ML20204K103
| ML20204K103 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/19/1999 |
| From: | HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20137D872 | List: |
| References | |
| NUDOCS 9903300240 | |
| Download: ML20204K103 (240) | |
Text
Docket No. 50-423 B17343 Attachment 5 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Spent Fuel Pool Rerack (TSCR 3-22-98)
Non-Proprietary Version of Licensing Report for Spent Fuel Rack Installation-at Millstone Nuclear Station Unit 3 j l
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9903300240 990319 ~i PDR ADOCK 05000423 : March 1999 P PDR j ,
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LICENSING REPORT for' SPENT FUEL RACK INSTALLATION I l
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MILLSTONE NUCLEAR STATION l UNIT 3 COMPANY PRIVATE This document has all proprietary information removed and has replaced those sections, figures, and tables with highlighting and/or notes to designate the removal of such information. This document is the property of Hoitec International and its Client. It is to be used only in connection with the performance of work by Holtec International 'or its designated subcontractors. Reproduction, publication or presentation, in whole or in part, for any other purpose by any party other than the Client is expressly forbidden.
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Table of Contents
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1.0 INTRODUCTION
1-1 2.0 OVERVIEW OF PROPOSED CAPACITY EXPANSION 2-1 2.1 General Description 2-1 2.2 Design Basis 2-2 2.3 Codes, Standards, and Practices for the Spent Fuel Pool Modification 2-3 2.4 Quality Assurance Program '8
- 2.5 Mechanical Design 2-8 3.0 FABRICATION, MATERIALS, AND HEAVY LOADS CONSIDERATIONS 3-1 3.1 Rack Fabrication 3-1 3.2 MaterialConsiderations 3-4 3.3 Heavy Load Considerations for the Proposed Reracking Operation 3-6 4.0 CRITICALITY SAFETY EVALUATION 4-1
!' 4.1 Design Bases 4-1 4.2 Summary of Criticality Analyses 4-5 j 4.3 Reference Fuel Storage Cells 4-10 l 4.4. Analytical Methodology 4-12 l
4.5 Region 1 Criticality Analyses and Tolerances 4-16 4.6 Region 2 Criticality Analyses and Tolerances 4-18 I 4.7 Region 3 Criticality Analyses and Tolerances 4-20 l 4.8 Abnormal and Accident Conditions 4-22 4.9 References 4-25 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5-1 5.1 Introduction 5-1 IlOLTEC INTERNATIONAL
- Millstone Point Unit 3 i
1 5.2 Syst:m Descriptioi 1 5-2 1 5.3 Discharge / Cooling Alignment Scenarios 5-3
. 5,4 Decay Heat Load, In-Core Hold Time, SFP Heat-Up Time 5-5 5.5 Local Pool Water Temperature 5-5 5.6 Fuel Rod Cladding Temperature 5-9 5.7 Results 5-10 5.8 References 5-11 6.0 STRUCTURALISEISMIC CONSIDERATIONS 6-1 6.1 Introduction 6-1 i
6.2 Overview of Rack Structural Analysis Methodology 6-1 6.3 Description of Racim 6-5 6.4 Synthetic Time-liistories 6-5 6.5 22-DOF Nonlinear Rack Model for Dynamic Analysis 6-6 6.6 Whole Pool Multi-Rack Methodology 6-12 I
6.7 Structural Evaluation of Spent Fuel Rack 6-16 i l
6.8 Seismic Analysis 6-21 1 6.9 Time IIistory Simulation Results . 6-25 6.10 Rack Structural Evaluation 6-35 6.11 Level A Evaluetion 6-44 6.12 Hydrodynamic Loads on Pool Walls 6-45 6.13 Conclusion 6-46 6.14 References 6-47 7.0 FUEL HANDLING AND CONSTRUCTION ACCIDENTS 7-1 7.1 Introduction 7-1 7.2 Description of Fuelifandling Accidents 7-1 7.3 Mathematical Model 7-4 SIIADED REGIONS CONTAIN IIOLTEC PROPRIETARY INFORMATION
. Millstone Point Unit 3 ii
7.4 Results 7-4 l 7.5 Rack Drop - 7-5 7.6 Gate Drop 7-5 7.7 Closure 7-6 7.8 References 7-7 l
8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8-1 8.1 Introduction 8-1 8.2 Description of Pool Structures 8-1 8.3 Definition of Loads 8-2 8.4 Analysis Procedures 8-3 8.5 Results of Analyses 8-7 8.6 Pool Liner 8-8 8.7 Bearig Pad Analysis 8-8 8.8 Conclusions 8-9 1
8.9 References 8-10 1 9.0 BORAL SURVEILLANCE PROGRAM 91 l
9.1 Purpose 9-1 9.2 COUPON SURVEILLANCE PROGRAM 9-2 9.3 In-Service Inspection (Blackness Tests) 9-5 9.4 References 9-6 10.0 INSTALLATION 10-1 10.1 Introduction 10-1 10.2 Rack Arrangement 10-4 10.3 Pool Survey and Inspection 10-5 10.4 Pool Cooling and Purification 10-5 l 10.5 Installation of New Racks 10-6 l.
l HCLTEC INTERNATIONAL Millstone Point Unit 3 iii
I 10.6 Safety, Radiation Protection,and ALARA Methods 10-7 i
10.7 Radwaste Material Control 10-8 l 11.0 RADIOLOGICAL EVALUATION 11-1 i 11.1 Solid Radwaste 11-1
( l1.2 Gaseous Releases Il-1 11.3 Personnel Doses Il-1 11.4 Anticipated Dose During Re-racking 11-2 12.0 ENVIRONMENTAL COST-BENEFIT ASSESSMENT 12-1 l
12.1 Introduction 12-1 12.2 Imperative for Reracking 12 1 12.3 Appraisalof Alternative Options 12-2 I
i 12.4 Cost Estimate 12-3 12.5 Resource Commitment 12-3 12.6 Environmental Considerations 12-4 12.7 References for Section 12 12-4 i
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- 1.0 ' INTRODUCTION
- l. 1 Millstone Point Unit 3 (MP3) is a Westinghouse Pressurized Water Reector (PWR) owned and operated by Northeast Utilities (NU). The plant is located at a three unit site in the town of Waterford,- Connecticut. A license was granted by the USNRC on January 31, 1986 and full commercial operation of the 1150 MWe plant began in 1986.
The MP3 reactor core contains 193 assemblies. During re-fueling, spent fuel is placed in the plant's !
2 pool; an L - shaped basin with a total nominal area of approximately 1,574 ft . The pool presently contains 756 storage cells which were installed during original plant construction. The twenty-one existing storage racks are of end-connected-construction (ECC). Each contains a 6 x 6 array. As is -
true for all ECC racks, the individual boxes are connected to each other at their extremities; there is no longitudinal inter-cell connection between the cells. The ECC racks employ a 0.06 incli wall storage cell at a pitch of 10.35 inches, with Boraflex serving as the neutron absorber.
4 This license application addresses installation of fifteen high-density racks in the MP3 pool. These fifteen high density racks have a maximum capacity of 1,104 storage cells. Additional storage ,
capacity is needed since MP3 will lose its full-core reserve discharge capacity at the end of its seventh cycle. Tables 1.1 and 1.2 demonstrate this. Table 1.1 shows the historic and projected ,
I discharges into the MP3 pool. Table 1.2 shows the current and post-modification storage capacities. ;
The new racks will extend the date of loss-of-full-core-reserve discharge capability approaching end oflicense (see Table 1.2).
Northeast Utilities plans to install fourteen modules initillly and the fifteenth rack (AS) at a later date. The analyses include the fifteenth rack. Figure 21 c/ Section 2 shows the planned layout. The existing fuel racks will not be moved. However, credit for Boraflex as a neutron absorber will be eliminated, j The new high density racks proposed for MP3 have been designed by Holtec International of l Marlton; New Jersey. The racks are free-standing and self-supporting. The principal construction HOLTEC INTERNATIONAL
-. Millstone Point Unit 3 1-1
materials for the new racks are ASME SA240-Type 304L stainless steel sheet and plate stock and SA564 (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 aluminum cermet manufactured under a U.S. patent and sold under the brand name Boral by AAR Advanced Structures, Livonia, Michigan.
The new racks are designed and analyzed in accordance with Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel Code. The material procurement and fabrication of the rack modules conforms to 10CFR50 Appendix B requirements. The racks proposed for the MP3 pool are identical in their anatomical details to racks recently provided by Holtec International to many PWR plants. Table 1.3 lists recently licensed PWR plants with racks similar to those proposed for MP3.
This Licensing Report documents the design and analyses performed to demonstrate that the new spent fuel racks satisfy all requirements of the governing codes and standards. The safety assessment of the proposed rack modules involves demonstration of thermal-hydraulic, criticality, and structural adequacy. Thermal-hydraulic adequacy requires that the fuel cladding withstand the imposed thermal stress and that the steady-state bulk pool temperature remain within prescribed limits. The criticality analyses show that the neutron multiplication factor (keff) for the stored fuel array is bounded by the MP3 limit of 0.945 (the USNRC limit is 0.95) under assumptions of 95% l probability and 95% confidence. Con:;equences ofinadvertent placeraent of a fuel assembly are also evaluated as part of the criticality analysis. The demonstration of structural adequacy of the rack modules shows that the free-standing modules and pool walls maintain the stored fuel within the configurations considered in the thermal-hydraulic and criticality analysis under all load conditions.
This document has been prepared for submission to the U.S. Nuclear Regulatory Commission for securing regulatory approval of the modification of the MP3 pool as proposed herein.
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HOLTEC INTERNATIONAL Millstone Point Unit 3 1-2 1'
Table 1.1 MP3 lilSTORIC AND PROJECTED FUEL DISCliARGE SCIIEDULE End-of-Cycle Bundles Permanently Total Number of Fuel Discharged Discharged Assemblies Date Discharged 1 75 75 10/87 2 85 160 5/89 3 79 239 2/91 4 68 307 7/93 5 109 416 4/95 6 85 501 3/99 /
7 84 585* 11/00 8 85 670 9/02 9 84 754 6/04 10 85 839 4/06 11 84 923 2/08 12 85 1,008 11/09 13 84 1,092 9/11 14 85 1,177 7/l3 15 84 1,261 4/15 16 85 1,346 2/17 17 84 1,430 12/18 18 85 1,515 9/20 19 84 1,599 7/22 20 85 1,684** 5/24 21 193 1,877 2/26
- Loss of Full-Core-Reserve with current storage capaciy t __
- Loss of Full-Core-Reserve with new racks instilled IlOLTEC INTERNATIONAL Millstone Point Unit 3 1-3
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I Tcbis1.2 AVAILAHLE STORAGE IN MP3 POOL AT PRESENT AND AFTER CAMPAIGN I EXTENSION Refueling Refueling Outage Discharge Size With Present Available Capacity Aj ;
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Tebl 1.3 PRESENTLY LICENSED PEER SITES WITII RACK DESIGNS SIMILAR TO TIIAT IN TIIIS APPLICATION Plant Docket Number Year Licensed Sequoyah 50-327 1994 50-328 Connecticut Yankee 50-213 1994 Fort Calhoun 50-285 , 1994 Salem 1 & 2 50-272 1994 50-311 Beaver Valley 50-334 1992 D. C. Cook 50-315 1992 50-316 Zion 1992 50-295 50-304 Three Mile Island 1 50-289 1990 IlOLTEC INTERNATIONAL Millstone Point Unit 3 1-5
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l 2.0 l
OVERVIEW OF PROPOSED CAPACITY EXPANSION i l
2.1 . General Description This section provides general information on the new storage modules proposed for the MP3 spent fuel pool. It also describes the basis for the detailed criticality, thermal-hydraulic, and seismic analyses presented in subsequent sections of this report.
The storage capacity expansion of the MP3 spent fuel pool features a two region arrangement. In the proposed scheme, a group of five modules will store the most reactive fuel (up to 5 weight % by I
. volume (w/o)) without any burnup limitation in a 3-out-of-4 configuration, with the fourth location blocked and empty of fuel. Fuel may be stored in these racks in a 4-out-of-4 configuration with an enrichment /burnup limitation. These racks will use a flux-trap design. The grouping of flux-trap racks is referred to as Region 1. The remaining ten racks do not use flux-traps and are collectively referred to as Region 2. Region 2 racks have an enrichment /burnup limitation on them. Figure 2.1 1
i shows the module layout, i The existing spent fuel storage racks are collectively referred to as Region 3. The existing racks are not moved or modified in any way by this rerack. As discussed in Section 4 of this report, the Region 3 racks will no longer credit Boraflex as a neutron absorber material.
Table 2.1 provides geometric and physical data for Region 1 and Regior, cells. The rack modules have five distinct sizes, denoted as types A, B, C, D, and E. Table 2.2 gives the number of cells in each of these rack types. As indicated in the table, the rerack would provide an additional 1,104 storage locations. The module dimensions and weights are presented in Table 2.3.
The proposed modules for the MP3 fuel pool are qualified as freestanding racks.
HOLTEC INTERNATIONAL Millstone Point Unit 3 2-1
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2.2 Design Basis
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This section describes the concepts and features that underlie.the design of the new MP3 rack modules." The key _ criteria are set forth in the classical USNRC memorandum entitled "OT Position for Review and. Acceptance of Spent Fuel Storage and Handling Applications", April 14,1570 as
. modified by amendment dated January 18,1979. The individual sections of this report expound on ;
the specific design bases derived from the above-mentioned "OT Position Paper". Nevertheless, a brief summary of the design bases for the MP3 racks are summarized in the following:
- a. Disposition: All new rack modules are required to be frw-+ nLg.
- b. ; Kinematic Stability: All free-standing modules must be kinematically stable (against tipping or overturning) when a seismic evrut that is 150% of the postulated SSE is imposed.
n c. Structural Compliance: All primary stresses in the rack modules must satisfy the limits postulated in Section III, subsection NF of the ASME Boiler and Pressure Vessel Code.- 4
- d. Thermal-Hydraulic Compliance: The spatial average bulk pool temperature is required to remain under 150'F in the wake of a normal refueling with single active failure of one train of spent fuel pool cooling. In addition to the limitations on the bulk pool temperature, the local water temperature in the MP3 pool must remain i subcooled (i.e., below the boiling temperature coincident with local elevated pressure conditions). j
- e. Criticality Comoliance: The flux-trap storage cells (Region 1) must be able to store j
' fresh Zircaloy clad fuel with 5 w/o initial enrichment in a 3 out-of-4 configurtion '
while maintaining the reactivity s 0.045. Region 2 cells must be able to store the Zircaloy clad fuel of 5.w/o enrichment and 39,000 MWD /MTU burnup while maintaining the reactivity s 0.945,
- f. Radiolocical Comoliance: The reracking of Millstone 3 must not lead to violation of the off-site dose limits, or adversely affect the area dose environment as set forth in
. the Millstone Unit 3 FSAR.
- g. Pool Structure: The ability of the reinforced concrete structure to satisfy the load combinations set forth in NUREG-0800, SRP 3.8.4 must be demonstrated.
- h. Rack Stress Fatigue: In addition to satisfying the primary stress criteria of Subsection
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HOLTEC INTERNATIONAL Millstone Point Unit 3 - 2-2
.1 NF, the alternating local stresses in the rack structure during a seismic event are also required to be sufficiently bounded such that the " cumulative damage factor" due to .;
- one SSE and five OBE events does not exceed 1.0. I
- i. Liner Integrity: The integrity of the liner under cyclic in-plane loading during a seismic event must be demonstrated.
- j. Bearina Pads: The bearing pads must be sufficiently thick such that the pressure on the liner continues to satisfy the ACI limits during and after a design basis seismic l event.
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- k. Accident Events: In the event of postulated drop events (uncontrolled lowering of a fuel assembly, for instance), it is necessary to demonstrate that the suberiticality of .
the rack structure and its thermal hydraulic adequacy are not compromised. l 1
\ l Construction Events: The field construction services required to be carried out for 1.
- executing the reracking must be demonstrated to be within the " state of proven art".
The foregoing design bases are further articulated in subsequent sections of this licensing l report.
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2.3 Codes, Standards, and Practices for the Spent Fuel Pool Modification ]
The design and fabrication of the rack modules is performed under a strict quality assurance program which meets 10CFR50 Appendix B requirements.
The following codes, standards and practices are used for all applicable aspects of the design, construction, and assembly of the spent fuel storage racks. Additional specific references related to detailed analyses are given in each section.
- a. Design Codes
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- 1. AISC Manual of Steel Construction, 8th Edition,1980 (provides detailed structural criteria for linear type supports).
- 2. ANSI N210-1976, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design).
i
- 3. American Society of Mechanical Engineers (ASME), Boiler and Pressure )
Vessel Code,Section III, Division 1,1995 Edition.
HOLTEC INTERNATIONAL i Millstone Point Unit 3 2-3
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- 4. ANSI /AISC-N690-1984 - Nuclear Facilities - Steel Safety Related Structure for Design, Fabrication and Erection.
, 5. - ASNT-TC-1 A, 1984 American Society for Nondestructive Testing (Recommended Practice for Personnel Qualifications).
- 6. ACI 349 Code Requirements for Nuclear Safety Related Concrete Structures.
- 1. ASME 'SA240 - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Fusion-Welded Unfired Pressure Vessels.
- 2. ASTM A262 - Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel. ,
- 3. ASME SA276 - Standard Specification for Stainless and Heat-Resisting Steel .
Bars and Shapes.
- 4. ASME SA479 - Steel Bars for Boilers & Pressure Vessels.
- 6. ASTM C992 - Standard Specification for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.
- 7. ASME SA312 - Specification for Seamless and Welded Austenitic Stainless l Steel Pipe.
- 8. ASME SA564 - Specification for Ilot Rolled and Cold-Finished. Age- '
Hardening Stainless and Heat Resisting Steel Bars and Shapes.
- 9. American Society of Mechanical Engineers (ASME), Boiler and Pressure >
- Vessel Code, Section Il-Parts A and C,1995.
l
- 10. ASTM A262 Practices A and E - Standard Recommended Practices for Detecting Susceptibility to Intergrannular Attack in Stainless Steels.
I 1. ASTM A380 - Recommended Practice for Descaling, Cleaning and Marking Stainless Steel Parts and Equipment.
HOLTEC INTERNATIONAL Millstone Point Unit 3 2-4
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- c. Weldine Codes
- 1. ASME Boiler and Pressure Vessel Code,Section IX -Welding and Brazing Qualifications,1995.
- 2. AWS Dl.1 - Welding Standards (1989).
- d. Ouality Assurance. Cleanliness. Packacine. Shinnine. Receivine. Storace. and Bandline Reauirements
- i. NQA-2-Part 2.2 1983 ' - Packaging, Shipping, Receiving, Storage, and Handling ofItems for Nuclear Power Plants (During Construction Phase).
- 2. NQA-1-1983 - Basic Requirements and Supplements. !
- 3. ASME Boiler and Pressure Vessel,Section V, Nondestructive Examination, .
1995 Edition.
- 4. ANSI - N45.2.6 - Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants (Regulatory Guide 1.58).
- e. Governinn NRC Desien Documents
- 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.
- 2. NRC Resolution of Generic Technical Activity A-36, July 1980, NUREG-0612, Control of Heavy Loads in Nuclear Power Plants.
- f. Other ANSI Standards (not listed in the precedine) -
- 1. ANSI /ANS 8.1 - 1983, Nuclear Criticality Safety in Operations with )
Fissionable Materials Outside Reactors. I
- 2. ANSI /ANS 8.7 - 1974, Guide for Nuclear Criticality Safety in the Storage of Fissile Materials.
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- 3. ANSI /ANS 8.11 - 1975, Validation of Calculation Me' hods for Nuclear Criticality Safety.
I HOLTEC INTERNATIONAL Millstone Point Unit 3 2-5 L
- g. Code-of-Federal Reuulations 1
l 1. 10CFR21 - Reporting of Defects and Non-compliance.
l 2. 10CFR50 - Appendix A - General Design Criteria for Nuclear Power Plants.
- 3. 10CFR50 - Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
l
- 4. 10CFR Part 20 - Radiation Protection Standards.
- 5. 29CFR Section 1910.401 - OSHA Standards for Commercial Diving Operations.
- h. Regulatory Guides
- 1. RG 1.13 - Spent Fuel Storage Facility Design Basis.
- 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 -
(endorses ANSI N45.2) - Quality Assurance Program Requirements, June,1972.
- 4. RG 1.29 - Seismic Design Classification.
]
- 5. RG 1.38 - (endorses ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants, March,1973.
- 6. RG 1.44 Control of the Use of Sensitized Stainless Steel.
- 7. RG 1.58 - (endorses ANSI N45.2.6) Qualification of Nuclear Power Plant inspection, Examination, and Testing Personnel. Rev.1, September,1980. i
- 8. RG 1.64 - (endorses ANSI N45.2.11) Quality Assurance Requirements for the Design of Nuclear Power Plants, October,1973. ;
- 9. RG 1.74 - (endorses ANSI N45.2.10) Quality Assurance Terms and Definitions, February,1974.
t l 10 RG 1.88 - (endorses ANSI N45.2.9) Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records. Rev. 2 October,1976.
l 11. RG 1.92 - Combining Modal Responses and Spatial Components in Seismic HOLTEC INTERNATIONAL Millstone Point Unit 3 2-6
Response Analysis.
- 12. RG 1.123.- (endorses ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement ofItems and Services for Nuclear Power Plants.
- 13. NRC Regulatory Guide 3.41 Rev., May 1977 - Validation of Calculation Methods for Nuclear Criticality Safety.
- 14. NRC . - Regulatory Guide 1.26 Rev. 3, Feb. 1976, Quality Group l
Classifications and Standards for Water, Steam and Radioactive Containing Components of Nuclear Power Plants.
- i. Branch Technical Position
- 1. CPB 9.1 Criticality in Fuel Storage Facilities.
L l
- 2. ASB 9 Residual Decay Energy for Light-Water Reactors for_ Long-Term l Cooling.
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! j. Standard Review Plan (NUREG-0800. July 1981)
- 1. SRP 3.7.1 - Seismic Design Parameters.
l
- 2. SRP 3.7.2 - Seismic System Analysis.
- 3. SRP 3.7.2 - Seismic Subsystem Analysis.
- 4. SRP 3.8.4 - Other Seismic Category 1 Structures (including Appendix D). 3 1 j l 5. SRP 9.1.2 - Spent Fuel Storage.
- 6. SRP 9.1.3 - Spent Fuel Pool Cooling and Cleanup System.
- k. Other MP3 Final Safety Analysis Report (FSAR).
i MP3 Technical Specification.- ;
I i~ ' NRC Bulletin 96-02," Movement of11eavy Leads Over Spent Fuel, Over Fuel in the i Reactor Core, or Over Safety-Related Equipment", April 11,1996.
HOLTEC INTERNATIONAL
[- - Millstone Point Unit 3 2-7 L-
G 2.4 Quality Assurance Program The governing quality assurance requirements for fabrication of the MP3 spent fuel racks are enunciated in 10CFR50 Appendix B The quality assurance program for design of the Millstone
' Unit 3 racks are described in Holtec's Nuclear Quality Assurance Manual, which has been reviewed T '
and approved by Northeast Utilities (NU). This program is designed to provide a flexible but highly controlled system for the design, analysis and licensing of customized components in accordance 1
- with various codes, specifications, and regulatory requirem:nts.
The manufacturing of the racks will be carried out by Holtec's designated manufacturer (U.S. Tool
& Die, Inc.). The Quality Assurance System enforced on the manufacturer's shop floor shall provide for all controls necessary to fulfill all quality assurance requirements with sufficient simplicity to make it functional on a day-to-day basis. UST&D has manufactured high density racks for over 60
' nuclear plants around the world._UST&D has been audited by the industry group NUPIC, and the
-QA branch of NMSS with most satis ractory results.
The Quality Assurance System that will be used by Holtec to install the racks is also controlled by the Holtec Nucicar Quality Assurance Manual and by NU's site-specific requirements.
2.5 Mechanical Design l
The Millstone Unit 3 rack modules are designed as cellular structures such that each fuel cell has a prismatic square opening with conformal lateral support and a flat horizontal bearing surface.
i Figures 2.2 and 2.3 show pictorial views of Region 1 and Region 2 modules, respectively. As can
-be inferred from these schematic representations, the high density modules for MP3 have been designed to simulate multi-flange beam structures resulting in excellent detuning characteristics with respect to the applicable seismic events.
HOLTEC INTERNATIONAL Millstone Point Unit 3 - 2-8
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l Figure 2.4 provides ar elevation view of the Region 1 and Region 2 racks Iccated in the Spent Fuel
)1 Pool. l l
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Millstone Point Unit 3 2-9
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IIOLTEC INTERNATIONAL Millstone Point Unit 3 2-10 i i
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Table 2.2 MODULE DATA FOR RERACK CAMPAIGN I l
! I MODULE NUMBER OF CELLS l 1.P. QTY. North-South East-West Total Per Total No. of Cells for Direction Direction Rack this Rack Type A 5 9 9 81 405 B 1 9 10 90 90 C 1 7 10 70 70 l
D 5 7 10 70 350 E 3 7 9 63 189 TOTAL: 15 - - - 1,104 l
l HOLTEC INTERNATIONAL Millstone Point Unit 3 2-11 l
Table 2.3 MODULE DIMENSIONS AND WElGHTS FOR RERACK CAMPAIGN I Dimension (inches)'
Module I.D. Shipping Weight in Pounds North-South East-West A 81.53 81.53 13,090 B 81.53 90.54 14,410 C 63.49 90.54 11,490 D 69.33 103.38 18,085 E 63.49 81.53 10,450 i
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Nominal rectangular planform dimensions.
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3.0 FABRICATION, MATERIALS, AND HEAVY LOADS CONSIDERATIONS l 3.1 Rack Fabrication The object of this section is to provide a self-contained description of rack module construction and to enable an independent appraisal of the adequacy of design.
3.1.1 Fabrication Objective The are four interrelated manufacturing requirements for MP3's high density storage racks, 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 flow 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 seismic input motion (OBE and SSE).
3.1.2 Rack Module for Region 1 This section describes the Region I fabrication sequence.
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. Figure 3.1 shows the box.
The minimum weld penetration is@af the @ inch box metal gage. A die is used to flare out one end of the box to provide the tapered lead-in (Figure 3.2). Three-quarter inch diameter holes IIOLTEC INTERNATIONAL Millstone Point Unit 3 3-1
=
i are punched on all four sides near the other end of the box to provide the requisite auxiliary flow holes. Each box constitutes a storage location. Each external side of the box is equipped with a stainless steel sheath that holds one integral Boral sheet (poison material).
Thh design objective calls for attaching Boral tightly on the box surface. This is accomplished by die forming the internal and external bo'x sheathings, as shown in Figure 3.3. The flanges of the
. sheathing are attached 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.
. Having fabricated the required number of composite box assemblies, they are joined together in a i
fixture using' connector elements in the manner shown in Figure 3.4. Figure 3.5 shows an elevation view of two storage cells of a Region I rack module. A representative connector element is also shown in the figure. Joining the cells by the connector elements results in a well- defined shear flow path and essentially makes the box assemblage into a multi-flanged beam-type structure. The
' " baseplate"is attached to the bottom edge of the boxes. The baseplate is ? 4 inch thick austenitic
]
stainless . steel plate stock which has @ inch diameter holes (except lin locations, which are rectangular) cut out in a pitch identical to the box pitch. The baseplate is attached to the cell assemblage by fillet welding the box edge to the plate.
In the final step, adjustable leg supports (shown in Figure 3.6) a're welded to the underside of the
- baseplate. The adjustable legs provide aginch vertical height adjustment at each leg location.
i Appropriate NDE (nondestructive examination) occurs on all welds including visual examination of '
sheathing welds, box longitudinal seam welds, box-to-baseplate welds, and box-to-box connection !
l welds; and liquid penetrant examination of support leg welds, in accordance with the design drawings.
1 3.1.3 Rack Module for Region 2 Region'2 storage cell locations have a single poison panel between adjacent box wall surfaces.
There are five significant components (discussed below) of the Region 2 racks: (1) the storage box
, HOLTEC INTERNATIONAL Millstone Point Unit 3 32 i!
1 l'
subassembly (2) the baseplate, (3) the neutron abso'ber r material, (4) the sheathing, and (5) the
/ support legs.
1 i 1, Storane cell bcx subassemb!v: As described for Region 1, the boxes are fabricated
[
l from two precision formed channels by seam welding inla machine equipped with copper chill bars and pneumatic clamps to minimize distortion due to welding heat
' input. Figure 3.1 shows the box.
. Each box has four lateral holes punched near its bottom edge to provide auxiliary flow holes. A sheathing is attached to each side of the box with the poison material installed in the sheathing cavity. The edges of the sheathing and the box are welded together to fonn a smooth edge. The box, with integrally connected sheathing, is referred to as the " composite box".
The composite boxes are arranged in a checkerboard array to form an assemblage of storage cell locations (Figure 3.7). Filler panels and corner angles are welded to the edges of boxes at the outside boundary of the rack to make the peripheral formed.
cells. The inter-box welding and pitch adjustment are accomplished by small longitudinal connectors. This assemblage of box assemblies is welded edge-to-edge as shown in Figure 3.7, resulting in a honeycomb structure with axial, flexural and torsional rigidity depending on the extent of intercell welding provided It can be seen from Figure 3.7 that two edges of each interior box are connected to the contiguous boxes resulting in a well-defmed path for " shear flow".
- 2. Baseolate: The baseplate provides a continuous horizontal surface for supporting the !
fuel assemblies. IThe baseplate has a @ inch diameter hole (except lift locations which are rectangular) in each cell location as described in the preceding section.
The baseplate is attached to the cell assemblage by fillet welds. g
- 3. The Neutron Absorber Material: As mentioned in the preceding section, Boral is ;
used as the neutron absorber material. I
- 4. Sheathing: As described earlier, the sheathing serves as the locator and retainer of the poison material.
)
- 5. Sunoort lens: As stated earlier, all support legs are the adjustable type (Figure 3.6). l The top position is made af austenitic steel material. The bottom part is made of 17:4 Ph series stainless steel to avoid galling problems. !
Each support leg is equipped with a readily accessible socket to enable remote leveling of the rack after its placement in the pool, p An elevation view of three contiguous Region 2 cells is shown in Figure 3.8.
HOLTEC INTERNATIONAL Millstone Point Unit 3 3-3
3.2 Material Considerations 3.2.1 Introduction
- Safe storage of nuclear fuel requires that the materials utilized in the fabrication of racks be of proven durability and be compatible with the pool water environment. This section provides the necessary information on this subject.
i i~
' 3.2.2! Structural Materials The following structural materials are utilized in the fabrication of the spent fuel racks:
- a. ASME SA240-304L for all sheet metal stock,
- b. Internally threaded support legs: ASME SA240-304L.
- c. Externally 1 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 R308L.
J 3.2.3 Poison Material In addition'to the structural and non-structural stainless material, the racks employ Boral*, a patented product of AAR Advanced Structures, 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 light-weight metal with high tensile strength which is !
protected from corrosion by a highly resistant oxide film. The two materials, boron carbide and aluminum, are chemically compatible and ideally suited for long-term use in the radiation, thermal and chemical environment of a nuclear reactor or the spent fuel pool. !
Boral's use in spent fuel pools as the neutron absorbing material can be attributed to the following reasons:
HOLTEC INTERNATIONAL i
Millstone Point Unit 3 4 m
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~
- i. The content and placement of boron carbide provides a very high removal cross
- l. . section for thermal neutrons.
L l l-ii. Boron carbide, in the form of fine particles, is homogeneously dispersed throughout L the central layer of the Boral panels.
l --
iii. ' The boron carbide and aluminum materials in Boral are totally unaffected by long-L. term exposure to radiation.
i.
- iv. The neutron absorbing central layer of Boral is clad with permanently bonded i
surfaces of aluminum.
l 'v. Boral is stable, strong, durable, and corrosion resistant.
Holtec International's QA program ensures that Boral is manufactured by AAR Brooks & Perkins under the control and surveillance of a Quality Assurance / Quality Control Program that conforms to l the requirements of 10CFR50 Appendix B, " Quality Assurance Criteria for Nuclear Power Plants".
As indicated in Table 3.1, Boral has been licensed by the USNRC for use in numerous BWR and PWR spent fuel storage racks and has been extensively used in overseas nuclear installations.
l l Boral Material Characteristics i 4
r l . Aluminum: Aluminum is a silvery-white, ductile metallic element that is the most abundant in the
- earth's crust. The 1100 alloy aluminum is used extensively in heat exchangers, pressure and storage ;
' tanks, chemical equipment, reflectors and sheet metal work. It has high resistance to corrosion in L
industrial and marine atmospheres.' The physical, mechanical and chemical properties of the 1100 alloy aluminum are listed in Tables 3.2 and 3.3.
l 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 HOLTEC INTERNATIONAL Millstone Point Unit 3 3-5
m i
loss of metal from general corrosion or pitting corrosion and the film remains stable between a pH range of 4.5 to 8.5.'
Boron Carbide:' The boron carbide contained in Boral is a fine granulated powder that conforms to -
1 ASTM C-750-80 nuclear grade Type III. The particles range in size between 60 and 200 mesh and the material conforms to the chemical composition and properties listed in Table 3.4.
3.2.4 L Compatibility with Coolant
]
l-All materials used in the construction of the MP3 racks have an established history ofin-pool usage.
_.Their physical, chemical and radiological compatibility with the pool environment has been established throughout the industry. As noted in Table 3.1, Boral has been used in both vented and l l ;
! unvented configurations in fuel pools with equal success. Austenitic stainless steel is the most widely used stainless anoy in nuclear power plants.
3.3 Heavy Load Considerations for the Proposed Reracking Operation l
l A 10-ton crane will be utilized for handling all heavy loads in the reracking operation. A remotely
- engageable lift rig, meeting NUREG-0612 stress criteria, will be used to lift the new modules. It consists of independently loaded lift rods with a " cam type" lift configuration. This ensures that failure of one traction rod will not result in uncontrolled lowering of the load; compliant with Section 5.1.6(1) of NUREG-0612. The remotely engageable lift rig also has the following
. attributes:
- a. The stresses in the lift rods.are self limiting inasmuch as an increase in the magnitude of the load reduces the eccentricity between the upward force 'and downward reaction (moment arm),
- b. It is impossible for a traction rod to' lose its engagement with the rig in locked position due to the load of the lifted rack pulling each traction rod in the downward direction, thus keeping it within its locking slots. Moreover, the locked configuration can be directly verified from above the pool water without the aid of an underwater camera due to the orientation of position locator flags atop each traction rod.
l HOLTEC INTERNATIONAL
. Millstone Point Unit 3 3-6 k
- c. The stress analysis of the rig is carried out and the primary stress limits postulated in ANSI 14.6 (1978) are shown to be met,
- d. The rig is load tested with 300h c#the maximum weight to be lifted. The test weight ,
is maintained in the air for one hour. All critical weld joints are liquid penetrant l examined, after the load test, to establish the soundness of all criticaljoints. )
)
Pursuant to the defense-in-deyh approach of NUREG-0612, the following additional measures of safety will be undertaken for the reracking operation.
)!
- i. The cranes and lifting devices used in the project will be given a preventive maintenance checkup and inspection per the MP3 procedures before beginning the reracking operation.
ii. Safe load paths will be developed for moving the new racks in the Fuel Building.
The "new" racks will not be carried over any regien of the pool containing fuel or safe shutdown equipment.
iv. I The rack upending will be carried out in an area which is not poolside and will be i qualified for a postulated rack drop from 6 feet elevation. Additionally, this area will not be overlapping to any safety related component,
- v. All crew members involved in the reracking operation will be given training in the use of the lifting, upending equipment, and all other aspects of the reracking operation.
In addition to the above design, testing, and operation measures, the consequences of a postulated rack drop were also considered on the integrity of the pool structure. The following analysis was performed.
- a. The heaviest rack module was postulated to free fall from the top of the water surface level to the pool floor,
- b. The fall of a rack is assumed to occur in its normal vertical configuration which ,
minimizes the retarding effect of water drag.
- c. The falling rack is assumed to impact the pool slab undergoing an elastic / plastic impact.
HOLTEC LNTERNATIONAL Millstone Point Unit 3 3-7 l
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l The results of these calculations show that the maximum additional load on the pool structure is less than the capacity of the slab. Therefore, the integrity of the pool structure under the postulated rack drop event is ensured.
l The fuel shufCe scheme developed for the spent fuel pool corresponding to the rack change-out presented in thd preceding section is predicated on the following criteria:
- 1. No heavy load (rack or rig) with a potential to drap on a rack shall be carried over or near active fuel. This shall be accomplished by shuffling fuel into racks that are not in the area of the safe load path.
L 2. All heavy loads are lined in such a manner that the C.G. of the lift point is aligned with the C.G. of the load being lifted.
- 3. Turnbuckles are utilized to " fine tune" the verticality of the rack being lined.
All phases of the reracking activity will be conducted in accordance with written procedures which will be reviewed and approved in accordance with MP3 procedures.
l The guidelines contained in NUREG-0612, Section 5 will be followed throughout the reracking i l
activity. The guidelines of NUREG-0612 call for measures to " provide an adequate defense-m-depth for handling of heavy loads near spent fuel..." and cite four major causes of load handling accidents, namely
- i. operator errors ii. rigging failure iii. lack of adequate inspection iv. inadequate procedures The MP3 rack expansion program ensures maximum emphasis on mitigation of the potential load L drop accidents by implementing measures that will eliminate a possible accident during 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.
~
HOLTEC INTERNATIONAL l
Millstone Point Unit 3 38 1:
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g Pperator errors: As mentioned above, MP3 plans to provida comprehensive traming to the.
i,stallation crew.
.Rianing failure: The lining <.i.e designed for handling and installation of the racks in the MP3 fuel pool has redundancies in ,ne lin legs, and lift eyes 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 - 1978, including compliance with'the primary stress criteria, load testing at 300% of maximum lift load, and dye examination of critical welds.
The MP3. rig' design is simi!. r to the rigs used in the rerack of numerous other plants, such as Sequoyah, Zion, Salem, Thru Mile Island Unit 1, D.C. Cook, and Connecticut Yankee.
Lack of adeauate insoection: The designer of the racks will~ develop a set of inspection points which have proven to have eliminated any incidence of re-work or erroneous installation in numerous prior rerack projects.' Inspection oflining equipment will be performed per NUREG-
- 0612.
Inadeauate procedures: MP3 plans a multitude of procedures to cover the entire rerack effort, such as' mobilization, rack handling, upending, lining, installation, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. Procedures for installation of new racks will be developed.
The series of operating procedures planned for MP3 rerack are the successors of the procedures implemented successfully in other projects.
In addition to the above, a complete inspection and preventive maintenance program of all the cranes and lifting equipment used in the project prior to the start of reracking are planned. Safe load paths will be developed as required by NUREG-0612.
Table 3.5 provides a synopsis of the requirements delineated in NUREG-0612, and our intended compliance.
t In summary, the measures implemented in MP3 reracking are identical to those utilized in all recent reracks in the U.S., none of which has experier.ced any mishaps or reportable condition.
HOLTEC INTERNATIONAL Millstone Point Unit 3 3-9 I
L - - -
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Table 3.1 BORAL EXPERIENCE LIST (Domestic and Foreign)
PRESSURIZED WATER REACTORS Vented Const- Mfg.
Plant Utility ruction Year Bellefonte 1,2 Tennessee Valley Authority No 1981 Donald C. Cook Indiana & Michigan Electric No 1979 Indian Point 3 NY Power Authority Yes 1987 Maine Yankee Maine Yankee Atomic Power Yes 1977 Salem 1,2 Public Service Electric & Gas No 1980 Sequoyah 1,2 Tennessee Valley Authority No 1979 Yankee Rowe Yankee Atomic Power Yes 1964/
1983 Zion 1,2 Commonwealth Edison Yes 1980 Company Byron 1,2 Commonwealth Edison Yes 1988 Company Braidwood 1,2 Commonwealth Edison Yes 1988 Company Yankee Rowe Yankee Atomic Electric Yes 1988 Three Mile Island GPU Nuclear Yes 1990 Sequoyah (rerack) Tennessee Valley Authority Yes 1992 Salem 1,2 Public Service Electric & Gas Yes 1994 Donald C. Cook American Electric Power Yes 1992 (rerack)
BOILING WATER REACTORS Browns Ferry 1,2,3 Tennessee Valley Authority Yes 1980 Brunswick 1,2 Carolina Power & Light Yes 1981 Clinton Illinois Power Yes 1981 HOLTEC I'ITERNATIONAL Millstone Point Unit 3 3-10
Tcbl:3.1 BORAL EXPERIENCE LIST (Domestic and Foreign)
' Cooper Nebraska Public Power Yes 1979 Dresden 2,3.- Commonwealth Edison Co. Yes 1981 Duane Arnold Iowa Electric Light and No 1979 Power J.A. FitzPatrick NY Power Authority No 1978 E.I. Hatch 1,2 Georgia Power Yes 1981 Hope Creek Public Service Electric & Gas Yes 1985 Ilumboldt Bay Pacific Gas and Electric Yes 1986 Lacrosse Dairyland Power Yes 1976 Limerick 1,2 PECO Nuclear No 1980 Limerick 2 PECO Nuclear Yes 1994 Monticello Northern States Power Yes 1978 Peach Bottom 2,3 PECO Nuclear No 1980 Perry 1,2 Cleveland Elec. Illuminating No 1979 Pilgrim Boston Edison No 1978 Susquehanna 1,2 Pennsylvania Power & Light No 1979 Vermont Yankee Vermont Yankee Atomic Yes 1978/
Power 1986 l
l Ilope Creek Public Service Electric & Gas Yes 1989 l
l Shearon Harris Carolina Power & Light Yes 1991 Pool B Duane Arnold Iowa Electric Light & Power Yes 1993 Pilgrim Boston Edison Company Yes 1993 LaSalle Unit 1 Commonwealth Edison Yes 1992 Company c IlOLTEC INTERNATIONAL Millstone Point Unit 3 3-11
Tcble 3.1 (continu:d)
FOREIGN INSTALLATIONS USING BORAL England 1 PWR Plant Nuclear Electric plc.
France 12 PWR PIants Electricite de France South Korea Ulchin 1,2 KEPCO Kori 4 KEPCO I Yonggwang 1,2 KEPCO South Africa
)
Koeberg 1,2 ESCOM Switzerland Beznau1,2 Nordostschweizerische Gosgen Kraftwerke AG Kernkraftwerk Gosgen-Daniken AG Taiwan Chinshan 1,2 Taiwan Power Cornpany Kuosheng 1,2 Taiwan Power Company Mexico Laguna Verde Comision Federal de Electricidad Units 1,2 l
l HOLTEC INTERNATIONAL Millstone Point Unit 3 3-12 L
i l
Tcble 3.2 1
,-- )
1100 ALLOY ALUMINUM PHYSICAL PROPERTIES Density 0.098 lb/cu. in.
2.713 gm/cc Melting Range 1190-1215 deg. F 643-657 deg. C Thermal Conductivity (77 deg. F) 128 Btu /hr/sq ft/deg. F/R 0.53 cal /sec/sq cm/deg. C/cm Coefficient of Thermal Expansion 13.1 x 10-6i n/in., F (68-212 deg. F) 23.6 x 10-6 cm/cm, C Specific heat (221 deg. F) 0.22 Btu /lb/deg. F !
0.23 cal /gm/deg. C Modulus of Elasticity 10x106 psi Tensile Strength (75 deg. F) 13,000 psi annealed 18,000 psi as rolled Yield Strength (75 deg. F) 5,000 psi annealed 17,000 psi as rolled Elongation (75 deg. F) 35-45% annealed 9-20% as rolled l Hardness (Brinell) 23 annealed 32 as rolled Annealing Temperature 650 deg. F 343 deg. C i
I IlOLTEC INTERNATIONAL Millstone Point Unit 3 3-13
Tr.ble 3,3 CHEMICAL COMPOSITION - -
1100 ALLOY ALUMINUM 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. others each l
I l
HOLTEC INTERNATIONAL l- Millstone Point Unit 3 3-14 i
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Trble 3.4 BORON CARBIDE CHEMICAL COMPOSITION, WElGHT %
Total boron 70.0 min.
B' isotopic content in natural 18.0 boron Boric oxide 3.0 max.
Iron 2.0 max.
Total boron plus total carbon 94.0 min.
BORON CARBIDE PHYSICAL PROPERTIES Chemical formula BC4 Boron content (weight) 78.28 %
Carbon content (weight) 21.72 %
Crystal Structure tombohedral Density 2.51 gm/cc 0.0907 lb/cu.in.
Melting Point 2450 C-4442 F Boiling Point 3500 C-6332 F Microscopic Capture cross-section 600 barn l
I l~
IlOLTEC INTERNATIONAL l
Millstone Point Unit 3 3-15 1 1
Table 3.5 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612)
Criterion Compliance
- 1. Are safe load paths defined for the movement of heavy Yes loads to minimize the potential ofimpact, if dropped on irradiated fuel and safe shutdown equipment? -
- 2. Will procedures be developed to cover: identification of Yes required equipment, inspection, and acceptance criteria required before movement ofload, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?
- 3. Will crane operators be trained and qualified? -Yes
- 4. Will special lifting devices meet the guidelines of ANSI Yes 14.6-19787 5, Will non-customer lifting devices be installed and used Yes in accordance with ANSI B30.9-19717
- 6. Will the cranes be* inspected and tested prior to use in Yes -
L rerack?
- 7. Does the crane nieet the intent of ANSI B30.2-1976 and Yes CMMA-707 I
l HOLTEC INTERNATIONAL Millstone Point Unit 3 3-16
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I I' i N BASEPLATE CONCENTRIC HOLE ELEVATION ~ VIEW OF REGION FIGURE 3.5; 1 RACK 84D $
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( NON-FLUX TRAL' CONSTRUCTION )
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g 4.0 CRITICALITY SAFETY EVALUATION
~
- 4.1 - ~ DESIGN BASES The high density spent fuel storage racks for Millstone Unit 3 are designed to assure that the
. ~ effective neutron multiplication factor, k,g, is equal to or less than 0.945 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with un-borated water at a temperature _within the operating range corresponding to the highest reactivity. Including all applicable uncertainties, the maximum k,gis shown to be less than or equal to 0.945 with a 95%
probability at a 95% confidence level [4.1.1]. Reactivity effects of abnormal and accident conditions have also beer evaluated to assure that under credible abnormal and accident conditions, the reactivity will not exceed 0.945.
L i
Applicable codes, standards, and regulations or pertinent sections thereof, include the following:
. . Code offederal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 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.
. L.I. Kopp," Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," June 1998.
e 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.
HOLTEC INTERNATIONAL
' Millstone Point Unit 3 4-1
I e
ANSI /ANS-57.2-1983, Design Requirements for Light Water Reactor Spent Fuel l
Storage Facilities at Nuclear Power Plants. '
l USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication factor, kett , including bias, uncertainties, and calculational statistics, shall be less l
than or equal to 0.95, with 95% probability at the 95% confidence level. In the present criticality l
safety evaluation, the design limit was assumed to be 0.945, which is more conservative than the limit specified in the regulatory guidelines.
u
- To ensure that the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made:
e Moderator is un-borated water at a temperature within the operating range that results in the highest reactivity.
6 The racks were assumed to be fully loaded with the most reactive fuel authorized to be stored in the racks without any control rods or burm ble poison, such as Integral Fuel Bumable Absorber (IFBA) rods.
- No soluble poison (boron) is assumed to be present in the pool water under normal operating conditions.
+ Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.
The effective multiplication factor of an infinite radial array of fuel assemblies was used except for the assessment of penpheral effects and certain abnormal / accident conditions where neutron leakage is inherent.
. . In-core depletion calculations assume conservative operating conditions, highest fuel and moderator temperature, and an allowance for the soluble boron concentrations during in-core operations.
- All Region 3 analyses assume that the Boraflex is replaced by water, and thus, no credit is taken for neutron absorption in the Botaflex panels.
l '
! The spent fuel storage racks are designed to accommodate the fuel assembly types listed in Table i-HOLTEC INTERNATIONAL
! Millstone Point Unit 3 4-2
r 2
4.1.1 with a maximum enrichment of 5 wt% "U. Although the two assembly types listed in i Table 4.1.1 are nearly identical, differing only in the guide and instrument tube dimensions, the L
Westinghouse 17x17 Vantage 5H (V5H) assembly was determined to have very slightly higher
{
reactivity. Therefore, the V5H assembly was used as the design basis fuel assembly.
Three separat, sarage regions are provided in the spent fuel pool. The independent acceptance criteria for storage in each of the regions are as follows: i l => Region 1 is designed to accommodate (1) new un irradiated fuel assemblies with a maximum 2
[ nominal enrichment of 5.0 wt% "U in a 3-out-of-4 arrangement with the fourth cell empty i
and blocked and (2) fuel assemblies in a 4-out-of-4 arrangement (unrestricted) with a 2
maximum nominal enrichment of 5.0 wt% "U which have accumulated a minimum burnup j of 8.0 mwd /kgU or fuel of initial enrichment and burnup combinations within the acceptable domain depicted in Figure 4.1.1.
=> Region 2 is designed to accommodate fuel assemblies with a maximum nominal enrichment 2
of 5.0 wt% "U which have accumulated a minimum burnup of 39.0 mwd /kgU or fuel of j initial enriclunent and burnup combinations within the acceptable domain depicted in Figure L 4.1.2.
t
=> Region 3 is designed to accommodate fuel assemblies with a maximum nominal enrichment 2
l of 5.0 wt% "U which have accumuleted minimum burnup and cooling times that fall within the acceptable domains depicted in Figure 4.1.3. ;
l The water in the spent fuel storage pool normally contains soluble boron which would result in a l l large sub-criticality margin under actual operating conditions. However, the NRC guidelines, I based upon the accident condition in which all soluble poison is assumed to have been lost, specify 11 t the limiting k, of 0.95 for normal storage be evaluated for the accident condition that i assumes the loss of soluble boron. The double contingency principle of ANSI N-16.1-1975 and of the April 1978 NRC letter allows credit for soluble boron under other abnormal or accident conditions, since only a single independent accident need be considered at one time. !
Consequences of abnormal and accident conditions have also been evaluated, where " abnormal" i HOLTEC INTERNATIONAL Millstone Point Unit 3 4-3
P 1 i l f j l
refers to conditions which may reasonably be expected to occur during the lifetime of the plant l ,and " accident" refers to conditions which are not expected to occur but nevertheless must be
)
protected against.
l l
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i t
HOLTEC INTERNATIONAL Millstone Point Unit 3 44 i
i
IL b
l-P 4.2
SUMMARY
OF CRITICALITY ANALYSES I . .. . .
E
' 4.2.1 ' Normal Ooeratina Conditionsi H
. The criticality analyses for each of the three separate regions'of the spent fuel storage pool are
- summarized in Tables 4.2.1,4.2.3, and 4.2.5, for the design basis storage conditions. For the .
l acceptance criteria defined in the previous section, the maximum k,y values are shown to be less than or equal to 0.945 (95% probability at the 95% confidence level) in each of the three regions.
L 4.2.1.1 . Region-.ll l
l Calculations have been performed to qualify the Region 1 racks for storage of new un-irradiated fuel assemblies with a maximum nominal enrichment of 5.0 wt% 2"U in a 3-out-of-4 3 arrangement with the fourth cell empty and blocked and in a'4-out-of-4 arrangement -
(unrestricted) with initial enrichment and burnup combinations within the acceptable domain depicted in Figure 4.1.1. The criticality analyses for Region 1 of the spent fuel storage pool are summarized in Table 4.2.1, and demonstrate that for the defm' ed acceptance criteria, the a maximum k,,is less than 0.93.
The data points shown in Figure 4.1.1 are tabulated in Table 4.2.2. For convenience, the l- minimum (limiting) burnup data may be described as a function of the nominal initial enrichment, E, in wt% 2"U by a bounding polynomial expression as follows:
i B = -0.6667xE + 12.093xE - 35.798.
where B is the minimum burnup in mwd /kgU and E is the enrichment inc vt% 2"U (for initial enrichments up to 5.0 wt% 2"U). Alternatively, because the data are nearly linear, linear interpolation between the points listed in Table' 4.2.2 is also acceptable, f
- - HO.LTEC INTERNATIONAL Millstone Point Unit 3 4-5 1
1 1
L ]
)
L4.2.1.1.1 Interface Between Storage Arrangements The two different storage arrangements that are available in the Region I racks (i.e.,3-out-of-4 and 4-out-of-4 ) may be utilized in any of the Region I racks, including both arrangements in a
- single rack', provided the following interface requirement is met; the row in the 3-out-of-4
- storage area bordering the interface between adjacent 3-out-of-4 and 4-out-of-4 storage areas must contain altemating cell blockers. The interface requirement is illustrated in Figure 4.2.1. A calculation was performed to demonstrate that such an arrangement is less reactive than either of the individual arrangements alone.
During the rack installation, cell blocking devices will be installed in a manner consistent with the aforementioned requirement. The interface requirement will be ensured through administrative procedures. A cell blocking device may be removed, provided all adjacent and diagonal fuel assemblies around the cell blocking device are removed beforehand.
- 4.2.1.2 Region 2 Calculations have been performed to qualify the Region 2 racks for storage of fuel assemblies 2
with a maximum nominal initial enrichment of 5.0 wt% "U which have accumulated a minimum bumup of 39.0 mwd /kgU or fuel of initial enriclunent and burnup combinations within the acceptable domain depicted in Figure 4.1.2. The criticality analyses for Region 2 of the spent fuel storage pool are summarized in Table 4.2.3, and demonstrate that for the defined acceptance criteria, the maximum k,,is less than 0.N.
The calculated maximum reactivity in Region 2 includes the reactivity effect of the axial
. distribution in burnup and provides an additional margin of uncertainty for the depletion calculations. The data' points shown in Figure 4.1.2 are tabulated in Table 4.2.4. For HOLTEC INTERNATIONAL Millstone Point Unit 3 4-6 l i
i j
convenience, the minimum (limiting) burnup data may be described as a function of the nominal initial enrichment, E, in wt% 235U by a bounding polynomial expression as follows:
B = -0.4608xE' + 6.641xE' - 34.854xE + 90.385xE - 83.40.
where B is the minimum burnup in mwd /kgU and E is the enrichment in wt% 235U (for initial enrichments from 2.0 to 5.0 wt% 235U). Fuel assemblies with enrichments less than 2.0 wt%
235 U will conservatively be required to meet the burnup requirements of 2.0 wt% 235U assemblies as shown in Fig 4.1.2. Alternatively, because the data are nearly linear, linear interpolation between the points listed in Table 4.2.4 is also acceptable.
4.2.1.3 Region 3 Calculations have been performed to qualify the existing Westinghouse designed racks, referred to herein as Region 3 racks, for storage of fuel assemblies with a maximum nominal initial 23 emichment of 5.0 wt% 'U which have accumulated minimum burnup and cooling times that fall within the acceptable domains depicted Figure 4.1.3. The criticality analyses for Region 3 of the spent fuel storage pool are summarized in Table 4.2.5 and demonstrate that the maximum k, )
is equal to 0.945, which conforms to the defined acceptance criterion.
l The calculated maximum reactivity in Region 3 includes the reactivity effect of the axial distribution in burnup and provides an additional margin of uncertainty for the depletion calculations. The data points shown in Figure 4.1.3 are tabulated in Table 4.2.6. For j a
convenience, the minimum (limiting) burnup data for each of the cooling times shown in Figure 4.1.3 may be described as a function of the nominal initial enrichment, E, in wt% 235U by bounding polynomial expressions as follows:
Cooling Time (years) Polynomial Expression L
i IIOLTEC INTERNATIONAL I Millstone Point Unit 3 4-7
!I
l O B = 0.1000xE 2 + 14.jgg6xg.17.5390 5 B = 0.4651xE2 + 10.0120xE- 11.3919 10 B = 0.6730xE2+ 7.8408xE - 8.3853.
l 20 B = 0.6151xE2 + 7.3547xE- 7.9121 j
where B is the minimum burnup in mwd /kgU and E is the enrichment in wt% 2"U (for initial enrichments from 2.0 to 5.0 wt% 2"U). Fuel assemblies with enrichments less than 2.0 wt%
2"U will conservatively be required to meet the burnup requirements of 2.0 wt% 2"U assemblies as shown in Fig 4.1.3. Alternatively, because the data are nearly linear, linear 1
interpolation between the points listed in Table 4.2.6 is also acceptable. )
I I
The burnup criteria identified above for acceptable storage in each of the three regions will be - !
1 implemented by appropriate administrative procedures to ensure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2.
4.2.2 Abnormal and Accident Conditions I Although credit for the soluble poison normally present in the spent fuel pool water is permitted
.under abnormal or accident conditions, most abnormal or accident conditions will not result in exceeding the limiting reactivity even in the absence of soluble poison. The effects on reactivity of credible abnormal and accident conditions are discussed in Section 4.8 and summarized in Tables 4.2.7 and 4.2.8. Strict administrative procedures to assure the presence of soluble poison l
will preclude the possibility of the simultaneous occurrence of the two independent accident conditions.
The inadvertent misplacement of a fresh fuel assembly has the potential for exceeding the limiting l reactivity, should there be a concurrent and independent accident condition resulting in the loss'of all soluble poison. Assuring the presence of soluble poison during fuel handling operations will HOLTEC INTERNATIONAL Millstone Point Unit 3 4-8 L
preclude the possibility of the simultaneous occurrence of the two independent accident conditions. The largest reactivity increase would occur if a fresh fuel assembly of 5.0 wt% 2"U enrichment were to be inadvertently loaded into an empty cell in Region 3 with the remainder of the rack fully loaded with fuel of the highest permissible reactivity. Under this accident condition, credit for the presence of soluble poison is permitted by the NRC guidelines .
Calculations indicate that 800 ppm soluble boron, that is to be required by the Technical Specifications during fuel handling operations, is more than adequate to assure that the limiting k, of 0.945 is not exceeded.
With the assumption that the Boraflex panels are replaced by water, the moderator temperature I coefficient of reactivity in Region 3 is positive. Therefore, an increase in the spent fuel pool temperature above the normal operating conditions (i.e., above 160 F), has the potential for exceeding the limiting reactivity in Region 3, should there be a concurrent and independent accident condition resulting in the loss of all soluble poison. The largest reactivity increase would occur if boiling took place in Region 3. with the remainder of the rack fully loaded with fuel of the highest permissible reactivity. Calculations indicate that 100 ppm soluble boron is more than adequate to assure that the limiting k, of 0.945 is not exceeded for temperatures greater than 160 F and boiling.
However, since the spent fuel pool cooling system is capable of maintaining fuel pool water temperature less than 160 F even with a single failure, this calculation is outside of the design basis, and no further action is necessary, 4 i
i t Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Reg. Guide 1.13 (Section 1.4, Appendix A).
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-9
E e
L 14.3' : REFERENCE FUEL STORAGE CELLS v.
!' 4.3.1 Reference Fuel Assembly'
' The design bat uel assembly is the Westinghouse 17x17 Vantage SH (V5H) assembly. Table 4.1.1 summarizes the fuel assembly design specificat.ons.
l l
- 4.3.2 L Region 1 Fuel Storane Cells ,
)
i Figure 4.3.1 'shows the calculational model of the nominal Region 1 spent fuel storage cell' containing a 17x17 V5H assembly. The Region I storage cells are composed of stainless steel boxes separated by a gap with fixed neutron absorber panels, Boral, on each of the box walls. 1
'The ick steel walls define the storage cells which have ch nominalinside' dimension. A . . inch stainless steel sheath supports the Boral panel and i defines the boundary of the flux-trap water-gap used to augment reactivity control. The cells are i located on a lattice spacing o nch in one direction and ch in the other direction. Stainless steel channels connect the storage cells'in a rigid structure and define-the flux-trap between the Boral panels, which are ' ~./ [ nch in one direction and Y ; .
~
. inch in the other direction. The Boral absorber has a thickness of nch and a l nominal B-10 areal density o The Boral absorber-panels are [ [ nches in width an inches in length. Boral panels are installed on all exterior walls facing other racks, as well as, non-fueled regions, i.e., the pool walls. The minimum gap between neighboring Region I style racks and between Region 1 and Region 2 style racks is 1.5 inches. Region 1 and Region 3 racks are not located adjacent to one another.
! 4.3.3 Region 2 Fuel Storane Cells i
Figure 4.3.2 shows the cak ulational model of the nominal Region 2 spent fuel storage cell l
L containing'a 17x17 V5H assembly. The Region 2 storage cells are composed of stainless steel
. walls with a single fixed neutron absorber panel, Boral, (attached by a inch stainless steel HOLTEC INTERNATIONAL Millstone Point Unit 3 4-10 l
p c
sheathing) centered on each side in i M inch channel. Stainless steel boxes are arranged in an alternating pattern such that the cennection of the box corners fonn storage cells between those of the stainless steel boxes. These cells are located on a lattice spacing o y;j M M inch. The hf.hk lick steel walls define a storage cell which has c SMk Inch nominal inside dimension. The Boral absorber has a thickness o:hfkh hch and a nominal B-10 areal E"
inches in width an$ry] inches in length. Boral panels are installed on all exterior walls facing other racks, as well as, non-fueled regions, i.e., the pool walls. The minimum gap between neighboring Region 2 style racks is 0.50 inches, while the minimum gap between Region 1 and Region 2 style racks is 1.5 inches. The minimum gap between Region 2 and Region 3 racks is 1.28 inches.
4.3.4 Recion 3 Fuel Storage Cells Figure 4.3.3 shows the calculational model of the nominal Region 3 spent fuel storage cell containing a 17x17 V5H assembly. The Region 3 storage cells are composed of stainless steel l boxes separated by a gap with fixed neutron absorber panels, Boraflex, on each of the box walls.
The lidhk thick steel walls define the storage cells which have ;
MONMM nch nominalinside dimension. A $p[Nh inch stainless steel sheath supports the Boraflex
! panel and defines the boundary of the flux-trap water-gap used to augment reactivity control.
l I
The cells are located on a lattice spacmg of b. @m! , The Boraflex absorber has a thiukness i
of Mnch and a nominal B-10 areal density of approximately W Yl The Borallex l absorber panels are /- inches in width. Ilowever, all Region 3 analyses assume that the l \
Boraflex is replaced by water, and thus, no credit is taken for neutron absorption in the Boraflex l
panels. The minimum gap between Region 3 and Region 2 style racks is 1.28 inches and the I minimum gap between Region 3 and Region I style racks is 76.09 inches. Region 3 and Region 1 racks are not located adjacent to one another.
I IlOLTEC INTERNATIONAL i Millstone 'oint Unit 3 4-11 l
4.4 ANALYTICAL METHODOLOGY
.4.4.1 Reference Desian Calculations -
1 I
The principal methods for the criticality analyses of the high density storage racks include the l following codes: (1) MCNP4a [4.4.1], (2) KEN 05a [4.4.2], and CASMO-3 [4.4.5-4.4.7].
~ MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los i
Alamos National Laboratory. KENO 5a is a three-dimensional multigroup Monte Carlo code developed at the Oak Ridge National Laboratory as part of the SCALE 4.3 package [4.4.3]. The
]
KENO 5a calculations used the 238-group SCALE cros-s-section library and NITAWL [4.4.4] for S
U resonance shielding effects (Nordheim integral treatment). Benchmark calculations, presented in Appendix 4A, indicate a bias of 0.0009 with an uncertainty of 0.0011 for McNP4a and 0.0030 - 0.0012 for KEN 05a, both evaluated with the 95% probability at the 95%
confidence level [4.1.1].
l Fuel depletion analyses during core operation were performed with CASMO-3, a two . ;
dimensional multigroup transport theory code based on capture probabilities [4.4.5 - 4.4.7].
Restarting the CASMO-3 calculations in the storage rack geometry yields the two-dimensional infinite multiplication factor (k ) for the storage rack. Parallel calculations with CASMO-3 for the storage rack at various enrichments enable a reactivity equivalent enrichment (fresh fuel) to be ' determined that provides the same reactivity in the rack as the depleted fuel. CASMO-3 was also used to determine the small reactivity uncertainties (differential calculations) of
. manufacturing tolerances and the reactivity effect of various decay times (for Region 3 only).
In the geometric models used for the calculations, e ;h fuel rod and its cladding were described explicitly and reflecting boundary conditions were used in the radial direction which has the effect of creating an infinite radial array of storage cells. Monte Carlo calculations inherently ;
include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of the MC'4P4a and KEN 05a calculated reactivities and to assure convergence, a minimum of 1 million neutron histories were accumulated in each calculation.
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-12
1 4.4.2 Fuel Burnuo Calculations and Uncertainties CASMO-3 was used for burnup calculations in the hot operating condition. CASMO-3 has been extensively benchmarked [4.4.7, 4.4.8] against cold, clean, critical experiments (including -
plutonium-bearing fuel), Monte Carlo calculations, reactor operations, and heavy element
[ . concentrations in irradiated fuel. In addition to burnup calculations, CASMO-3 was used for evaluating the small reactivity increments (oy differential calculations) associated with manufacturing tolerances, for determining temperature effects, and the reactivity effects of decay time.
In the CASMO-3 geometric models, ea.:h fuel rod ard its cladding were described explicitly and reflective boundary conditions were usei betweer. storage cells. These boundary conditions have the effect of creating an infinite array of storage cells.
Conservative ase::nptions of moderator and fuel temperatures and the average operating soluble bocon concentrations were used to assure the highest plutonium production and hence cccurvatively high values of reactivity during burnup. Since critical experiment data with spent fuel is not available for determining the uncertainty in depletion calculations, an allowance for uncertainty in reactivity 'was assigned based upon other considerations. Assuming the uncer-tainty in depletion calculations is less than 5% of the total reactivity decrement, a burnup dependent uncertainty in reactivity for burnup calculations was assigned. Thus, the burnup uncertainty varies (increases) with burnup. This allowance for burnup uncertainty was included in determination of the acceptable Imrnup versus enrichment combinations, and is believed to be a conservative estimate, t The majority of the uncertainty in depletion calculations derives from uncertainties in fuel and moderator temperatures and the effect of reactivity control r -^ods (e.g., soluble boron). For depletion calculations, bounding values of these operating parameters weie assumed to . .,. ire conservative results in the analyses.
HOLTEC INTERNATIONAL Millstone Point (Jr.L a 4-13 L.. .
i 4.4.3 Effect of Axial Burnuo Distribution j
Initially, fuel loaded into the reacter will burn with a slightly skewed cosine power distribution.
' As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower regions. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of high neutron leakage. Consequently, it is expected that over most of the burnup history, fuel assemblies with distributed burnups will exhibit a slightly lower reactivity than that calculated for the uniform 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 oflarge regions of significantly reduced burnup.
Among others, Turner [4.4.9] has provided generic analytic results of the axial burnup effect based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnups at values less than about 27 mwd /kgU with small positive reactivity effects at higher burnup values. Because of the 24 decay of 'Pu, the effect of the axial burnup distribution becomes larger when cooling times are considered. For the present criticality analyses, the reference calculations utilized representative axial burnup distributions previously calculated for Millstone Unit 3. Burnup-equivalent enrichments were determined with CASMO-3 for each of 24 axial zones and used in three-dimensional Monte Carlo calculations. Results of these calculations, therefore, inherently include the effect of the axial distribution in burnup. Comparison of these results to results of calculations with uniform axial burnup allows the reactivity effect of the axial burnup distribution to be quantified. This reactivity effect is included, where applicable, in the calculation of the ,
i maximum k,y values. For Region 3, where credit for cooling time is considered, calculations were performed to determine the reactivity effect at each of the cooling times.
1 I
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-14 L'_
L l- 4.4.4 Long-Term Changet in Reactivity l L
Since the fuel racks in Region 3 are intended to contain spent fuel for long periods of time, i 1
i Econsideration was given to the long-term changes in reactivity of spent fuel. Calculations i
- confirm that reactivity continuously decreases as the spent fuel ages. Early in the decay period, L ' Xenon grows from Iodine decay (reducing reactivity) and subsequently decays, with the l
reactivity reaching a maximum at about 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. To assure conservatism in the restart calculations, the Xe-135 is set to zero. The decay of Pu-241 (13-year half-life) and growth of l
l Am-241 substantially reduce the reactivity during long term storage. Figure 4.1.3 illustrates the l reduction in reactivity during long term storage. For Region 3 racks, credit is taken for this l long-term reduction in reactivity, and includes the increased effect of the axial burnup distribution. However, for Regions 1 and 2, no credit is taken for this long-term reduction in
! reactivity, other than to indicate an increasing suberiticality margin.
i 4
l 1
! I i
l l
l l
l l
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-15
4.5 REGION 1 CRITICALITY ANALYSES AND TOLERANCES 4.5.1 Nominal Design Case For the nominal storage cell design in Region 1, the criticality safety analyses are summarized in Table 4.2.1. These data confirm that the maximum reactivity in Region 1 remains conservatively less than the regulatory limit (ke rr 0.95). An independent calculation with the KEN 05a code I provides confirmation of the validity of the reference MCNP4a calculations.
4.5.2 Uncertainties Due to Bumun I I
For storage in the 3-outef-4 arrangement, consideration of fuel burnup is not necessary, and thus, burnup related uncertainties are not applicable. However, for unrestricted storage in the 4-out-of-4 arrangement, fuel burnup is required. CASMO-3 was used for the depletion analysis j and the restart option was used to analytically transfer the spent fuel into the storage rack I configuration at a reference temperature of 4 C (corresponding to the highest reactivity, see Section 4.8.1). Calculations were also made for fuel of several different initial enrichments and interpolated to define the burnup-dependent equivalent enrichments , at each bumup. MCNP4a calculations were then made for the equivalent enrichment to establish the limiting k,,r value, which includes all applicable uncertainties. At the limiting burnups required for Region 1 storage, !
the effect of the axial distribution in burnup is negative, and thus, is not included. These calculations were used to define the boundary of the acceptable domain shown in Figure 4.1.1.
t The (reactivity) equivalent enrichment is the fresh un-burned fuel enrichment that yields the same reactivity as the depleted fuel, both evaluated in the storage rack configuration. The equivalent enrichment may then be used in three-dimensional MCNP4a or KENO 5a calculations.
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-16
I I
4.5.3 Uncertainties Due to Tolerances The reactivity effects of manufacturing tolerances are tabulated, along with the tolerances, in
! Table 4.5.1. The individual tolerances were conservatively calculated for the design basis fresh l
unburned fuel assembly.
4.5.4 Eccentric Fuel Positionine The fuel assembly is assumed to be normally located in the center of the storage rack cell.
However, calculations were also made with the fuel assemblies assumed to be in the corner of the l storage rack cell (four-assembly cluster at closest approach). These calculations indicated that the reactivity effect is small and negative. Therefore, the reference case in which the fuel assemblies are centered is controlling and no uncertainty for eccentricity is necessary.
4.5.5 Water-Gan Snacina Between Racks l
The minimum water-gap between racks, which is 1.5 inches between neighboring Region 1 style l racks and also 1.5 inches between Region 1 and Region 2 style racks, constitutes a neutron flux-trap for the storage cells of facing racks. The racks are constructed with the base plates extending beyond the edge of the cells which assures that the minimum spacing between storage racks is maintained under all credible conditions. This water-gap flux-trap i: !arger e an those i between Region 1 cells, and thus, will act to reduce the reactivity below the cited maximum. l l
l l
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-17
r1 1
]
4.6 REGION 2 CRITICALITY ANALYSES AND TOLERANCES
. 4.6.1 ' Nominal Desian Case For the nominal storage cell design in Region 2, the criticality safety analyses are summarized in Table 4.2.3. These data confirm that the maximum reactivity in Region 2 remains conservatively less than the regulatory limit (k,, 0.95). An independent calculation with the KEN 05a code provides confirmation of the vaiidity of the reference MCNP4a calculations.
l 4.6.2 Uncertainties Due to Burnuo I l
CASMO-3 was used for the depletion analysis and the restart option was used to analytically
)
transfer the spent fuel into the storage rack configuration at a reference temperature of a C j (corresponding _to the highest reactivity, see Section 4.8.1). Calculations were also made for fuel i
of several different initial enrichments and interpolated to define the burnup-dependent equivalent l 1
' enrichments , at each burnup. MCNP4a calculations were then made for the equivalent enrichment to establish the limiting k,y value, which includes all applicable uncertainties and the l effect of the axial burnup distribution. These calculations were used to define the boundary of the acceptable domain shown in Figure 4.1.2.
J L '4.6.3 . Uncertainties Due to Tolerances The reactivity effects of manufacturing tolerances are tabulated, along with the tolerances,-in
- Table 4.6.1. The individual reactivity allowances were conservatively calculated for the design basis fresh unburned fuel assembly.
' t The (reactivity) equivalent enrichment is the fresh un-burned fuel enrichment that yields the same reactivity as the depleted fuel, both evaluated in the storage rack configuration. The equivalent enrichment may then be used in three-dimensional MCNP4a or KEN 05a calculations.
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-18 r
r,
[
~
[ 4.6.4 Eccentric Fuel Pcj:tioning j
l . .
4
' The fuel assembly is assumed to be normally located in the center of the storage rack cell,
. Ilowever, calculations were also made with the fuel assemblies assumed to be in the corner of the ]
storage rack cell (four-assemb'j cluster at closest approach). These calculations indicated that the reactivity effect is small and negative. Therefore, the reference case in which the fuel assemblies are centered is controlling and no uncertainty for eccentricity is necessary.
4.6.5 Water-Gao Soacine Between Racks The minimum water-gap between racks, which is 0.50 inches between neipbhoring Region 2 style racks and 1.5 inches between Region I and Region 2 style racks, constitutes a neutron flux-l trap for the storage cells of facing racks. The racks are constructed with the base plates 1
extending beyond the edge of the cells which assures that the minimum spacing between storage l racks is maintained under all credible conditions. Region 2 style .xks do not contain water gaps, and thus, this water-gap finx-trap will act to reduce the reactivity below the cited maximum.
l l
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-19 L
i 4.7 REGION 3 CRITICALITY ANALYSES AND TOLERANCES
!' 4.7.1 Nominal Desien Case For the nominal storage cell design in Region 3, the criticality safety analyses are summarized in Table 4.2.5. These data confirm that the maximum reactivity in Region 3 remains conservatively less than the regulatory limit (km 0.95). Independent calculations with the MCNP4a and KENO 5a codes provide confirmation of the validity of the reference CASMO-3 calculations.
4.7.2 Uncertainties Due to Burnun CASMO-3 was used for the depletion and decay time analyses and the restart option was used to analytically transfer the spent fuel into the storage rack configuration at a reference temperature of 160 F (corresponding to the highest reactivity, see Section 4.8.1). Calculations were also made for fuel of several different inkial enrichments and interpolated to define the burnup-dependent equivalent enrichments, at each burnup. KEN 05a calculations were then made for the equivalent enrichments to determine the effect of the axial burnup distribution. These calculations l
were made for each of the cooling times. CASMO-3 calculations were used for the establish the limiting k, value, which includes all applicable uncertainties and the effect of the axial burnup distribution. These calculations were used to define the boundary of the acceptable domains shown in Figures 4.1.3. i 4.7.3 Uncertainties Due to Tolerances The reactivity effects of manufacturing tolerances were calculated for various burnups and each l- of the defined cooling times with the design basis fuel assembly. For conservatism, the largest I reactivity effect for each tolerance was used to establish the corresponding reactivity allowance.
These values are tabulated, along with the tolerances, in Table 4.7.1.
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-20 L
L: _ _ _ . -.
m~
!4 l 1
4.7.4 : Eccentric Fuel Positionine -
The fuel assembly is assume'dt' o be normally located in the center of the storage rack cell. l However, calculations were also made with the fuel assemblies assumed to be in the comer of the storage rack cell (four-assembly cluster at closest approach). Because no credit is taken for the
~ Boraflex panels in the Region 3 racks, these calculations determined that the reactivity effect is small and positive. Therefore, the positive uncertainty' associated with fuel eccentricity is included in the determination of the maximum reactivity in Table 4.2.5.
y l
l l
1 HOLTEC INTERNATIONAL Millstone Point Unit 3 4-21
' 4.8 - ; ABNORMAL AND ACCIDENT CONDITIONS l
4.8.1 Temocrature and Water Density Effects l
1
-4.8.1.1 Region 1 and 2, i The moderator temperature coefficient of reactivity in Region 1 and Region 2 is negative.
Therefore, a moderator temperature of 4 C (39 F) was assumed for the reference calculations,-
which assures that the true reactivity will always be lower over the expected range of water:
temperatures. Temperature effects on reactivity have been calculated (CASMO-3) and the results _
1 are shown in Table 4.8.1. In addition, the introduction of voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown in Table 4.8.1. !
With soluble boron present, the temperature coeflicients of reactivity would differ from those listed in Table 4.8.1. However, the reactivities would also be substantially lower at all temperatures with
- soluble boron present. The data in Table 4.8.1 is pertinent to the higher-reactivity unborated case.
)
i i
Since the Monte Carlo codes, MCNP4a and KENO 5a, cannot handle temperature dependence, all
, MCNP4a and KENO 5a calculations were performed at 20*C and a positive temperature
-correction fact'or (the value of Ak between calculations at 20 C and 4 C) was applied to the results.
4.8.1.2 Region 3 With the assumption that the Boraflex panels are replaced by water, the moderator temperature coefficient of reactivity in Region 3 is positive. Therefore, a moderator temperature of 160 F was assumed for the reference calculations (for normal conditions). Temperatures above 160 F are accident conditions, during which credit for soluble boron is allowed. Temperature effects on reactivity have been calculated (CASMO-3) and the results are shown in Table 4.8.2. In addition, L the introduction of voids in the water internal to the storage cell (to simulate boiling) increased reactivity, as shown in Table 4.8.2. Calculations indicate that 100 ppm soluble boron is more than adequate to ' assure that the limiting kaof 0.945 is not exceeded for temperatures greater than 160 j HOLTEC INTERNATIONAL 4 e i Millstone Point Unit 3. 4-22
]
o
_a_
' F and boiling. However, since the spent fuel pool cooling system is capable of maircining fuel pool water temperature less than 160 F, this condition is outside of the design basis, and no further
" action is necessary.
With soluble boron present, the temperature coefficients of reactivity would differ from those listed in Table 4.8.2. However, the reactivities would also be substantially lower at all temperatures with l soluble boron present. The data in Table 4.8.2 is pertinent to the higher-reactivity unborated case.
I The CASMO-3 calculations were performed at 4 C and a positive temperature correction factor !
' (the value of Ak between calculations at 4 C and 160'F) was applied to the results.
4.8.2 . Lateral Rack Movement Lateral motion of the storage racks under seismic conditions could potentially alter the spacing between racks. In Region 1, the minimum water gap between racks (1.5 inches, as limited by the base plate extensions) is larger than the corresponding design water-gap spacing (0.79 inches in
, one direction and 1.244 inches in the other direction) intemal to the racks. Consequently, there will be no positive effect on reactivity.
Region 2 storage cells do not use a flux-trap, and thus, the calculated maximum reactivity does L
~
4
_ not rely on spacing between racks. Nevertheless, the minimum water gap between Region 2 racks (0.50 inches, as limited by the base plate extensions) and the Boral panels, which are l installed on al! exterior walls of Region 2 racks, assure that the reactivity is always less than the design limitation. Furthermore, soluble poison would assure that a reactivity less than the design
, limitation is maintained under all accident or abnormal conditions. 7 The minimum distance between Region 3 and Region 1 racks is 76.09 inches, and thus, lateral rack moment is of no concern. The minimum water gap between Region 3 and Region 2 racks is 1.28 inches. which is comparable to the water-gap spacing (1.26 inches) internal to the Region 3 HOLTEC INTERNATIONAL Millstone Point Unit 3 4-23
racks. In addition, the Region 2 racks have Boral panels installed on all exterior walls (Region 3 racks are assumed to be unpoisoned). Furthermore, soluble poison would assure that a reactivity less than the design limitation is maintained under all accident or abnormal conditions.
4.8.3 Abnormal Location of a Fuel Assembly 235 The abnormal location of a fresh un-irradiated fuel assembly of 5.0 wt% U enrichment could, in the absence of soluble poison, result in exceeding the regulatory limit (k,y 0.95). This could occur if a fresh fuel assembly of the highest permissible enrichment were to be inadvertently loaded into either a Region 2 or Region 3 storage cell. Calculations confirmed that the highest reactivity, including uncertainties, for the worst case postulated accident condition (fresh fuel assembly in Region 3) would exceed the limit on reactivity in the absence of soluble boron. Soluble boron in the spent fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. Calculations indicate that the 800 ppm soluble boron, that is to be required by the Technical Specifications during fuel handling operations, is more than adequate to assure that the limiting k,y or0.945 is not exceeded.
4.8.4 Dropoed Fuel Assembly For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel in the rack of more than 12 inches, including the potential deformation under seismic or accident conditions. At this separation distance, the effect on reactivity is insignificant.
Furthermore, the soluble boron in the pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.
HOLTEC INTERNATIONAL Millstone Point Unit 3 - 4-24
p
4.9 REFERENCES
[4.1.1] . M. G. Natrella, Fmperimental Statistics, National Bureau of Standards Handbook ;
e 91, August 1963. .I i
.[4.4.1] J.F. Briesmeister, Editor,"MCNP - A General Monte Carlo N-Particle Transport Code, I Version 4A," LA-12625, Los Alamos National Laboratory (1993).
[4.4.2] L.M. Petrie and N.F. Landers, " KENO Va - An Improved Monte Carlo Criticality Program with Supergrouping," Volume 2, Section F11 from " SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR-0200, Rev. 4, January 1990.
[4.4.3] " SCALE 4.3: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation For Workstations and Personal Computers, Volume 0," CCC-545, ORNL-RSICC, Oak Ridge National Laboratory (1995).
[4.4.4] N.M. Greene, L.M. Petrie and R.M. 'Vestfall, "NITAWL-II: Scale System Module for Perfonning Shielding and . Working Library Production," Voieme 1, Section F1 from
" SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR 0200, Rev. 4, January 1990.
'[4.4.5] A. Ahlin and M. Edenius,"CASMO - A Fast Transport Theory Depletion Code for LWR Analysis," ANS Transactions, Vol. 26, p. 604,1977.
[4.4.6] 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.
[4.4.7] M. Edenius and A. Ahlin, "CASMO-3: New Features, Benchmarking, and
~ Advanced Applications," Nucl. Sci. Eng., 100 (1988).
[4.4.8] E. Johansson," Reactor Physics Calculations on Close-Packed Pressurized Water Reactor Lattices," Nuclear Technalogy, Vol. 68, pp. 263-268, February 1985.
~[4.4.9] 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 Millstone Point Unit 3 4-25
Table 4.1.1 l
Fuel Assembly Specifications Fuel Rod Data '
Westinghouse Westinghouse I Assembly type Standard Vantage-5II Fuel pellet outside diameter, in. l % n ;1] lp g; g gl Cladding thickness, in. lqlg@l lAsgp;]
Cladding outside diameter, in. lW! Nl lNCj$f2l Cladding material y y Pellet density, % T.D. 1%!I lust l41 Maximum nominal enrichment, 5.0 5.0 wt%2 "U Fuel Assembly Data Fuel rod array 17 x 17 17 x 17 Number of fuel rods p
.$_ h@
l Fuel rod pitch, in. {ip wl [gpq {
Number of control rod guide and lngl lt gl l
instrument thimbles Thimble outside diameter, in. l$& , ul 7 lWs4sj Thimble thickness, in. l:Re@M1 l N 5;%
l Active fuel Length, in. l W Ol l'inil IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-26
p i
L>.
l Table 4.2.1 i I
f Summary of the Criticality Safety Analyses for Region 1
. Storage Arrangement 3-out-of-4 4-out-of-4 l
Design Basis Burnups at 5.0 wt%2 rU 0 8.0 mwd /kgU Uncertainties Bias Uncertainty (95%/95%) i 0.0011 i 0.0011 l Calculational Statistics: (95%/95%,2.0xo) 0.0911 i 0.0015 Depletion Uncertainty N/A i 0.0028 Fuel Eccentricity negative negative Manufacturing Tolerances (Table 4.5.1) 0.0111 i 0.0111
)
Statistical Combination of Uncertaintiest i 0.0112 t 0.0116 f l
l Reference k,y (MCNP4a) 0.9122 0.9132 Total Uncertainty (above) 0.0112 0.0116 )
Axial Burnup Distribution N/A negligible Calculational Bias (see Appendix A) 0.0009 0.0009 Temperature Correction to 4 C (39 F) 0.0015 0.0015 Maximum k,, 0.9258 0,9272tt Regulatory Limiting k,, 0.9500 0.9500 I
t The value used for the MCNPe a(or KEN 05a) statistical uncertainty is 2.0 times the estimated standard deviation.
Each final k value calculated by MCNP4a (or KENMd is the result of averaging a minimum of 200 cycle k values, and thus, is based on a minimum sample The K multiplier, for a one-sided statistical tolerance I with 95% probability at the 95% confidence level, corresponding to a sampic size of 200, is 1.84. However, for this analysis a value of 2.0 was assumed for the K multiplier, which is larger (more conservative) than the value corresponding to a sample size of 200.
l t Square root of the sum of the squaces.
- tt KENO 5a verification calculation resulted in a maximum k,of 0.9270. i l
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-27 l
t
I Table 4.2.2 Burnup-Enrichment Limits in Region 1 NominalInitial Fuel Enrichment Minimum Fuel Burnup ;
2 (wt% "U) (mwd /kgU) 3.7 0.00 4.0 1.91 4.5 5.12 5.0 8.00 l
l l
I I
I i
i i
HOLTEC INTERNATIONAL !
Millstone Point Unit 3 4-28 i
L i
Table 4.2.3 ,
l Summary of the Criticality Safety Analyses for Region 2
. Design Basis Burnup at 5.0 wt% *U 39.0 mwd /kgU Uncertainties Bias Uncertainty (95%/95%) t 0.0011 3 Calculational Statistics * (95%/95%, 2.0xo) i 0.0013 Depletion Uncertainty i 0.0142 Fuel Eccentricity ~ negative Manufacturing Tolerances (Table 4.6.1) i0.0059 Statistical Combination of Uncertaintiest i 0.0155 Reference k,,(MCNP4a) 0.9142 Total Uncertainty (above) 0.0155 l Axial Burnup Distribution 0.0110 Calculational Bias (see Appendix A) 0.0009 Temperature Correction to 4*C (39 F) 0.0020 Maximum k,, 0.9436tt Regulatory Limiting k,, 0.9500
.
- The value used for the MCNP4a (or KENO 5a) statistical uncertainty is 2.0 times the estimated standard deviation.
Each fir at k value calculated by MCNP4a (or KENO 5a)is the result of averaging a minimum of 200 cycle k )
( values, and thus, is based on a minimum sample size of 200. The K multiplier, for a one-sided statistical tolerance i with 95% probability at the 95% confidence level, corresponding to a sample size of 200, is 1.84. Ilowever, for this
- - analysis a value of 2.0 was assumed for the K multiplier, which is larger (more conservative) than the value corresponding to a sample size of 200.
t Square root of the sum of the squares.
tt KENO 5a verification calculation resulted in a maximum k,g of 0.9449.
HOLTEC INTERNATIONAL
' Millstone Point Unit 3 4 29 L
r Table 4.2.4 Burnup-Enrichment Limits in Region 2 NominalInitial Fuel Enrichment Minimum Fuel Bumup 2
(wt% "U) (mwd /kgU) 2.0 3.48 2.5 10.04 I
3.0 15.92 l l
3.5 21.48 i I
4.0 26.83 4.5 33.75 !
5.0 39.00 IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-30
L' Table 4.2.5 Summary of the Criticality Safety Analyses for Region 3 Cooling Time (years) 0 5 10 20 Design Basis Burnup (mwd /kgU) 55.41 49.90 47.31 43.91 at 5.0 wt%2 "U Uncertainties Depletion Uncertainty - 0.0182 0.0186 0.0189 0.0189-1 Fuel Eccentricity 0.0017 0.0017 0.0017 0.0017 l Manufacturing Tolerances gg F T y y gg gpg pigggg (Table 4.7.1)
Statistical Combination of Uncertaintiest MQ (khh ((] %Q Reference k. (CASMO-3) 0.8796 0.8705 0.d643 0.8650 l
Total Uncertainty (above) 0.0196 0.0200 0.0203 0.0203 Axial Burnup Distribution 0.0298 0.0386 0.0445 0.0438 Temperature Correction to 160 F 0.0160 0.0160 0.0160 0.0160 Maximum k,, 0.945 0.945 0.945 0.945 Regulatory Limiting k,, 0.950 0.950 0.950 0.950 t Square root of the sum of the squares.
IiOLTEC INTERNATIONAL
' Millstone Point Unit 3 4-31
l Table 4.2.6 Burnup-Enrichment Limits in Region 3 for Various Decay Tiines Minimum Fuel Burnup (mwd /kgU) l NominalInitial
- Fuel Enrichment 0 5 10 20
! 2 (wt% "U) (years decay time) (years decay time) (years decay time) (years decay time) 2.0 10.64 1 9.64 9.05 8.47 2.5 18.51 16.55 15.42 14.32 l 3.0 25.62 22.66 21.08 19.56 l l
3.5 32.58 28.44 26.50 24.59 1
4.0 40.33 35.39 32.82 30.63 l 4.5 47.95 42.67 40.03 37.25 1
)
5.0 55.41 49.90 47.31 43.91 l l
4 l
l' l
IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-32 i
l
r l
l Table 4.2.7 Reactivity Effects of Abnormal and Accident Conditions in Regions 1 and 2 Abnormal / Accident Conditions Reactivity Effect i
Temperature Increase (above 4 C) Negative (Table 4.8.1)
Void (boiling) Negative (Table 4.8.1)
Assembly Drop (on top of rack) Negligible Lateral Rack Movement Negligible l Misplacement of a fresh fuel assembly Positive - controlled by less than 800 ppm soluble boron (a minimum 800 ppm soluble boron is to be required by Technical ;
Specifications during fuel movement)
.a 1
j l
l i
l HOLTEC INTERNATIONAL Millstone Point Unit 3 4-33 l-
Table 4.2.8 Reactivity Effects of Abnormal and Accident Conditions in Region 3 Abnormal / Accident Conditions Reactivity Effect
~
Temperature Increase (above 160 }') Positive (Table 4.8.2) - controlled by 100 ppm soluble boron (however, outside of design basis)
Void (boiling) Positive (Table 4.8.2) - controlled by 100 ppm soluble boron (however, outside of design basis)
Assembly Drop (on top of rack) Negligible Lateral Rack Movement Negligible Misplacement of a fresh fuel assembly Positive - controlled by less than 800 ppm soluble boron (a minimum 800 ppm soluble boron is to be required by Technical Specifications during fuel movement)
IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-34 l
l l
L-
Table 4.5.1 -
Reactivity Effects of Manufacturing Tolerances in Region 1 Tolerance Reactivity Effect, Ak Minimum Boralloading ([fMP 'MFAMEiffnominal) 10.0029 Minimum Boral width [23EEE@ nominal) 10.0018 Maximum box I.D. l[ nominal) i0.0102 Maximum box wall thickness [3Bt gg nominal) i0.0007 Density tolerance @7MINM230!!NN$31pominal) 10.0018 I Enrichment ECW "W"EdW nominal) 10.0017 Total (statistical sum)t 10.0111 l
1
- t Square root of the sum of the squares.
I IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-35
Table 4.6.1 l
Reactivity Effects of Manufacturing Tolerances in Region 2 Tolerance Reactivity Effect, Ak !
Minimum Boral loading ([iWND$W"WYf] nominal) i0.0045 Minimum Boral width (FM~J7Ef3 nominal) 10.0023 Minimum box 1.D. CTMMnominal) i0.0017 Maximum box wall thickness ety$itygty!!j' nominal) i0.0002 Density ([$s3Egag3M??"?!] nominal) i0.0013 j i
Enrichment ((Tfff215HfFUSGiominal) i0.0021 l l
Total (statistical sum)t i0.0059 i
i l
i I
i i
I l
1 t Square root of the sum of the squares.
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-36
Table 4.7.1 Reactivity Effects of Manufacturing Tolerances in Region 3 )
Tolerance Reactivity Effect, Ak Minimum box 1.D. igisC47?FP]1ominal) i0.0004 Minimum pitch $@5CWWEEnominal) 10.0030 Minimum box wall thickness (ET3?nWata nominai) i0.0026 Minimum sheathing thickness (I((gI[ nominal) i0.0033 Density (['@NsFi?RW73 nominal) i0.0032 Enrichment (TFJiETdF9ZIf773 nominal) 10.0039 Total (statistical sum)t 0.0072 i
I l
t Square root of the sum cf the squares.
IIOL.TEC INTERNATIONAL Millstone . Point Uni:3 4-37
I' l Table 4.8.1 Reactivity Effects of Temperature and Void in Regions 1 and 2 i
Reactivity Effect, Ak l
Case Region 1 Region 2 !
I 4 C (39'F) reference reference
- 20'C (68'F) -0.0015 -0.0020 j
60 C(140 F) -0.0084 -0.0094 l 120*C(248 F) -0.0241 -0.0253 120 C w/10% void -0.0527 -0.0508 l
I IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-38
Table 4.8.2 Reactivity Effects of Temperature and Void in Region 3 Case Reactivity Effect, Ak 4 C (39 F) -0.0160 20 C (68*F) -0.0122-40 C(104 F) 0.0074 65'C (149 F) -0.0014 71.1 C (160 F) reference 90'C (l94 F) +0.0045 120'C (248 F) +0.0123 120 C w/10% void +0.0163 l
l i
L. HOLTEC INTERNATIONAL Millstone Point Unit 3 4-39 o
{~;
8 I I s e s 9 4 4 1 8 -------- I-----------------4--------+--------
4 1 6 1 1 1 l 4 , 3 6 I e I t i l i a l i i 1 6 I 4 t 1 i i I I I I I i 6 i
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3.6 3.8 4 4.2 4.4 4.6 4.8 5 Initial Fuel Enrichment (wt% U-235) Figure 4.1.1 Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enriclunent to Permit Unrestricted Storage in Region 1 IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-40
x r 40 , , , , , l
.,-. , , t. -
l
., f , , ,
35 ----------4,----------l----------l------ r---------4,-
, l . ,
i f , , , t' .. I , 30 - - - - - - - - - - , .---------,. ----------r---------,--- q l A.ccrPTABLU, DOMAIN l, l l
, 1
. I , , ,
25 ----------,'---------,'---------'r--- S= , iic l l l l 6 20 ----------4--------- i ,
------ --i----------*---------4----------
, , i cu , . , , ,
9 EI , , , , , e , , , , a , z=3 15
----------4.,-------
, ----------*---------4,----------
m t ., .
., ., ., I
, , , , ,f 10 --------- ---------s----------i----------'---------4----------
l l l UNACCEPf, ABLE DOM,AIN I , , , 5
-------j,---------i--------------------t---------i,----------
1 4 , i I 0 ' ' ' 2 2.5 3 3.5 4 4.5 5 Initial Fuel Enrichn.ent (wt% U-235) . Figure 4.1.2 Minimum Required Fuel Assembly Bumup as a Function of Nominal ; Initial Enrichment to Permit Unrestricted Storage in Region 2 (Fuel 2 assemblies with enrichments less than 2.0 wt% "U will conservatively be required to meet the burnup requirements of 2.0 wt% 2,5U assemblies). HOLTEC INTERNATIONA L Millstone Point Unit 3 4-41
f I. 1 i 60 , , , , , 6 . , , 1 0 years co.oh.ng - . i i 5 years cooling - - i i e i 55 - 10yehficddlidg :1- - i " - - " - " ""-""[--""-"j""""- 20 years eqoling -- l l l l 8 . , l ,
- 50 ----------}-------->----------l---------L-------a- ----- --
c , , , . / l l ; l , ,' ,e
, , , 8 , #
45 2-------- s--------- i--------- '----- - /- - *'- '
------ -- A d, CEPTABLEl, DOMAIN l, -l l ,' );
, (, , , - ,
, , , / ,, # /
c 40 --------- 2,-------- s,--------- ,i---------
--- ,s-,,s,-
/- ------ ,
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.p 35 --------- '----------l---------
, '- --- ,- 47'- - /- - ' - - - - - - - - - -
e ,
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a - 1 30 ----------.,---------,,-----
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=-
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8 , - , m<
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- Figure 4.1.3 Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enrichment to Permit Unrestricted Storage in Region 3 (Fuel assemblies with enrichments less than 2.0 wt% 235 U will conservatively be required to meet the burnup requirements of 2.0 wt%235U assemblies).
HOLTEC INTERNATIONAL Millstone Point Unit 3 4-42 l r' t i'
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RECION I 4-00T-0F-4 (UNRESTRICTED) ORREC10N2STORACE I Figtee 4.2.1 Illustration of the Interface Requirement Between 3-out-of-4 and 4-out-of-4 (Unrestricted) Storage in Region 1 SIIADED REGIONS ARE IiOLTEC PROPRIETARY INFORMATION Millstone Point Unit 3 4-43
Reflective Bounchnv Condition
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Boral Panel Box Wall Reflective Boundary Condition
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0 0 0 0 0 0 00 0 0 0 0 0 0 0 0 0 00000 00000000000 0 0 0 0 0 0 0 0 0 00 0 00 0 0 0 0 00000000000000000 0 0 0 0 0 0 0 0 0 0 0 00 0 0 0 0 , Reflective Boundary Condition Figure 4.3.2 A Two-Dimensional Representation of the Calculational Model Used for the Region 2 Rack Analysis. This Figure was Drawn with the Two-Dimensional Plotter in MCNP4a llOLTEC INTERNATIONAL Millstone Point Unit 3 4-45
Reflective Boundary Condition Water Gap j O 0 0 0 0 0 0 0 0 0 0 0 0 0D 0 0 ) 0 0 0 0 0 0 0 0 0 0 0 0 0 00 0 0 0 0 0 0,0 00 0 0 0 0 00 00 0 0 0 0 000 0 0.0 0 0 0 0 0 00 0 0 u 0 0 0 0'0 0 0lO 0 0 0 0 0 0 0 0 e i 0 0 00 0 00000 0 0 0 0 00 00 0 0 0 0 0 0 0 0 0 0 0"O00 00 0 $
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- Region 3 Rack Analysis. This Figure was Drawn with the Two-Dimensional Plotter in MCNP4a IIOLTEC INTERNATIONAL Millstone Point Unit 3 4-46 e
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APPENDIX 4A: BENCHMARK CALCULATIONS
\
4A.1 INTPEinICTION AND
SUMMARY
Benchmark calculatici s - ave 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 KENO 5a [4A.2] - -- - - uses group-dependent cross sections. For the KENO 5a 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 2 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. KENO 5a computes and prints the " energy of the avera;;e lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KENO 5a, the number of fissions in each group may be collected and the EALF determined (post-processing). l 1 t Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. Rese 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. Holtec Intemational Appendix 4 A, Page 1 1
~
J
, Figures 4A.1 and 4A.2 show the calculated k,, for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO 5a,' respectively (UO2 fuel only). The scatter in the data (even for comparatively minor variat ion in critical parameters) represents experimental erro? 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 B&W 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 KENO 5a). The total bias (systematic error, or mean of the deviation from a k,, of exactly'1.000) for the two methods of analysis are shown in the table below. Calculational Bias of MCNP4a and KENO 5a MCNP4a 0.0009 i 0.0011 KENO 5a 0.0030 i 0.0012 The bias and standard error of the bias were derived directly from the calculated k,n values in Table 4A.1 using the following equations", 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= k, (4A.1) t A classical example of experimental error is the corrected enrichment in the PNL experiments, first as r.n 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 rtastics, for example, reference (4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO 5a. 1 i lioltec International . Appendix 4A, Page 2 j
1 k,* -( k,)2 /n
,2, , ..: ,.i (4.,.2) n (n-1)
Blas = (1 - k )
- K og (4A.3) i where ki are the calculated reactivities of n critical experiments; o, is the unbiased i 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- E ), is the actual bias which is added to the MCNP4a and KENO 5a results. The second term, Kog, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 arut 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 KENO 5a are 2.04 and 2.05, respectively. The bias values are used to evaluate the maximum k,, values for the rack designs. KENO 5a 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 KENO 5a (SCALE) calculations. 4A.2 Effect of Enrichment
' 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 rod 4A.4 show the calculated k,y values (Table 4A.1) as a function of the fuel enrichment reported for the critical experiments. Linear 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 KENO 5a). Thus, there are no corrections to the bias for the various enrichments.
Holtec Internationali Appendix 4A, Page 3 i
- As.,further. confirmation of the absence of.any trends with enrichment, a typical .
configuration was calculated with both MCNP4a and KENO 5a for various enrichments. i The cross-comparison of calculations with codes of comparable sophistication is suggested l I in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of k,7 for the two independent codes as evidenced by the 45' slope of the curve. Since it is very unlikely that avo 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. 1 I 4A.3 Effect of 'B Londina I l l t 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 nevealed. Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A.1) , and shows the reactivity worth (Ak) 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 , I unusually large experimental errors. B&W made an effort to repoit some of their experimental errors. Other laboratories did not evaluate their experimental errors. 1 I To further confirm the absence of a significant trend with 'B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO 5a (as suggested in Reg. Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry. l 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,* line, within an expected 95% probability , limit). I i t The reactivity worth of the absorber panels was determined by repeating th calculation with the absorber analytically removed and calculating the incremental (Akphange in reactivity due to the absorber.
' Holtec Intemational Appendix 4A, Page 4
e
- 4A.4 Miscellaneous and Minor Parameters 4A.4.1 - Reflector Material and Spacing PNL has performed a number of critical experiments with thick steel and lead reflectors.t Analysis of these critical experiments are listed in Table 4A.5 (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 Fuel Pellet Dinnwtar and f attica Pitch 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. (0.4% 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. I 4A.4.3 Soluble Boron Concentation Effects 1 Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO 5a) 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 j slightly conservative. j I t Parallel experiments with a depleted uranium reflector were also j. ; formed but not l included in the present analysis since they are no: nertinent to the Holtee rack design, i HoltecIntemational Apneadix 4A, Page 5 1 ,
I
)
l 4A.5 MOX Fuel . . . The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for j UO2fuel. However, a number of MOX critical experiments have been analyzed and the l results are shown in Table 4A.7. Results of these analyses are generally above a (, of 1.00, indicating that when Pu is present, both MCNP4a and KENO 5a 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 KENO 5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO 5a. It is also possible that the overprediction in (a for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This { l possibility is supported by the consistency in calculated (, over a wide tr.nge of the ( spectral index (energy of the average lethargy causing fission). l I
\
l Holtec Intemational . Appendix 4A, Page 6
i l i
'I - 4A 6 References ,
[4A.1] J.F. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N- { Particle Transport Code, Version 4A; Los Alamos National l Laboratory, LA-12625-M -(1993). I
. [4A.2] _ SCALE 4.3, "A Modular Code System for Performing Standardized '
Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/RS, Revision 5, Oak Ridge National .t
.. . Laboratory, September 1995. - - -
i 1 [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 Laboratoty, September 1994. [4A.4] W.C. Jordan et al., " Validation of KENOV.a", CSD/TM-238, Martin hbrietta 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, An Introduction to Mathematical l Statistics and its Applications, Prentice-Hall,1986. ' [4A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, I 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 & Wilcox Company, November 1991. [4A.9] LW. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984. Holtec Intemational Appendix 4A, Page 7
[4A.10]. J.C. Manaranch,e 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 2 Subcritical Clusters of 2.35 w/o and 4.31 w/o "U Enriched UO2 I 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 Suberitical - 2 Clusters of 2.35 w/o and 4.31 w/o "U Enriched UO2 Rods in Water with Uranium or lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December,1981. [4A.13] S.R. LLinan et al., Critical Separation Between Suberitical 2 Clusters of 4.31 w/o "U Enriched UO 2Rods 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 % 2"U Enriched UO2 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 1%5. [4A.18] M.G. Natrella, nrnerimental Statistics, National Bureau of Standards, Handbook 91, August 1%3. l Holtec Intemational . Apper. dix 4A, Page 8
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rt a i el pu A sr e xc 4 4 8 5 9 5 6 9 9 el a x P 3 0 6 i u N 6 9 2 6 6 1 3 1 1 6 6 8 7 4 4 8 2 8 3 2 9 2 5 3 0 ec g d n C 0 0 0 0 5 6 2 1 1 r ah e e p 0 0 0 M 0 0 0 l yg t p l l n % A 6 6 6 6 7 6 7 6 6 a uin i 0 0 0 0 0 0 0 s 0 0 0 0 0 0 0 0 u m 5 a 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 n er ut O 0 0 0 e N i i i i i C i i i f o d E 6 9 6 5 6 7 N 7 3 yi n 8 S n uK 3 0 0 8 9 9 6 9 9 0 0 0 5 9 9 4 0 0 6 9 9 3 1 0 0 0 0 i l t i e d o a. 1 0 0 1 0 1 0 1 1 ib n i i t s s ta a l u c 2 1 1 1 1 1 1 1 0 1 0 1 1 1 oe pr ee l c 0 0 0 0 1 0 1 0 1 0 1 0 1 0 1 ht re a 4 a 0 0 0 0 0 0 0 0 0 0 0 gw C P N 0 i 0 i 0 0 0 0 0 0 0 0 ny , k i i i i i i i t e r C 8 3 9 6 0 6 8 4 3 sh et M 9 a 5 0 8 0 7 0 9 9 0 0 3 0 0 0 9 9 6 0 3 0 g g m 0 1 0 1 0 1 9 0 0 1 0 1 0 1 9 0 0 1 0 um s s 1 h 1
)
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% 7 4
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) )
7
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c 6 1 6 1 6 1 4 A 4 A A 4 A 4 A 4 A A 4 r o f er o emu i n A A A ( ( ( ( ( 4( ( sh ehpl l a - e 4 ( 4( 4 5 5 5 5 5 5 5 t n r ( 8 8 8 8 8 8 8 d t o _ f e 3 0 3 0 3 0 3 3-3 3-3 3 3 3 3 3 ns aie xA i t e 8 8 8 3- - 3- 3- 3- t a _ R 5- 5- 5- P P P P P P P sFe . . r s n _ A A C Ls r _ L L L A A A A A e oi e Ah N N C C C C C C C r s r a N N t . n P P P W W W W W W W ETeb I s
- c et " e t
3 4 5 6 7 8 9 0 1 2 t l 5 5 5 5 5 5 5 6 6 6 o o N H
l I Table 4A.2 COMPARISON OF MCNP4a AND KENO 5a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS Calculated k, i la Enrichment MCNP4a KENO 5a 3.0 0.8465 i 0.0011 0.8478 i 0.0004 3.5 0.8820 i 0.0011 0.8841 0.0004 3.75 0.9019 i 0.0011 0.8987 i 0.0004 ! 4.0 0.9132 i 0.0010 0.9140 0.0004 4.2 0.9276 i 0.0011 0.9237 i 0.0004 > 4.5 0.9400 0.0011 0.9388 i 0.0004
' j l
l
.l l l
I i I t Based on the GE 8x8R fuel assembly. Holtec Intemational ppendix 4A; Page 14 l l
h. 1 l Table 4A.3 i
]
MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS
)
Ak MCNP4a l Worth of Calculated EALF' Ref. Experknent Absorber k, (eV) 4A.13 PNb2615 Boral Sheet 0.0139; 0.9994i0.0012 0.1165
'4A.7 B&W-1484 Core XX 0.0165 1.0008 i 0.0011 0.1724 .
4A.13 PNb2615 1.62% Boron-steel 0.0165 0.9996 i 0.0012 0.1161
]
4A.7 B&W-1484 Core XIX 0.0202- 0.9961 i 0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.0243 0.9994 i 0.0010 0.1544 4A.7 B&W-1484 Core XVII 0.0519 0.9962 i 0.0012 0.2083 ! 1 4A.11 PNb3602 Boral Sheet 0.0708 0.9941 i 0.0011 0.3135 ' 4A.7 B&W-1484 Core XV 0.0786 0.9910 i 0.0011 0.2092 1 4A.7 B&W-1484 Core XVI 0.0845 0.9935 i 0.0010 0.1757 4A 7 B&W-1484 Core XIV 0.1575 0.9953 i 0.0011 0.2022 4A.7 B&W-1484 Core XIII 0.1738 1.0020 i 0.0011 0.1988 4A.14 PNb7167 Expt 214R flux trap 0.1931 0.9991 i 0.0011 0.3722
'EALF is the energy of the average lethargy causing fission.
Holtec Internationa ppendix 4A, Page 15
l I l i Table 4A.4 1 COMPARISON OF MCNP4a AND KENO 5a 3 CAlfULATED REACTIVITIESt FOR VARIOUS B LOADINGS l Calculated k,, 10
' B, g/cm2 MCNP4a KENO 5a I 0.005' 1.0381 0.0012 1.0340 i 0.0004 j I
0.010 0.9960 i 0.0010 0.9941 i 0.0004 1 0.015 0.9727 0.0009 0.9713 i 0.0004 0.020 0.9541 0.0012 0.9560 i 0.0004 0.025 0.9433 t 0.0011 0.9428 0.0004 0.03 0.9325 0.0011 0.9338 0.0004 0.035 0.9234 i 0.0011 0.9251 0.0004 0.04 0.9173 i 0.0011 0.9179 0.0004 ! I t Based on a 4.5% enriched GE 8x8R fuel assembly. Holtec International '. dppendix 4A, Page 16
t Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORSt i Separation, Ref. Case E, wt% cm MCNP4a h KENO 5a 4 4A.11 Steel 2.35 1.321 0.9980 i 0.0009 0.9992 i 0.0006 Reficctor 2.35 2.616 0.9968 i 0.0009 0.9964t0.0006 2.35 3.912 0.9974 i 0.0010 0.9980 i 0.0006 i 2.35 = 0.9962 i 0.0008 0.9939 i 0.0006 4A.11 Steel 4.306 1.321 0.9997 i 0.0010 1.0012 i 0.0007 Reflector 4.306 2.616 0.9994 i 0.0012 0.9974 i 0.0007 ! 4.306 3.405 0.9969 0.0011 0.9951 i 0.0007 4.306 - 0.9910 i 0.0020 0.9947 i 0.0007 4A.12 lead 4.306 0.55 1.0025 i 0.0011 0.9997 i 0.0007 Reflector 4.306 1.956 1.0000 i 0.0012 0.9985 i 0.0007 4.306 5.405 0.9971 i 0.0012 0.9946 i 0.0007 t i Arranged in order ofincreasing reflector-fuel spacing. Holtec International ~ Appendix 4 A, Page 17
r" i i I l l l i l i Table 4A.6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated Q Boron Concentration, ; Reference Experiment ppm MCNP4a KENO 5a l 4A.15 PNIe4267 , 0 0.9974 i 0.0012 - l l 4A.8 B&W-1645 886 0.9970 0.0010 0.9924 i 0.0006 4A.9 B&W-1810 1337 1.0023 i 0.0010 - I 4A.9 B&W-1810 1899 . 1.0060 i 0.0009 - 4A.15 PNIe4267 2550 1.0057 t 0.0010 - 1 I I f HoltecInternationa: Appendix 4A, Page 18
l Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNNa KENO $a Reference Case' h EALF" 4 EALF" PNIe5803 MOX Fuel - Exp. No. 21 1.0041i0.00ll 0.9171 1.0046 i 0.0006 0.8868 {4A.16) - MOX Fuel- Exp. No. 43 1.0058 1 0.0012 0.2968 1.0036 1 0.0006 0.2944 MOX Fuel- Exp. No.13 1.0083 i 0.0011 0.1665 0.9989 1 0.0006 0.1706 MOX Fuel- Exp. No. 32 1.0079 1 0.0011 0.1139 0.9966 1 0.0006 0.1165 WCAP. Saxton @ 0.52* pitch 0.9996i0.00ll 0.8665 1.0005i0.0006 0.8417 3385-54 {4A.17] Saxton 0 0.56" pitch 1.0036 i 0.0011 0.5289 1.0047t0.0006 0.5197 Saxton @ 0.56" pitch borated 1.0008 1 0.0010 0.6389 NC NC Saxton @ 0.79* pitch 1.0063i0.00ll 0.1520 0.1555 1.0133 i 0.0006 Note: NC stands for not calmhrut t Arranged in order ofincreasing lattice spacing. tt EALF is the energy of the average lethargy causing fission. 1 Holtec Intemational Appendix 4 A, Page 19 l
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0.84 iiii iii, iiii ii iii, iii ,,,, . iiii iiii iiii 0.84 0.86 0.88 0.90 0.92 0.94 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KEN 05A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS
I:
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/ i I 1.04 l = 1.03 ~ _ inns n/m n . ,., r
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F i i L 5.0 THERMAL-HYDRAULIC CONSIDERATIONS ; l l 5.1 Introduction l This section provides a summary of the methods, models, analyses, and numerical results used to L demonstrate compliance of the reracked Millstone Point Unit 3 (MP3) Spent Fuel Pool (SFP) and ! the Spent Fuel Pool Cooling and System (SFPCS) with the provisions of Section III of the USNRC OT Position Paper for Review and Acceptance of Spent Fuel Storage and IIandling Applications", (April 14,1978). The methods used here are similar to methods of thermal-hydraulic analysis that i have been used in other rerack licensing projects.-
' The thermal-hydraulic qualification analyses for the rack arrays may be broken down into the following categories:
)
- 1. Evaluation of the maximum decay heat load for the postulated discharge scenarios.
ii. Evaluation of the in-core hold times required to prevent exceeding the maximum j temperature limit, as a function of component cooling water temperature. ., I l iii. Evaluation of the postulated loss-of-forced cooling scenarios to establish that pool boiling will not occur. iv. Determination of the maximum temperature difference between the pool local temperature and the bulk pool temperature at the instant when the bulk temperature reaches its maximum value, to establish that nucleate boiling at any location around the fuel is not possible with forced cooling available,
- v. Evaluation of the maximum temperature difference between the fuel rod cladding temperature and the local pool water temperature to establish that departure from HOLTEC INTERNATIONAL Millstone Point Unit 3 5-1
m nucleate boiling (DNB) at any' location around the fuel is not possible with forecd l cooling available. i l ~ A previous licensing submittal [5.1.1] has addressed items i through iii, above. The previous submittal is incorporated, by reference, into this document which addresses items iv and v. The
; following sections present the plant system description, analysis assumptions, a synopsis of the analysis methods employed, and final results.
5.2 . System Description The fuel pool cooling and purification system removes decay heat from spent fuel stored in the fuel l pool and provides adequate cooling of water in the fuel pool. Two 100% capacity fuel pool coohng j pumps and two 100% capacity fuel pool coolers are provided to ensure 100-percent redundant l cooling capacity. This portion of the system is Seismic Category I and Safety Class 3. The spent fuel pool water flows from the fuel pool discharge through either of the two fuel pool cooling pumps and through the tubeside of either fuel pool cooler, and then returns to the fuel pool. Table 5.2.1 lists the performanco chameteristics of the fuel pool cooling system. Cooling for the fuel pool coolers is provided by the reactor plant component cooling water system. Each pipe which enters the fuel pool has a vent hole drilled into the pipe to act as an anti-siphoning device or terminates at an elevation above these vent holes. These provisions prevent siphoning of the fuel pool water to less than 10 feet above the spent fuel. One pump and one cooler are sufficient to maintain the bulk pool temperatures to a maximum of 150 F for any long-term period. The bulk i peak temperature of the spent fuel pool is limited to 200*F for structural qualification of the spent fuel pool. Following a design basis accident with loss of power, the reactor plant component cooling water j system is not available to cool the spent fuel pool coolers until approximately four hours after the accident. Power from the emergency generators is not immediately available due to loading i considerations.- Pool cooling will be reinitiated at this time. I HOLTEC INTERNATIONAL Millstone Point Unit 3 5-2
Redundant safety grade fuel pool temperature indication is provided on the main control board. i Redundant safety class 3 level instruments are located in the fuel pool and can be read from the control room. They are set to provide indication before the water level falls below 23 feet above the top of the fuel racks. Piping penetrations are at least 11 feet above the top of the spent fuel so that - failure ofinlets, outlets, or accident piping leaks cannot reduce the water below this level. Normal makeup water to the spent fuel pool is the primary grade water system. Should primary grade water be unavailable, makeup water can be provided from the refueling water storage tank, a Seismic Category I source. Both of these systems connect to the spent fuel pool throt:gh the non-nuclear safety purification system. Water can also be provided from the hose station of the fire l protection system near the spent fuel pool. As an additional safety feature, a Seismic Category I, Safety Class 3 flow path is provided from the service water system. The fuel pool has redundant safety grade low level lights and temperature indicators provided in the main control room. Non-safety grade level indication is provided locally and high and low level l
- alarms are provided both locally and in the control room.
Local temperature indicators are provided on each fuel pool cooler outlet. Fuel pool cooler outlet high temperature is alarmed locally. Fuel pool cooler outlet flow is indicated, and low flow alarmed, locally, Fuel pool cooler instrumentation is non-safety grade. The fuel pool cooling pumps have control switches and indicating lights in the main control room. The discharges of all pumps have local pressure indicators. Upon high temperature at the pool, the plant will respond per procedural requirements. The cooling pumps can be operated manually l l' either from the control room or the switchgear. The purification pumps are operated locally. l 5.3 Discharge / Cooling Alignment Scenarios ! - The Millstone Unit 3 spent fuel pool is designed to meet the following post-reactor shutdown fuel discharge scenarios. HOLTEC INTERNATIONAL Millstone Point Unit 3 5-3 , 1
o s l Case 1: Scheduled Full-Core Offload p One full core (193 assemblies) is off-loaded to the pool after one year of operation at full power.
- Case 2: Unscheduled Full-core Offload 1
L
- One full core (193 assemblies) is offloaded to the pool after a previous outage lasting for 10 days l . .
and followed by 36 days of operation at full power. Case 3: Partial Core Discharge This case is for a partial core discharge of up to 97 assemblies into the pool followed by loss of cooling for 4 hours. The temperature and decay heat loads in the pool at the start of loss of cooling correspond to the time at 600 hours after reactor shutdown. Component Cooling Water (CCP) temperature is assumed to be at an operating high temperature of 95 F. In Case 1 and Case 2 discharge scenarios, it must be demonstrated that peak bulk pool temperatures do not exceed 150 F temperature limit when normal cooling is operational with CCP supplied to . fuel pool heat exchanger. One fuel pool pump and one heat exchanger are assumed to be normally ! available for removing decay heat from the Millstone Unit 3 fuel pool for all scenarios. The two 100% capacity fuel pool cooling purnps and two 1004 capacity fuel pool coolers are able to provide completely redundant cooling capacity. i The CCP system, following a design basis accident, is not available to cool the fuel pool for four hours. In the event ofloss of pool cooling, it must be demonstrated that the bulk pool temperature shall not exceed 200 F during this four-hour post LOCA heat up of the pool, l s l HOLTEC INTERNATIONAL Millstone Point Unit 3 5-4
5.4 - Decay Heat Load, In-Core Hold Time, SFP Heat-Up Time I 1 Section 4.0 of a previous licensing submittal [5.1.1] contains a description of the solution methodology used to evaluate the decay heat loads, in-core hold time requirements, and SFP heat- ' up. times for the MP3 SFP and SFPCS. Please refer to that document for a discussion of the solution methodology for these evaluations. Note that for conservatism reference [5.1.1] assumed a higher end of cycle discharge size than assumed in Table 1.2 herein. 5.5 Local Pool Water Temperature In this section, a summary of the methodology for evaluating the local pool water temperature is presented. A single conservative evaluation for a bounding amalgam of conditions is performed. The result of this single evaluation is a bounding temperature difference between the pool bulk temperature and the maximum local water temperature. In order to determine an upper bound on the maximum local water temperature, a series of ! conservative assumptions are made. The most important of these assumptions are: )
~ '_. For calculation of hydraulic resistance parameters, all racks are assumed to be Holtec designed Region 2 style racks. The lack of flux traps in this rack design minimizes the total !
flow area per stored assembly, thereby maximizing the hydraulic resistance and resultant temperatures. l 3 e With a full core discharged into the racks farthest from the coolant water inlet, the remaining cells in the spent fuel pool are postulated to be occupied with previously discharged fuel. e Y The hottest assemblies, located together in the pool, are assumed to be located in pedestal cells of the racks. These cells have a reduced ' vater entrance area, caused by the pedestal blocking the baseplate hole, and a correspondingly increased hydraulic resistance. HOLTEC INTERNATIONAL Millstone Point Unit 3 5-5
e. No downcomer flow is assumed to exist between the rack modules.
- All rack cells are conservatively assumed to be 50% blocked at the cell outlet to account for drop accidents resulting in damage to the upper end of the cells. This cell blockage is conservative, since structural evaluations have shown that only about 10% of the cell is
. blocked subsequent to the impact of dropped objects.
e The We'stinghouse 17x17 Std. assembly, which is most resistive to axial fluid flow, is assumed to populate the ' tire storage region. Thus, the hydraulic resistance to heat transfer is maximized. The inlet piping which returns cooled pool water from the SFPCS terminates above the level of the fuel racks. It is not apparent from heuristic reasoning alone that the cooled water delivered to the pool would not bypass the hot fuel racks and exit through the outlet piping. To demonstrate adequate cooling of hot fuel in the pool, it is therefore necessary to rigorously quantify the velocity i i field in the pool created by the interaction of buoyancy driven flows and water injectionkgress. A i Computational Fluid Dynamics (CFD) analysis for this demonstration is required. Tlv objective of this study is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled I
'i conditions in the pool is met for all postulated fuel discharge / cooling alignment scen,irios. The ;
l local thermal-hydraulic analysis is performed such that partial cell blockage and sligh! fuel assembly variations are bounded. An outline of the CFD approach is described in the following. i There are several significant geometric and thermal-hydraulic features of the MP3 SFP 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 assembli.s 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 the Vestinghouse 17x17 fuel assembly rod arrays in each rack cell. This situation shall HOLTEC INTERNATIONAL Millstone Point Unit 3 5-6
be modeled as a porous solid region in which fluid flow is governed by the classical Darcy's Law: BP p
=- V, - Cp lv l V L g
BX, K(i) 2 where SP/6Xi is the pressure gradient, K(i), V, and C are the corresponding permeability, velocity and inertial resistance parameters and p is the fluid viscosity. The permeability and inertial resistance parameters for the rack cells loaded with Westinghouse 17x17 fuel were determined based on the friction factor correlations for the laminar flow conditions typically encountered due to the low buoyancy induced velocities and the small size of the flow channels. The MP3 pool geometry requires an adequate portrayal oflarge 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 between the fuel rack outline and pool wall
. liner clearance knov n 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 7 .ixes 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 1 1 cells to capture the bulk and local features of the flow field. The distributed heat sources in the spent fuel pool racks are modeled by identifying distinct heat generation zones considering full-core discharge, bounding peak effects, and presence of background decay heat from old discharges. Three heat generating zones were modeled. The first consists of background fuel from previous discharges, the remaining two zones consist of fuel from a bounding full-core-discharge scenario. The two full core discharge zones are differentiated by one zone with higher than average decay heat generation and the other with less than average decay heat generation. The background decay heat load is determined such that the total decay heat load in the
' pool is equal to the calculated decay heat load limit. This is a conservative model, since all of the j fuel with higher than average decay heat is placed in a contiguous area. A uniformly distributed heat generation rate was applied throughout each distinct zone.
- HOLTEC INTERNATIONAL Millstone Point Unit 3 5-7
I-The CFD analysis was performed on the industry standard FLUENT [5.5.4] fluid flow and heat transfer moc 'ing program. The FLUENT code enabled buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying Reynolds
. Stresses to the mean bulk flow quantities with the following turbulence modeling options:
(i) k-e Model (ii) RNG k-c Model (iii) _Reynolds Stress Model l The k-e Model is considered most appropriate for the MP3 CFD analysis. The k-e turbulence model is a time-tested, general purpose turbulence model. This model has been demonstrated to give good results for the majority of turbulent fluid flow phenomena. The Renormalization Group (RNG) and Reynolds Stress models are more advanced models that were developed for situations where the k-c Model does not provide acceptable results, such as high viscosity flow and supersonic shock. The flow regime in the bulk fluid region is such that the k-e Model will provide acceptable results. Rigorous modeling of fluid flow problems requires a solution to the classical Navier-Stokes equations of fluid motion [5.5.1]. The governing equations (in modified form for turbulent flows I with buoyancy effects included) are written as: I ap,u, + ap,{u',u'j) B . ' a u, au~. s i at Ox, " & .# < ax, Ox, , BP OP,ku'ou'i)
- p, p (T - T ) g, + g where u, are the three time-averaged velocity components, p(u'i u';) are time-averaged Reynolds stresses derived from the turbulence induced fluctuating velocity components u'i, p, is the fluid density at temperature T , p is the coefficient of thermal expansion, u is the fluid viscosity, gi 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 , and a turbulent velocity scale k as shown below [5.5.2]:
~
HOLTEC INTERNATIONAL Millstone Point Unit 3 ' 5-8 l
I p{u', u',) = p k 6e - p, + 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 j methodology is known as the k-e model for turbulent flows as described by Launder and Spalding [5.5.3]. Some of the major input values for this analysis are summarized in Table 5.5.1. At elevation view 1 of the assembled CFD model is presented in Figure 5.5.1. Figures 5.5.2 and 5.5.3 present converged temperature contours and converged velocity vectors, respectively. 5.6 Fuel Rod Cladding Temperature ; i In this section, the method to calculate the temperature of the fuel rod cladding is presented. Similar to the local water temperature calculation methodology presented in the preceding section, this evaluation is performed for a single, bounding scenario. The maximum fuel cladding superheat j above the local water temperature is calculated. 1 The maximum specific power of a fuel array qi can be given by: q, = q F,, where: F,y = Radial peaking factor q = Average fuel assembly specific power, Btu /hr The peaking factors are given in Table 5.5.1. The maximum temperature rise of pool water in the most disadvantageously placed fuel assembly, defined as the one which is subject to the highest local pool water temperature, was computed for a bounding case. Having determined the maximum local water temperature in the pool, it is now possible to determine the maximum fuel cladding temperature. A fuel rod can produce F, times the average heat emission rate over a small length, where F, is the axial rod peaking factor. The axial heat distribution in a rod is generally a maximum HOLTEC INTERNATIONAL Millstone Point Unit 3 5-9
p in the central region, and tapers off at its two extremities. Thus, peak cladding heat flux over an l ' infinitesimal area is given by the equation:
, q F., F,
.. A, ! where A, is th6 total cladding external heat transfer area in the active fuel length region. Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards from bottom to top under laminar flow conditions. Rohsenow and Hartnett [5.6.1] . report a Nusselt-number based heat transfer correlation for laminar flow in a heated channel. The film temperature driving force (AT,) at the peak cladding flux location is calculated as follows: hr = Nu Kw ATr =hri where, hr is the water side film heat transfer coeflicient, D, is sub-channel hydraulic diameter, Kw is water thermal conductivity and Nu is the Nusselt number for laminar flow heat transfer. In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding - has a crud deposit resistance R (equal to 0.0005 ht-ft2 - FiBtu), which covers the entire surface. Thus, including the temperature drop across the crud resistance, the cladding to water local temperature difference (AT,) is given by: l A T = A Tr + R, q, 5,7 - Results I Section 5.0 of a previous licensing submittal [5.1.1) contains a summary presentation of the results ; of evaluations of the decay heat loads, in-core hold time requirements, and SFP heat-up times for ! the MP3 SFP and SFPCS. Please refer to that document for a discussion the results of these evaluations. A summary of the results of the local water and fuel cladding evaluation is presented below. i HOLTEC INTERNATIONAL Millstone Point Unit 3 5-10
L Consistent with our approach to make conservative assessments of temperature, the local water temperature calculations are performed for a pool with decay heat generation equal to the maximum calculated decay heat load limit. Thus, the local water temperature evaluation is a calculation of the temperature increment over the theoretical spatially uniform value due to local hot spots (due to the presence of a highly heat emissive fuel bundle). l The CFD study has analyzed a single bounding local thermal-hydraulic scenario. In this scenario, a i l bounding full-core discharge is considere( in which the 193 assemblies are located in the pool, farthest from the cooled water inlet, while the balance of the rack cells are postulated to be occupied by fuel from old discharges. The maximum temperature difference between the SFP bulk temperature and the peak local water temperature is conservatively calculated to be 41.67 F. The maximum temperature difference between the fuel cladding and the local water is calculated to be 36.31"F. Applying both calculated
- temperature differences to the bulk maximum normal operating pool temperature of 150 F [5.1.1]
yields a conservatively bounding 191.67 F maximum local water temperature and a conservatively ; bounding 227.98 F peak cladding temperature. Both the maximum local water and fuel cladding temperatures are lower than the 239.45 F local boiling temperature on top of the racks. Thus, boiling does not occur anywhere within the MP3 SFP. , i l I i HOLTEC INTERNATIONAL Millstone Point Unit 3 5-11
5.8 References
'[5.1.1] " Licensing Report for Reclassification of Discharge in Millstone Point Unit 3 Spent Fuel Pool,"llottec Report HI-971843, Revision 2.
[5.5.1] Batchelor, G.K., "An Introduction to Fluid Dynamics," Cambridge University Press,1967. l- [5.5.2] Hinze, J.O., " Turbulence," McGraw Hill Publishing Co., New York, NY,1975. L [5.5.3] Launder, B.E., and Spalding, D.B.," Lectures in Mathematical Models of Turbulence", Academic Press, London,1972. l [5.5.4] "QA Documentation and Validation of the FLUENT Version 4.32 CFD Analysis Program," ! Holtec Report HI-961444. [5.6.1] Rohsenow, N.M., and Hartnett, J.P., " Handbook of Heat Transfer", McGraw Hill Book Company, New York,1973. i t l llOLTEC INTERNATIONAL Millstone Point Unit 3 5-12 L_
E 1 i l Table 5.2.1 FUEL POOL COOLING AND PURIFICATION SYSTEM PRINCIPAL COMPONENT DESIGN CIIARACTERISTICS l Fuel Pool Cooling Pumps Quantity 2 Capacity (gpm) 3,500 llead (ft) 115 1 Design pressure (psig) 200 Design temperature (*F) 200 l Fuel Pool lleat Exchangers Quantity 2 Design heat load per exchanger (Btu /hr) 27.7 x 106 Reactor plant component cooling water flow per exchanger (gpm) 1,800 Reactor plant component cooling water inlet temperature (*F) 95 Reac'.or plant component coolir g water outlet temperature ( F) 126 Fuel pool cooling flow (gpm) 3,500 Fuel pool water inlet temperature ( F) 140 Fuel pool water outlet temperature (* F) 125 Tubeside design pressure (psig) 150 Design temperature ( F) 200 1 i llOLTEC INTERNATIONAL l Millstone Point Unit 3 5-13 l I
l TABLE 5.5.1 PRIMARY DATA FOR LOCAL TEMPERATURE EVALUATION Fuel Rod Outer Diameter 0.374 in. Rack Cell Inner Dimension 8.80 in. ! Active Fuel Length 144 in. SFPCS Water Flow Rate 3500 gpm i Fuel Radial Peaking Factor 1.70 Fuel Total Peaking Factor 2.60 SFP Length (North-South) Neglecting Southwest Area 355.82 in. Opposite Cask Pit SFP Width (East-West) 452.41 in. East Wall Minimum Rack-to-Wall Gap 4.17 in. ; West Wall Minimum Rack-to-Wall Gap 6.31 in. ; North Wall Minimum Rack-to-Wall Gap 3.17 in. i i' Minimum Rack-to-Floor (Bottom Plenum) Height 4.25 in. Rack Cell lleight (including baseplate) 170.0 in. SFP Floor Liner Elevation i 1 ft. & 3.25 in. SFPCS Inlet Pipe Elevation 46 ft. & 4 in. I SFPCS Inlet Pipe Diameter 12 in. Sch. 408 SFPCS Outlet Pipe Truncation Elevation 44 ft. & 5 in. SFPCS Outlet Pipe Diameter 10 in. Sch. 40S SFP Low Water Alarm Water Elevation 48 ft. & 11 in. l l IlOLTEC INTERNATIONAL Millstone Point Unit 3 5-14
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( ! l l- 6.0 STRUCTURAL / SEISMIC CONSIDERATIONS l L 6.1.' Introduction L This section considers the structural adequacy of the new maximum density spent fuel racks under l ( all loadings postulated for normal, seismic, and accident conditions at MP-3. The proposed pool J
. layout is shown in Figure 2.1, chapter 2.
i 1 As discussed in Chapter 1, the reracking of MP-3 involves the addition of fifteen new high density i racks to the existing capacity. At the time of the original rack installation, the state-of-the-art limited the seismic evaluation to single rack 3-D simulations. As we discuss in this chapter, it is now possible to model the entire assemblage of new rack modules in one comprehensive simulation known as the 3-D Whole Pool Multi-Rack (WPMR) analysis. In order to maintain continuity with the previous analysis methods, both single rack and WPMR analyses have been performed to ! establish the structural margins of safety in the MP-3 racks. The analyses undertaken to confirm the structural integrity of the racks are performed in compliance
.w ti li 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.
6.2 Overview of Rack Structural Analysis Methodology 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 motion of closely spaced adjacent structures. 1 HOLTEC INTERNATIONAL Millstone Point Unit 3 6-1 _ . . . . . - ..m..... . . . . - . . . .. . .. .. _. . .. -
1 U l Linear methods, such as modal analysis and response spectrum techniques, cannot accurately i 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. Both Whole Pool Multi-Rack (WPMR) and Single Rack analysis are used in this project to simulate a Lthe 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. l' Background of Analysis Methodology 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- i 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-off and 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 l 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 apriori. 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 ollen maximizes the computed rack displacements as well as the equivalent elastostatic stresses.
HOLTEC INTERNATIONAL > Millstone Point Unit 3 6-2
L 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.
The three-dimensional single rack dynamic model introduced by Holtec International personnel 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. I However, Single Rack analysis is still a valuable tool to examine the behavior of a rack under i different load conditions. It is used here as a first step in evaluating the racks. WPMR analysis i builds upon the Single Rack model. The worst case loads and stresses that result from either of these two models are used to determine the structural adequacy of the racks. ) The 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the array - i
. of variables as follows: '
Interface Coefficient of Friction 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.
HOLTEC INTERNATIONAL Millstone Point Unit 3 6-3
W l Rack Beam Behavios Rack elasticity, relative to the rack b'ase, is included in the model by introducing linear springs to represent the elastic bending action, twisting, and extensions. (
- Impact Phenomena Compression-only gap elements ure used to provide for opening and closing of L ~ interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface.
i l i These interface gaps are modeled using nonlinear spring elements. The term " nonlinear spring" is l a generic term used to denote the mathematical representation of the condition where a restoring force is not linearly proportional to displacement. Fuel Loading Scenarios The fuel assemblies are conservatively assumed to rattle in unison which j L obviously exaggerates'the contribution ofimpact against the cell wall. Fluid Couplinn Holtec International extended Fritz's classical two-body fluid coupling model to j l 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 [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, and Shearon Harris - plants [6.2.1,6.2.3] and, subsequently, in numerous other rerack projects. The WPMR analyses i 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.
6.3 Description of Racks HOLTEC INTERNATIONAL Millstone Point Unit 3 l 6-4
n The storage capacity expansion of the MP-3 spent fuel pool features a two region arrangement. In the proposed scheme, five modules will store the most reactive fuel (up to 5 % w/o) without any
)
i burnup limitation. These racks will use a flux-trap design. The grouping of flux-trap racks is I referred to as Region 1. The remaining new racks do not use flux-traps and are collectively referred l . to as Region 2. Region 2 racks have an enrichment /burnup limitation. 3 l l l 6.3.1 Fuel Weights For the dynamic rack simulations, the dry fuel weight is conservatively taken to be 1700 lbs. The actual fuel assembly weight is approximately 1482 lbs. The higher fuel weight value of 1700 lbs is used to account for rod control cluster assemblies (RCCAs) being stored along with fuel assemblies. Therefore, the analyses conservatively consider an RCCA to be stored along with an assembly at < every location. l 6.4 Synthetic Time-Histories q 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 i which were generated over a 20 second duration for SSE and OBE events, respectively. These artificial time-histories are used in all non-linear dynamic simulations of the racks. Results of the correlation function of the three time-histories are given in Table 6.4.1. Absolute 1 values of the correlation coefficients are shown to be less than 0.15, indicating that the desired l HOLTEC INTERNATIONAL Millstone Point Unit 3 6-5
i' statistical independence of the three data sets has been met. 1 i 6.5 . 22-DOF Nonlinear Rack Model for Dynamic Analysis l 6.5.1 General Remarks The single rack 3-D model of the MP-3 racks has been prepared with due consideration of the following characteristics, which are typical of high-density modules designed by Holtec International.
.'i.- As a continuous structure, the rack possesses an infinite number of degrees-of-freedom, of which the cantilever beam type modes are most pronounced unde.-
seismic excitation if the rack is of the honeycomb construction gents. (The MP-3 racks, like all prior Holtec designs, are of the honeycomb type.) ii. The fuel assemblies are " nimble" structures with a relatively low beam mode fundamental frequency. iii. The interstitial gap between the storage cells and the stored fuel assemblies leads to a rattling condition in the storage cells during a seismic event. ) iv. The lateral motion of the rack due to seismic input is resisted by the pedestal-to-pool slab interfacial friction and is abetted or retarded by the fluid coupling forces , produced by the proximity of the rack to other structures. (The fluid coupling forces ! are distinct from the nonconservative forces such as fluid " drag" which are, by NRC l regulations, excluded from the analysis). The construction of a 3-D single rack dynamic model consists of modeling the rack as a multi-degree-of-freedom structure such a manner that the selected DOFs capture all macro-motion modes of the rack, such as twisting, overturning, lift-off, sliding, flexing, and combinations thereof. i
~ Particular attention must be paid to incorporating the potential for the friction-resisted sliding of the rack on the liner, lift-off and subsequent impact of the pedestals on the slab, collision of the rack with adjacent structures, and most important, rattling of the fuel in the storage cells. The dynamic model must also provide for the ability to simulate the scenarios of partially loaded racks with arbitrary loading patterns.
HOLTEC INTERNATIONAL Millstone Point Unit 3 - 6-6
i As the name implies, the single rack (SR) dynamic model is a 3-D structural model for one rack in the pool. The rack selected for the SR analysis in this p.uject is the one with the most mass, or most non-square cross section (i.e., aspect ratio). The dynamic model of this rack, i.e., its structural stiffness characteristics, rattling effect of the stored fuel, etc., can be prepared with extreme l diligence in the manner described in the following, resulting in an excellent articulation of the rack I l structure. Even the fluid coupling effects between the fuel assemblies and the storage cell can be modeled with acceptable accuracy [6.5.2]. If the rack is adjacent to a wall, the fluid coupling effects between the rack and the wall can also be set down deterministically because the wall is a fixed
. structure. Such a definitive situation does not exist, however, when the neighboring structure to the subject rack is another free-standing rack. During a seismic event, the subject rack and the neighboring rack will both undergo dynamic motions which will be governed by the interaction among the inertia, fluid, friction, and rattling forces for each rack. The fluid coupling forces between two racks, however, depend on their relative motious. Because the motion of the neighboring rack is undefined, it is not possible to characterize the hydrodynamic forces arising from the fluid coupling between the neighboring rack and the subject rack. This inability to accurately model the inter-rack fluid coupling effects is a central limitation in the single rack analysis.
)
. To overcome this limitation intrinsic to the single rack solutions, an artificial boundary condition, referred to as the "out-of-phase" assumption, has been historically made to bound the problem.
In the opposed-phase motion assumption, it is assumed that all racks adjacent to the subject rack are { vibrating 180 out-of-phase, resulting in a plane of symmetry between the subject rack and the I neighboring rack across which water will not flow. Thus, the subject rack is essentially surrounded by a fictitious box with walls that are midway to the adjacent racks. Impact with the adjacent rack is assumed to have occurred if the subject rack contacts the " box wall". In summary, in the opposed-phase motion analysis the analyst makes the election that the adjacent racks are moving at 180 out-of-phase from the subject rack at all times during the seismic event. This is an artificial technical construct, albeit one that is known to predict rack-to-rack impact conservatively. .
. Therefore, to maintain consistency with past analyses, an array of single rack 3-D simulations were l carried out, principally to compare the results (viz., rack-to-rack impact, maximum primary stress l levels, pedestal loads, etc.) with the more definitive WPMR analysis. The description below provides the essentials of the 22 DOF model for a single rack.-This model is used in both 3-D single j l_ HOLTEC INTERNATIONAL i
Millstone Point Unit 3 6-7
I[ rack simulations and as the building block for the more complicated WPMR analyses, described later in this chapter,
. The dynamic modeling of the rack structure is prepared with special consideration of all 'I
'nonlinearities and parametric variations. Particulars of modeling details and assumptions for the rack analysis are given in the following
! a. 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 captu.ed in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom and simulate rack stiffness. ,;, b.' 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- i 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. q l
- c. 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-l phase within a rack. This exaFgerates computed dynamic loading on the rack structure and, therefore, yields conservative results.
- d. . 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.
- e. Fluid damping and form drag are conservatively neglected.
- f. Sloshing is found io be negligible at the top of the rack and is, therefore, neglected in i the analysis of the rack.
- g. 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
. HOLTEC INTERNATIONAL Millstone Point Unit 3 6-8
7 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,
- h. . Pedestals are modeled by gap elements in the vertical direction and as " rigid links" for transferring horizontal stress. Each pedestal support is lini.ed 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.
. i. 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 coefficiems are based on the nominal gap ir order to provide a conservative measure of fluid resistance to gap closure.
- j. The model for the rack is considered supported, at the base level, on four pedestal.
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 arbitmry rocking and sliding motions. 6.5.2 Element Details Figure 6.5.1 shows a schematic of the dyna.nic 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 rutraint elements. Table 6.5.1 provides a complete listing of each of the 22 degrees-of-freedom for a rack model. Six transitional 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* in Figure 6.5.1) are described by ten horizontal transitional 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, HOLTEC INTERNATIONAL Millstone Point Unit 3 6-9
1 t I are provided at cach fuel mass. !
' Figure 6.5.3 pravides a 2-D schematic elevation of the storage rack model, discussed in more detail . i J 2ect'.on 6.5.3. This _ view shows the vertical location of the five storage masses and some of the support pedestal spring members.
Figure 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]. Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model. Figure 6.5.5 depicts the inter-rack impact springs (used to develop potential impact loads between racks or between rack and wall). The approximate spring contact location at rack top and bottom and the numbering of each impact spring used in the model are shown in Figure 6.8.1 and Figure
- 6.8.2.
'6.5.3 Fluid Coupling Effect 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. If one body (mass mi) vibrates adjacent to a second body (mass m2), and both bodies are submerged in frictionless fluid, then Newton's equations of motion for the two bodies are: . (mi+ M ) $ +i Mn $ = applied ii 2 forces on mass mi + 0 (X,2)
M 2iX + i(m2 + M ) X 22 = applied 2 forces on mass m2 + 0 (X 2*) 5 ,i and $ denote2 absolute accelerations of masses mi and m2, respectively, and the notation 2 O(X ) denotes nonlinear terms. i M , Mu, M , 2iand M are ii 22 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 g l fluid adds mass to the body (M to mass mi), and an inertial force proportional to acceleration of ii HOLTEC INTERNATIONAL Millstone Point Unit 3 6-10
the adjacent body (mass m2). Thus, accelerauon of one body affects the force field on another. This force field is a function ofinter-body gap, reaching large values for small gaps. Lateral motion of a
' fuel assembly inside a storage location encounters this effect. 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 computed assuming that all racks adjacent to the rack being analyzed are vibrating in-phase or 180 out of phase. The WPMR analyses do not require any assumptions with regard to phase. 6.5.4 Stiffness Element Details Table 6.5.2 lists all spring elements used in the 3-D 22-DOF single rack model. It helps to explain the stiffness details. In the table, the following coordinate system applies: x= Ilorizontal axis along plant North y= liorizontal axis along plant West z= Vertical axis upward from the rack base if the simulation model is restricted to two dimensions (one horizontal motion plus one vertical l l motion, for example), for the purposes of model clarification only, then Figure 6.5.3 describes the I
- 1. 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 *.: rattling fuel mass at the rack top. Support pedestal spring rates Ks are modeled by type 3 gap 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 Tale 6.5.2 are shown in Figure 6.5.3 as Kr. The spring elements depicted in Figure 6.5.4 represent type I elements. Friction at support / liner interface is modeled by the piecewise linear friction springs with suitably llOLTEC INTERNATIONAL Millstone Point Unit 3 6-11
large stiffness K, up to the limiting lateral load pN, where N is the current compression load at the interface between support and liner. At every time-step during transient analysis, the current value of N (either zero if the pedestal has lifted off the liner, or s 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. 6.6 Whole Pool Multi-Rack Methodology 6.6.1 General Remarks The single rack 3-D (22-DOF) models for the new racks outlined in the preceding subsection are used as a first step to evaluate the structural integrity and physical stability of the rack modules.
- However, prescribing the motion of the racks adjacent to the module being analyzed is an l
l assumption in the single rack simulations which cannot be defended on the grounds of conservatism. For closely spaced racks, demonstration of the kinematic compliance is further verified by including all modules in one comprehensive simulation using a Whole Pool Multi-Rack ! (WPMR) model. The WPMR analysis builds on the Single Rack model by simultaneously modeling all racks; a coupling effect results due to the multi-body motion. l l 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: i
- a. Prepare 3-D dynamic models suitable for a time-history analysis of the new maximum density racks. These models include the assemblage of all new rack modules in the pool. Include all fluid coupling inte: actions 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. HOLTEC INTERNATIONAL Millstone Point Unit 3 6-12
- b. Perform 3-D dynamic analyses on various physical conditions (such as coefficient of i
friction and extent of cells containing fuel assemblies). Archive appropriate ; displacement and load outputs from the dynamic model for post-proces;ing. l
- 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.
6.6.2 Multi-Body Fluid Coupling During the seismic event, all racks in the pool are subject to the input excitation simultaneously, i The motion of each free-standing module would be autonomous and independent of others as long l as they do not impact each other and no water is present in the pool. While the scenario ofinter-l 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. [6.5.2,6.5.3], 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 i 1 inherently inaccurate. The Whole Pool Multi-Rack model removes this intrinsic limitation of the ' 1 rack dynamic models by simulating the 3-D motion of all modules simultaneously. The fluid j coupling effect, therefore, encompasses interaction between every set of racks in the pool, i.e., the motion of one rack produces fluid forces en 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. l The derivation of the fluid coupling matrix [6.6.2] relies on the classical inviscid fluid mechames 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 [6.6.2]. 6.6.3 Coefficients of Friction To eliminate the last significant element of uncertainty in rack dynamic analyses, muhiple IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-13
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 nonnal distribution characteristics. The assigned values are then held constant during the 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. I w- fD nt of friction ( ) between the pedestal supports and the pool floor is indeterminate.
. Rabinowicz [6.5.1], results of 199 tests performed on austenitic stainless steel plates suomerged in water show a mean value of 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),0.8 (upper limit), and for random friction values clustered about a mean of 0.5. The bounding values of = 0.2 and 0.8 have been found to envelope the upper limit of module response in previous rerack projects.
6.6.4 Governing Equations of hiotion Using the structural model discussed in the foregoing, equations of motion corresponding to each degree-of-freedom are obtained using Lagrange's Formulation [6.6.1]. The system kinetic energy includes contributions from so!id structures and from trapped and surrounding fluid. The final I system of equations obtained have the matrix form: l l j d'q (M] _ g ,, __ =M+M t it is noted that DYNARACK 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. IlOLTEC INTERNATIONAL Millstone Point Unit 3 6-14 l
E.: where: l [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). q - the nodal displacement vector relative to the pool slab displacement (the term wi9 q indicates the second derivative with respect to time,
. i.e., acceleranon)
[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 The above column vectors have length 22n. The equations can be rewritten as follows:
= [Ml' [Q] + [Ml' [G]
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 DYNARACK [6.2.4]. 6.7 Structural Evaluation of Spent Fuel Rack 6.7.1 Kinematic and Stress Acceptance There are two sets of criteria to be satisfied by the rack modules:
- a. Kinematic Criteria 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:
e 1.5 times the upset seismic loading condition is applied. I e 1.1 times the faulted seismic loading condition is applied. ' HOLTEC INTERNATIONAL Millstone Point Unit 3 6-15
- b. Stress Limit Criteria Stress limits must'not be exceeded under the postulated load combinations provided herein.
6.7.2 Stress Limit Evaluations The stress limits presented below apply to the rack structure and are derived from the ASME Code, i Section 111, Subsection NF [6.7.1]. Parameters and terminology are in accordance with the ASME Code.~ Material properties are obtained from the ASME Code Appendices [6.7.2], and are listed in Table 6.3.1. (i) Normal and Unset Conditions (Level A or Level B) l
- a. Allowable stress in tension on a net section is:
F, = 0.6 S, Where, S, = yield stress at temperature, and F, is equivalent to primary membrane j stress. '
- b. Allowable stress in shear on a net section is:
F, = .4 Sy
- c. Allowable stress in compression on a net section kA.
- F. = S, 47 g 444r, kA/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.
A=' 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) HOLTEC INTERNATIONAL - Millstone Point Unit 3 ' 6-16
p n
= ~ 1/2'(clamped at both ends) r= radius of gyration of component
- d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is:
l l F, = 0.60 S y (equivalent to primary bending)
- e. Combined bending and compression on a net section satisfies:
L , C., f , , C , /s, 3 F, D, Fs, D, Fs, where: f, = Direct compressive stress in the section f,, = Maximum beuding stress along x-axis f,y = Maximum bending stress along y-axis C., = 0.85
= 0.85 C,y D, = 1 - (f/F,,)
Dy = 1 - (f/P,y)
= 2 P,,,,, (x E)/(2.15 (ki/r)2,,,)
E= Young's Modulus { and subscripts x.y reflect the particular bending plane. l
- f. Combined flexure and compression (or tension) on a net section:
I" + + <LO 0.6 S, Fs, Fs, The above requirements are to be met for both direct tension or compression. I
- g. Welds I
Allowable maximum shear stress on the net section of a weld is given by: l F = 0.3 S, L where S, is the weld material ultimate strength at tempeia'ure. For fillet weld legs in contact with base metal, the shear stress on the gross section is Hmitca to 0.4Sy, HOLTEC INTERNATIONAL Millstone Point Unit 3 - 6-17
where S, is the base material yield strength at temperature. (ii).. Level D Service Limits Section F-1334 (ASME Section III, Appendix F) [6.7.2], states that the limits for the Level D condition are the minimum of 1.2 (S,/FJ or (0.7S,/F,) times the corresponding limits for l
' the Level A condition. S, is ultimate tensile stress at the specified rack design temperature. I 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. j J Exceptions to the above general multiplier are the following: a) Stresses in shear shall not exceed the lesser of 0.72S, or 0.42S,. In the case of the Austenitic Stainless material used here,0.72S, governs. 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. 1 d) For welds, the Level D allowable maximum weld stress is not'specified in Appendix F - f of the ASME Code. An appropriate limit for weld tiroat stress is conservatively set here ! as: F, = (0.3 S.) x factor where: factor = (Level D shear stress limit)/(Level A shear stress limit) 6.7.3 Dimensionless Stress Factors 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 HOLTEC INTERNATIONAL Millstone Point Unit 3 6-18
y limiting value of each stress factor is 1.0, based on the allowable strengths for each level, for L-evels A, B, and D (where 1.2S; < .7S ).; 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) 1
'R=
2 Ratio of gross shear cn a net section in the x-direction to its allowable value RA = -! Ratio of maximum x-axis bending stress' to its allowable value for the section R4= ~ Ratio of maximum y-axis bending stress to its allowable value for the section
-R 3= Combined flexure and compressive fac:or (as defined in the foregoing) . R[= Combined flexure ar,d tension (or compre_. vn) factor (as defined in the foregoing)
R= 7 Ratio of gross shear on a net section in the y-direction to its allowable value 6.7.4 Loads and Leading Combinations for Spent Fuel Racks The applicable loads and their combinations which must be considered in the seismic analysis of rack modules is 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 o D + L + T, + E ' D t L + T, + E Level B D + L + T, + Pr ! D .+ L + T, + E' Level D l D + L + T, + F, h bMd WMW Ch M *h must be demonstrated. HOLTEC INTERNATIONAL Millstone Point Unit 3 6-19 o
p Where: l 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 F, = Impact force from accidental drop of the heaviest load from the maximun, possible height. E= Operating Basis Earthquake (OBE) E' = Safe Shutdown Earthquake (SSE) T=o Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition) T, = Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions) T, and T oproduce 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 temperature difference I between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary (thermal) stresses. l 6.8 Seismic Analysis
. 6.8.1. Acceptance Criteria Only the SSE event based cases are selected for dynamic simulations if the maximum stress factors obtained from these cases are below the limit prescribed for i
OBE events. The maximum stress factor limit for OBE events is one half of the ; stress factor limit for SSE events Therefore, if the stress factors obtained from the SSE cases are less than 0.5 ther aey also meet the OBE stress factor limits and ! hence no further OBE runs are required. HOLTEC INTERNATIONAL Millstone Point Unit 3 6-20
m 6.8.2 Parametric Simulations Consideration of the parameters described earlier results in a number of scenarios for both the WPMR and the single rack' analyses. Using the criterion presented in 6.8.1, the number of scenarios can be conservatively reduced.- This analysis considers only SSE simulations since the results from
. these simulations meet the above acceptance criteria. Although not essential, one additional simulation (Run No. 7) is performed for comparison between the SSE and OBE results. This additional run is a re-run of most bounding SSE simulation with the OBE seismic time histories.
' The table below presents a complete listing of the simulations discussed herein. The Whole Pool Multi-Rack model considers all 11fleen new racks in the pool. In addition to this basic model, an interim configuration is also considered for the scenario when only the nine racks closest to the pool's West wall (see figure 2.1) are installed. This interim configuration is selected because of the large fluid gap, due to the absence of remaining new racks in the pool, weakens the hydrodynamic
. effect and, therefore, yields large rack displacements and pedestal loads. Rack number 3,4,5,8,9 and 10 (see Figure 6.8.1) are not considered in this model. The rack numbering scheme used to identify the racks for whole' pool multi rack (WPMR) simulation is introduced in Figures 6.8.1 or 6.8.2. Single rack analyses are performed to investigate the structural adequacy of the rack when subjected to an array of different fuel loading patterns (for example Fully loaded, partially loaded, etc.) and interface coefficient of frictions. Single rack simulations are also used to confirm the WPMR results and to determine the potential for rack overturning. In the evaluations, one rack from each region was chosen for the single rack analysis. Rack C1 (Region 2) and Rack D5 (Region 1) were selected, as they are the most slender, i.e. they have the highest aspect ratios in their respective regions. In addition to these single rack simulations, two single rack runs that exhibit the greatest displacement are re-run with severe earthquake conditions (1.5xSSE) for the purpose of checking the potential for rack overturning. Run no. 20 and 33 are such runs.
l LIST OF RACK SIMULATIONS (. Run Model Load Case COF Event 1 WPMR Full Pool 0.2 SSE !. HOLTEC INTERNATIONAL Millstone Point Unit 3 6-21 u
LIST OF RACK SIMULATIONS 1 Hun Model Load Case COF Event 2 WPMR Full Pool 0.8 SSE 3 WPMR Full Pool Random SSE 4 WPMR Interim Configuration 0.2 SSE 5 WPMR Interim Conliguration 0.8 SSE 6 WPMR Interim Configuration Random SSE 7 WPMR Full Pool Random OBE SINGLE RACK RUNS (Rack Cl) 8 Single Rack Fully Louded 0.2 SSE 9 Single Rack Fully Loaded 0.8 SSE 10 Single Rack Fiilly Loaded 0.5 SSE 11 Single Rack Nearly Empty 0.2 SSE 12 Single Rack Nearly Empty 0.8 SSE 13 Single Rack Nearly Empty 0.5 SSE 14 Single Rack IIalf-Full Rack (symmetric 0.2 SSE about diagonal) 15 Single Rack Half-Full Rack (symmetric 0.8 SSE about diagonal) 16 Single Rack Italf-Full Rack (symmetric 0.5 SSE , about diagonal) 17 Single .ack Italf-Full Rack (symmetric 0.2 SSE about short axis) 18 Single Rack lialf-Full Rack (symmetric 0.8 SSE about short axis) 19 Single Rack Half-Full Rack (symmetric 0.5 SSE about short axis) i IIOLTEC INTERNATIONAL i-Millstone Point Unit 3 6-22
LIST OF ltACK SIMULATIONS Run Model Load Case COF Event 20 Single Rack Case with max. 0.5 1.5 x SSE Displacement l IlOLTEC INTERNATIONAL Millstone Point Unit 3 6-23
P !;~ SINGLE RACK RUNS (Rack DS) 21 Single Rack Fully Loaded 0.2 SSE 22 Single Rack Fully Loaded 0.8 SSE 23 Single Rack Fully Loaded 0.5 SSE 24 Single Rack Nearly Empty 0.2 SSE 25 Single Rack Nearly Empty 0.8 SSE 26 Single Rack Nearly Empty 0.5 SSE 27 - Single Rack IIalf-Full Rack (symmetric about 0.2 - SSE diagonal) 28 Single Rack IIalf-Full Rack (symmetric about 0.8 SSE diagonal) 29 Single Rack lialf-Full Rack (symmetric about 0.5 SSE diagonal) 30 Single Rack IIalf-Full Rack (symmetric about 0.2 SSE short axis) 31 Single Rack . lialf-Full Rack (symmetric about 0.8 SSE l short axis) 32 Single Rack IIalf-Full Rack (symmetric about 0.5 SSE short axis) 33 Single Rack Case with max. displacement 0.5 1.5 x SSE Where: Random = Gaussian distribution with a mean of 0.5 coeflicient of friction (upper and lower limits of 0.2 and 0.8). Note that run no. 20 and 33 are re-runs of run no.10 and 23 except that the racks in these runs are simulated as an isolated rack in the pool as required by subsection 6.7.1. HOLTEC INTERNATIONAL Millstone Point Unit 3 6-24
i j 6.9 Time History Simulation Results The results from the DYNARACK runs may be seen in the raw data output files. 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. Further reduction in this bulk ofinformation is provided in this section by extracting the worst case values from the parameters ofinterest; 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. For each table, the COF column refers to the interface coefficient of friction discussed in subsection 6.2.1. The " Rack" column denotes racks by number (applicable to the DYNARACK model) for whole pool multi rack runs and by letter (applicable to the pool layout drawing) for single rack runs. I 6.9.1 Rack Displacements A tabulated summary of the maximum displacement for each simulation is provided below. Note that all of the maximum displacements occurred at the tops of the storage racks, as expected from swaying, bending, and tipping behavior. The location / direction terms defined as follows: uxt,uyt = displacement of top comei . ack, relative to the slab, in the North-South and East- l West directions, respectively. The maximum displacements for every simulation, I including the single rack tipover analyses, occurred at the top of the racks shown in the last table column. 1 RACK DISPLACEMENT RESULTS l Run Model COF Max. Location / Rack Displacement (xorV) 4 (inches) i Direction I 1 WPMR (full) 0.2 0.747 Top 13 2 WPMR (full) 0.8 0.75 Top 9 3 WPMR (full) Random 0.743 Top 5 4 WPMR (interim) 0.2 0.645 Top 2 IIOLTEC INTERNATIONAL, Millstor.s Point Unit 3 6-25
1 i RACK DISPLACEMENT RESULTS Run Model COF Max. Location / Rack Displacement (x or y) (inches) Direction 5 WPMR (interim) 0.8 1.03 Top 6 6 WPMR (interim) Random 0.745 Top 6 7 WPMR (full) Random 0.422 Top 13 SINGLE RACK RUNS (Rack Cl) 8 single rack (full) 0.2 0.38 Top Cl 9 single rack (full) 0.8 0.4193 Top Cl 10 single rack (full) 0.5 0.4254 Top Cl 11 single rack (nearly empty) 0.2 0.0719 Top Cl 12 single rack (nearly empty) 0.8 0.0714 Top Cl 13 single rack (nearly empty) 0.5 0.073 Top Cl 14 single rack (half) 0.2 0.235 Top Cl 15 single rack (half) 0.8 0.2851 Top Cl 16 single rack (half) 0.5 0.283 Top Cl 17 single rack (half-short axis) 0.2 0.2153 Top Cl 18 single rack (half-short axis) 0.8 0.2371 Top Cl 19 single rack (half-short axis) 0.5 0.2387 Top C1 20 single rack (overturning) 0.5 0.492 Top C1 SINGLE RACK RUNS (Rack D5) 21 single rack (full) 0.2 0.266 Top D5 22 single rack (full) 0.8 0.382 Top D5 23 single rack (full) 0.5 0.4062 Top D5 24 single rack (nearly empty) 0.2 0.0848 Top D5 25 single rack (nearly empty) 0.8 0.107 Top D5 l HOLTEC INTERNATIONAL Millstone Point Unit 3 6-26 l
RACK DISPLACEMENT RESULTS Run Model COF M a x. Location / ' Rack l Displacement h or V) l 1 i l (lnches) Direction 26 single rack (nearly empty) 0.5 0.1098 Top D5 1 l 27 single rack (half) 0.2 0.283 Top D5 28 single rack (half) 0.8 0.546 Top D5 29 single rack (half) 0.5 0.514 Top D5 30 single rack (half-short axis) 0.2 0.1594 Top D5 31 single rack (half-short axis) 0.8 0.217 Top D5 32 single rack (half-short axis) 0.5 0.2086 Top D5 33 single rack (overturning) 0.5 1.02 Top D5 l l The table shows that the maximum rack displacement is only 1.03 inches which occurs during, run No. 5. This small displacement indicates that rack overturning is not a concern. 4 l l 6.9.2 Pedestal Vertical Forces Pedestal number 1 for each rack is located in the northeast corner of the rack. Numbering increases counterclockwise around the periphery of each rack. The following bounding vertical pedestal forces are obtained for each run: IIOLTEC INTERNATIONAL l Millstone Point Unit 3 6-27
r 1 I MAXIMUM VERTICAL LOADS Run Model COF Event Max. Vertical Rack Load I 1 WPMR (full) 0.2 SSE 127000 10 2 WPMR (full) 0.8 SSE 146000 12 3 WPMR (full) Random SSE 147000 6 4 WPMR (interim) 0.2 SSE 125000 1 5 WPMR (interim) 0.8 SSE 143000 7 , i 6 WPMR (interim) Random SSE 145000 1 7 WPMR (full) Random OBE 124000 11 SINGLE RACK RUNS (Rack Cl) 8 single rack (full) 0.2 SSE 101000 Cl l 9 single rack (full) 0.8 SSE 108000 Cl 10 single rack (full) 0.5 SSE 109000 Cl 11 single rack (nearly empty) 0.2 SSE 16400 Cl 12 single rack (nearly empty) 0.8 SSE I8100 Cl 13 single rack (nearly empty) 0.5 SSE 18100 Cl 14 single rack (half) 0.2 SSE 60600 C1 15 single rack (half) 0.8 SSE 61400 Cl 16 single rack (hall) 0.5 SSE 61500 Cl 17 single rack (half-short axis) 0.2 SSE 54400 Cl 18 single rack (half-short axis) 0.8 SSE 67000 Cl l 19 single rack (half-short axis) 0.5 SSE 67000 C1 20 single rack (overturning) 0.5 1.5 x SSE - N/A N/A SINGLE RACK RUNS (Rack D5) HOLTEC INTERNATIONAL
- Millstone Point Unit 3 6-28
n , MAXIMUM VERTICAL LOADS Run Model COF Event Max. Vertical Rack Load 21 single rack (full) 0.2 SSE 104000 D5 22 single rack (full) 0.8 SSE 113000 D5 23 single rack (full) 0.5 SSE 111000 D5 24 single rack (nearly empty) 0.2 SSE 20900 D5 25 single rack (nearly empty) 0.8 SSE 27900 D5 26 single rack (nearly empty) 0.5 SSE 27800 D5 27 single rack (half) 0.2 SSE 67700 D5 28 single rack (half) 0.8 SSE 84400 D5 29 single rack (half) 0.5 SSE 78800 D5 30 single rack (half-short axis) 0.2 SSE 60800 D5 31 single rack (half-short axis) 0.8 SSE 70000 D5 , I 32 single rack (half-short axis) 0.5 SSE 69800 D5 l 33 single rack (overturning) 0.5 1.5 x SSE N/A N/A As may be seen, the highest pedestal load of 147,000 lbs. occurs in run 3 of the WPMR model. The effect of this load is evaluated in the bearing pad analysis. l l I 6.9.3 Pedestal Friction Forces l The maximum (x or y direction) shear load bounding all pedestals in the simulation are reported below and are obtained by inspection of the complete tabular data.
~
IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-29 l r
MAXIMUM IIORIZONTAL LOADS Run Model COF Event Max. Rack IIorizontal Load 1 WPMR (full) 0.2 SSE 23800 11
)
2 WPMR (full) 0.8 SSE 51700 6 ) i 3 WPMR (full) Random SSE 46900 6 l l 4 WPMR (interim) 0.2 SSE 23500 1 5 WPMR (interim) 0.8 SSE 43200 11 6 WPMR (interim) Random SSE 39800 11 1 7 WPMR (full) Random OBE 33300 13 l SINGLE RACK RUNS (Rack Cl) 8 single rack (full) 0.2 SSE 17900 Cl l 9 -- single rack (full) 0.8 SSE 35600 Cl l l 10 single rack (full) 0.5 SSE 30600 Cl i lI single rack (nearly empty) 0.2 SSE 2800 Cl 12 single rack (nearly empty) 0.8 SSE 5380 Cl 13 single rack (nearly empty) 0.5 SSE 5420 Cl 14 single rack (half) 0.2 SSE 11300 Cl i 15 single rack (half) 0.8 SSE 18500 Cl 16 single rack (half) 0.5 SSE 18800 Cl 17 single rack (half-short axis) 0.2 SSE 8890 Cl 18 single rack (half-short axis) 0.8 SSE $800 Cl 19 single rack (half-short axis) 0.5 SSE 22800 C1 20 single rack (overturning) 0.5 1.5 x SSE N/A N/A SINGLE RACK RUNS (Rack D5) IlOLTEC INTERNATIONAL Millstone Point Unit 3 6-30
)
MAXIMUM IIORIZONTAL LOADS ; Fun Model COF Event Max. Rack IIorizontal Load 21 single rack (full) 0.2 SSE 17600 D5 , 22 single rack (full) 0.8 SSE 31000 D5 23 single rack (full) 0.5 SSE 31200 D5 24 single rack (nearly empty) 0.2 SSE 3960 D5 25 single rack (nearly empty) 0.8 SSE 8640 D5 26 single rack (nearly empty) 0.5 SSE 8530 D5 l 27 single rack (half) 0.2 SSE 12700 DS 28 single rack (half) 0.8 SSE 19700 D5 29 single rack (half) 0.5 SSE 19000 D5 30 single rack (half-short axis) 0.2 SSE 10600 D5 31 single rack (half-short axis) 0.8 SSE 24900 D5 ; 32 single rack (half-short axis) 0.5 SSE 19400 D5 33 single rack (overturning) 0.5 1.5 x SSE N/A N/A l l l The largest pedestal load of 51,700 lbs occurs in run 2 of the WPMR model. The effect of this load is evaluated in the liner fatigue analysis. I l l 6.9.4 Rack Impact Loads ; e A freestanding rack, by definition, is a structure subject to potential impacts during a seismic event. I I ! 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 loLds. ! i IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-31
i l 6.9.4.1 Rack to Rack Impacts There is no rack to rack impact at rack top between any two racks during any of the seismic events.
- However, rack to rack impacts at the baseplate of the structure are predicted between Holtec racks.
There are no impacts between Ifoltec racks and Westinghouse racks during any of the seismic events. The maximum instantaneous impact forces at the baseplate are summarized below from all simulations performed. MAXIMUM RACK-TO-RACK (BASEPLATE) IMPACT Run Model Max. Impact Load (kips) 4 WPMR 20.67 l 21 Single 6.28 It may be noted that all impact loads occurred only at the bottom of the racks. i 6.9.4.2 Rack to WallImpacts Racks did not impact the pool walls under any simulation. 1 l 6.9.4.3 Fuel to Cell Wall Impact Loads A review of all simulations performed allows determination of the maximum instantaneous impact load between fuel assembly and fuel cell wall at any modeled impact site. The maximum fuel / cell j wall impact load values are reported in the following table. i HOLTEC INTERNATIONAL Millstone Point Unit 3 6-32
FUEL-TO-CELL WALL IMPACT H_u_!! Model COF Event Max. Impact Rack Load (ths) 1 WPMR (full) 0.2 SSE 802 15 2 WPMR (full) 0.8 SSE 752 15 3 WPMR (full) Random SSE 697 15 4 WPMR (interim) 0.2 SSE 630 15 , 5 WPMR (interim) 0.8 SSE 592 2 6 WPMR (interim) Random SSE 592 2 7 WPMR (full) Random OBE 432 9 SINGLE RACK RUNS (Rack Cl) 8 l single rack (full) 0.2 SSE 460 Cl 9 single rack (full) 0.8 SSE 443 Cl 10 single rack (full) 0.5 SSE 426 Cl 11 single rack (nearly empty) 0.2 SSE 510 Cl 12 single rack (nearly empty) 0.8 ' SSE 510 Cl 13 single rack (nearly empty) 0.5 SSE 510 Cl 14 single rack (half) 0.2 SSE 530 Cl 15 single rack (half) 0.8 SSE 533 Cl 16 single rack (half) 0.5 SSE 533 Cl 17_ single rack (half-short axis) 0.2 SSE 517 C1 18 single rack (half-short axis) 0.8 SSE 523 Cl 19 single rack (half-short axis) 0.5 SSE 523 C1 20 single rack (overturning) 0.5 1.5 x SSE N/A N/A SINGLE RACK RUNS (Rack DS) 21 single rack (full) 0.2 SSE 453 D5 IlOLTEC INTERNATIONAL Millstone Point Unit 3 6-33
1 l l l l FUEL-TO-CELL WALL IMPACT Run Mod.e_I COF Event Max. Impact Rack Load (Ibs~) 22 single rack (full) 0.8 SSE 462 D5 23 single rack (full) 0.5 SSE 462 D5 24 single rack (nearly empty) 0.2 SSE 488 D5 25 single rack (nearly empty) 0.8 SSE 450 D5 26 single rack (nearly empty) 0.5 SSE 450 D5 27 single rack (half) 0.2 SSE 473 D5 ~ 28 single rack (half) 0.8 SSE 478 D5 29 single rack (half) 0.5 SSE 480 D5 30 single rack (half-short axis) 0.2 SSE 520 D5 31 single rack (half-short axis) 0.8 SSE 477 D5 32 single rack (half-short axis) 0.5 SSE 477 D5 33 single rack (overturning) 0.5 1.5 x SSE N/A N/A The maximum Fuel-to-Cell Wall Impact is recorded to be 802 lbs. during run no.1. The structural integrity of the cell wall under the impact of this load must be evaluated. The discussion of this evaluation is provided in Section 6J 0.3. g gw r ,.'w2
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ng :g qfeg y, Y Mu & Umb.a,:a Mu%L $_En EY$ 6.10 Rack Structural Evaluation 1 6.10.1 Rack Stress Factors l l l With time history results available for pedestal normal and lateral interface forces, the maximum values for the previously defined stress factors can be determined for every pedestal in the array of racks. 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. With this information in hand, the maximum stress in the limiting rack cell (box) can be evaluated. From all of the simulations, the bounding stress factors for each run, in either cellular or the pedestal region, are summarized below : It is evident from the DYNARACK results for the stress factors that the maximum stresses occur at the cellular region to the baseplate (CRB) interface. The compressive stress in the CRB is principally due to the flexural motion of the rack module. In order to account for the possible reduction in the section modulus of the CRB section due to the localized compressive stress, we assume that the maximum compressive stress occurs over the entire CRB section. With this i e tremely conservative assumption, the stress magnifier per NF-3222.2 can be calculated and applied to the stress factors of the DYNARACK results. The table below IlOLTEC INTERNATIONAL Millstone Point Unit 3 6-35
1 incorporates the slenderness magnifier on the CRB section compressive stn .Jues called from the DYNARACK runs. l l~ M AXIMUM STRESS FACTORS Run Model COF Event Stress Factor Stress Factor i j Cell (CRB) Type / Rack 1 WPMR (full) 0.2 SSE 0.367 R6/5 2 WPMR (full) 0.8 SSE 0.403 R6/5 3 WPMR (full) Random SSE 0.401 R6/5 4 WPMR (interim) 0.2 SSE 0.377 R6/2 l l 5 WPMR (interim) 0.8 SSE 0.378 R6/7 6 WPMR (interim) Random SSE 0.383 R6/7 7 WPMR (full) Random OBE 0.522 R6/10 SINGLE RACK RUNS (Rack Cl) 8 single rack (full) 0.2 SSE 0.312 R6/C1 9 single rack (full) 0.8 SSE 0.340 R6/Cl 10 single rack (full) 0.5 SSE 0.343 R6/Cl 1I single rack (nearly empty) 0.2 SSE 0.051 R6/Cl 12 single rack (near!y empty) 0.8 SSE 0.055 R6/Cl l 13 single rack (near .,pty) 0.5 SSE 0.055 R6/Cl l 14 single rack > ' a f) 0.2 SSE 0.189 R6/Cl ! 15 single rack (half) 0.8 SSE 0.191 R6/Cl 16 single rack (half) 0.5 SSE 0.191 R6/Cl 17 single rack (ha'f-short axis) 0.2 SSE 0.168 R6/Cl 1 I8 single rack (half-short axis) 0.8 SSE 0.202 R6/Cl l 19 single rack (half-short axis) 0.5 SSE 0.202 R6/Cl 20 single rack (overturning) 0.5 1.5 x SSE N/A N/A HOLTEC INTERNATIONAL Millstone Point Unit 3 6-36
l SINGLE RACK RUNS (Rack D5) 21 single rack (full) 0.2 SSE 0.196 R6/D5 22 single rack (full) 0.8 SSE 0.208 R6/D5 23 single rack (full) 0.5 SSE 0.209 R6/D5
-24 single rack (nearly empty) 0.2 SSE 0.040 R6/D5 25 single rack (nearly empty) 0.8 SSE 0.050 R6/D5 26 single rack (nearly empty) 0.5 SSE 0.050 R6/D5 27 single rack (half) 0.2 SSE 0.129 R6/D5 28 single rack (half) 0.8 SSE 0.154 R6/D5 29 single rack (half) 0.5 SSE 0.144 R6/D5 t 30 single rack (half-short axis) 0.2 SSE 0.117 R6/D5 31- single rack (half-short axis) 0.8 SSE 0.129 R6/D5-32 single rack (half-short axis) 0.5 SSE 0.131 R6/D5 33 single rack (overturning) 0.5 1.5 x SSE N/A N/A Thus, the maximum stress factor in either pedestal or cellular region for SSE and OBE are 0.403 and 0.522, respectively. An evaluation of the stress factors for all of the simulations performed, leads to the conclusion that all stress factom are less than the mandated limit of 1.0 for the load cases examined. The : tress allowables are indeed satisfied for the load levels considered for every
, limiting la. ttion in every rack in the array. 6.10.2 Pedestal Thread Shear Stress The complete post-processor results give thread stresses under faulted conditions for every pedestal for every rack in the pool. The average shear stress in the engagement region is given below for the limiting pedestal in each simulation. ( llOLTEC INTERNATIONAL Millstone Point Unit 3 6-37 4
L TIIREAD SIIEAR STRESS j Run Model COF Event Stress Raek (PJD
~'
1 WPMR (full) 0.2 SSE 6162 10 2 WPMR (full) 0.8 SSE 7083 12 3 WPMR (ful;) Random SSE 7132 6 4 WPMR (interim) 0.2 SSE 6066 1
~
5 WPMR (interim) 0.8 SSE 6938 , 6 WPMR (interim) Random SSE 7035 1 7 WPMR (full) Random OBE 6016 11 SINGLE RACK RUNS (Rack Cl) 8 single rack (full) 0.2 SSE 4900 Cl
~9 single rack (full) 0.8 SSE 5240 Cl 10 . single rack (full) 0.5 SSE 5289 Cl
~
1I single rack (nearly empty) 0.2 SSE 796 Cl 12 single rack (nearly empty) 0.8 SSE 878 Cl 13 single re.ck (nearly empty) 0.5 SSE 878 Cl 14 single rack (half) 0.2 ; SSE 2940 Cl 15 single rack (balf) 0.8 SSE 2980 Cl 16 single rack (half) 0.5 SSE 2984 Cl 17 single rack (half-short axis) 0.2 SSE 2640 Cl 18 single rack (half-short axis) 0.8 SSE 3251 Cl 19 single rack (half-short axis) 0.5 SSE 3250 C1 20" single rack (overturning) 0.5 1.5 x SSE N/A N/A SINGLE RACK RUNS (Rack D5) l 21 single rack (full) 0.2 SSE 5046 D5 l I l IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-38
THREAD SHEAR STRESS Run Model COF Event Stress Rack t iPlil 22 single rack (full) 0.8 SSE 5483 D5 1 23 single rack (full) 0.5 SSE 5386 D5 24 single rack (nearly empty) 0.2 SSE 1014 D5 25 single rack (nearly empty) 0.8 SSE 1354 D5 26 single rack (nearly empty) 0.5 SSE 1349 D5 27 single rack (ha10 0.2 SSE 3285 D5 28 single rack (half) 0.8 SSE 4095 D5 29 single rack (half) 0.5 SSE 3823 D5 30 single rack (half-short axis) 0.2 SSE 2950 D5 31 single rack (half-short axis) 0.8 SSE 3396 D5 32 single rack (half-short axis) 0.5 SSE 3387 D5 33 single rack (overturning) 0.5 1.5 x SSE N/A N/A The ultimate strength of the female part of the pedestal is 66,200 psi. The yield stress for this material is 21,300 psi. The allowable shear stress for Level B (OBE) conditions is 0.4 times the yield stress which gives 8,520 psi and the allowable shear stress for level D is 0.72 times the yield stress which gives 15,336 psi. The maximum calculated shear stress value for the SSE is 7,132 psi and 6,016 psi for the OBE w. .a are less than their respective allowable values. Therefore, thread shear stresses are acceptable under all conditions. 6.10.3 Local Stresses Due to Impacts Impact loads at the pedestal base (discussed in subsection 6.9.2) produce stresses in the pedestal for which explicit stress limits are prescribed in the Code. However, impact loads on the cellular region of the racks, as discussed in subsection 6.9.4.3 above, produce stresses which attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses. IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-39
Even though limits on secondary stresses are not prescribed in the Code for class 3 NF structures, evaluations must be made to ensure that the localized impacts do not lead to plastic deformations in the storage cells which affect the suberiticality of the stored fuel array.
- a. Imoact Loadine Between Fuel Assembly and Cell Wall Local cell wall integrity is conw s :nively 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 3,187 lbf, including a safety factor of 2.0) is much greater than the highest calculated impact load value (of 802 lbf, see subsection 6.9.4.3) obtained froin any of the rack analyses. Therefore, fuel impacts do not represent a significant concern with respect to fuel rack cell deformation.
- b. Imoacts Between Adiacent Racks As may be seen from subsection 6.9.4.1, the bottom of the storage racks 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 i baseplate length of about 63 inches by its 3/4-inch thickness) . The resulting compressive stress from the highest impact load of 20,670 lbs. distributed over 47.25 sq. inches is only I
438 psi, which is negligible. This is a conservative computation, since the simulation assumes a local impact site. Therefore, any deformation will not effect the configuration of the stored fuel. Impact between the racks in the cellular region containing active fuel is shown not to occur. I 1 HOLTEC INTERNATIONAL Millstone Point Unit 3 6-40
6.10.4 Assessment of Rack Fatigue Margin Deeply submerged high density spent fuel storage racks arrayed in close proximity to each other in l
- a free-standing configuration behave primarily as a nonlinear cantilevered structure when subjected i
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 i friction-induced lateral forces act simultaneously in x and y directions with the requirement that l their vectorial sum does not exceed N, where is the limiting interface coefficient of friction and j i N 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 maximum friction force, F, that the interface can exert. In I other words, the lateral force at the pedestal / liner interface, F, is given by F s; p N (r) i i where N (vertical thrust) is the time-varying function of T. F does not always equal N; 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
.1 earthquake event.
l 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 if the amplitude of the stress cycles is j 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.
HOLTEC INTERNATIONAL Millstone Point Unit 3 6-41
.~.
L l Because the locations of maximum stress (viz., the pedestal / rack baseplatejunction) 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.
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, U, can be determined using the classical Miner's rule U = I "'-N, where ni is the number of stress intensity cycles of smplitude o,, and N, is the permissible number of cycles corresponding to ai from the ASME fatigue curve for the material of constructien. 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 6.10.1 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
- peruits establishing a set of alternating stress intensity ranges versus cycles for an SSE and an OBE event. Following ASME Code guidelines for computing U, it is found that U =0.92 due to the combined effect of one SSE and twenty OBE events. This is below the ASME Code limit of 1.0.
6.10.5 Weld Stresses i HOLTEC INTERNATIONAL Millstone Point Unit 3 6-42 i
)
[ l l l l Weld locations subjected to significant seismic loading are at the bottom of the rack at the i baseplate-to-cell connection, at the top of the pedestal support at the baseplate connection, and at l i j ' cell-to-cell connections. Bounding values of resultant loads are used to qualify the connections. Table 6.9.1 provides the comparison of calculated stress vs. allowable stress.
- a. Baseolate-to-Rack Cell Welds l -
i The highest predicted weld stress for SSE is calculated from the set of forces Fx, Fy and Fz at the Cell Baseplate interface when R6 (defined above in 6.10.1) is maximum. The weld between the cell and the baseplate is checked to determine that the maximum weld stress under SSE event is 11,520 psi. This value is less than the permissible allowable value of 35,748 psi.
- b. Basentate-to-Pedestal Welds The weld between baseplate and support pedestal is checked to determine that the maximum stress under the SSE and the OBE event are 6,975 psi and 4,194 psi respectively. These calculated stress values are well below the SSE and OBE allowable of 35,748 psi and 19,860 psi, respectively.
- c. Cell-to-Cell Welds Cell-to-cell connections are formed by 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. ; l 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 ;
l weld. Table 6.9.1 gives results for the maximum allowable load that can be transferred by 1 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
- HOLTEC INTERNATION AL
, - Millstone Point Unit 3 6-43
upper bound value is very conservatively obtained by applying the bounding rack-to-fuel impact load from any simulation in two orthogonal directions simultaneously, 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 the allowable load. l l 6.11 Level A Evaluation i The Level A condition is not a governing condition for spent fuel racks since the general level ofloading is far less than Level B loading. To illustrate this, the heaviest spent fuel 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 MAXIMUM PEDESTAL LOAD I l Dry Weight of Largest Holtec Rack (Region 1) = 18,050 lbf Dry Weight of 70 Fuel Assemblies = 119,000 lbf Total Dry Weight = 137,050 lbf Total Buoyant Weight (0.87 x Total Dry Weight) = 119,233.5 lbf Load per Pedestal = 29,808 lbf I The stress allowables for the normal condition is the same as for the upset condition, which resulted in a maximum pedestal load of 147,000 lbs. Since this load (and the corresponding stress thro'ughout the rack members) is much greater than the 29,808 lb load calculated above, the seismic condition c 'trols over normal (Gravity) condition. Therefore, no further evaluation is performed. 6.12 Hydrodynamic Loads on Pool Walls The maximum hydrodynamic pressures (in psi) that develop between the fuel racks and the spent fuel pool walls develops for the case of the rack that exhibits the largest displacement. This has l HOLTEC INTERNATIONAL Millstone Point Unit 3 - 6-44
e: p been done for both the SSE and OBE cases. The results for these worst case conditions are shown i. in the table below. l l Case Maximum Pressure (psi) SSE 7.92 Oi3E 4.31 i i These 1 y> mdynamic pressures were considered in the evaluation of the Spent Fuel Pool structure. i l i l I l l HOLTEC INTERNATIONAL Millstone Point Unit 3 6-45
p 6.13 Conclusion ) l Thirty-three discrete freestanding dynamic simulations of maximum density spent fuel storage racks 1 have been perfonned to establish the structural margins of safety. Of the thirty -three parametric analyses, four simulations consisted of modeling all 15 fuel racks in the pool in one comprehensive t l Whole Pool Multi Rack (WPMR) model. Three additional runs were performed for interim j
- configuration case. The remaining twenty-six runs were carried out with the classical single rack 3-D model. The parameters varied in the difTerent runs consisted of the rack / pool liner interface coefficient of friction, extent of storage locations occupied by spent nuclear fuel (ranging from nearly empty to full) and the type of seismic input (SSE or OBE). Maximum (maximum in time !
and space) values of pedestal vertical, shear forces, displacements and stress factors (normalized stresses for NF class 3 linear type structures) have been post-processed from the array of runs and l summarized in tables in this chapter. The results show that: (i) All stresses are well below their corresponding "NF" limits.
-(ii) There is no rack-to-rack or rack-to-wall impact anywhere in the cellular region of the rack modules (iii) The rack overturning is not a concern.
l An evaluation of the fatigue expenditure in the most stressed location in the rpost heavily loaded rack module under combined effect of one SSE and twenty OBE events showr, that the Cumulative Damage Factor (using Miner's rule) is below the pennissible value of 1.0. In conclusion, all evaluations of struebral safety, mandated by the OT Position Paper [6.1.2] and the contemporary fuel rack structwal analysis practice have been carried out. They demonstrate consistently large margins of safety in all new storage modules. l l l HOLTEC INTERNATIONAL Millstone Point Unit 3 6-46
6.14 References [6.1.1] 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, Janvary,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. [6.5.1] Rab:.uwicz, E., Triction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company,
- 1976.
i [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 ofImmersed Solids," Journal of Engineering for Industry, Trans.'of the ASME, February 1972, pp 167-172. l l {6.6.1) Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering," McGraw Hill,1976.- [ HOLTEC INTERNATIONAL Millstone Point Unit 3 6-47
[6.6.2] Paul, B., " Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment", (Proprietary), NUSCO/Iloltec Report i11-88243. [6.7.1] ASME Boiler & Pressure Vessel Code, Section III, Subsection NF,1995 Edition. [6.7.2] ASME Boiler & Pressure Vessel Code, Section III, Appendices,1995 Edition. [6.7.3j USNRC Standard Review Plan, NUREG-0800 (Section 3.8.4, Rev. 2,1989). [6.10.1] Chun, R., Witte, M. and Schwartz, M., " Dynamic Impact Effects on Spent Fuel Assemblies," UCID-21^46, Lawrence Livermore National Laboratory, October 1987.
~
HOLTEC INTERNATIONAL Millstone Point Unit 3 6 48
t l Table 6.2.1 PARTIAL LISTING OF FUEL RACK APPLICATIONS USING DYNARACK I I PLANT DOCKET NUMBER (s) YEAR Eso 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 Oyster Creek USNRC 50-219 1984 Pilgrim USNRC 50-293 1985 V.C. Summer USNRC 50-395 1984 Diablo Canyon Units 1 & 2 USNRC 50-275,50-323 1986 l Byron Units 1 & 2 USNRC 50-454,50-455 1987 Braidwood Units 1 & 2 USNRC 50-456,50-457 1987 I Vogtle Unit 2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 I 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 l Three Mile Island Unit 1 USNRC 50-289 1991 James A. FitzPatrick USNRC 50-333 1990 Shearon Harris Unit 2 USNRC 50-401 1991 Hope Creek USNRC 50-354 1990 Kuosheng Units 1 & 2 Taiwan Power Company 1990 Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units 1 & 2 Comision Federal de Electricidad 1991 Zion Station Units 1 & 2 USNRC 50-295,50-304 1992 IIOLTEC INTERNATIONAL Millstone Point Unit 3 6 49 i
Table 6.2.1 PARTIAL LISTING OF rUEL RACK APPLICATIONS USING DYNARACK Sequoyah USNRC 50-327,50-328 1992 1 LaSalle Unit 1 USNRC 50-373 1992 Duane Arnold Energy Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit 1 USNRC 50-220 1993 Beaver Valley 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 Yonggwang Units 1 & 2 KINS 1996 Kori-4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 ! Angra Unit 1 Brazil 1996 Sizewell B United Kingdom 1996 I IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-50
Tcble 6.3.1 RACK MATERIAL DATA (200 F) (ASME - Section II, Part D) { i Young's Modulus Yield Strength Ultimate Strength Material E S, S, (psi) (psi) (psi) l SA240; 304L S.S. 27.6 x 10' 21,300 66,200 'l SUPPORT MATERIAL DATA (200 F) SA240 Type 304L (upper 27.6 x 10' 21,300 66,200 part of support feet) SA-564-630 (lower part of 27.6 x 10 6 106,300 140,000 support feet; age hardened at i100 F) l , IIOLTEC INTERNATIONAL r Millstone Point Unit 3 6-51
l l Table 6.4.1 i l' TIME-HISTORY STATISTICAL CORRELATION RESULTS OBE l Datal to Data 2 0.090 ) Datal to Data 3 0.016 Data 2 to Data 3 0.008 ) { SSE Datal to Data 2 0.118 Datal to Data 3 -0.021 Data 2 to Data 3 -0.127 l l Datal corresponds to the time-history acceleration values along the X axis (North) Data 2 corresponds to the time-history acceleration values along the Y axis (West) l Data 3 corresponds to the time-history acceleration values along the Z axis (Vertical) l l l l l IIOLTEC INTERNATIONAL
-Millstone Point Unit 3 6-52
r-I i Table 6.5.1 Degrees-of-freedom DISPLACEMENT ROTATION LOCATION (Node) U, U, U, 0, 0, 0, 1 pi P2 P3 44 4s As 2 p, p, p, q,o q,, q,2 Node 1 is assumed to be attached to the rack at the bottom most point. Node 2 is assumed to be attached to the rack at the top most point. l Refer to Figure 6.5.1 for node identification. 2' pi3 p,4 3' pi3 p3. 4' pi7 pi, S' pi, p2e l' P2s P22 where the relative displacement variables q, are defined as: p, = q,(t) + U,(t) i = 1.,7,13,15,17,19,21
= q,(t) + U,(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 l
I
~
IIOLTEC INTERNATIONAL l Millstone Point Unit 3 6-53
1 l l Table 6.5.2 (DYNARACK) NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION l ELEMENTS l l
- 1. Nonlinear Springs (Type 3 Gap Elements - 520 Total) '
Node Loc. Number Description l 1 Support S1 Z compression-only element ! 2 Support S2 Z compression-only element 3 Support S3 Z compression-only element 4 Support S4 Z compression-only element 5 2,2* X rack / fuel assembly impact element between nodes 2 and 2' 6 2,2* X rack / fuel assembly impact element between nodes 2 and 2' 7 2,2* Y rack / fuel assembly impact element between nodes 2 and 2' 8 2,2* Y rack / fuel assembly impact element between nodes 2 and 2' 9-360 Impact elements corresponding to the rattling masses at nodes 1",3',4' and 5'(similar to elements 5 thru 8) 361-520 Bottom and Inter-rack impact elements Top Cross section of Rack (around edge) IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-54
Table 6.5.2 (DYNARACK) NUMBERING SYSTEM FOR GAP ELEMENTS AND FRICTION ELEMENTS II. Linear Springs (Type 1 Elements - 90 Total) Rack No. Number Description 1 1 Rack beam bending element (x-z plane) 2 1 Rack shear deformation element (x-z plane) 3 1 Rack beam bending element (y-z plane) 4 1 Rack shear deformation element (y-z plane) 5 1 Rack beam axial deformation element 6 1 Rack beam torsional deformation element l 7-12 2 Similar to elements 1 thru 6 13-18 3 Similar to elements 1 thru 6, continue to Rack 15 Ill. Piece-wise Linear Frictinn Springs (Type 2 Elements - 120 Total) Rack No. Number Description 1 1 Pedestal 1, X direction 2 1 Pedestal 1, Y direction 3 1 Pedestal 2, X direction 4 1 Pedestal 2, Y direction 5 1 Pedestal 3, X direction 6 1 Pedestal 3, Y direction 7 1 Pedestal 4, X direction 8 1 Pedestal 4, Y direction 9-16 2 Similar to elements 1 thru 8 17-24 3 Similar to elements 1 thru 8, continue to Rack 15 IIOLTEC INTERNATIONAL Millstone Point Unit 3 6-55
r-Table 6.9.1 COMPARISON OF BOUNDING CALCULATED LOADS / STRESSES VS/ CODE l ALLOWABLES AT IMPACT LOCATIONS AND WELDS Item / Location Calculated Allowable 802 3,187* Fuel assembly / cell wall impact, Ibf. I1,520 (SSE) 35,748 (SSE) Rack Cell to base.olate weld, psi 6,975 (SSE) 35,748 (SSE) Female pedestal to baseplate weld, 4,194 (OBE) 19,860 (OBE) psi 7,796 Cell to cell welds, Ibf. 2,268**
- 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 802 lbs
, ** Based on the fuel assembly to cell wall impact load simultaneously applied in two orthogonal directions. HOLTEC INTERNATIONAL Millstone Point Unit 3 6-56 L-.._--
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- 7.0 - FUEL HANDLING AND CONSTRUCTION ACCIDENTS l 7.1 Introduction .
i 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 1 the proposed racks under postulated fuel assembly drop scenarios germane to MP3. ' Two scenarios are postulated in addition to the fuel assembly drop: 1) dropping the heaviest rack in - the pool from the maximum possible height during rack installation and 2) dropping a pool gate onto the racks. The first accident requires showing that the pool structure can withstand a rack drop so as to prevent a rapid loss of water. The gate scenario requires showing that the drop will not
- damage either the fuel assemblies themselves or the poison material on the racks.
7.2 Description of Fuel Handling Accidents In the evaluation of fuel handling accidents, the concern is with the damage to the storage racks. The configuration of the fuel assemblies, rack cell s.ize, spacing, and neutron absorber material must l remain consistent with the configurations used in the criticality evaluations. Maintaining these ' designed configurations will ensure that the results of the criticality evaluations remain valid. Radiological concerns due to fuel damage are not an issue, since the fuel handling design basis i accident mnsiders the worst case condition of a falling assembly, which remains unchanged. Tlus condition is a fuel assembly falling onto another assembly. Fuel damage subsequent to a fuel assembly drop.is primarily influenced by the weight and design of the fuel assembly, the drop height (which determines the kinetic energy upon impact), and the orientation of the falling assembly. Since none of these parameters are changed under the proposed modification, the number of fuel rods damaged during a fuel assembly drop remains consistent with the previously analyzed fuel handling design basis accident. I HOLTEC INTERNATIONAL L Millstone Point Unit 3 7-1 t
During the previously evaluated design basis event the kinetic energy of the falling assembly is maximized by selection of the greatest drop distance (from the Handling Machine to the floor of the _ pool). A drop event considering a falling assembly striking the top of the stomge cell represents a significantly reduced drop height and a corresponding reduction in the kinetic energy of the falling H assembly. The new storage configuration does not change the elevation of the top of stored fuel. Therefore, the new configuration does not represent a significant change in the kinetic energy.of a
' fuel assembly directly striking the top of stored fuel. A falling fuel assembly striking the top of the racks and causing sufficient deformation to also strike the top of a stored assembly is also possible, but is even less limiting. The falling assembly would impart far less kinetic energy to the stored - j assembly than a direct impact, since a significant portion of the kinetic energy of the falling assembly would be absorbed by damage to the racks. Therefore, the radiological consequences j resulting from a fuel drop accident continue to be bounded by the previously evaluated design basis accident.
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. Inasmuch as the new racks are of honeycomb construction, the deformation produced by the impact is expected to be ! confined to the region of collision. However, 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, j the ; lastic deformation of the rack cell wall should not extend more than 18.125 inches (downwards) from the top of the rack. This will ensure that the configurations considered in the criticality evaluatiens are not compromised. 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 through air; resulting in 75,600 lbs-in of kinetic energy. The impactor weight corresponds to the weight of a fuel assembly along with a Rod Control Cluster Assembly (RCCA) and the fuel handling tool. This weight was chosen to bound the drop energies that result from a 2,200 lb buoyant weight (MP3 Tech Spec) dropped from l HOLTEC INTERNATIONAL Millstone Point Unit 3 7 30 inches. This results in an impact energy of 66,000 lbs-in. Also, the analyzed fuel drop bounds the analyzed fuel drop on Westinghouse racks (2,112 lbs. buoyant weight dropped 30 inches). This results in an impact energy of 63,360 lbs.-in. During normal fuel handling, a fuel assembly cannot reach a height greater than 30 inches above the racks. Therefore, the case considered here envelopes the existing design basis. It is readily apparent from the description of the rack modules 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 a rack periphery cell in the manner shown in Figure 7.2.1. In order to maximize the penetration into the top of the rack by the falling assembly, the rack is considered empty (i.e., without assemblies or RCCAs). Exclusion of the stored fuel from the model eliminates the possibility of sharing the kinetic energy with the rack, thus maximizing rack damage (e.g., depth of penetration). Finally, the fuel assembly 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. Figure 7.2.2 depicts the impacted rack in plan view. 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". If the baseplate is pierced, then the fu. l assembly might damage the pool liner and/or 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 rack beseplate 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). HOLTEC INTERNATIONAL Millstone Point Unit 3 7-3
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 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 Mathematical Model 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 7.3.1 contains the results for the three drop events. In the second step of the solution, an clasto-plastic finite element model of the impacted region on Holtec's computer Code PLASTIPACT (Los Alamos National Laboratory's DY.NA3D 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 conservatively assumed to be rigid. The physical properties of material types undergoing deformation in the postulated impact events are summarized ! in Table 7.3.2. I i 7.4 Results 7.4.1 Shallow Drop Events l Figure 7.4.1 provides an isometric of the finite element model utilized in the shallow drop impact analysis. l HOLTEC INTERNATIONAL Millstone Point Unit 3 7-4 , l J
Dynamic analyses show that the top of the impacted region undergoes severe localized deformation. Figure 7.4.2 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 6.64 inches, which is below the design limit of I 8.125 inches. Figure 7.4.3 shows the plan view of the post-collision geometry. Approximately 10% of the cell opening in the impacted cell is blocked.
-7.4.2 Deep Drop Events 4
The deep drop scenario (Figure 7.4.4) wherein the impact region is located above the support pedestal (Figure 7.4.4a) is found to produce a negligible deformation on the baseplate. The maximum Von Mises stress occurs in a localized region at is limited to only 25 ksi. Insignificant plastic strain occurs in the liner. Therefore, it is concluded that the pool liner will not be damaged. The deep drop condition through an interior cell (Figure 7.4.4b) does produce some deformation of. I the baseplate and localized severing of the baseplate / cell wall welds (Figure 7.4.5). However, the fuel assembly support surface is lowered by a maximum of 2.9 inches, which is less than the distance of 4-5/8 inches from the baseplate to the liner. Therefore, the pool liner will not be
.l damaged.
7.5 Rack Drop The drop of a rack during the reracking process was also postulated. This evaluation considered a l . rack to be dropped to the bottom of the pool from a height of 40 feet. The analysis of damage to the 1 liner and underlying concrete was determined by neglecting any bearing pads at the impact site and considering that the pedestal directly strikes the unprotected liner. It was determined that the pool
- floor would not suffer structural damage.
i l 7.6 Gate Drop
' A drop of the spent fuel pool canal gate was also analyzed. The analysis considered a drop of the liOLTEC INTERNATIONAL Millstone Point Unit 3 7-5
4 5000 lb (dry weight) from a conservatively assumed height of 36" onto empty spent fuel storage racks and onto the spent fuel pool liner. The actual gate carrying height above the racks is 21".
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The MP3 technical specifications prohibit the gate from travel over spent fuel. l l
)
The existing Westinghouse racks and the new Holtec racks were evaluated. The Region 2 type Holtec rack was selected for the drop evaluation since they contain less material and weld connections than the Region 1 type racks. In both cases a perinheral row of cells was chosen to ensure that the gate transmitted its entire energy into the racks most vulnerable location. i 1 The results demonstrate that a potential gate drop would penetrate the rack cell for a distance of 5 inches for the Westinghouse racks and 7.45 inches for the Holtec racks, causing local damage and deformation. However the damage is limited to the upper cellular region of the rack and does not 1 extend to the rack cells in the active fuel zone. The drop would also not damage the poison
]
material (Boral) in the Holtec racks. The racks would therefore remain functional with respect to storage of spent fuel in cells adjacent to those potentially impacted by the gate drop. It was also shown that the gate would not pierce the spent fuel pool liner. l i I At this time NU will not license to allow fuel to be under the safe load path of a gate during gate l 1 movement. It should be noted that the gate drop issues do not need to be addressed until the new racks are installed. The Canal gate is not located close to the existing racks, 7.7 Closure The fuel assembly drop accident events postulated for the pools were analyzed and found to produce localized damage well within the design limits for the racks. The configuration of the fuel 1
- and poison (Boral) is not compromised from the configurations analyzed in the criticality l
4 evaluations discussed in Section 4.0. Therefore, there are no criticality concems for these accidents. A construction accident event wherein the heaviest rack falls from a 40' height onto the pool floor ; was also considered. Analyses show that the pool structure will not suffer structural damage. HOLTEC INTERNATIONAL i
- Millstone Point Unit 3 7-6
,t
y 7.8 References 1 [7.1] "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978. i i 1 I l l l I l I IIOLTEC INTERNATIONAL Millstone Point Unit 3 7-7
r-TABLE 7.3.1 i IMPACT EVENT DATA 1 Impactor Drop Impact Weight (lbs) Impactor lieight Velocity , 4 Case (inches) (inch /sec)
- 1. Shallow drop event 2,100 Fuel Assembly 36 155 l 2. Deep drop event 2,100 Fuel Assembly 204.375 355
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- 3. Construction event lieaviest Rack Rack Module 480 300
- 4. Gate Drop 5,000- Gate 36 144 1
i l l I HOLTEC INTERNATIONAL Millstone Point Unit 3 7-8
i l Table 7.3.2 Material Definition Material Type Density Elastic Stress Strain Name (pcf) Modulus (psi) I (psi) First Yield Failure Elastic Failure Stainless Sii240- 490 2.760e+07 2.130e+04 6.620e+04 7.717e-04 3.800e-01 l Steel 304L Stainless SA240- 490 2.760e+07 2.500e+04 7.100e+04 7.717e-04 3.800e-01 Steel 304 ; 1 Stainless SA564- 490 2.760e+07 1.063e+05 1.400e+05 3.851e-02 3.800e-01 j Steel 630 ! i Concrete 4000 150 3.605e+06 4.000e+03 2.022e+04 1.110e-03 5.500e-02 I 1 IIOLTEC INTERNATIONAL Millstone Point Unit 3 7-9
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' Project 71144
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8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 8.1 Introduction The MP3 Spent Fuel Pool (SFP) is a safety related, seismic category I, reinforced concrete structure. This section present 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 "dng factored load combinations mandated by SRP 3.8.4 [8.1.2] based on the " ultimate strength" design metimi 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 Figure 2.1 with all storage locations occupied by fuel assemblies. l The highly loaded wall sections adjoining the floor slabs are carefully examined. 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 the American j Concrete Institute (ACI). 8.2 Description of Pool Structures i The analyzed reinforced concrete structure model is isolated from the remainder of the Fuel Building reinforced concrete structure and include; three pools: the Spent Fuel Pool (SFP), the Cask Pit (CP), and the Transfer Canal (TC). The vertical reinforced concrete walls of the pools are supported at different elevations on a very massive reinforced concrete mat, i The threr pools are located in the area delimited by the G and H Fuel Building column lines (parallel to the East direction) and column lines 52.8 and 50.6 (parallel to the North direction)
- and are separated by reinforced concrete walls of various thicknesses. The walls are supported at HOLTEC INTERNATIONAL Millstone Point Unit 3 81
~different elevations by a massive on-grade reinforecd concrcte slab, which extends down to the soil elevation -3'-3". Figure 8.2.1 shows these major structural features of the pool.
1-
..The thicknesses of the walls surrounding SFP are: 6'-0" at North and East, and 6'-6" at South and on the West side a 6'-6" thick wall (the Canal Wall) separates the SFP from the TC, and in the South-East corner isolation from the CP is realized with two walls, 6'-0" and 5'-5", located respectively at the North and the West side of the CP. The continuity of the SFP West (Canal wall) and CP West walls is interrupted by the existence of the fuel gate openings. The SFP on-grade mat upper elevation is located at elevation 11'-3" and has a thickness of 14'-6".
The thicknesses of the walls surrounding the TC are: 6'-0" at West, North and South. The thickness of the mat is 12'-6" and its upper elevation is located at 9'-3". A sump is located on the south sit cf the TC and consequently the mat lowers to the elevation of 9'-3". The CP mat upper elevation is located at 25'-9", but its Pit floor elevation is only 4'-9". The walls of the CP are 5'-0" and 7'-0 thick along the East and South side, respectively. 8.3 Definition of Loads Pool structural loading involves the following discrete components: 8.3.1 Static Loading (Dead Loads and Live Loads)
- 1) Dead weight of the modeled concrete structure is calculated considering a density of 150 lb/ft'
- 2) Dead weight of the Fuel Building reinforced concrete upper structure;
- 3) Live Loads such as cranes transmitted by upper building structure;
- 4) Hydro-static water pressures wMeh vary linearly abmg the height of the walls.
HOLTEC INTERNATIONAL Millstone Point Unit 3 8-2 j L U
- 8.3.2 l Seismic Induced Loads -
- 1) ' The inertial loads generated by seismic events.
- 2) Hydrodynamic inertia loads due to the contained water mass and sloshing loads (considered in-accordance with [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.-
8.3.3 Thermal Loading 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 (150 F ) and the accident temperature (200 F). 8.4 Analysis Procedures 8.4.1 Finite Element Analysis Model 1 l 1 The finite element model encompasses the entire Spent Fuel Pool and two other reinforced I concrete structures located immediately adjacent to the Spent Fuel Pool (the Cask Pit, and the Transfer Canal). The interaction with the rest of the Fuel 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 the SFP which is 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 Millstone Point Unit 3 8-3
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 13,209 nodes,7,252 solid type finite-elements,3,692 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. The dynamic behavior of the water mass contained in the SFP during a seismic event is modeled according to the guidelines set in TID-7024 [8.3.1]. Neglecting the possibility of water contained in the Transfer Canal is conservative. The loading which wou!d be induced by the hydrostatic pressure would tend to offset the equivalent pressures on the other side of the wall (in the Spent Fuel Pool). The effect of hydrostatic pressure on only one side of this all more than offsets any loading which would be induced from water sloshing. To simulate the interaction between the modeled region and the rest of the Fuel 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) i; 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 including 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 [8.1.3]. 8.4.2 Load Application The structural region isolated from the Fuel Building is numerically investigated using the finite element method. The pool walls and their supporting reinforced concrete mat are represented by a 3-D finite-element model. IIOLTEC INTERNATIONAL Millstone Point Unit 3 8-4
- _ _ _ _ _ _ _ _ - 4
h The individual loads considered in the analysis are greaped 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 Building upper structure, crane deadicad, and the hydro-static pressure of the contained water), live loads (crane 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 load 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 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) methmi. 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 32 Hz. Three independent orthogonal acceleration spectra are applied to the model. The acceleration spectra are ccasidered to act simultanepusly in three-directiom. The SRSS method is used to sum the similar quantities calculated for each direction. Results for individual load cases are combined using the factored load combinations discussed below considering two scenarios: first, when the Spent Fuel Pool and the Cask Pit are full of water; second, when only the Spent Fuel Pool is full of water. The corr.bined stress resultants are HOLTEC INTERNATIONAL
' Millstone Point Unit 3 8-5 1
I
compared with the ultimate moments and shear capacities of all structural elements pertinent to the Spent Fuel Pool and Cask Pit, which are calculated in accordance with the ACI 349-85 to develop the safety factors. 8.4.3 Load Combinations 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 '*To) 1 - Load Combination No. 6 = 1.2*D + 1.9*E - Lo d s' aination No. 7 =- 1.2*D - 1.9*E For ' Factored Load Conditions" the following load combinat;ons are: - Load Combination No. 8 = D + L + To + E' - Load Combination No. 9 = D + L + To - 2 - Load Combination No.10 = D + L + Ta + 1.25*E - Load Combination No.11 = D + L + Ta - 1.25+E - Load Combinadon No.12 = D + L + Ta + E' l - Load Combination No.13 = D + L + Ta - E' HOLTEC INTERNATIONAL L Millstone Point Unit 3 8-6 L
where:
.D=
dead loads; L= live loads; To = thermal load during normal operation; Ta = thermal load under t.ccident condition; JE = OBE 'earthquake induced loads; E' = SSE earthquake induced loads, l 8.5 Results of Analyses 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 formulas for bending loading are computed using appropriate concrete and reinforcement strengths. For MP3, the concrete and reinforcement allowable strengths are: concrete fl = 5,000 psi reinforcement f, = 60,000 psi l Table 8.5.1 and 8.5.2 shows results from potentially limiting load combinations for f.he bending j strength and shear of the slab and walls, respectively. They demonstrate that the structural capacity is not exceeded. In the tables, a limiting safety margin is defined for each section; the allowable bending moment and shear force defined by A CI divided by the calculated bending moment or shear force (from th; finite element analyses). The major regions of the pool mucture consist of ten concrete walls
' delimiting the SFP and Cask Pit.' 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 procedure is repeated for all the potential limiting load combinations I
HOLTEC INTERNATIONAL j
-Millstone Point Unit 3 - 87
r 8.6 Pool Liner l- . The pool liner is subject to in-plate strains due to movement of the rack support feet during the l . seismic event. Analyses are performed to establish that the liner will not tear or rupture under limiting loading conditions in the pool, and that there is no fatigue problem under the condition of 1 SSE event plus 20 OBE events. These analyses are based on loadings imparted from the most highly loaded pedestal in the pool assumed to be placed in the most unfavorable position. 8.7- Bearing Pad Analysis 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 (8.1.3) limit on bearing pressures. Section 10 of the code i gives the design bearing strength as f, = @ (.85 fl)e e where $ = .7 and fl is the specified concrete strength for the spent fuel pool. c = 1 except when the supporting surface is wider on all sides than the loaded area. In that case, c = (A2 /Ai )', but not
.more than 2. 'A, is the actual loaded area, and A2 is an area greater than Ai and is defined in [8.1.3]. ~ Using a value of c ) 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 MP3 the compressive strength is, fl = 5,000 psi, and the allowable static bearing pressure is f, = 2,975 psi assuming no concrete confinement.
The bearing pad selected is 1" thick, austenitic stainless steel plate stock. Most rack pedestals are !
.J l . located away from leak chases. However, in the most limiting configuration, the bearing pad is centered o' ver a leak chase.
4 HOLTEC INTERNATIONAL
' Millstone Point Unit 3 8-8 V
p An ANSYS finite element simulation of the model is presented in Figure 8.7.1. The model permits the bearing pad to deform and lose contact with the liner, if th3 conditions of elastostatics so dictate.
~
The slab is modeled as an elastic foundation which supports the liner. A vertical force of 221,000 lbs'is applied to the model. This load is chosen to bound the factored results of the rack time-
- history sim'ulations.
The average pressure at the pad to liner interface is computed and compared against the stress limit. Calculations show that the average pressure at the slab / liner interface is 2,564 psi which is below
. the ACI allowable of 2,975 psi, providing a factor of safety of 1.16.
The stress distribution in the bearing pad is also evaluated. The maximum bending stress in the bearing pad under the peak vertical load is 21,747 psi. With a material yield strength of 25,000 psi at 200'F, the factor of safety is 1.15. l Thcrefore, the bearing pad design devised for MP3 is deemed appropriate for the prescribed loadings. 8.8 Conclusions
- Regions affected by loading the fuel pool completely with high density racks are examined for l
structural integrity. ' tI 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. l HOLTEC INTERNATIONAL Millstone Point Unit 3 8-9 m _ I
8.9 References [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. J [8.1.2] NUREG-0800, SRP-3.8.4, Rev. l., July 1981. [8.1.3] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit Michigan. ! [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. l HOLTEC INTERNATIONAL Millstone Point Unit 3 8-10
0 , t Table 8.5.1 BENDING STRENGTil EVALUATION i Critical Load Combinations Location L.imiting Safety Margin (see Section 8.4.3) ; Canal Wall 18.01 12 l Cask Pit West Wall 13.75 13 I Pool East Wall 20.06 13 ( Pool North Wall 22.16 11
-s Cask Pit North Wall 19.56 12
~l Pool South Wall 17.77 12-i l
l l I l l HOLTEC INTERNATIONAL l^ Millstone Pois t Unit 3 8 11
Table 8.5.2 SHEAR STRENGTH EVALUATION Critical Load Combinations j Location Limiting Safety Margin (see Section 8.4.3) { Canal Wall l 4.73 12 Cask Pit West Wall 1.97 12 Pool East Wall 3.39 12 Pool North Wall 2.68 12 Cask Pit North Wall 2.47 13 Pool South Wall 3.6 13 l l
)
)
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9.0 BORAL SURVEILLANCE PROGRAM 9.1 ' Purpose
]
I i
~ Boral'", the neutron absorbing material incorporated in the spent fuel storage rack design to assist m a i
controlling system reactivity, consists of finely divided particles of boron carbide (B4 C) uniformly I distributed in type 1100 aluminum powder, clad in type 1100 aluminum and pressed and sintered in - I a hot-rolling process. Tests simulating the radiaiion, thermal and chemical environment of the spent 1 I fuel pool have demonstrated the stability and chemical inertness of Boral (References [9.1.1]-
]
[9.1.3]). The accumulated dose to the Boral over the expected rack lifetime is estimated to be about
.3 x 10' to 1 x 10" rads depending upon how the racks are used and the number of full-core off-loads that may be necessary.
Based upon the accelerated test programs, Boral is considered a satisfactory material for reactivity control in spent fuel storage racks and is fully expected to fulfill its design function over the lifetime of the racks. Nevertheless, it is prudent to establish a surveillance program to monitor the integrity
]
and performance of Boral on a continuing basis and to assure that slow, long-term synergistic j efTects, if any, do not become significant. Furthermore, the April 14,1978 USNRC letter to all .
. power reactor licensees (Reference [9.1.4]),~ specifies that
" Methods for verification of long-term material stability and mechanical integrity of special poison materials utilized for neutron absorption should include actual tests."
The purpose of the surveillance program is to characterize certain properties of the Boral with the objective of providing data necessary to assess the capability of the Boral panels in the racks to j continue to perform their intended function. The surveillance program is also capable of detecting 1 the onset of any significant degradation with ample time to take such corrective action as may be l necessary. HOLTEC INTERNATIONAL Millstone Point Unit 3 9-1 x ,
In response to the need for a comprehensive Boral survdllance program to assure that the suberiticality requirements of the stored fuel array are safely maintained, a surveillance program has been developed incorporating certain basic tests and acceptance criteria. The Boral surveillance program depends primarily on representative coupon' samples to monitor performance of the absorber material without disrupting the integrity of the storage system. The principal parameters to be measured are the thickness (to monitor for swelling) and boron content. 9.2 COUPON SURVEILLANCE PROGRAM 9.2.1 Coupon Description The coupon measurement program includes coupons suspended on a mounting (called a " tree"), placed in a designated cell, and surrounded by spent fuel. Coupons will be removed from the array on a prescribed schedule and certain physical and chemical properties measured from which the stability and integrity of the Boral in the storage cells may be inferred. i
)
Each surveillance coupon will be approximately 4 inches wide and 8 inches long. The coupon surveillance program will use a total of 8 test coupons. In mounting the coupons on the tree, the coupons will be positioned axially within the central 8 feet of the fuel zone where the gamma flux is expected to be reasonably uniform. Each coupon will be carefully pre-characterized prior to insertion in the pool to provide reference initial values for comparison with measurements made after irradiation. The surveillance coupons will be pre-characterized for weight, length, width and thickness. In addition, two coupons will be preserved as archive samples for comparison with subsequent test coupon measurements. Wet chemical analyses of samples from the same lot of Boral will be available from the vendor for comparison. 1 HOLTEC INTERNATIONAL i Millstone Point Unit 3 9-2
9.2.2 Surveillance Coupon Testing Schedule The coupon tree"is surrounded by freshly discharged fuel assemblies at each of the first five refuelings following installation of the racks to assure that the coupons will have experienced a slightly higher radiation dose than the Boral in the racks. Beginmng with the fifth load of spent fuel, the fuel assemblies will remain in place for the remaining lifetime of the racks. The scheduled coupon management schedule is shown in Table 9.1. At the time of the first fuel off-load following installation of the coupon tree, the (8) storage cells surrounding the tree shall be loaded with freshly-discharged fuel assemblies that had been among the higher specific power assemblies in the core. Shortly before the second reload, the coupon tree is removed and a coupon removed for evaluation. The coupon tree is then re-installed and, at reload, again surrounded by freshly discharged fuel assemblies. This procedure is continued for the third, fourth, and fillh off-loading of spent fuel (except that a coupon is not pulled at the fourth refueling). From the fifth cycle on, the fuel assemblies in the (8) surrounding cells remain in place. Evaluatien of the coupons removed will provide information of the effects of the radiation, thermal and chemical environment of the pool and by inference, comparable information on the Boral panels in the racks Over the duration of the coupon testing program, the coupons will hav. accumulated more radiation dose than the expected lifetime dose for normal storage cells. Coupons which have not been destructively analyzed by wet-chemical processes, may optionally be returned to the storage poc! and re-mounted on the tree. They will then be available for subsequent investigation of defects, should any be found. 9.2.3 Measurement Program The coupon measurement program is intended to monitor changes in physical properties of the Boral absorber material by performing the following mea urements on the pre-planned schedule: Visual Observation and Photography, HOLTEC INTERNATIONAL Millstone Point Unit 3 9-3
Neutron Attenuation, Dimensional Measurements (length, width and thickness), Weight and Specific Gravity, and Wet-chemical analysis (Optional). The most significant measurements are thickness (to monitor for swelling) and neutron attenuation' (to confirm the concentration of Boron-10 in the absorber material). In the event loss of boron is observed or suspected,'the data may be augmented by wet-chemical analysis (a destructive gravimetric technique for total boron only). l i 9.2.4 Surveillance Coupon Acceptance Criteria Of the measurements to be performed on the Boral surveillance coupons, the most important are (1) the neutron attenuation measurements (to verify the continued presence of the boron) and (2) the thickness measurement (as a monitor of potential swelling). Acceptance criteria for these measure-ments are as follows: i i A decrease of no more than 5% in Baron-10 centent, as determined by neutron attenuation, is acceptable. (This is tantamount to a requirement for no loss in boron within the accuracy of the measurement.) An increase in thickness at any point should not exceed 10% of the initial thickness at that point. Changes in excess of either of these two criteria requires investigation and engineering evaluation which may include early retrieval and measurement of one or more of the remaining coupons to i
)
i Neutron attenuation measurements are a precise instrumental method of chemicai analysis for Boron
-10 content using a non-destructive technique in which the per centage of thermal neutrons trans mitted through the panel is measured and compared with pre- deter mined calibration data. Boron-10 is the nuclide of principal interest since it is the isotope responsible for neutron absorption in the Boral panel.
HOLTEC INTERNATIONAL Millstone Point Unit'3 9-4
provide corroborative evidence that the indicated changc(s) is real._ If the deviation is determined to be real, an engineering evaluation shall be performed to identify further testing or any corrective action that may be necessary. The remaining measurement parameters serve a supporting role and should be examined for early indications of the potential onset of Boral degradation that would suggest a need for further attention and possibly a change in measurement schedule. These include (1) visual or photographic evidence of unusual surface pitting, corrosion or edge deterio'ation, r or (2) unaccountable weight
. loss in excess of the measurement accuracy.
9.3 In-Service Inspection (Blackness Tests) In-service inspection involves directly testing the Boral panels in the storage racks by neutron , loggingt (sometimes called " Blackness Testing"). This technique is able to detect areas of significant boron loss or the existence of gaps in the Boral, but cannot determine other physical
. properties such as those measured in the coupon program.
In the event that the surveillance coupon program hows a confirmed indication'of degradation, blackness testing may be one of the techniques emp.oyed to investigate the extent of degradation, if any, in the racks. t ' Neutron logging, is a derivative of well-logging methods successfully used in the oil indust y for many years.
'~
HOLTEC INTERNATIONAL j Millstone Point Unit 3 9-5
9.4 References [9.1.1) " Spent Fuel Storage Module Corrosion Report", Brooks & Perkins Report 554, Jurie 1,1977 [9.1.2] " Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools", Brooks & Perkins Report 578, July 7,1978 l 1 [9.1.3] "Boral Neutron Absorbing / Shielding Material - Product Performance Report", Brooks & Perkins Report 624, July 20,1982 [9.1.4] USNRC Letter to All Power Reactor Licensees transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling l Applications", April 14,1978 l l HOLTEC INTERNATIONAL
' Millstone Point Unit 3 96
p Table 9.1 i COUPON MEASUREMENT SCIIEDULE r
-(
hj-l
- s
'i ,-,. a m t , z,e.a i
1 i Remove coupons for evaluating within 1 or 2 months preceding the next refueling. The first ! coupon is scheduled to be removed just prior to the end of Cycle 8. l 3
** i Place freshly discharged fuel in the 8 surrounding cells at the beginning of the 1st,2nd,3rd,4th, and l'
5th refueling cycles after completion of reracking. l l IlOLTEC INTERNATIONAL Millstone Point Unit 3 9-7
i i
'10.0 INSTALLATION 10.1 Introduction .
1
. All construction work at Millstone 3 will be performed in compliance with NUREG-0612 (refer to j Section 3.0), applicable Quality Assurance procedures, and site-specific project procedures. l Crane and fuel bridge operators are to be adequately trained in the operation-of load handling machines per_ the requirements of ANSI /ASME B30.2, latest revision, and the Northeast Utilities training program. Consistent with past practices, videotaped aided training will be given to th installation team, all of whom will be required to successfully complete a written examination prior to the commencement of work.
The lifting device designed for handling and installation of the new racks at Millstone 3 is remotely . l 1 l engageable. The lifting device complies with the provisions of ANSI 14.6-1978 and NUREG-0612, including compliance with the primary stress criteria, load testing at a multiplier of maximmn j 1 working load, and nondestructive examination of critical welds. J An intensive surveillance and inspection program shall be maintained throughout the installation phase of the rerack project. A complete set of operating procedures which cover the entire gamut of operations pertaining to the rack installation will be used. Similar procedures have been utilized and successfully implemented by Holtec International on previous rack installation projects. These procedures assure ALARA practices are followed and provide detailed requirements to assure equipment, personnel, and plant safety. The following is a list of procedures which will be used to implement the construction phase of the rerack project. l HOLTEC INTERNATIONAL j Millstone Point Unit 3 - 10-1 i
r-A.' Installation /Handline, Removal Procedure: This procedure provides direction for the handling / installation of the new high density modules. , This procedure delineates the steps necessary for receiving a new high density rack on site, the proper' method for unloading and uprighting the rack, staging the rack prior to installation, and installation of the rack. This procedure also provides for the installation of new rack bearing pads, adjustment of the new rack pedestals ano performance of the as-built field . survey. Any pool modifications that may be necessary such as protrusion tnmcation are also described in the
. procedure.
B. Receipt Inspection Procedure: This procedure delineetes the steps necessary to perform a thorough receipt inspection of a new rack module afler its arrival on site. The receipt inspection includes dimensional measurements, cleanliness inspection, visual weld examination, and verticality measurements. j 1 C. Cleanine Procedure: This procedure provides for the cleaning of a new rack module, ifit is required, in order to meet the requirements of ANSI 45.2.1, Level C. Permissible cleaning agents, methods and ~ limitations on materials to be employed are provided. D. Pre-Installation Dran Test Procedure: This procedure stipulates the requirements for performing a functional test on a new rack module prior to installation into the spent fuel pool. The procedure provides direction for inserting and !. Lwithdrawing a " dummy" fuel assembly into designated cell locations, and establishes an acceptance criteria in terms of maximum kinetic drag force. L HOLTEC INTERNATIONAL Millstone Point Unit 3 10-2
E. Post-In.s_tallation Dran Test Procedure:
)
, i This procedure stipulates the requirements for performing a functional test on a new rack module following installation into the spent fuel pool or cask pit. The procedure will provide direction for inserting and withdrawing a " dummy" fuel assembly into designated cell locations, and establishes I
. an acceptance criteria in terms of maximum kinetic drag force. F. Underwater Divine Procedure: Underwater diving operations may be required to support the new rack installation. This procedure ) i describes the method for introducing a diver into the spent fuel pool or cask pit, provides for J 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 l i Standard 29CFR-1910, Subpart T, and establishes contingencies in the event of an emergency. j i G. ALARA Procedure: a Consistent with the site's ALARA Program, this procedure provides details to minimize the total man-rem received during the rerack project, by accounting for time, distance, and shielding. Additionally, a pre-job checklist is established in order to mitigate the potential for an overexposure. H. Liner Inspection Procedure: 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. 1 i HOLTFC INTERNATIONAL 1 l
- Millstone Point Unit 3 10-3 l
I. Leak Detection Procedure: This procedure describes the method to test the spent fuel pool liner for potential leakage using a vacuum box. This procedure may be applied to any suspect area of the pool liner. J. Underwater Weldine Procedured
' In the event of a positive leak test result, an underwater welding procedure will be implemented which will provide for the placement of a stainless steel repair patch over the area in question. The procedure contains' appropriate qualification records documenting relevant variables, parameters, und limiting conditions. The weld procedure is qualithd in accordance with AWS D3.6-93, Specification for Underwater Welding or may be qualified to an alternate code accepted by Northeast Utilities.
K. Job Site Storace Procedure: This procedure establishes the requirements for safely storing a new rack module on-site, in the j event that long term job-site storage is necessary. This procedure provides environmental restrictions, temperature limits, and packaging requirements. 10.2 Rack Arrangement The existing Millstone Unit 3 rack arrangement consists of 21 racks, representing 756 cell locations. l The new proposed rack arrangement consists of 15 free-standing Holtec racks providing a total of 1,104 storage locatiou in the fuel pool. Of these 1,104 cell locations, five racks consisting of 350 cells are designated as Region I storage, and the remaining 10 racks containing 754 cells are designated as Region 2 storage. i A schematic depicting the spent fuel pool in the new maximum density configuration can be seen in
. Figure 2.1.
HOLTEC INTERNATIONAL Millstone Point Unit 3 - 10-4 L
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'10.3 Pool Survey and Inspection L l L 'A pool inspection shall be performed to determine if any 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' $
L . to installation. In the event that protrusions are found which would pose any interference to the installation l process, it is anticipated.that underwater diving operations and mechanical cutting methods would be employed to remove the protrusions. 1 1 10.4 Pool Cooling and Purification 10.4.1 Pool Cooling The spent fuel 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, will pool cooling be terminated in a manner or for a duration which would I create a violation of the Millstone 3 Technical Specification Existing procedures are in place to control actions regarding the shutdown of the spent fuel pool cooling system, and to ensure that the pool bulk temperature will always remain within required limits. 4 10.4.2 Pool Purification The existing spent fuel pool filtration system shall be operational in order to maintain pool clarity. Additio .ly, an underwater vacuum system shall be used as necessary to supplement fuel pool purification. A 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. l HOLTEC INTERNATIONAL Millstone Point Unit 3 10-5
p i L 10.5 Installation of New Racks The new high density racks, supplied by Holtec International, shall be delivered in the horizontal < position. A new rack module shall be removed from th: 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. The new rack lifting device shall be installed into the rack and each lift rod successively engaged. Thereafter, the rack shall be transported to a pre-leveled surface.where the appropriate quality control receipt inspection shall be performed. In preparing the spent fuel pool for rack installation, the pool floor shall be inspected and any debris which may inhibit the installation of bearing pads will be removed. After pool floor preparation, new rack bearing pads shall be positioned in preparation for the module which is to be installed. Elevation measurements will then be performed in order to gage the amount of adjustment required, if any, for the new rack pedestals. The new rack module shall be lifted with the 10-ton 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 after installation.
The rack modules shall be lowered into the spent fuel pool using another 10-ton crane. A hoist with equivalent capacity may be attached to this crane for installation activities in order to eliminate contamination of the min hook during lifting operations 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 or a height to allow for clearance over floor projections. HOLTEC INTERNATIONAL Millstone Point Unit 3 - 10-6
i' i i Elevation readings shall be taken to confirm that the module is level and as-built rack-to-rack and l
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rack-to-wall offsets shall be recorded. The lifling device shall be disengaged and removed from the i fuel pool under Health Physics direction. Post-installation free path verification will be performed using an inspection gage in order to ensure that no cell location poses excessive resistance to the insertion or withdrawal of a bundle. This test shall confirm final acceptability of a new rack module. ) 9 l l 10.6 Safety, Radiation Protection, and ALARA Methods - 10.6.1 Safety During the construction phase of the rerack project, personnel safety is of paramount importance, outweighing all other concerns. All work shall be carried out in strict compliance with applicable approved procedures. 10.6.2 Radiation Protection Health Physics shall' provide necessary coverage in order to provide radiological protection and monitor dose rates. The Health Physics 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 Health Phyr.ics. Health Physics 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 ALARA 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. HOLTEC INTERNATIONAL Millstone Point Unit 3 10-7 i
Time Each member of the project team will be properly trained and will be provided appropriate education and understanding of critical evolutions. Additionally, daily pre-job briefings will be y employed to acquaint each team member with the scope of work to be perfonned and the proper j means of executing such tasks. Such pre-planning devices reduce worker time within the l radiologically controlled area and, therefore, project dose, l Distance { Remote tooling such as lift fixtures, pneumatic grippers, a support leveling device and a lift rod 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. Shieldine During the course of the rerack project, 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 Radwaste Material Control Radioactive waste generated from the rerack effort shall include vacuum filter bags, miscellaneous tooling, underwater appurtenances and protective clothing. < Vacuum filter bags shall be removed from the pool and stored as appropriate in a suitable container HOLTEC INTERNATIONAL Millstone Point Unit 3 10-8
in order to maintain low dose rates. Contaminated tooling shall be properly stored per Radiation Protection direction throughout the project. At project completion, an effort will be made to decontaminate tooling to the most practical extent possible. l l 1
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l I ! i I HOLTEC INTERNATIONAL Millstone Point Unit 3 10-9
I1.0. RADIOLOGICAL EVALUATION 11.1 Solid Radwaste 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 l 1 primarily by the requirement for water clarity, and the resin is normally changed about once a year.
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I During re-racking operations, a small amount of additional resins may be generated by the pool cleanup system on a one-time basis. I 11.2 Gaseous Releases 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. 11.3 Personnel Doses During normal operations, personnel working in the fuel storage area are exposed to radiation from the spent fuel pool. Radiological conditions are dominated by the most recent batch of discharged spent fuel. The radioactive inventory of the older fuel is insignificant compared to that from the recent offload. Analysis shows that the rerack will not significantly change radiological conditions. 4 Therefore the rack expansion project falls within the existing design basis of Millstone's Spent Fuel Pool. HOLTEC INTERNATIONAL Millstone Point Unit 3 11-1
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' 11.4 : Anticipated Dcse During Re-rac'k ing. - All of the operations involved in re-racking will utilize detailed proccdures prepared with full con- - sideration' of ALARA principles. Similar operations have been performed in a numbcr of facilities in the past, and there is every reason to believe that re-racking can be safely and efficiently accom-plished at MP3, with low radiation exposure to personnel.
Total dose for the re-racking operation is estimated to be between 2 and 5 person-rem, as indicated
'in Table 11.4.1. While individual' task efforts and doses may differ from those in Table 11.4.1, the total is believed to be a reasonable estimate for planning purposes. Table 11.4.2 shows previousjob es:posures that Holtec International has experienced during actual rack installations. Divers will be used where necessary, and the estimated person-rem burden includes a figure for their possible dose.
The existing radiation protection program at MP3 is adequate for the re-racking operations. Where there is a potential for significant airborne activity, continaous air monitors will be in operation. Perscnnel will wear protective clothing as required and, if necessary, respiratory protective equiptr at. Activities will be governed by a Radiation Work Permit, and personnel monitoring equiprnent will be issued to each individual. As a minimum, this will include thermoluminescent dosimeters (TLDs) and self-readirg dosirreters. Additionti personnel monitoring equipment (i.e., extremity TLDs or multiple TLDs) may be utilized as required. Work, personnel traflic, and the movement of equipment will be monitored and controlled to minimize contamination and to asse e that dose is maintained ALARA. 1 I HOLTEC INTERNATIONAL Millstone Point Unit 3 11-2 _ _ _ _ _ = _ .
r 1 l Table 11.4.1 ; j PRELIMINARY ESTIMATE OF PERSON-REM DOSE DURING RE-RACKING ' Estimated Number of Person-Rem Step Personnel llours Doset Clean and vacuum pool 3 25 0.3 to 0.6 Remove underwater 4 80 0.4 to 0.8 appurtenances Installation of new rack modules 5 55 0.7 to 1.3 Move fuel to new racks 2 150 0.8 to 1.5 Total Dose, person-rem 2 to 5 i i I t Assumes minimum does rate of 2-1/2 mrem /hr (expected) to a maximum e45 mrern/hr except for pool vacuuming operations, which assume 4 to 8 mrem /hr, and diving operations, which assume 20 to 40 mrem!hr. l IlOLTEC INTERNATIONAL l l Millstone Point Unit 3 11-3 1 I
r i l Table 11.4.2 SPENT FUEL RERACK EXPOSURE ; Plant Job Exposure (Man-Rem)
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TMl 5.9 D.C. Cook 2.2 i Ft. Calhoun 2.5 Zion 13.0* Salem Unit 1/ Unit 2 4.5/1.0 Limerick 2.0 Duane Arnold 5.5 ! l Connecticut Yankee 7.5 Sequoyah 2.5 l
- N.B. Hydrolasing was not permitted to maintain Boron concentration levels in the pool. Existing I l
racks were removed and steam cleaned in the decon pit. 1 1 l I J l IIOLTEC INTERNATli$/ AL Millstone Point Unit 3 11-4
n . L g L i 12.0 ENVIRONMENTAL COST-BENEFIT ASSESSMENT l 12.1 Introduction i Article V of the U'SNRC OT Position paper [12.1] requires the submittal of a cost / benefit analysis l for the chosen fuel storage capacity enhancement method. This section abstracts the analyses and evaluations made by NU before selecting reracking as the most viable alternative. r 12.2 Imperative for Reracking The specific need to increase the limited existing storage capacity of the MP-3 spent fuel pool is based on the continually increasing inventory in the pool, the prudent requirement to maintain full-core off-load capability, and a lack of viable economic attematives. In particular: 1
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- a. NNECO has no current contractual arrangements with fuel reprocessing facilities, l4 nor is this technology economically viable in the U.S.
- b. NUSCO (on behalf of Millstone Unit 3) has executed a disposal contract with the Department of Energy (DOE) pursuant to the Nuclear Waste Policy Act of 1982, I L but DOE has no plans to provide disposal facilities prior to 2010.
- c. Adoption of this proposed spent fuel storage expansion would not necessarily extend the time period that spent fuel assemblies would be stored on site. Spent fuel will be sent offsite for final disposition under existing legislation, but (as indicated above) the government
- - facility is not expected to be available to begin to receive fuel for at least 12 years.
l Reference is made to Tables 1.1 and 1.2 of Section I wherein the current and projected fuel discharges in the MP-3 spent fuel pool are tabulated. It is seen that the MP-3 fuel pool will lose the
- capacity to discharge one full core (193 fuel assemblies) in 2000.
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HOLTEC INTERNATIONAL
, l Millstone Point Unit 3 12-1
12.S .- -Appraisal of A.lternative Options i NU has determined that wet storage expansion is by far the most viable option for the MP-3 pool in
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comparison to other alternatives. ! The key considerations in evaluating the alternative options were: Safety: minimize the number of fuel handling steps j Economy: minimize total installed and O&M cost ~ Security: protection from potential saboteurs, natural phenomena Non-intrusiveness: minimize required modification to existing systems ! Maturity: extent ofindustry experience with the technology L ALARA: minimize cumulative dose due to handling of fuel. Wet storage expansion was found by NNECO to be the most attractive option with respect to each of the foregoing criteria. In particular: l a. There are no operational commercial interim s,torage facilities available for l l NNECO's needs in the United States, nor are there expected to be any in the foreseeable future. ]
- b. While plans are being formulated by DOE for construction of a spent fuel repository pursuant to the Nuclear Waste Policy Act of 1982, this facility is not expected to be available to accept spent fuel any earlier than 2010. Furthermore, DOE's Acceptance Priority Rankings suggest that Millstone-3's spent fuel would be ren.oved substantially later than 2010.
- c. Dry storage could be a technically feasible alternative to wet storage. However, the least '
expendve type of dry storage has been evaluated to entail a capital expenditure that i2 approximately S.3 times as large as that associated with wet storage. Other problems with
' dry storage include substantial incremental fuel movements, storage located away from the secured boundary of the site, incremental security requirements and operation and maintenance expenses, plant modifications to support the use of dry storage cask systems, and potential repackaging of fuel to meet repository requirements.
HOLTEC INTERNATIONAL Millstone Point Unit 3 12-2
- To summarize, the only acceptable option for Millstone Unit 3 is to increase its onsite wet fuel storage capacity. The alternatives have either little pro ' experience, or they are cost prohibitive.
- 12.4 Cost Estimate
- The proposed construction contemplates the reracking of the MP-3 spent fuel pool using free-standing, high density, poisoned spent fuel racks. The engineering and design is completed for full reracking of the MP-3 pool.' This rerack will provide sufficient MP-3 pool storage capacity to maintain a full core off-load capability to approximately the end oflicense.
The total capital cost is estimated to be Epproximately $10 million as detailed below. Cost estimates do not include cost of capital, overhead, or project contingencies. They are for the
. purpose of comparison only.
Engineering, design, project management $2 million Rack fabrication $5 million Rack installation $3 million As described in the preceding section, many alternatives were considered prior to proceeding with wet storage expansion, which is not the only technical option available to increase on-site storage caoacity. Wet storage expansion does, however, provide a definite cost advantage over other technologies. 12.5 Resource Commitment The expansion of the MP-3 spent fuel pool capacity is expected to require the following primary resources: Stainless steel: 250 tons
. Boral neutron absorber: 60 tons, of which 50 tons is Boron Carbide powder and 10 tons are aluminum.
I-HOLTEC INTERNATIONAL
' Millstone Point Unit 3 12-3 L._ /
m The re,irements 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. i
~ 12.6 Environmental Considerations -
This rerack is not expected to increase the m .ximum bulk pool temperature above the previously j licensed value. Therefore, the cooling water demand on the Long Island Sound and the water vapor emission to the environment should remain unchanged. L 12.7 References for Section 12
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[12.1] OT Positiu Paper for Review and Acceptance of Spent Fuel Storage and llandling Applications, USNRC (April,1978). 1 [12.2] Electric Power Research Institute, Report No. NF-3580, May,1984. [12.3] " Spent Fuel Storage Options: A Critical Appraisal", Power Generation Technology, Sterling Publishers, pp. 137-140, U.K. (November,1990). 4 l HOLTEC INTERNATIONAL Millstone Point Unit 3 12-4}}