe collectively referred to herein as the fuel pool structure. This section describes the analysis to demonstrate structural adequacy of the pool structure, as required by Section IV of the USNRC OT Position Paper [8.1.1].
The pool regions are analyzed using the finite element method. Results for individual load components are combined using factored load combinations mandated by SRP 3.8.4 [8.1.2]
based on the "ultimate strength" design method of the American Concrete Institute (ACI 318)
[8.1.3]. It is demonstrated that for the critical bounding factored load combinations, structural integrity is maintained when the pools are assumed to be fully loaded with spent fuel racks, as shown in Figures 1.2 and 1.3 with all storage locations occupied by fuel assemblies.
The regions examined in the SFPs are the floor slabs, and the highly loaded wall sections adjoining the slabs. Both moment and shear capabilities are checked for concrete structural integrity. Local punching and bearing integrity of the slab in the vicinity of a rack module support pedestal pad is evaluated. All structural capacity calculations are made using design formulas meeting the requirements of ACI 318.
8.2 The SFPs are located inside the Fuel Handling Building and are supported by a two way, reinforced'concrete slab. The minimum thickness of the slab is 12.0 feet, including grout. The SFPs are separated by reinforced concrete walls and transfer canals.
Figure 1.1 shows the layout of the majority of the Fuel Handling Building. A plan of the building area of concern is shown in Figure 8.2.1, which shows the major structural dimensions of the Holtec International 8-1 Report HI-971760
pools. The floor liner plate of the SFPs are located at elevation 246.0 The spent fuel area operating floor is at elevation 286.0.
8.3 Pool structural loading involves the following discrete components:
8.3.1
- 1) Dead weight of pool structure includes the weight of the Fuel Handling Building concrete upper structure.
- 2) Maximum dead weight of rack modules and fuel assemblies in the fully implemented storage configuration, as shown in Figures 1.2 and 1.3.
- 3) Dead weight of a shipping cask including yoke of 250 kips.
- 4) The Cask Crane, Auxiliary Crane and Spent Fuel Handling Machine (Refueling Platform) are designed to move along the N-S direction. The dead weight and the rated liftweight of these cranes are considered as live load.
- 5) The hydrostatic water pressure.
8.3.2
- 1) Vertical loads transmitted by the rack support pedestals to the slab during a SSE or OBE seismic event.
Holtec International 8-2 Report HI-971760
- 2) Hydrodynamic inertia loads due to the contained water mass and sloshing loads (considered in accordance with TID-7024 [8.3.1]) which arise during a seismic event.
- 3) Hydrodynamic pressures between racks and pool walls caused by rack motion in the pool during a seismic event.
- 4) Seismic inertia force of the walls and slab.
8.3.3 Thermal loading is defined by the temperature existing at the faces of the pool concrete walls and slabs. Two thermal loading conditions are evaluated: The normal operating temperature and the accident temperature.
8.4 8.4.1 The finite element model encompasses the two SFPs, the Fuel Transfer Canal, the Cask Loading Pool, and adjacent transfer canals and building structure. The interaction with the rest of the Fuel Handling Building reinforced concrete, which is not included in the finite-element model, is simulated by imposing appropriate boundary conditions. The structural area of interest for the reracking project includes only two pools which are involved in the fuel storage capacity increase. However, by augmenting the area of interest, by considering in the constructed finite-element model and numerical investigation the additional areas described above, the perturbation induced by the boundary conditions on the stress field distribution for the area of interest is minimized. A finite element 3D view of the structural elements considered in the numerical investigation is shown in Figure 8.4.1.
Holtec International Report HI-971760
The preprocessing capabilities of the STARDYNE computer code [8.4. 1] are used to develop the 3-D finite-element model. The STARDYNE finite-element model contains 13,353 nodes, 3,564 solid type finite-elements, 7,991 plate type finite-elements and 24 hydro-dynamic masses. Figure 8.4.1 depicts an isometric view of the three-dimensional finite element model without the water and concentrated masses (racks, cask, etc.).
The dynamic behavior of the water mass contained in the SFPs and Transfer Canal during a seismic event is modeled according to the guidelines set in TID-7024.
To simulate the interaction between the modeled region and the rest of the Fuel Handling Building a number of boundary restraints were imposed upon the described finite-element model.
The behavior of the reinforced concrete existing in the structural elements (walls, slab and mat)
~
is considered elastic and isotropic. The elastic characteristics
~ of the concrete are independent of the reinforcement contained in each structural element for the case when the un-cracked cross-section is assumed. This assumption is valid for all load cases with the exception of the thermal loads, where for a more realistic description of the reinforced concrete cross-section the assumption of cracked concrete is used. To simulate the variation and the degree of cracking patterns, the original elastic modulus of the concrete is modified in accordance with Reference
[8.4.2].
8.4.2 The structural region of concern, from column lines 43 to 73 and from line L to N, is isolated from the Fuel Handling Building. This region is n'umerically investigated using the finite element method. The pool walls and their supporting reinforced concrete slab are represented by a 3-D finite-element model.
Holtec International 8-4 Report HI-971760
The individual loads considered in the analysis are grouped in five categories: dead load (weight of the pool structure, dead weight of the rack modules and stored fuel, dead weight of the reinforced concrete Fuel Handling Building upper structure, the hydro-static pressure of the contained water), live loads (weights of the Cask Crane, Auxiliary Crane, and SFHM and their maximum suspended loads), thermal loads (the thermal gradient through the pool walls and slab for normal operating and accident conditions) and the seismic induced forces (structural seismic forces, interaction forces between the rack modules and the pool slab, seismic loads due to self-excitation of the pool structural elements and contained water, and seismic hydro-dynamic interaction forces between the rack modules and the pool walls for both OBE and SSE conditions). The dead and thermal loads are considered static acting loads, while the seismic induced loads are time-dependent.
The material behavior under all type of loading conditions is described as elastic and isotropic representing the uncracked characteristics of the structural elements cross-section, with the exception of the thermal load cases where the material elasticity modulus is reduced in order to simulate the variation and the degree of the crack patterns. This approach [8.1.3]
acknowledges the self-relieving nature of the thermal loads. The degree of reduction of the elastic modulus is calculated based on the average ultimate capacity of the particular structural element.
The numerical solution (displacements and stresses) for the cases when the structure was 7
subjected to dead and thermal loads is a classical static solution. For the time-dependent seismic induced loads the displacement and stress field are calculated employing the spectra (shock) method. This method requires a prior modal eigenvector and eigenvalues extraction.
Natural frequencies of the 3-D finite-element model are calculated up to the rigid range, considered as greater than 34 Hz. Three independent orthogonal acceleration spectra are applied to the model. The acceleration spectra are considered to act simultaneously in three-directions. The SRSS method is used to sum the similar quantities calculated for each direction.
Holtec International 8-5 Report HI-971760
Results for individual load cases are combined using the factored load combinations discussed below. The combined stress resultants are compared with the ultimate moments and shear capacities of all structural elements pertinent to the SFPs, which are calculated in accordance with the ACI 318 to develop the safety factors.
8.4.3 The various individual load cases are combined in accordance with the NUREG-0800 Standard Review Plan [8.1.2] requirements with the intent to obtain the most critical stress fields for the investigated reinforced concrete structural elements.
For "Service Load Conditions" the following load combinations are:
- Load Combination No. 1 = 1.4* D + 1.7*L
- Load Combination No. 2 = 1.4* D + 1.7*L + 1.9*E
- Load Combination No. 3 = 1.4* D + 1.7*L - 1.9*E
- Load Combination No. 4 = 0.75* (1.4* D + 1.7*L + 1.9*E +1.7*To)
- Load Combination No. 5 = 0.75* (1.4* D + 1.7*L - 1.9*E + 1.7*To)
- Load Combination No. 6 = 1.2*D + 1.9*E
- Load Combination No. 7 = 1.2*D - 1.9*E For "Factored Load Conditions" the following load combinations are:
Holtec International 8-6 Report HI-971760
- Load Combination No. 8 = D + L + To + E' Load Combination No. 9 = D + L + To - E' Load Combination No. 10 = D + L + Ta + 1.25*E
- Load Combination No. 11 = D + L + Ta - 1.25*E
- Load CombinationNo. 12=D + L + Ta + E' Load Combination No. 13 = D + L + Ta -
E'here:
D= dead loads; live loads; TQ = thermal load during normal operation; Ta = thermal load under accident condition; OBE earthquake induced loads; SSE earthquake induced loads.
8.5 The STARDYNE computer code is used to obtain the stress and displacement fields for the 1 individual load cases.
The STARDYNE postprocessing capability is employed to form the appropriate load combinations and to establish the limiting bending moments'and shear forces in various sections of the pool structure. A total of 13 load combinations are computed. Section limit strength Holtec International 8-7 Report HI-971760
formulas for bending loading are computed using appropriate concrete and reinforcement strengths. For Harris, the concrete and reinforcement allowable strengths are:
concrete 4,000 psi f,'einforcement f = 60,000 psi Table 8.5.1 shows results from potentially limiting load combinations for the bending and shear strength of the slab and walls. For each section, we define the limiting safety margins as the limited strength bending moment or shear force defined by ACI for that structural section divided by the calculated bending moment or shear force (Gom the finite element analyses). The major regions of the pool structure consist of the four concrete walls and floors delimiting each of the SFPs. Each area is searched independently for the maximum bending moments in different bending directions and for the maximum shear forces. Safety margins are determined from the calculated maximum bending moments and shear forces based on the local strengths. The procedures are repeated for all the potential limiting load combinations. Therefore, limiting safety margins are determined. Table 8.5.1 demonstrates that the limiting safety margins for all sections are above 1.0, as required.
8.6 2aoLLiner The pool liners are subject to in-plate strains due to movement of the rack support feet during the seismic event. Analyses are performed to establish that the liner willnot tear or rupture under limiting loading conditions in the pool. These analyses are based on loadings imparted Rom the most highly loaded pedestal in the pool assumed to be positioned in the most unfavorable position. Bearing strength requirements are shown to be satisfied by conservatively analyzing the most highly loaded pedestal located in the worst configuration with respect to underlying leak chases.
8.7 Conclusions Holtec International 8-8 Report HI-971760
Regions affected by loading the fuel pool completely with high density racks are examined for structural integrity under bending and shearing action. It is determined that adequate safety margins exist assuming that all racks are fully loaded with a bounding fuel weight and that the factored load combinations are checked against the appropriate structural design strengths. It is also shown that local loading on the liner does not compromise liner integrity under a postulated fatigue condition and that concrete bearing strength limits are not exceeded.
Holtee International 8-9 Report HI-971760
8.8 I cferuwm
[8.1.1] OT Position for Review and Acceptance of Spent Fuel Handling Applications, by B.K. Grimes, USNRC, Washington, D.C., April 14, 1978.
[8.1.2] NUREG-0800, SRP-3.8.4, Rev. 1., July 1981.
[8.1.3] ACI 318-95 and ACI 318R-95, "Building Code Requirements for Structural Concrete and Commentary," American Concrete Institute, 1995.
[8.3.1] "Nuclear Reactors and Earthquakes, U.S. Department of Commerce, National Bureau of Standards, National Technical Information Service, Springfield, Virginia (TID 7024).
[8.4.1] STARDYNE User's Manual, Research Engineers, Inc., Rev. 4.4, July 1996.
[8.4.2] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit Michigan.
Holtec International 8-10 Report HI-971760
Table 8.5.1 BENDING AND SHEAR STRENGTH EVALUATION Pool Critical Flexure Critical Shear Location Limiting Load Combinations Limiting Load Combinations Safety Margin (see Section 8.4.3) Safety Margin (see Section 8.4.3)
North Wall 1.97 1.31 South Wall 3.51 2.20 East Wall 1.72 1.10 West Wall 1.05 10 1.06 Pool Floor Slab 2.1 2.71 D North Wall 2.32 3.43 South Wall 1.30 10 1.08 3,7 East Wall 1.48 2,6 1.07 37 West Wall 1.05 1.06 Pool Floor Slab 2.01 1.64 Holtec International 8-11 Report HI-971760
Figure 8.2.1; Pool Structure Dimensions HI-971760
1 L9 1
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90 9.1 No significant increase in the volume of solid radioactive wastes is expected from operating with the expanded storage capacity. The necessity for pool filtration resin replacement is determined primarily by the requirement for water clarity, and the resin is normally expected to be changed about once a year. During racking operations, a small amount of additional resins may be generated by the pool cleanup system on a one-time basis.
9.2 Gaseous releases from the fuel storage area are combined with other plant exhausts. Normally, the contribution from the fuel storage area is negligible compared to the other releases and no significant increases are expected as a result of the expanded storage capacity.
9.3 During normal operations, personnel working in the fuel storage area are exposed to radiation from the spent fuel pool. Operating experience has shown that area radiation dose rates originate primarily &om radionuclides in the pool water. As expected, subsequent to the removal of transhipped fuel from the shipping casks, Harris has experienced increases in the pool water radionuclide concentrations due to sloughing of crud and other contaminants associated with fuel handling. Additionally, radionuclide concentration increases are also experienced subsequent to the discharge of fuel Rom the Harris Unit 1 reactor. These two conditions represent the the previously analyzed conditions for pool water radionuclide concentrations and will not be significantly changed by the capacity expansion of storing spent fuel in pools C and D Therefore, no additional evaluations for pool water radionuclides are required for the proposed change.
Holtee International 9-1 Report HI-971760
Radiation dose rates in accessible areas around the SFPs will be determined for comparison with existing zone designations. Any changes required to the zone designations will be identified and included in an update to the Harris FSAR, ifnecessary.
Operating experience has also shown that there have been negligible concentrations of airborne radioactivity in the Spent Fuel Pool area. No increase in airborne radioactivity is expected as a result of the expanded storage capacity.
9.4 All of the operations involved in racking will utilize detailed procedures prepared with full consideration of ALARAprinciples. Similar operations have been performed in a number of facilities in the past, and there is every reason to believe that racking can be safely and efficiently accomplished at Harris, with low radiation exposure to personnel. The Harris racking project represents lower radiological risks due to the fact that the pools currently contain no spent fuel.
'otal dose for the racking operation is estimated to be between 2 and 3 person-rem, as indicated in Table 9.4.1. While individual task efforts and doses may differ from those in Table 9.4.1, the total is believed to be a reasonable estimate for planning purposes. Divers will be used only ifnecessary, but the estimated person-rem burden includes a figure for their possible dose.
The existing radiation protection program at Harris is adequate for the re-racking operations.
Where there is a potential for significant airborne activity, continuous air monitors will be in operation. Personnel will wear protective clothing as required and, if necessary, respiratory protective equipment. Activities will be governed by a Radiation Work Permit, and personnel monitoring equipment will be issued to each individual. As a minimum, this will include Holtec International 9-2 Report Hl-971760
thermoluminescent dosimeters (TLDs) and self-reading dosimeters. Additional personnel monitoring equipment (i.e., extremity TLDs or multiple TLDs) may be utilized as required.
Work, personnel traffic, and the movement of equipment will be monitored and controlled to minimize contamination and to assure that dose is maintained ALARA.
Holtee International 9-3 Report HI-971760
Table 9.4.1 PRELIMINARYESTIMATE OF PERSON-REM DOSE DURING RACKING Estimated Number of Person-Rem Step Personnel Hours Dose Clean and vacuum pool 25 1.5 to 2.0 Remove underwater 0.4 to 0.8 appurtenances Installation of new rack modules 20 0.1 to 0.2 Total Dose, person-rem 2to3 Holtec International 9-4 Report HI-971760
ioo 10.1 The construction phase of the Harris Spent fuel pool rack installation willbe executed by Carolina Power & Light. CPkL will also be responsible for specialized services, such as underwater diving and welding operations, ifrequired. All construction work at Harris willbe performed in compliance with NUREG-0612 (refer to Section 3.0), and site-specific procedures.
Crane and fuel bridge operators are to be adequately trained in the operation of load handling machines per the requirements of ANSVASME B30.2, latest revision, and the plant's specific training program.
The lifting devices designed for handling and installation of the new racks and removal of the old racks at Harris are remotely engageable. The lifting devices comply with the provisions of ANSI N14.6-1978 and NUREG-0612, including compliance with the primary stress criteria, load testing at a multiplier of maximum working load, and nondestructive examination of critical welds.
An intensive surveillance and inspection program shall be maintained throughout the rack installation phase of the project. A set of inspection and QC hold points willbe implemented which have been proven to eliminate any incidence of rework or erroneous installation in numerous previous rack installation campaigns in Pools A and B.
Holtec International and CPAL have developed a complete set of operating procedures which cover the entire gamut of operations pertaining to the rack installation effort. Similar procedures have been utilized and successfully implemented by Holtec International on previous rack installation projects. These procedures assure that ALARApractices are followed and provide detailed requirements to assure equipment, personnel, and plant safety. The following is a list of procedures which will be available for use in implementing the rack installation phase of the Holtec International 10-1 Holtec Report HI-971760
This procedure provides direction for the handling/installation of the new high density modules.
The procedure delineates the steps necessary to receive a new high density rack on site, and the proper method for unloading and uprighting the rack, staging the rack prior to installation, and installation of the rack. The procedure also provides for the installation of new rack bearing pads, adjustment of the new rack pedestals and performance of the as-built field survey.
~
This procedure delineates the steps necessary to perform a thorough receipt inspection of a new rack module after its arrival on site. The receipt inspection includes dimensional measurements, cleanliness inspection, visual weld examination, and verticality measurements.
C.
This procedure provides for the cleaning the requirements of a new rack module, of ANSI 45.2.1, Level C.
on materials to be employed are provided.
ifit is required, in order to meet Permissible cleaning agents, methods and limitations D.
This procedure stipulates the requirements for performing a functional test on a new rack module prior to installation into Pools C or D. The procedure provides direction for inserting and withdrawing a "dummy" fuel assembly into designated cell locations, and establishes an acceptance criteria in terms of maximum kinetic drag force.
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I This procedure stipulates the requirements for performing a functional test on a new rack module following installation into Pools C or D. The procedure willprovide direction for inserting and withdrawing a "dummy" fuel assembly into designated cell locations, and establishes an acceptance criteria in terms of maximum kinetic drag force.
Underwater diving operations may be required to assist in the positioning of new rack modules.
This procedure describes the method for introducing a diver into Pools C or D, provides for radiological monitoring during the operation, and defines the egress of the diver from the fuel pool following work completion. Furthermore, this procedure requires strict compliance with OSHA Standard 29CFR-1910, Subpart T, and establishes contingencies in the event of an emergency.
Consistent with Holtec International's ALARAProgram, this procedure provides details to minimize the total man-rem received during the rack installation project, by accounting for time, distance, and shielding. Additionally, a pre-job checklist is established in order to mitigate the potential for an overexposure.
In the event that a visual inspection of any submerged portion of the Spent Fuel Pool liner is deemed necessary, this procedure describes the method to perform such an inspection using an underwater camera and describes the requirements for documenting any observations.
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This procedure describes the method to test the Spent Fuel Pool liner for potential leakage using.'
vacuum box. This procedure may be applied to any suspect area of the pool liner.
In the event of a positive leak test result, an underwater welding procedure willbe implemented which willprovide for the placement of a stainless steel repair patch over the area in question.
The procedure contains appropriate qualification records documenting relevant variables, parameters, and limiting conditions. The weld procedure is qualified in accordance with AWS D3.6-93, Specification for Underwater Welding or may be qualified to an alternate code accepted by CP&L and Holtec International.
This procedure establishes the requirements for safely storing a new rack module on-site, in the event that long term job-site storage is necessary. This procedure provides environmental restrictions, temperature limits, and packaging requirements.
10.2 Pools C and D at Harris have been previously unused. The new rack arangement has been prepared to maximize flexibilityin the number and type (PWR vs. BWR) fuel assemblies stored.
The new rack arrangement for Pool C consists of a mixture of free-standing PWR and BWR Holtec racks.,
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A breakdown of the number of racks and storage cells in the first campaign and completely filled configuration of Pool C is as follows:
First Campaign Filled Pool PWR Cells 360 927 BWR Cells 1320 10 2763 19 Total 1680 14 3690 30 Pool D will store a maximum of 1025 PWR assemblies in 12 rack modules. Racks willbe added to the pools on an as needed basis. A schematic plan view depicting the Spent Fuel Pools in the new maximum density configuration can be seen in Figure 1.1.
10.3 A pool inspection shall be performed to determine ifany items attached to the liner wall or floor will interfere with the placement of the new racks or prevent usage of any cell locations subsequent to installation.
In the event that protrusions are found which would pose any interference to the installation process, it is anticipated that underwater diving operations and mechanical cutting methods I
would be employed to remove the protrusions.
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iO.4 10.4.1 The pool cooling system shall be operated in order to maintain the pool water temperature at an acceptable level. It is anticipated that specific activities, such as bearing pad elevation measurements, may require the temporary shutdown of the Spent Fuel Pool cooling system. At no time, however, willpool cooling be terminated in a manner or for a duration which would create a violation of the Harris Technical Specification or procedures.
Prior to any shutdown of the Spent Fuel Pool cooling system, the duration to raise the pool bulk temperature to 137'F will be determined. A margin temperature of 112'F is chosen such that the cooling system may be restarted prior to reaching this temperature. This will ensure that the pool bulk temperature will always remain below 137'F.
10.4.2 The existing Spent Fuel Pool filtration system shall be operational in order to maintain pool clarity. Additionally, an underwater vacuum system shall be used as necessary to supplement fuel pool purification. The vacuum system may be employed to remove extraneous debris, reduce general contamination levels prior to diving operations, and to assist in the restoration of pool clarity following any hydrolasing operations.
10.5 The new high density racks shall be delivered in the horizontal position. A new rack module shall be removed &om the shipping trailer using a suitably rated crane, while maintaining the horizontal configuration, and placed upon the upender and secured. Using two independent overhead hooks, or a single overhead hook and a spreader beam, the module shall be uprighted into vertical position.
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The new rack lifting device shall be installed into the rack and each liftrod successively engaged.
Thereafter, the rack shall be transported to ayre-levelled surface where the appropriate quality control receipt inspection shall be performed.
In preparing Pool C or D for the initial rack installation, the pool floor shall be inspected and any debris which may inhibit the installation of bearing pads willbe removed. New rack bearing pads shall be positioned in preparation for the rack modules which are to be installed. Elevation measurements willthen be performed in order to gage the amount of adjustment required, ifany, for the new rack pedestals.
The new rack module shall be lifted with the Auxiliary Crane and transported along the safe load path. The rack pedestals shall be adjusted in accordance with the bearing pad elevation measurements in order to achieve module levelness aAer installation.
It is anticipated that the rack modules shall be lowered into the Pools C and D using the Cask Handling Crane. A hoist with sufficient capacity willbe attached to the Auxiliary Crane for installation and removal activities in order to eliminate contamination of the main hook during lifting oper'ations in the pools. The rack shall be carefully lowered onto its bearing pads.
Movements along the pool floor shall not exceed six inches above the liner, except to allow for clearance over floor projections.
Elevation readings shall be taken to confirm that the module is level and as-built rack-to-rack and rack-to-wall offsets shall be recorded. The lifting device shall be disengaged and removed &om the fuel pool under Radiation Protection direction.
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10.6.1 During the rack installation phase of the project, personnel safety is of paramount importance, outweighing all other concerns. Allwork shall be carried out in strict compliance with applicable approved procedures.
10.6.2 Radiation Protection shall provide necessary coverage in order to provide radiological protection and monitor dose'rates. The Radiation Protection department shall prepare Radiation Work permits (RWPs) that will instruct the project personnel in the areas of protective clothing, general dose rates, contamination levels, and dosimetry requirements.
In addition, no activity within the radiologically controlled area shall be carried out without the knowledge and approval of Radiation Protection,- Radiation Protection shall also monitor items removed from the pool or provide for the use of alarming dosimetry and supply direction for the proper storage of radioactive material.
10.6.3 The key factors in maintaining project dose As Low As Reasonably Achievable (ALARA)are time, distance, and shielding. These factors are addressed by utilizing many mechanisms with respect to project planning and execution.
Each member of the project team willbe properly trained and willbe provided appropriate education and understanding of critical evolutions. Additionally, daily pre-job briefings willbe employed to acquaint each team member with the scope of work to be performed and the proper Holtec International 10-8 Holtec Report HI-971760
4 means of executing such tasks. Such pre-planning devices reduce worker time within the radiologically controlled area and, therefore, project dose.
Remote tooling such as lift fixtures, pneumatic grippers, a support levelling device and a liftrod disengagement device have been developed to execute numerous activities from the pool surface, where dose rates are relatively low. For those evolutions requiring diving operations, diver movements shall be restricted by an umbilical, which will assist in maintaining a safe distance from irradiated sources. By maximizing the distance between a radioactive sources and project personnel, project dose is reduced.
During the course of the rack installation, primary shielding is provided by the water in the Spent Fuel Pool. The amount of water between an individual at the surface (or a diver in the pool) and an irradiated fuel assembly is an essential shield that reduces dose. Additionally, other shielding, may be employed to mitigate dose when work is performed around high dose rate sources.
10.7 Radioactive waste generated from the rack installation effort shall include vacuum filter bags, miscellaneous tooling, and protective clothing.
Vacuum filter bags shall be removed from the pool and stored as appropriate in a suitable container in order to maintain low dose rates.
t Contaminated tooling shall be properly stored per Radiation Protection direction throughout the project. At project completion, an effort willbe made to decontaminate tooling to the most practical extent possible.
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ENVIRONMENTALCOST/BENEFIT ASSESSMENT Introduction Article V of the USNRC OT Position Paper [11.1] requires the submittal of a cost/benefit analysis for the chosen fuel storage capacity enhancement method. This section provides justification for selecting rack installation in Pools C and D as the most viable alternative.
11.2 Im erative for Increased Stora e Ca acit The specific need to increase the limited existing storage capacity at the Harris facility is based on the continually increasing inventory in Pools A and B due to core offloads at Harris and transhipments from the Robinson and Brunswick plants, the prudent requirement to maintain full-core offload capability, and a lack of viable economic alternatives.
Based on the current number of stored assemblies and estimated discharge and transhipment rates, the Harris fuel pool is projected to lose the capacity to discharge one full core in 2001.
This projected loss of storage capacity in the Harris pool would affect CP&L's ability to operate the reactors. CP&L does not have an existing or planned contractual arrangement for third party fuel storage or fuel reprocessing.
11.3 A raisal of Alternative 0 tions CP&L has determined that rack installation at the Harris pools is by far the most viable option for increasing spent fuel storage capacity in comparison to other alternatives.
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The key considerations in evaluating the alternative options are:
Safety: minimize the number of fuel handling steps Economy: minimize total installed and 08rM cost Security: protection from potential saboteurs, natural phenomena Non-intrusiveness: minimize required modification to existing systems Maturity: extent of industry experience with the technology ALARA:minimize cumulative dose due to handling of fuel Rack installation was found by CPS'o be the most attractive option in respect to each of the foregoing criteria. An overview of the alternatives is provided in the following.
Rod Consolidation Rod consolidation has been shown to be a potentially feasible technology. Rod consolidation involves disassembly of spent fuel, followed by the storage of the fuel rods from two assemblies into the volume of one and the disposal of the fuel assembly skeleton outside of the pool (this is considered a 2:1 compaction ratio). The rods are stored in a stainless steel can that has the outer dimensions of a fuel assembly. The can is stored in the spent fuel racks. The top of the can has an end fixture that matches up with the spent fuel handling tool. This permits moving the cans in an easy fashion.
Rod consolidation pilot project campaigns in the past have consisted of underwater tooling that is manipulated by an overhead crane and operated by a maintenance worker. This is a very slow and repetitive process.
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The industry experience with rod consolidation has been mixed thus far. The principal advantages of this technology are: the ability to modularize, compatibility with DOE waste management system, moderate cost, no need of additional land and no additional required surveillance. The disadvantages are: potential gap activity release due to rod breakage, potential for increased fuel cladding corrosion due to some of the protective oxide layer being scraped off, potential interference of the (prolonged) consolidation activity which might interfere with ongoing plant operation, and lack of sufficient industry experience.
On-Site Cask Stora e Dry cask storage is a method of storing spent nuclear fuel in a high capacity container. The cask provides radiation shielding and passive heat dissipation. Typical capacities for PWR fuel range from 21 to 37 assemblies that have been removed from the reactor for at least five years. The casks, once loaded, are then stored outdoors on a seismically qualified concrete pad. The pad will have to be located away from the secured boundary of the site because of site limitations. The storage location will be required to have a high level of security which includes frequent tours, reliable lighting, intruder detection, (E-field), and continuous visual monitoring The casks, as presently licensed, are limited to 20-year storage service life. Once the 20 years has expired the cask manufacturer or the utility must recertify the cask or the utility must remove the spent fuel from the container.
There are several plant modifications required to support cask use. Tap-ins must be made to the gaseous waste system and chilled water to support vacuum drying of the spent fuel and piping must be installed to return cask water back to the Spent Fuel Pools. A seismic concrete pad must be made to store the loaded casks. This pad must have a security fence, surveillance protection, a diesel generator for emergency power and video surveillance.
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Finally, the cask facility must have equipment required to vacuum dry the cask, backfill it with helium, make leak checks, remachine the gasket surfaces ifleaks persist, and assemble the cask on-site. For casks which have closure gaskets, the space between the inner and outer lid must be continuously monitored to check for inner seal failure.
Presently, no MPC cask has been licensed. Because of the continued uncertainty in the government's policy, the capital investment to develop a dry storage system is considered to be an inferior alternative for Harris at this time.
Modular Vault D Stora e Vault storage consists of storing spent fuel in shielded stainless steel cylinders in a horizontal configuration in a reinforced concrete vault. The concrete vault provides radiation shielding and missile protection. It must be designed to withstand the postulated seismic loadings for the site.
A transfer cask is needed to fetch the storage canisters from the fuel pool. The plant must provide for a decontamination bay to decontaminate the transfer cask, and connection to its gaseous waste system and chilled water systems. A collection and delivery system must be installed to return the pool water entrained in the canisters back to the fuel pool. Provisions for canister drying, helium injection, handling, and automatic welding are also necessary.
The storage area must be designed to have a high level of security similar to that of the nuclear plant itself. Due to the required space, the vault secured area must be located outside the secured perimeter. Consideration of safety and security requires it to have its own video surveillance system; intrusion detection, and an autonomous backup diesel generator power source.
Some other concerns relating to the vault storage system are: inherent eventual "repackaging" for shipment to the DOE repository, the responsibility to eventually decommission the new facility, Holtec International 11-4 Holtec Report HI-971760
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large "footprint" (land consumption), potential fuel handling accidents, potential fueVclad rupture due to high temperature and high cost.
E At the present time, no MPC technology based vault system has yet been offered for licensing to the USNRC. Therefore, this option is considered to be unavailable at this time.
Horizontal Silo Stora e A variation of the horizontal vault storage technology is more aptly referred to as "horizontal silo" storage. This technology suffers from the same drawbacks which other dry cask technologies do, namely,
- i. No fuel with cladding defects can be placed in the silo.
ii. Concern regarding long-term integrity of the fuel at elevated temperature.
iii. Potential for eventual repackaging at the site.
iv. Potential for fuel handling accidents.
- v. Relatively high cumulative dose to personnel in effecting fuel transfer (compared to rack installation).
vi. Compatibility of reactor/fuel building handling crane with fuel transfer hardware.
vii. Potential incompatibility with DOE shipment for eventual off-site shipment..
viii. Potential for sabotage.
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11.3.1 Alternative 0 tion Summa An estimate of relative costs in 1997 dollars for the aforementioned options is provided in the following:
Rack Installation: $ 12 million Horizontal Silo: $ 35-45 million Rod consolidation: $ 25 million Metal cask (MPC): $ 68-100 million Modular vault: $ 56 million The above estimates are consistent with estimates by EPRI and others [11.2, 11.3].
To summarize, there are no acceptable alternatives to increasing the on-site spent fuel storage capacity of Harris. 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 facto throw-away nuclear fuel cycle. Since the cost of spent fuel reprocessing is not offset by the salvage value of the residual uranium, reprocessing represents an added cost for the nuclear fuel cycle which already includes the NWPA Nuclear Waste Fund fees. In any event, there are no domestic reprocessing facilities. Third, at over $ '/a million per day replacement power cost, shutting down the Harris reactor is many times more expensive than simply installing racks in the existing Spent Fuel Pools.
11.4 Cost Estimate The proposed construction contemplates installation of storage modules in Harris Pools C and D using free-standing, high density, poisoned spent fuel racks. The engineering and design is completed for rack installation in the pools. This rack installation project willprovide sufficient pool storage capacity to maintain full-core offload capability until the end of the current plant license.
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The total capital cost is estimated to be approximately $ 12 million as detailed below.
Engineering, design, project management: $ 2 million Rack fabrication: $ 7 million Rack installation: $ 3 million As described in the preceding section, many alternatives were considered prior to proceeding with rack installation, which is not the only technical option available to increase on-site storage capacity. Rack installation does, however, enjoy a definite cost advantage over other technologies.
11.5 Resource Commitment The expansion of the Harris Spent Fuel Pool capacity is expected to require the following primary resources:
Stainless steel 250 tons Boral neutron absorber: 20 tons, of which 15 tons is Boron Carbide powder and 5 tons are aluminum.
The requirements for stainless steel and aluminum represent a small fraction of total world output of these metals (less than 0.001%). Although the fraction of world production of Boron Carbide required for the 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 alternatives.
Experience has shown that the production of Boron Carbide is highly variable and depends upon need and can easily be expanded to accommodate worldwide needs.
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11.6 Environmental Considerations Prior to the proposed modification, Pools C and D were maintained full of water with levels consistent with those of Pools A and B. Although water was allowed to be exchanged between all four pools at various times, there was no heat load associated with Pools C and D. Therefore, the bulk pool temperatues in Pools C and D have always been maintained at or below the temperatures in Pools A and B. Due to the heat load arising from the spent fuel inventory, the pool cooling system willbe connected to Pools C and D to provide adequate heat removal capabilities. The maximum normal bulk pool temperature willbe realized when the capacity is maximized for Pools C and D, but will still be s 137'F t.
Maintaining four pools (instead of the previous two pools) in the Fuel Handling Building with bulk pool temperatures F137'F t willresult in an increase in the pool water evaporation rate.
This pool water evaporation increase has been determined to increase the relative humidity of the Fuel Building atmosphere by less than 10%. This increase is within the capacity of both the normal and the ESF Ventilation Systems. The net result of the increased heat loss and water vapor emission to the environment is negligible The 137 F limit is consistent with that currently in the Harris FSAR and procedures for pools A and B. CP&L is in the process of re-evaluating systems and components to allow for an increase the allowable bulk pool temperature.
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11.7 References
[11.1] OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, USNRC (April 1978).
[11.2] Electric Power Research Institute, Report No. NF-3580, May 1984.
[11.3] "Spent Fuel Storage Options: A Critical Appraisal", Power Generation Technology, Sterling Publishers, pp. 137-140, U.K. (November 1990).
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