ML20235Z109
| ML20235Z109 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 03/31/1989 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13303B064 | List: |
| References | |
| NUDOCS 8903150148 | |
| Download: ML20235Z109 (363) | |
Text
{{#Wiki_filter:_ - - -. - ._ -_ - _ t i ATTACHMENTETO UCENSEAMENDMENT , 78 A 6 SPENT FUELPOOL RERACKING LICENSING REPORT O
. SOUTHERN CALIFORNIA EDISON SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 MARCH 1989 l
71mm mis
.i 4
TABLE OF CONTENTS y (. 1.- INTRODUCTION Page 1-1 1.1. Purpose 1-1 q 1.2 Proposed Facility Modification 1 1.2.1- Present Facility Description 1-1 1.2.2 Proposed Modification. 4 1 1.3 Interfaces with other Organizations 1-4 J 1.4 Summary of Report _ 1 9 1.5' Conclusions 1 " 1.6. References 1-7 '
- 2.
SUMMARY
OF RACK DESIGN 2.1 2.1' Existing Racks . 2.1-1 1 2.2 New High Density Racks' 2.2-1 3 2.3 References 2.3-1
- 3. NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS 3.1-1 3.1 . Neutron Multiplication Factor. 3.1-1 3.1.1 Normal Storage 3.1-1 3.1.2 Postulated Accidents 3.1-4' 3.1.3 Calculation Methods .
3.1-6 3.1.~3.1 Criticality Analysis for Region I 3.1-11 3.1.3.2 Criticality Analysis for Region II 3.1-13 3.1.3.3 Criticality Analysis for Checkerboard or Alternating Row Loading for Region II 3.1-16 3.1.3.4 Fuel Reconstitution Station in Region II 3.1-19
;- 3.1. 4 - . Acceptance Criteria and Methodology for Criticality 3.1-20 3.2 Decay Heat Calculations for the Spent Fuel Pool (Bulk)- 3.2-1 3.2.1 Spent Fuel Pool Cooling System Design 3.2-1 3.2.2 Decay Heat Analysis 3.2-3 3.2.2.1- Basis 3.2-3 3.2.2.2 Model Description 3.2-4 3.2.2.3 Bulk Fool Temperature Results . 3.2-7 3.2.2.4 Spent Fuel Pool Heat Load and Cooling System Summary 3.2-8 3.2.3- Spent Fuel Pool. Makeup 3.2-12 3.2.3.1 Normal Makeup 3.2-12 3.2.3.2 Additional Makeup 3.2-12 3.2.3.3 Makeup Water Quantities 3.2-13
'3.2.4 Purification System Evaluation 3.2-14 3.2.5 Component Cooling Water and Saltwater Cooling Systems Evaluation 3.2-14 3.2.6 Fuel Handling Building HVAC 3.2-15 3.3 Thermal-Hydraulic Analyses for the Spent Fuel Pool (Localized) 3.3-1 3.3.1 Basis 3.3-1 3.3.2 Model Description 3.3-2 3.3.3 Cladding Temperature 3.3-5 3.4 ' Potential Fuel and Rack Handling Accidents 3.4-1 3.4.1 Rack Mishandling 3.4-1 O TMFWOO57 i
= _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _-_ _ ._ i
l TABLE OF CONTENTS (cont) PJlLE9 3.4.2 Temporary Construction Crane Drop 3.4-2 3.4.3 Loss of Pool Cooling (Storage Rack Drop) 3.4-3 3.4.4 Conclusions 3.4-3 3.5 Technical Specification Changes 3.5-1 3.6 References 3.6-1
- 4. MECHANICAL, MATERIAL, AND STRUCTUFAL CONSIDERATIONS 4.1-1 4.1 Description of Structure 4.1-1 s 4.1.1 Description of Fuel Handling Building 4.1-1 4.1.2 Description of New Spent Fuel Racks 4.1-2 4.1.2.1 Design of New Spent Fuel Racks 4.1-2 4.1.2.2 Fuel Handling 4.1-9 4.2 Applicable Codes, Standards, and Specifications 4.2-1 4.2.1 Fuel Handling Building - Spent Fuel Pool Analysis 4.2-1 4.2.2 Spent Fuel Racks - Design and Fabrication 4.2-2 4.3 Seismic Inputs 4.3-1 4.4 Loads and Load Combinations 4.4-1 4.4.1 Spent Fuel Pool 4.4-1 4.4.1.1 Loads 4.4-1 4.4.1.2 Load Combinations 4.4-4 4.4.2 Spent Fuel Racks 4.4-6 4.4.2.1 Loads 4.4-6 4.4.2.2 Load Combinations 4.4-7 4.5 Design and Analysis Procedures 4.5-1 9 4.5.1 4.5.1.1 Analysis Procedures for the Spent Fuel Pool Spent Fuel Pool Structure Finite Element 4.5-1 4.5-1 Analysis 4.5.1.2 Liner and Anchorage Analysis 4.5-4 4.5.1.3 Foundation Stability and Soil Bearing 4.5-4 4.5.2 Design and Analysis Procedures for Spent Fuel Storage Racks 4.5-5 4.5.2.1 Analysis Overview 4.5-5 4.5.2.2 Seismic Model 4.5-8 4.5.2.3 Time History Evaluation 4.5-26 4.6 Structural Acceptance Criteria 4.6-1 4.6 1
, Structural Acceptance Criteria for Spent Fuel Pool Structure 4.6-1 4.6.1.1 Criteria 4.6-1 4.6.1.2 Material Properties 4.6-1 4.6.1.3 Results 4.6-2 4.6.2 Structural Acceptance Criteria for Spent Fuel Storage Racks 4.6-11 4.6.2.1 Criteria 4.6-11 4.6.2.2 Stress Limits for Specified Conditions 4.6-12 4.6.2.3 Results for Rack Analysis 4.6-15 4.6.3 Spent Fuel Handling Machine (SFHM) Uplift Analysis 4.6-16 4.6.4 Fuel Assembly Drop Accident Analysis 4.6-16 4.6.4.1 Statement of Problem 4.6-16 TMFWOO57 ii
L TABLE OF CONTENTS (cont) Page 4.6.4.2 Model Definition 4.6-18
; ' 4.6.4.3 Drop Analysis Results 4.6-19' 4.6.5- Other Equipment Drop Analysis 4.6-20 4.6.6 Rack Displacements 4.6-25 4.6.7- Rack Location Verification 4.6-27 4.7 -Materials, Quality Control, and Special l Construction Techniques 4.7-1 4.7.1 . Construction Materials 4.7-1 4.7.2 Neutron Abscrber Material 4.7-2 4.7.3 Quality Assurance 4.7-7' 4.7.4 Special Construction Considerations 4.7-8 4.7.4.1 Removal / Installation Sequencing '
4.7-8 4.7.4.2 Safe Load Paths 4.7-13 4.7.4.3 Temporary 1 Construction Gantry Crane 4.7-15 4.7.4.4 Postulated Construction Load Drops 4.7-16 4.7.4.5 Use of Cask Pool and Cask Pool Cover 4.7-18 4.7.4.6 Control of Heavy Loads Evaluation 4.7-20 4.8 Boraflex Testing and Inservice Surveillance 4.8-1 4.8.1 Program Intent 4.8-1 4.8.2- Description of Specimens 4.8-1 4.8.3 Specimen Evaluation 4.8-2
-4.9 References 4.9 i
- 5. COST / BENEFIT, RADIOLOGICAL AND ENVIRONMENTAL ASSESSMENT 5.1-1 5 .1. Cost / Benefit and Environmental Assessment 5.1-1 O 5.1.1 5.1.2 Need for Increased Storage capacity Alternatives 5.1-1 5.1-2 5.1-10 5.1.3 Total Rerecking Cost 5.1.4 Resources Committed 5.1-12 5.1.5 Thermal Impact on the Environment 5.1-13 5.2 Radiological-Evaluation 5.2-1 5.2.1 Solid Radwasta 5.2-1 5.2.2 Gaseous Releases 5.2-2 5.2.3 Personnel Exposure 5.2-3 5.2.4 Radiation Protection During Reracking Activities 5.2 5.2.4.1 General Description of Protective Measures 5.2-5 5.2.4.2 Anticipated Exposures During Reracking 5.2-13 5.2.4.3 Exposure Controls During Diving Operation 5.2-14 5.2.5 Disposal of Rack and Other Materials 5.2-19 5.2.6 Radiological Impact on the Environment 5.2-20 5.3 Accident Evaluation 5.3-1 5.3.1 Spent Fuel Handling Accidents 5.3-1 5.3.1.1 Fuel Assembly Drop Accident 5.3-1 5.3.1.2 Cask Drop Analysis 5.3-3 5.3.1.3 Abnormal Location of a Fuel Assembly 5.3-4 5.3.1.4 Fuel Handling Accidents During Construction 5.3-5 5.3.2 Fuel Decay 5.3-6 5.3.3 Loads Over Spent Fuel 5.3-6 5.3.4 Loss of Spent Fuel Pool Cooling Flow 5.3-6 O TMFWOO57 iii I
TABLE OF CONTENTS (cont) Pace 5.3.5 Radiological Evaluation of Test Equipment Drop Onto the Racks . 5.3-7 5.3.6 Radiological Evaluation of Gate Drop Onto the Racks 5.3-8 5.3.7' Shielding Evaluation 5.3-9 5.3.8 Seismic Events 5.3-9 5.4 References 5.'4-1
- 6. RESPONSE TO REQUESTS FROM NRC RADIATION PROTECTION BRANCH DURING THE JUNE 3, 1988 MEETING.IN ROCKVILLE, MARYLAND- 6.1-1 0
4 TMFWOO57 iv
TABLE OF. CONTENTS (cont) Pace TABLES 2.1-1 Fuel Assembly Data . 2.1-3 3.1 Benchmark Critical Experiments 3.1-21
'3.1-2 Comparison of Phoenix Isotope Prediction to Yankee: Core 5 Measurements . .
. 3.1-22 3.1-3: Benchmark Critical Experiments, Phoenix Comparison 3.1-23 3.1-4 Data for U Metal and UO2 Critical Experiments 3.1-24 3 . 2-1 ' Principle Parameters of the Fuel Pool Cooling Systems- .
3.2-16 3.2-2 Anticipated Normal Offload Schedule 3.2-17 3.2-3 Postulated Full Core Offload Schedule 3.2-18 3.2-4 Anticipated Normal Refueling Heat Loads 3.2-19 3.2-5 Postulated Full' Core Offload Heat Loads 3.2-20 3.2-6 . spent Fuel Pool Heat Exchanger Design Specification'(per Heat; Exchanger)- 3.2-21 3.3 Normal Operation Results Local. Rack Thermal-Hydraulic Analysis 3.3-6 3.3-2 Flow. Blockage Analysis Results 3.3-7 4.1-1 Rack Data-(Each Unit)- 4.1-11
.4.4-1 Loads and Load Combinations for Spent Fuel Racks 4.4-8 ~
4.5-1 . Summary of' Seismic Analysis Bounding Cases 4.5-37 l4.6-1 Current Evaluation Results for the' Spent Fuel Pool' Walls and Basemat 4.6-28 Comparison of Governing Results for the Original O 4.6-2 Design Versus the Current Evaluation for the Spent Fuel Pool '4.6-29 4.6-3 Comparison of Modal-Characteristics for the
-Lumped Parameter Model'Versus the Current Evaluation- 4.6-30 4.6-4 Minimum Margin to Allowable Region I~ 4.6-31 4.6-5 Minimum Margin to Allowable Region II 4.6-32 4.6-6 Rack Gap Spacing Results .
4.6-33 5.1-1 Approximate Spent Fuel Storage Needs SONGS - Unit 2 . 5.1-14 5.1-2 Annual Power Replacement Costs Attributed to SONGS' Unit 2 5.1-15 5.1-3 Annual Power Replacement Costs Attributed to SONGS Unit 3 5.1-16
-5.2-1 Normal'and Maximum Isotopic Inventories of the Spent Fuel Pool Purification System Ion Exchanger [ Ci) '
5.2-22 5.2-2 Fuel Handlang Building Normal and Refueling Operation Airborne Radioactivity Concentrations 5.2-23 5.2-3 Dose Rates in the Vicinity of the Spent Fuel Pool (mrem /h) 5.2-25 5.2-4 Estimated Radiation Doses for Construction Activities 5.2-26 6-1 Radionuclides in the Spent Fuel Pool Water 6-14 6-2 Airborne Radionuclides in the Fuel Handling Building 6-15 TMFWOO57 v
TABLE OF CONTENTS (cont) 9 FIGURES q 1-1 Leak Chase System' 2.2-1 Spent Fuel Storage Rack Arrangement 2.2-2 Fuel Assemblies in High Density Spent Fuel Racks 2.2 Rack Location in Spent Fuel Pool 3.1-1 Region I Cell Layout L 3.1-2 Region II Cell Layout 3.1 Units 2 E 3 Fuel Minimum Burnup vs. Initial Enrichment for Region II Racks 3.1-4 Unit 1 Fuel Minimum Burnup vs. Initial Enrichment'for-Region II Racks 3.1-5 Fuel Storage Patterns for Region II Racks 3.1-6 Fuel Storage Patterns ~for Region II Racks Reconstitution Station-3.2-1 Spent Fuel Pool Cooling Loop 3.2-2 Proposed Unit 2' Spent Fuel Pool Cooling Piping 3.2-3 Shutdown Cooling for Spent Fuel Pool 3.3-1 Spent Fuel-Pool Natural Recirculation Model (Elevation View) 3.3-2 Epent Fuel Pool' Natural Recirculation Model (Plan View)' 3.3-3 Spent Fuel Rack Inlet Flow Area (Plan View) 4.1-1 Fuel Storage Rack (Region I) 4.1-2 Region I Rack Cross-Section 4.1-3 Region II Fuel Storage Rack 4.1-4 Region II Rack Cross-Section 4.1-5 Region II Rack Top-View 4.3 SONGS 2 and 3 Fuel Building Pool Floor Horizontal NS' O 4.3-2 (NE368 C4) for 4% Damping DBE Spectra SONGS.2 and 3 Fuel Building Pool Floor. Horizontal EW (NE368 C4) for 4% Damping DBE Spectra 4.3-3 SONGS 2 and 3 Fuel Building Pool Floor Vertical (NE360 C4) for 4% Damping DBE Spectra 4.3-4 SONGS 2 and 3 Fuel Building Pool Floor Horizontal NS (NE391 C4) for-2% Damping OBE Spectra 4.3-5 SONGS 2 and 3 Fuel Building Pool Floor Horizontal EW (NE406 C4) for 2%-Damping OBE Spectra 4.3-6 SONGS 2 and 3 Fuel Building Pool Floor Vertical (NE396 C4) for 2% Damping OBE Spectra 4.5-1 Isometric View of SONGS 2 and 3 Fuel Handling Building 4.5-2 Acceleration Time History N-S DBE O to 40 Sec. 4.5-3 Acceleration Time History N-S DBE 40 to 80 Sec. 4.5-4 Acceleration Time History E-W DBE O to 40 Sec. 4.5-5 Acceleration Time History E-W DBE 40 to 80 Sec. 4.5-6 Acceleration Time History Vert. DBE O to 40 Sec. 4.5-7 Acceleration Time History Vert. DBE 40 to 80 Sec. 4.5-8 Structural Models Regions I & II 4.5-9 Effective Structural Models Regions I EII 4.5-10 3-D Nonlinear Seismic Model Region I & II 4.5-11 Nonlinear Seismic Model (2-D View of 3-D Model) Region I 4.5-12 Nonlinear Seismic Model (2-D View of 3-D Model) Region II TMFWOO57 vi
TABLE OF CONTENTS (cont). FIGURES 4.5-13 Nonlinear' Seismic Model Partial Fuel, Quadrant Loading Region I 4.5-14 ' Nonlinear Seismic Model Partial Fuel, 4 Row Loading Region I 4.5-15 Multiple Rack Model Full / Full Region I 4.5-16 Multiple Rack Model Empty / Full Region I-4.5-17 Multiple Rack-Model Full / Full Region II 4.5-18 . Multiple Rack Model Empty / Full Region II 4.5-19 Multiple Rack Model'(Plan View) Region II 4.5-20 Cell to Cell Wald Finite Element Model Region II 4.6-1 SONGS'2 and 3 Free Vibration Analysis (All Standard Fuel in SFP) Elevation Looking North Mode No. 4 4.6-2 SONGS 2 and 3' Free Vibration Analysis (All Standard Fuel-in SFP) Elevation Looking East-Mode No. 5 4.6-3 SONGS 2 and 3 Free Vibration Analysis (All Standard Fuel in SFP) Elevation Looking East Mode No.,6 4.6-4 SONGS 2 and 3 Fuel-Building Raw Floor Spectra El. 17' 6" DBE Horizontal (N-S) for 2% Damping 4.6-5 SONGS 2 and 3 Fuel Building Raw Floor Spectra El. 17' 6"
'DBE Horizontal (E-W) for 2% Damping.
4.6-6 SONGS 2 and 3 Fuel Building Raw Floor Spectra El. 17' 6" DBE Vertical for.2% Damping 4.6-7 SONGS 2 and=3 Fuel Building Raw Floor Spectra El. 114' 0" DBE Horizontal (N-S) for 2% Damping 4.6-8 SONGS 2 and 3 Fuel Building Raw Floor Spectra El. 114'- N 0" DBE Horizontal (E-W) for 2% Damping 4.6-9 SONGS 2 and 3 Fuel Building Raw Floor Spectra El. 114' 0" DBE Vertical for 2% Damping 4.6-10 Drop Heights Over Racks 4.7-1 Fuel Handling Building Unit 2 4.7-2 Spent Fuel-Pool (Unit 2) Original Condition 4.7-3' Spent Fuel Pool-(Unit 2) Final Configuration 4.7-4 Rerack Sequencing of Spent Fuel Pool (Unit 2) Original Condition 4.7-5 Proposed Rerack Sequencing Step 1
'4.7-6 Proposed Rerack Sequencing Step 2 4.7-7 Proposed Rerack Sequencing Step 3 4.7-8 Proposed Rerack Sequencing Step 4 4.7-9 Safe Load Paths Fuel Handling Building (Unit 2) 4.7-10 Temporary Construction Gantry Crane 4.7-11 Plan (Cask Pool Cover) 1 4.7-12 Cask Pool cover Installation 4.7-13 Fuel Handling Building (Unit 2) Lifting Equipment 4.7-14 Temporary Cask Pool Storage Rack TMFWOO57 vii
. _ _ . - _= _ _-___ . _ _ _ _ _ _
- 1. INTRODUCTION 1.1 PURPOSE Southern California Edison Company (SCE) plans to install free standing, high density, spent fuel racks in the spent fuel pools (SFPs) of San Onofre Nuclear Generating. Stat!.on Units 2 and 3 (SONGS 2&3). The purpose of the new racks is to increase the-amount of spent fuel that can be stored in each existing SFP from-800 to 1572 elements. Therefore, this report supports the SCE request that a License Amendment be issued to the SONGS 2&3 Facilities Operating Licenses, NPF-10 and NPF-15(1,2),
respectively, to include installation and use of free standing
. racks that meet the criteria contained herein. Additional storage is needed because the Federal Repository will not be f- available as initially scheduled.
k 1.2 PROPOSED FACILITY MODIFICATION 1.2.1 PRESENT FACILITY DESCRIPTION The spent fuel racks are located in the SFP in the fuel handling building (FHB) which is a Seismic Category I reinforced concrete structure. The FHB is of heavy shear wall construction with a l concrete slab, steel frame, composite-action roof system and a wall thickness that provides nuclear shielding and protection TMFWOO57 1-1
1 r against design wind' loading and tornado-generated" missiles. Separate?FHBs are provided for Unit 2 and for Unit 3. . A License Amendment has been granted that allows for the storage of Unit 1-l I
-fuel assemblies in'both the Unit 2 and Unit 3 SFPs(3),
Each FHB contains a SFP in which 15 spent fuel storage racks ara 1 located. These racks currently are designed _to provide underwater storage locations for up to'800 fuel assemblies per. unit (3-2/3 cores). Each of the 15 racks is bolted to beams
-which'are anchored to the SFP floor. The apent fuel storage racks are designed to provide protection against damage to the ;
fuel and to prevent' fuel. assemblies.from being inserted into other than the prescribed locations. They are constructed entirely of stainless steel. Units 2 and 3 have separate and independent SFP cooling and () purification systems. Each cooling system.is designed to provide continuous cooling for spent fuel assemblies stored in the fuel pool. Under normal operating conditions with one train operating, the cooling system i maintains the pool temperature at 140F or less, with a maximum of 2-2/3 cores stored in the SFP. During full core discharge, assuming 7 days decay of a Unit 2 or 3 full core, 90 days decay for the most recent Unit 2 or 3 one-third core refueling, 90 days decay time for a Unit 1 full core and the remaining spaces filled. with Unit 2 or 3 fuel, the maximum pool temperature'is maintained at 140F or less. With these assumptions, the maximum heat load TMFWOO57 1-2 O
1 7 f l: i l- : ) in the fuel pool from 800 fuel assemblies is 40.3 x 10 6 BTU /h, - l and with both fuel pool cooling trains in service, the maximum 7 fuel pool water temperature is calculated to be 140F or-less. I ( Each fuel pool purification system includes purification equipment designed to remove soluble and insoluble foreign matter from the SFP water and dust from the pool surface. This maintains the fuel pool water purity and clarity, permitting visual observation of underwater operations. At a temperature of 140F or greater, the purification system does not operate. Any leakage from the fuel pool cooling system is detected by reduction in pool inventory. The leak chase channels and liner weld seams are shown in figure 1-1. A sump with a high level alarm, is provided to collect system leakage. Makeup to the SFP is from the seismic category I refueling water storage tank. The N SFP normal makeup capacj47 of 150 gal / min exceeds normal system leakage and evaporative losses. Several alternate backup paths for makeup cre available. The failure of portions of the SFP cooling systems, or of other systems due to seismic loading which are not designed as Seismic category I but are located close to essential portions of the system, will not affect the performance of essential functions.
l with both fuel pool cooling trains in service, the maximum fuel ( pool water temperature is calculated to be 140F or less. Each fuel pool purification system includes purification equipment designed to remove soluble and insoluble foreign matter i from the SFP water and dust from the pool surface. This j maintains the fuel pool water purity and clarity, permitting visual observation of underwater operations. At a temperature of 140F or greater, the purification system does not operate. Any leakage from the fuel pool cooling system is detected by reduction in pool inventory. The leak chase channels and liner weld seams are shown in figure 1-1. A sump with a high level alarm, is provided to collect system leakage. Makeup to the SFP is from the Seismic Category I refueling water storage tank. The ! SFP normal makeup capacity of 150 gal / min. exceeds normal system i leakage and evaporative losses. Several alternate backup paths for makeup are available. The failure of portions of the SFP cooling systems, or of other systems due to seismic loading which are not designed as Seismic Category I but are located close to essential portions of the system, will not affect the performance of essential functions. TMFWOO57 1-3
q 1.2.2 ' PROPOSED MODIFICATION With-the existing 800 storage locations in the present spent fuel ! storage racks,-Unit ~2 and Unit 3 will each retain.its. full-core reserve storage capacity through cycle 6 which is-scheduled to begin in 1991 and 1992, respectively. To increase the capacity ll to store discharged fuel assemblies at SONGS 2&3, SCE plans to replace the existing storage racks with frea standing high density spent fuel storage racks. This. higher density storage of fuel assemblies expands the capacity of each existing pool to approximately 1572 assemblies, extending the full-core reserve storage capability for. Unit 2 and Unit 3 through cycle 11 operation which is scheduled to begin in 2001 and 2002,
.respectively. .
1.3 INTERFACES WITH OTHER ORGANIZATIONS Southern' California Edison has overall responsibility for this ) l modification. Westinghouse (H) has designed the new free standing high density spent fuel storage racks. Additionally, H is responsible for the fabrication of the racks and the evaluation of the racks including accident conditions. Bechtel Power Corporation (BPC) is responsible for the building structural analysis, the evaluation of the SFP cooling system and TMFWOO57 1-4
t h e r e l a t e d a c c i d e n t e v a l u a t i o n s .- The installer, who-will be [ selected at a later date, will be responsible for the installation of the new racks.. _ 1.4
SUMMARY
OF REPORT l. l This' report follows the guidance of the NRC Position Paper entitled, "OT Position for Review and. Acceptance of Spent Fuel. Storage and Handling Applications," dated April 14,-1978, as amended by-the NRC letter dated January.18, 1979(4). Sections 3.0 through 5.0 of this report are consistent with the section format and content of the NRC OT Position Paper, Sections III through V. Section 2.0 provides a summary description of the main features of the' existing and the'new racks. The nuclear and thermal-hydraulic aspects in section 3.0 of this report address the neutron multiplication factor, consideration of normal storage and handling of spent fuel as well as postulated accidents with respect'to criticality and the ability of.the SFP cooling system to maintain sufficient cooling. Section 4.0 describes the mechanical, material, and structural aspects of the new racks. It contains information concerning the capability of the storage racks and SFP to withstand the effects of' natural phenomena and other design loading conditions. TMFWOO57 1-5
Special. construction aspects are:also included in this section. () Movement of spent fuel'storedfin.the SFP during removal of the
-present racks and: installation of_the new racks is also.
addressed. The environmental aspects in section'5.0 of this report-concern . I thermal and radiological releases from theifacility.under normal i and accident conditions. This section also addresses the occupational radiation exposures, including. exposures resulting from the.insbe.1)ation of the new racks, generation,of radioactive. waste, need for expansion, commitment of. material and nonmaterial resources, and a cost / benefit assessment. FI 1.5' CONCLUSIONS tJ on the basis of the information and evaluations presented in this report, SCE concludes that the proposed modifications'of the
-SONGS 2&3 spent fuel storage' facilities provide safe spent fuel storage, and.that~the proposed modification is 'in'conformance with NRC requirements. The installation and use of high density
-fuel storage racks will have no impact on the health and safety of the general public.
TMFWOO57 1-6
l I'
1.6 REFERENCES
.i -
- 1. San Onofre Nuclear Generating Station Unit 2 Facility Operating' License NPF-10, Docket'No. 50-361..
~
- 2. San Onofre Nuclear Generating ~ Station Unit 3 Facility Operating License NPF-15, Docket No. 50-362.
- 3. ' San Onofre Nuclear Generating Station Unit'2 Facility Operating License NPF-10, Amendment 63, dated-June 22, 1988, Docket No. 50-361; and San Onofre Nuclear Generating Station Unit 3 Facility Operating License NPF-15, Amendment 52, dated June 22, 1988, Docket No. 50-362.
- 4. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes, April 14, 1978, "OT Position
'for Review and Acceptance of Spent Fuel Storage and Handling Applications," as amended by the NRC letter dated January 18, 1979.
TMFWOO57 1-7
1
- 2.
SUMMARY
OF RACK DESIGN 2.1' EXISTING RACKS - The existing spent fuel storage racks are designed to maintain a minimum edge to edge spacing of 4.07 inches between fuel assemblies. This separation results in an effective neutron multiplication factor (Ke'ff) of less than 0.95 with the highest anticipated enrichment of 4.1%(1) . Structural deformations are limited and edge-to-edge spacing is maintained to preclude the possibility of criticality. The spent fuel ~ storage racks are anchored at their base to prevent base lateral movement when the racks are loaded with the designed number of fuel assemblies under anticipated loading conditions, (O f including the design basis earthquake (DBE). Two sizes of spent fuel storage racks are used. One size contains 64 storage locations in an'8 by 8 square array (102 inches x 102 inches). The other size of fuel storage rack is ! I similar in design, but contains only 32 storage locations in an 8 x 4 array (102 inches x 51 inches) . In both sizes, the spent fuel storage locations are spaced 12.75 inches by 12.75 inches on l centers. Each storage location consists of a square box 8.81 inches in outside dimension, with minimum 0.120-inch thick stainless steel walls. The spent fuel assembly is located within the stainless steel box. Type 304L stainless steel is used for the square boxes and for all of the principal structural grid i TMFWOO57 2.1-1
members and bracing. Therefore, material compatibility between-I' components of the SFP and the Zircaloy-4 and stainless steel clad & fuel'~ assemblies is established. For further information on the existing spent fuel storage racks,. see subsection 9.1.2 of the SONGS 2&3 UFSAR(2), Westinghouse 14 x 14 fuel assemblies from SONGS Unit 1 and Combustion Engineering (C-E) 16 x 16 fuel assemblies from SONGS
-2&3 may be stored in the existing Unit 2 and Unit 3 racks. The I outline dimensions and related data for the C-E 16 x 16 fuel is provided in subsection 4.2.2 and figures 4.2-6, 4.2-7, and 4.2-8 of the SONGS 2&3 UFSAR, as well as table 2.1-1 herein. The outline dimensions of the H 14 x 14 fuel assemblies is shown in figures 9.1-7 and 9.1-8 of the. SONGS 2&3 UFSAR as well as table 2.1-1.
v i l TMFWOO57 2.1-2
-_____-____-___D
?
'i
~ ,. 8
' f)
.:\Q
' Table-2.1-1 FUEL' ASSEMBLY DATA .
E c:s: Fuel Assembly Overall Length (in.) '138.5 1176.8 1 Upper End Fitting. Square Size (in.)_ 7.763 8.030-Lower End' Fitting l Square Size (in.) 17.763 8.135 Spacer' Grid Square' Size (in.) 7.763 '8.134' o Number of Spacer. Grids 7.0 '11.0' Number of Rods Per Assembly; 180.0- 236.0l in.)' O.422 0.382'
' Fuel FuelRod. RodInner Outer: Diameter Diameter ( (in. ) 0.389 0.332 0.506-
,' Fuel' Rod Pitch (in.) 0.556 150 (approx)
' Active Fuel Length (in.) .
120.0-Fuel: Rod Overall Length (in.) 126.7 161.2-161.9-Fuel. Assembly Weight _(lb) 1150.0 1451.0-t
+O TMFWOO57 2.1-3 l
2.2 NEW'HIGH DENSITY RACKS
/~'\ !
-(_) !
-The new spent fuel storage racks will provide for storage of new.
and spent fuel assemblies in appropriate regions of the SFP, while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from excess mechanical or thermal loadings. Westinghouse 14 x 14 fuel assemblies from SONGS Unit 1 and C-E 16 x 16 fuel assemblies from SONGS 2&3 may be stored in either Unit 2 or Unit 3 racks. The plan view spent fuel storage pool rack arrangement for SONGS Unit 2 is shown in figure 2.2-1. Unit 3 rack arrangement is opposite hand. Figures 2.2-2 and 2.2-3 show elevation views of the arrangement of the racks in the SFP. () Fuel will be stored in two regions within each pool. (312 locations) consists of two high density fuel racks, each Region I with 12 x 13 cells (125.5 inches by 135.9 inches) which has spacing obtained by utilizing a neutron absorbing material (Boraflex) and is used for full core off load (217 fuel assemblies), plus 95 locations. Region II (1260 locations), which also uses Boraflex, has six high density fuel racks, each with 14 x 15 cells (124.82 inches by 133.67 inches) and provides normal storage for spent fuel assemblies. Region I will be used to store non-irradiated, 4.1 w/o (or less) U235 enriched fuel and l fuel which has not achieved a pre-determined burnup. Region II l 1 is designed to accommodate irradiated fuel which meets the predetermined burnup. Placement of fuel in Region II is TMFWOO57 2.2-1 s
4
-determined by burnup calculations and is controlled b adn'inistratively.
) Fuel which does not meet the burnup criterion may._befplaced in Region-II.in a checkerboard or alternating. row
, array 1(paragraph 3.1.3.3). In these cases, fuel storage '
surrounding the assembly will be controlled administratively'to prevent inadvertent insertion:in an unapproved configuration
'(Kegg'> 0 . 95). No physical: barrier is necessary between-the two' regions. The racks meet the requirements of the NRC "OT
- Position for Review and Acceptance of Spent' Fuel Storage and Handling Applications,". dated April 14, 1978,.and modified January'18, 1979, with the exception that, for Region II storage, credit is taken for fuel burnup based on the proposed Revision 2 of USNRC Regulatory Guide l.13, Section 6. See subsection 4.1.2 for detailed rack description.
p t A summary of major design criteria'is provided below: A. The racks are designed in accordance with the NRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978 and revised January 18, 1979, with the exception that credit is taken for burnup for all storage locations other than those in Region I which is used for full core discharge. (Region II credit for burnup allowed in proposed Revision 2 to USNRC Regulatory Guide 1.13.)
~
TMFWOO57 2.2-2
4 B. The racks are designed to meet the. nuclear requirements of
-ANSI-57.2-1983. The effective multiplication factor, 7]
Keff, in the SFP is less than or equal to i 0.95,.' including uncertainties and under-credible'
; conditions.
C. The racks are designed to allow coolant flow such that boiling in the water channels between the-fuel assemblies in the rack does not occur. Maximum fuel cladding temperatures are calculated for various pool cooling t conditions (see section 3.3). D. The racks are designed to Seismic Category I requirements. They will not impact each other or the pool walls under seismic conditions, and are classified as ANS Safety Class 3 and ASME Code Class 3 (Component Support Structures, Subsection NF of Section III). The structural. evaluation and seismic analyses are performed using the specified loads and load combinations'in table 4.4-1. E. The racks are designed to withstand loads which may result from fuel handling accident or from the maximum uplift force of the spent fuel handling machine (SFHM) of 6000 pounds without violating the criticality acceptance criterion. l l TMFWOO57 2.2-3 L 1 l
s
+ . .
t P , F.: ;Eachistorage.positionLin:the racksLis designed'to-support-and~ guide-the fuel assembly;in a-manner..that will preclude
.the; possibility.of.applicat' ion of excessive lateral," axial and' bending l loads.to fuel assembliesLduring: fuel l assembly, handling and' storage operations.
r-F- 2G. The materials'used in-construction'of--the racks;are compatible'withLthe storage' pool environment and will:not-contaminate the fuel assemblies. r
.H. The: spent fuel racks consist of two designs varying with storage capability.-TRegion I racks are designed to' store
' fresh' fuel-(and thus burned fuel can'also be stored).
Region II racks are designed to store burned fuel whose-d initial" enrichment vs. discharge burnup meet ^the. criteria of figures'3.1-3 and 3.1-4. ' New and burned fuel which does not meet the enrichment vs. burnup criteria of figures'3.1-3 and 3.1-4 can be stored in Region II if a checkerboard or alternating row pattern is used. I. Both H 14 x 14 fuel assemblies from SONGS Unit 1 and C-E 16 x 16 fuel assemblies from SONGS 2&3 may be stored in either Unit 2 or Unit 3 racks. TMFWOO57 2.2-4
7____-.
. l'.
~
J. .The rack seismic / structural analyses are done in
.accordance with-USNRC SRP 3.8.4, Revision l', July 1981,.
"Other:. Category I' Structures, Appendix D,: Technical Position on Spent Fuel Racks".
\
i i O TMFWOO57 2.2-5 i
, u
' 7 7 - 8 r L 6 2 r 8 4 3 J 1 3 0
- 3) 0 4
- 3 2 1(
t t3 0 P 1 y> l .
- 5 5 9 1 1 _
l l g g 6 X iR X i R _ nn 3 4 4 ic a ic 1 1 _ pp a SS _ CC TT CC _ i n n
) .
T 0 i U )e 4. 85 2 5 5 ( 1 p 1 l 2 O g 0 8 1 8 X R X i R T Y a t A. Imta as I
- 3. 4 4 N 1 3 1 1 _ U L E O ro s
t r n N _ R Or r F Pi it oi t a c o a 0 O N E LM coL o _ L 0 8 2 N3 ( _ OU3 A F N T igal eg L Nn t ar e 5 5 5 r o 4 2 1 1 l _ L L A S P(EU ot t S l X i S S X i 4 8 R R W 0 ~ 4 2 6 3 2 ~ 3 4 4 L G 2 1 1 O 1 1
. . O :
P 4N : l l I _ NN
>> OO I I N -- .
G GG EE RR 0 9 3 1 3 _ M R 1 E 5 x X I _ W 5 3 R Ow7 2 3 1 2 1 2 1 _ L O O P
< 4& l t
) ) P_ - 0 r
. P 1
0 T j 5 kF 5 5 + L J 1 5 2 2 0 1 0 1 1 0 9 1 7 6 m$3$y zc0b $ og $ $ g e :>45z f PW I , $E nE hyom
, n
- Ex 3$EO @$ i n $$ n*
UNIT 2 OR 3 CE 16 x 16 FUEL A55Y. gm CELL A55Y. k,) ," c4 n '
$ ! .3
. ! ,-4 3-4 n i , 1 g, I I I l !. l l
31.04" ii t RE :. .
,, , 'l I ml E.iri I 'l *i* t ***8 .._L-,_', , " ,
- Lt ._ _.L., *_!_*,.L_Lli I . , . 1 l '
l UNIT 1 i i l I i l [ ' i
! 1 < i ,
1' I I t W 14 x 14 Q j. FUEL A55Y q hy,. _ _, l-) 1 J -
.! l .I .. j l
', 8
, j ,
l ! i q l l l l ! _! I, .l l l l l yl 1 I I, I k I 1
' ! l, , l g:
.l ., .
l l l i d 4
- J i d . .'l i i
J l l .J
.l I -
I i ! , , ,
'l I i l ! l ! l ! l l
~ ~
]
~
190.00" .l i l i -j ['j l 'l ]' j g REF. l
]1 .
176.8" l , I '
, g i ! !
\ l l ! ' t i gj_, . -. _
s _ q l 1 g ; *,.. ... . d -\ s 138.5" 1 l i ; ! j l l
~
l
~ ~
I I Ifj l ] ; li, l l
'I l i 1
I! i l , ' i I
~
I l l
~
I i l~ h ]_ h ]' l l
~l I !- ? E J i
; I !
- 1. ,
l I I I! . l , i l ji [ ll l~ l~i l]!- ' E [ ; . I
~l I lid' io -
[
'l
';i l 1 1: ! l
, ! l ; 1 1 i
;l I!"i i '
!~
1 Fi !~l i l F i Il j"; l i, N ' i l~l i I l g i ' ' ' l i d,u,, i,-- W n l ,, n .: i i '
.n a i
,. , i i i i i . i o r 4 i vo i i I t i
" j i~ i I j L" i j l Lg i i " 1l I i_
i N BASE PLATE b 6 .00" REF k SUPPORT PAD A55Y k POOL FLOOR REGION I RACK SHOWN. REGION II RACK HEIGHT 15 IDENTICAL. SAN ONOFRE NUCLEAR GENERATING STATION jq NOTE: BRIDGE PLATES NOT SHOWN IJnits 2 & 3 FUEL ASSEMBLIES IN HIGH DENSITY SPENT FUEL RACKS FIGURE 2.2-2
SPENT FUEL POOL 7~t -(UNIT 2) L)L
', 5 k~
f
}YJ
] e.
{* .s;
- OPERATING FLOOR E L. 63'- 6" o m
-TECH SPEC /
WATER LEVEL _ _ _ _ - - _ _
'3 ..
d w
*i '.' 's *- =,
.;.. :. A *s .,
w g.a *: A i .*. m
* ~
o 8' -
*:.t Y
- o m: h..e:k.'+
i 9 : 7'N # 0 .. m v. TOP OF RACKS "~-
! y J j 2
e
'd
*i :'l "
BOTTOM t k ,g. ____.. _ ______ i# 4'.- OF POOL EL.1716"
.'s
.. a',8 . .h g.' ..A4 b.
- .f* *: ..
ii _^ __ _r _ _ - - SECTION VIEW LOOKING NORTH SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 RACK LOCATION IN SPENT FUEL POOL FIGURE 2.2-3
.s .
.J
- 2. 3- REFERENCES '
O l1. Technical Specification, NUREG-0741'Section 5.3.1, Amendment
.No. 39, dated' December 2, 1985.
. San Onofre Nuclear Generating Station-Un'its 2 and 3 Updated 2.
Final Safety Analysis' Report, Docket Nos. 50-361 and'50-362.
.2 O .
O TMFWOO57 2.3-1
- 3. NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS PT k- 3.1 NEITTRON MULTIPLICATION FACTOR 3.1.1 NORMAL STORAGE Criticality of fuel assemblies in the spent fuel storage racks is prevented by the design of the racks which limits fuel assembly interaction.. This is done by fixing the minimum separation between assemblies, inserting neutron absorbing material (Boraflex) between assemblies and using administrative controls when.necessary.
The design basis for preventing criticality in the SFPs is that, including uncertainties, there is a 95% probability at a 95% () confidence level that the effective multiplication factor (Kerg) of the fuel assembly array will be less than or equal to 0.95 when the storage racks are fully loaded with spent fuel, including a half core of new fuel, and flooded with unborated water as recommended in ANSI-57.2-1983 and in OT Position Paper. The following are the conditions that are assumed in meeting this design basis: A. The fuel assemblies contain the highest enrichment l authorized (4.1 w/o) without any control rods or any burnable poison and are at their most reactive point in O TMFWOO57 3.1-1 1
t > life. The C-E 16 x'16-fuel assembly is more reactive than j( the H 14 x 14 fuel from SONGS 1 as stored in SONGS 2 or 3
% ,) '
. pools.
These fuel assemblies are conservatively modeled with water replacing the assembly grid volume and no U234 or
. U236-in the fuel pellet. .oU235 N burnup is assumed for-fuel stored in Region I.
B. The storage cell nominal geometry is shown on figure 3.1-1 for Region I and figure 3.1-2 for Region II. C.. The. moderator'is pure water at the temperature and density within the design limits of the pool which yields the
- largest. reactivity. No dissolved boron is assumed in the
'T water for normal condition.
' [V D. The nominal case calculation is infinite geometry in lateral.and axial extent. However, Boraflex sheets are not necessary on the periphery of all the rack arrays (see paragraphs 4.1.2.1.2.3 and 42 1.2.1.1.2) and because calculations show that this finite array is less reactive than the nominal case infinite array. Therefore, the nominal case of an infinite array of neutron absorbing l cells is a conservative assumption.
i TMFWOO57 3.1-2 l Z_________. _ _ _ _ _ . _ _ _ _ _ _ . _ _ _
l 'l ' [ ': E. Mechanical' uncertainties and biases due to mechanical 7" ; tolerances during rack fabrication are treated by using
'" worst case" conditions. The items included in the analysis.are:
L e Boraflex thickness l' e .: Stainless steel thickness e1 Cell ID e Center-to-center spacing e Asymmetric assembly positioning e Boraflex shrinkage e Boraflex edge deterioration The tolerances for these seven items are stacked such that the Boraflex, asse'bly m spacing and water gap between
/# . assemblies is minimized.. As a result the Boraflex %) material was reduced in' thickness by 0.007 inch and,in length / width by 0.06 inch for fabrication tolerances, 3%
for shrinkage and 0.125 inch for edge deterioration. The assembly center-to-center spacing is reduced by 0.06 inch. The stairTAss steel thickness and cell ID are increased by 0.004 inch and 0.025 inch respectively to reduce the water gap between assemblies. [ Note that because the Region II rack design does not have a water gap between cello (see figure 3.1-2), the worst case 2 conditions are also obtained by reducing the steel l f thickness and the cell ID.] w i TMFF0057 3.1-3 l
The uncertainty and bias associated with the calculation
. ll ' method are discussed in subsection 3.1.3.
) F. Credit is taken for the neutron absorption in full length structural materials and in solid materials added specifically for neutron absorption. A minimum B10 loading is assumed in the Boraflex sheets and B C 4 particle self-shielding is included as a bias in the reactivity calculation. (See paragraph 4.1.2.1.2 for Boraflex description.) 3.1.2 POSTULATED ACCIDENTS Most credible accident conditions will not result in an increase h in Keff of the rack. Examples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembiy has more than 12 inches of water separating it from the active fuel height of stored assemblies which precludes interaction). However, accidents can be postulated which would increase reactivity if the presence of boron was not assumed. Therefore, for accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, indejandent, concurrent events to ensure O TMFWOO57 3.1-4
protection against a criticality accident. Therefore, for () accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. This assumption does not violate item C in subsection 3.1.1. The presence of approximately 2000 ppm boron (conservative compared to 2350 ppm minimum) in the pool water will decrease reactivity by about 30% Ak. In perspective, this is more negative reactivity than is present in the Boraflex sheets so Keff for the rack would be less than 0.95 even if the Boraflex sheets were not present. For all postulated accidents involving misplacement and damage to fuel assemblies, should there be a reactivity increase, Keff would be less than or equal to 0.95 due to the combined effects of the dissolved boron and the Boraflex () sheets. The " optimum moderation" accident is not a problem in spent fuel storage racks because the presence of Boraflex sheets removes the conditions necessary for " optimum moderation". The Keff continually decreases as moderator density decreases to values less than 1.0 gm/cm 3, O TMFWOO57 3.1-5 l I
l t-
- 3.1.3: CALCULATIONAL METHODS '
O q Region I The criticality calculation method and cross-section values are-verified by comparison with~ critical experiment data (tables 3.1-1, 3.1-3, and 3.1-4) for assemblies similar to those'for-which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and . s_ uncertainty will apply to rack conditions which' include strong neutron absorbers, large water gaps and low moderator densities. The design method which ensures the criticality safety.of fuel assemblies in;the spent fuel storage rack uses the.AMPX system of codes (1,2) for cross-section generati~n o and KENO IV(3) for
. reactivity determination.
The'227 energy group cross-section library that is'the common starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENDF/B-V(1) data. The NITAWL(2) program includes, in this library,-the self-shielding
- resonance cross-sections that are appropriate for each particular
. geometry. The Wordheim Integral Treatment is used. Energy and-
. spatial weighting of crens-sections is performed by the XSDRNPM.
program (2) which is a one-dimensional S N transport theory code. These multigroup crocs-section sets are then used as input'to KENO IV(3) which is a' three-dimensional (3-D) Monte Cr rlo 4 criticality theory program designed for reactivity calculations. O TMFWOO57 3.1-6 m.a.-_--__.-___-______-.-_--m___.m_____.m___-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . _ _ . . . _ _ _ _ _ _ _
j A set of 33 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The experiments range from water moderated, oxide fuel arrays separated by various materials (B4 c, steel, water, etc.) that simulate Light Water Reactor (LWR) fuel shipping and storage conditions (4) to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials (5) (plexiglas and air) that demonstrate the wide range of applicability of the method. Table 3.1-1 summarizes these experiments. The average Keff of the benchmarks is 0.992. The standard deviation of the bias value is 0.0008 Ak. The 95/95 one sided tolerance limit factor for 33 values is 2.19. Thus, there is a 95% probability with a 95% confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0018 Ak. Region II Spent fuel storage, in the Region II spent fuel storage racks, is achievable by means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with fuel depletion. A series of reactivity calculations are performed to generate a set of enrichment-fuel assembly discharge burnup ordered pairs which all yield the equivalent Kerf when the fuel is stored in the Region II racks. O TMFWOO57 3.1-7
1
- Figure 3.1-3 shows the constant Kegg contour. generated'for the SONGS Unit 2 or 3 Region II racks..when storing the'C-E 16 x 16 fuel. Note in' figure 3.1-3 the endpoint at 0 MWD /mtu where the i
. enrichment is 1.85 w/o.and at 27,000 MWD /mtu where the enrichment
. is'4.10 w/o. The interpretation of the endpoint data is.as
~
follows: the reactivity of'the Region.II racks containing fuel . at 27,000 MWD /mtu burnup which had an initial ~ enrichment of 4.10 w/o is equivalent to the reactivity of the Region II racks containing fresh fuel having an initial enrichment'of.1.85 w/o.- It is important.to recognize that the curve in figure 3.1-3 is based on a constant rack reactivity for that region and not on a constant fuel assembly reactivity. Rack reactivity is the reactivity of the fuel assemblies in the fuel rack geometry. Fuel assembly reactivity is the reactivity of the fuel assemblies outside of the rack geometry in only water. Figure 3.1-4 shows the constant Keff contour generated for the SONGS Unit 2 or 3 Region II racks when storing H 14 x 14 fuel. The end points'in this figure are at 0 MWD /mtu where the enrichment is 2.40 W/o and at 20,500 MWD /mtu where the enrichment is 4.10 w/o. The data points on the reactivity equivalence curve were generated with a transport theory computer code, PHDENIX (6), -) i i O TMFWOO57 3.1-8
- _ _ _ - - - - - _ _ _ . .- b
- -_- -___ _ _ = _ - - _ _ _ _ - __ . _ _ . _ _ _ _ _ _ _ __ - _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1
~
PHOENIX is a' depletable, two-dimensional (2-D) , multigroup, y discrete' ordinates, transport theory code. A 25 energy group
}L
- nuclear data. library based on a modified version of the British" WIMS(7) library-is'used with PHOENIX.
A. study was done to examine fuel reactivity as a. function of time following discharge from the reactor. Fission product decay was.
. accounted for using CINDER (8), CINDER-is a point-depletion computer code used to determine fission product. activities. The' fission products'were permitted to decay'for 30. years:after discharge. The fuel'reactiv'ity was found to' reach a maximum at approximately.100 hours after-discharge. At this point in time, the major fission product poison, Xe135, has nearly completely decayed away. Furthermore, the fuel reactivity was found to decrease continuously from 100 hours to 30 yearsLfollowing
() discharge. Therefore, the most reactive point in time for a fuel y assembly after discharge from the reactor is conservatively approximated by neglecting the Xe135, I, In the Region II calculational method, the PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a reactor. In addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX. Comparisons between v neasured and' predicted uranium and plutonium isotopic fuel compositions are shown in table 3.1-2. The measurements were made on fuel discharged from Yankee Core 5(9). The data in table (y 5 l I; O TMFWOO57 3.1-9 h _m__ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ . _ . _ _ _ . _ _ _ _ _ _ _ . _ _ I
3.1-2 shows that.the agreement between PHOENIX predictions and s measured isotopic compositions is good. . k(') \ The agreement between reactivities computed with PHOENIX and the results of 81 critical benchmark experiments is summarized in-table 3.1-3. Key parameters describing each of the 81 experiments are given in table 3.1-4. These reactivity comparisons again show good agreement between experiment and PHOENIX calculations. A study was completed which evaluated the impact of the fuel assembly axial burnup distribution on the fuel' rack reactivity calculations. For these calculations the most skewed burnup distribution was obtained from 3-D calculations. The results show that the effect of non-uniform axial burnup distribution is statistically' insignificant and can be ignored in spent fuel storage pool calculations. Since the burnup history is not known exactly for fuel assemblies which will be discharged in the future, the fuel assembly isotopic (Pu, U, etc) content and distribution has some uncertainty. The reactivity bias of 0.01 Ak at 30,000 MWD /mtu is included to account for the content and' distribution uncertainty of the fuel assembly isotopic content which are a function of the fuel assembly burnup history. This bias increasus as a function of the discharge burnup and is included in the SONG 2 burnup credit curve as an increase in the required discharge Enrnup. TMFWOO57 3.1-10
An uncertainty associated with the burnup-dependent reactivities (). computed with PHOENIX is accounted for in the development of the Region II burnup requirements. A bias of 0.01 Ak at 30,000 MWD /mtu is considered to be very conservative since comparison between PHOENIX results and the Yankee Core experiments and 81 l benchmark experiments indicates closer agreement. 1 3.1.3.1 Criticality Analysis for Reaion I The spent fuel storage racks are described in subsection 4.1.2. I The minimum B10 loading in the Boraflex sheets is 0.026 gm B 10/ cm 2, Uncertainties and biases due to mechanical tolerances during ' construction will have an effect on the reactivity. The most () important effect on reactivity of the mechanical tolerances is the possible reduction in the water gap between the Boraflex sheets of adjacent cells. The worst combination of mechanical tolerances is that which results in the maximum reduction in the water gap. The analysis, for the effect of mechanical tolerances assumed a worst case of a rack with the minimum water gap between the storage cells. The reactivity increase of this configuration, in addition to asymmetric positioning of fuel assemblies, is included in the worst case Keff of the rack. For normal operation and using the method in the above sections, the maximum Keff for the Region I rack is determined in the following manner. A TMFWOO57 3.1-11
f; -8 .
~
N' [.
. w i'
Kworst.'+ Beethod .+;B p art + [(ksworst)2
.K eff
' ' =-
ff Q .. L + : (ksmethod)231/2 - L-
'where::
E L 'Kworst. =L worst. case KENO-IV'Keff' includes' fuel assembly positioning,;Boraflex-edge deterioration.off0.125-
' inches-and 3% shrinkage, and mechanical and material' l
tolerances which result'in spacing between assemblies' less than nominal. Bmethod
- method bias determined from benchmark critical comparisons.
() Bp art- = bias to account'for poison particle self-shielding. ksworst = .95/95 uncertainty-in the worst case KENO-IV Kerr. ksmethod = 95/95 uncertainty in the method bias. Substituting calculated values in~the order listed above, the result is:
, Keff =
0.9094 + 0.0083 + 0.0011 + [(0.0048)2 + (0.0018)2 31/2
= 0.9239 O TMFii0057 3.1-12
= _ _ - - _ _ _ - _ .
The value of Kegg for Region I from this analysis is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. Therefore, the acceptance criterion for i criticality is met. 3.1.3.2 Criticality Analysis for Reaion II The nominal and maximum Keff for storage of spent fuel in Region II is determined using the methods described for Region I in addition to the methods described for Region II. The actual l conditions for this determination are defined by the zero burnup intercept points in figures 3.1-3 and 3.1-4. The KENO-IV computer code is used to calculate the storage rack multiplication factor with an equivalent fresh fuel enrichment of 1.85 w/o for the C-E 16 x 16 fuel assembly and 2.40 w/o for the E h 14 x 14 fuel assembly. Combinations of fuel enrichment and discharge burnup yielding the same rack multiplication factor as at the zero burnup intercept are determined with PHOENIX. The spent fuel racks are described in subsection 4.1.2. The minimum B10 loading in the Boraflex sheets is 0.016 gm B 10/cm2 . The C-E fuel assembly has an equivalent " fresh fuel" enrichment of 1.85 w/o U235 and the H fuel assembly has an equivalent " fresh fuel" enrichment of 2.40 w/o U235, O TMFWOO57 3.1-13
l ! l The maximum Keff under normal conditions was determined with a
/7 " worst case" KENO-IV model, in the same manner as for the Region l U I storage racks. For the Region II racks, the water gaps are reduced from the nominal value to their minimum value. Thus, the .
" worst case" KENO-IV model of the Region II storage racks contains minimum water gaps with considerations for asymmetrically placed fuel assemblies. The uncertainty l j
ase,ociated with the reactivity equivalence methodology was included in the development of the burnup requirements. j l Based on the analysis described above, the following equation is used to develop the maximum Keff for the storage of spent fuel in the SONGS Unit 2 or 3 Region II spent fuel storage racks: Keff = Kworst + Bmethod + Bp art + [(ksworst)2 + (ksmethod)2 3g 1 where: Kworst = worst case KENO-IV Keff that includes fuel assembly positioning, poison edge deterioration of 0.125 inches and 3% shrinkage, material tolerances, and mechanical tolerance which can result in spacing between assemblies less than nominal Bmethod = method bias determined from benchmark critical comparisons Bp art = bias to account for poison partical self-shielding TMFWOO57 3.1-14
l J L. ksworst = 95/95 uncertainty in the worst case KENO-IV Keff ksmethod.= 95/95 uncertainty in-the method bias'. Substituting calculated values in the order listed above, the
~
I result ~for C-E 16 x 16 fuel is: Keff.= 0.9335 + 0.0083 + 0.0018 + [(0.0027)2 + (0.0018)2 31/2
= 0.9468.
Using the same equation as above to develop the maximum Keff for the storage'of H 14 x 14 spent fuel in the SONGS 2&3 Region II-spent fuel storage racks and substituting calculated values, the result is: O Keff = 0.9168 + 0.0083 + 0.0018 + [(0.0042)2 + (0.0018)231/2
= 0.9315 The maximum.Keff for Region II for either fuel's configuration is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. Therefore, the acceptance criteria for criticality'are met for storage of spent fuel at an equivalent " fresh fuel" enrichment of 1.85 w/o U235 with C-E fuel and 2.40 w/o U235 for H fuel.
O TMFWOO57 3.1-15 '
3.1.3.3 Criticality Analysis for Checkerboard or Alternating 1 Row Loadina for Reaion II New or burned fuel which does not meet the enrichment versus the discharge burnup criteria of figures 3.1-3 and 3.1-4 may be stored in Region II in a checkerboard pattern or alternating row. pattern if certain conditions are met (figure 3.1-5). I i To simplify the following discussion, the following definitions apply. Fuel Type 1 is defined to be new or burned fuel assemblies which would normally be stored in Region I and does not meet the enrichment versus discharge burnup criteria of figures 3.1-3 and 3.1-4. Fuel Type 2 is defined to be fuel which ) would normally be stored in Region II because it meets the criteria of figures 3.1-3 and 3.1-4. O Fuel Type 1 assemblies may be stored in Region II if: A. Initial enrichment 5 4.1 w/o U235, B. Checkerboard or alternating row pattern is used. C. One completely empty row of storage boxes seperates Fuel Type 1 assemblies from: l
- 1. Fuel Type 2 assemblies stored in Region II, and,
- 2. Fuel Type 1 assemblies stored in Region I.
i O TMFWOO57 3.1-16
't ' , , .
The' maximum'Keff for the above conditions was determined with SCEs NRC' approved NITAWL/ KENO-IV/s methodology (18), 1The analysis method is described here. The Keff.= 0.9468'
-(paragraph 3.1.3.2) for storage of 1.85 w/o fresh 16 x 16 C-E fuel assemblies (including allfassumptions and uncertainties)-is-the base case Kegf to/from whichak's shallibe added/ subtracted.
The Ak's for a checkerboard or alternating ; row pattern were generated as follows. First storage of 1.85 w/o fresh 16 x: 16 C-E fuel was modeled as was done in paragraph.3.1.3.2 above to genera'te a Base Case-2 Keff. Next the checkerboard / alternating row-patterns were modeled to' generate Keff's. Then, Ak' = Keff(checkerboard / alternate row) - Keff (Base Case 2)
' Calculational and model uncertainties in generating these Ak's cancel out.
The maximum Keff for storage of Fuel Type 1 in the Region II racks is: Keff = Keff (Base Case) + Ak
= 0.9468 + ddc
.The results of the adc analyses are:
/
HO TMFWOO37 3.1-17 l L _ -_-__-
_ _ _ _ _ _ _ _ _ _ _ ~ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Keff(Base Case 2) = 0.93780-Keff(checkerboard) = 0.93582 Keff(alternating row) = 0.93617 Therefore, , 1 Ak(checkerboard) = -0.00198 Ak(alternating row)' = -0.00163 : Finally, for storage of Fuel Type 1 in Region II'in a checkerboard pattern: Keff = 0.9468 - 0.00198 = 0.94482 For storage of Fuel Type 1 in en alternating row pattern: O Keff = 0.9468 - 0.00163 = 0.94517 The value of Keff from these analyses-is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. Therefore, the acceptance criterion for criticality is met. l TMFWOO57 3.1-18
3.1.3.4 Fu'l e Reconstitution Station in Reaion II A AKeff analysis similar to that described in paragraph 3.1.3.3 above was done to confirm that a fuel reconstitution station for Fuel Type 1 could be established in Region II. The reconstitution station consists of a repeating pattern (see figure 3.1-6) (at least eight rows separate repetitions) of: A. Completely empty row, j B.. Row of Fuel Type 1 in every other storage box with empty boxes between, and C. Completely empty row. The results of the Ak analyses are: Keff (Base Case 2) = 0.93780 Keff (Recon Station) = 0.92490 Therefore, Ak (Recon Station) = -0.0129 Finally, for the fuel reconstitution station in Region II: Keff = 0.9468 - 0.0129 = 0.9339 O TMFWOO57 3.1-19
h i The value'of Kegg from these analyses isoless than.0.95, ( h . including all uncertainties at 95/95 probability / confidence' level.. Therefore, the acceptance criterion 'for criticality :iis met. 3.1.4- ' ACCEPTANCE CRITERIA AND METHODOIDGY FOR CRITICALITY The acceptance criterion for criticality isLthat the neutron multiplication. factor in the SFP shall.'be less1than or equal to 0.95 including all uncertainties under-all conditions. The analytical methods employed herein conform.with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurizer Water Reactor Plants,"'Section 5.7, Fuel Handling () . System'; ANSI.57.2-1983, " Design Objectives for LWR Spent Fuel-Storage Facilities at Nuclear Power' Stations," Section 6.4.2; ANSI N16.9-1975, " Validation.of Calculational Methods for Nuclear Criticality Safety," NRC Standard Review Plan, Section 9.1.2,
" Spent Fuel Storage"; Land the NRC guidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling
-Applications".
l TMFWOO57 3.1-20 { l
i! l)l{ l ; lI 88556065850933140290000596701 2112222222212222221222221223418902 O ' f f 00000000000000000000000000000000000000000000000000600000000
+iiTT7++*iTTiiiii7TiiiTTiiTiT+i11 133 e
I 766415931873242855109587820188239 509195883264983498223014225426461 899989899999888888999999999999099 999999999999999999999999999999099 000000000000000000000000000000100 ep-e l p b u0 074000000347525t77402000000000000 36 97 411199289327 l1 1 152 314163 o-S8
)
5 4 ( S T N mmmmemmmn E ll uuuuuuuuu M I eennnnnnnnn eeiiiiiiiii R g tt mmmmmmmmm 1 E al ssuuuuuuuuu ll lllllll ssssss
- P l a ti sssss ssaaaaaaaaa ss ssssss _
1 X ar nnnnn eeddddddddd aaaaaa _
. E re iiiii llll ll gggggg 3 , at pa rrrppppprrnntttttttttll eeeeeeeee llilil L eM eee eeiiaaaaaaaaa aaaasa e A S tttCCCCCtt aaa44444 aatt aarrrrrrrrrrrrrrreeceee oooooooaoiiiiiil l C w w w 88888 w w s sbbbbbbbbb a a a a a a p p linlll pp p b I O a T T I R
C r n nnn K o i iii R t c rrrrrrrrrrrrrrrrrrrrr eeeeeeeeeeeeeeeeeeeeeeeeeeeeaeaaa f f fff fff A e ttttttttttttttttttttt rrrrrrrrrrrr M H l f e aaaaaaaaaaaaaaaaaaaaaaoaaaaaaaaaa w w w w w w w w w w w w w w w w w w w w wbtbbbbb pb pp p C R N E B t n 5_ e 3, m2 666666666666666666666 hU 444444444444444444444222222222222 c 222222222222222222222333333333333 io 999999999999 r/ E nw ssssssssssss rrrrrrrrrrrr n eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee cccccccccccccccccccccdddddddddddd iiiiiiiiii o iiiiiiiiiii nannnnnnnnnn tttttttttttttttLtttttiliiiiiiiiii li at tttttttttttttttttttttllllllllllll aaaaaaaaaaaaaaaaaaaaayyyyyyyyyyyy rp ei llllllllll!llllllllll cccccccccccc nr dddddddddddddddddddddllllllllllll ' ec Gs ooooeeooeoooeeoooooooaaaaaaaaaaaa e rrrrrrrrrrrrrrrrrrrrrtttttttttttt eeeeeeeeeee D mmmmmmmmmmm O O O U U U UUUUUUUUUUUU 123456789012345678901234561890123 O 111111111122222222223333 d
< 's4 s
.? -
(_- ~ Table 3.1-2(9) COMPARISON OF PHOENIX ISOTOPIC PREDICTION TO YANKEE CORE 5 MEASUREMENTS Quantity (Atom Ratio) % Difference' U235/U -0.67
.U236/U -0.28 U238/U -0.03 PU239/U. +3.27
~PU240/U. +3.63 PU241/U -7.01 PU242/U -0.20 PU239/U238' +3.24 MASS (PU/U) +1.41 FISS-PU/ TOT-PU -0.02 O
. Percent difference is average differencu of ten comparisons for each isotope.
t, O TMFWOO57 3.1-22
'1
; )
e l ' Table 3.1-3 BENCHMARK CRITICAL EXPERIMENTS PHOENIX COMPARISON I Description of Number of PHOENIX Keff Using Experiments Experiments Experiment Bucklinas UO2 Al clad 14 0.9947 SS clad 19 0.9944 Borated H 2 O 7 0.9940 Subtotal 40 0.9944 U-Metal Al clad 41 1.0012 O TOTAL 81 0.9978 ms p I
.)
i
. TMFWOO57 3.1-23
__..a__. . _ . . . _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _
f 4 V Table 3.1-4 j l i DATA FOR U METAL AND UO2 CRITICAL EXPERIMENTS I (Sheet-1 of 2) j Fuel Pellet Clad Clad Lattice , Case Cell A/O H20/U Density Diameter Material 00 Thickness Pitch Soren' i Number Type U-235 Ratio' (G/CC) (CM) Clad (CM) (CM) (CM) PPM
-1 Hexa 1.328 3.02 7.53 1.5265 Aluminum 1.6916 .07110 2.2050 0.0 2 Hexa 1.328 3.95 7.53 1.5265 Aluminum 1.6916 .07110 2.3590 0.0 3 Hexa 1.328 4.95 7.53 1.5265 Aluminum 1.6916 .07110 2.5120 0.0 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1506 .07110 1.5580 0.0 5 Hexa 1.328 4.89 7.52 .9855 Aluminum 1.1506 .07110 1.6520 0.0 6 Hexa 1.328 2.88 10.53 .9728 Aluminum 1.1506 .07110 1.5580 0.0 7 Hexa 1.328 3.58 10.53 .9728 Aluminum 1.1506 .07110 1.6520 0.0 8 Hexa 1.328 4.53 10.53 .9728 Aluminum 1.1506 .07110 1.8060 0.0 9 Sousee 2.734 2.18 10.18 .7620 55-304 .8594 .04085 1.0287 0.0 10 Square 2.734 2.92 10.18 7620 55-304 .8594 .04085 1.1049 0.0 11 Savaro 2.734 3.86 10.18 .7620 55-304 .8594 .04085 1.1938 0.0 12 Sausre 2.734 7.02 10.18 .7620 55-304 .8594 .04085 1.4554 0.0 13 Souare 2.734 8.49 10.18 .7620 55-304 .8594 .04085 1.5621 0.0 14 Square 2.734 10.38 10.18 .7620 55-304 .8594 .04085 1.6891 0.0 15 Square 2.734' 2.50 10.18 .7620 55-304 .8594 .04085 1.0617 0.0 16 Square 2.734 4.51 10.18 .7620 55-304 .8594 .04085 1.2522 0.0 17 Squaro 3.745 2.50 10.27 .7544 55-304 .8600 .04060 1.0617 0.0 18 Square 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 0.0 A 19 Sausre 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 0.0
/ 20 Savare 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 456.0
( 21 22 Square 3.745 Souare 3.745 4.51 4.51 10.37 10.37
.7544
.7544 55-304 55-304
.8600 .04060
.8600 .04060 1.2522 1.2522 709.0-1260.0 23 Savare 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 1334.0 24 Souare 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 1477.0 25 Souare 4.069 2.55 9.46 1.1278 55-304 1.2090 .04060 1.5113 0.0 26 Scuare 4.069 2.55 9.46 1.1278 55-304 1.2090 .04060 1.5113 3392.0 27 Square 4.069 2.14 9.46 1.1278 55-304 1.2090 .04060 1.4500 0.0 28 Souare 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 0.0 29 Square 3.037 2 64
. 9.28 1.1268 55-304 1.1701 .07163 1.5550 0.0 30 Square 3.037 8.16 9.28 1.1268 55-304 1.2701 .07163 2.1980 0.0 31 Square 4.069 2.59 9.45 1.1268 55-304 1.2701 .07163 1.5550 0.0 32 Souare 4.069 3.53 9.45 1.1268 55-304 1.2701 .07163 1.6840 0.0 33 Sousee 4.069 8.02 9.45 1.1268 55-304 1.2701 .07163 2.1980 0.0 j 34 Souare 4.069 9.90 9.45 1.1268 55-304 1.2701 07153 2.3f.10 0.0 35 Square 2.490 2.64 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 1677.0 36 Hexa 2.097 2.06 10.38 1.5240 Aluminum 1.6916 .07112 .2.1737 0.0 37 Hexa 2;006 3.09 10.38 1.5240 Aluminum 1.6916 .07112 2 AOS2 0.0 38 %xa 2.096 4.12 10.38 1.5240 a l um inurr. 1.$916 07112 2.4 462 0.0 39 Hexa 2.096 6.14 10.38 1.5240 Aluminum 1.6916 .07112 2.3031 0.0 40 Haua 2.095 8.20 10.38 1.5240 Aluminum 1.6916 .07112 3.3755 0.0 41 Hexa 1.307 1.01 18.90 1.5240 Aluminum 1.651f .07112 2.1?42 0.0 42 ' Hexa 1.307 1.51 18.90 1.5240 Aluminum 1.691G .07112 2.4 034 0.0 43 Hexa 1.307 2.02 18.90 1.5240 Aluminum 1.6915 .07112 2.f162 0.0 i
'Os i
l G TMFWOO57 3.1-24
q b Table 3.1-4 DATA FOR U METAL AND UO2 CRITICAL EXPERIMENTS (Sheet 2 of 2) Fuel Pellet Ctad Clad Lattice Case ( ;11 A/O H20/U Density Diameter Material DD Thickness Pitch Boron Number (ype U-235 matto. (G/CC) (CM) Clad (CM) (CM) (CM) PPM-
'44 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 45 Hexa 1.307- 4.02 18.90 1.5240 Aluminum. 1.6916 .07112 3.3249 0.0 46 Hexa 1.160- 1.01- 18.90- 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 47 Hexa 1.160 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 48 Hexa 1.160 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 49 Howa 1.160 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896- 0.0 50 Hexa 1.160 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 51 Hexa .1.040 1.01 18.90 1.5240 Aluminum 1.6916 07112 2.1742 0.0 52 Hexa 1.040 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 53- Hexa 1.040 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 54 Hexa 1.040 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 55 Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916 .07112- 3.3249 0. 0 -
56 Hexa 1.307 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 57 Hexa 1.307 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 58 Hexa 1.307 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 59 Hexa 1.307 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 60 Hexa 1.307 4.02 18.90 9830 Aluminum 1.1506 .07112 2.1742 0.0 61 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5E26 0.0. 62 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 [m 63 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 ( $4 Hexa 1.160 4.02 16.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0
\ 65 Hexa 1.160 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 66 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 67 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 -
68 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 69 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 70 Hexa 1.040 1.33 18.90 19.050 Aluminum 2.0574 .07520 2.8687 0.0 71 Hexa 1.040 1.58 18.90 19.050 Aluminum' 2.0574 .07620 3.0086 0.0 72 Hexa 1.040 1.83 18.90 19.050 Aluminum 2.0574 .07620 3.1425 0.0 73 Hexa 1.040 2.33 18.90 19.050 Aluminum 2.0574 .07620 3.3942 0.0 74 Hexa 1.040 2.33 18.90 19.050 Aluminum 2.0574 .07620 3.6284 0.0
. 75 Here t.040 3.83 18.90 19.050 A1ui/inum 2.0574 .07620 4.0566 0.0 76 Hexa 1.310 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.6160 0.0 77 Nexa 1.310 3.01 18.88 1.5240 Aluminum 1.6916 .07112 2.9900 0.0 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.6160 0.0 79 Hexa 1.159 3.01 18.85 1.5240 Aluminum 1.6916 .07112 2.9900 0.0-80 Hess 1 712 2.03 18.St .9830 Aluminum 1.1506 .07112 1.7250 0.0 81 Hean 1.312 3.02 18.88 .9830 Aluminum 1.1506 .07112 1.9610 0.0
( ( TMFWOO57 3.1-25
H O _ 10.40 Ref. Center to Center 8.640 Ref. Square
= cell Inside 4 Diameter I
'I l
-memmumamm -
.110 Stock ~
8 Call Thickness ~ l- -l I , I e g
. 020 Stock.
Wrapper ' Thickness ener p Boreflex , i ' i _ m 10.40 Ref. l.110 Stock *
- Center I to
*127 -
c,yggy_ indiumuullEEEEEE- f6N Center Depth , , I I e e l ,I
. _ _ _ _ +, . _ _ _ _ _ _ _
_____.+.___._ ,
,I I I I l I
- 1.0
- Cell to cell (water) gap Min (Typical)
SAN ONOFRE i NUCLEAR GENERATING STATION Units 2 & 3 REGION I O- CELL LAYOUT { FIGURE 3.1-1
.I
8.630.Ref. Square O i
=
Diameter Typical cell Inside
. I I
-.110 Stock l, l
= Cell e Thickness 1i 8.630 Ref. g Non-cell Inside ~ i Dian ter Typical 4 . _._._. ...
I I Boraflex l o -.
I . _ _
'.020 Stock l .090 Ref. l 8.85 Ref.
Wrapper Cavity
' Center .
Thickness Depth l to
' "t*#
-Wrapper I Spacing
_ _._ _ _ _ _ _ .j_ _ _ _ _ _ __ ._ 4 O i. I s I, I 8.85 Ref. Cene.er to Center Spacing
~
SAN ONOFRE NUCLEAR GEfJERATING STATION Units 2 & 3 R E GI O N !! O CELL LAYOUT FIGURE 3.12
40 , , , , , , , , tO I *I 'l I l- 1 I I j U , _ _! _ _ .. . .J _ _ _ _ .L _ _ _ _ l. . . _ _ I_ _ . ..l_. ._J__ I- 1 I I 1 l l [__L_. l'
- 35 I I i i I I I / .i I i- 1 I I I e l ,
I I 'l 1- l l l
.__i__.. . .l _ _ __4__ __p.. __i__.._q._ . a__ __p_.
1 I I I I I I i i l' I I I I / l- 1 3 I I I I I I I
'l I
E 'l *I I l i I s ,_ _I_ _ ..l__ __,_ __r_. __,__. . y__ ._3__ _ _ , . O l i I I I I I S 25 i ACCEPTABLE l i i i / i- i- i 1
.l. I i 1 .I I I i .1 1 i 1 I l-
_,_ _ __,__ _ _ 7_ _ _ . , . _ . .
, _ _ .. 7-- -
7-- - r-- co i 1. 1 l' / I I I i .
.g i i t t Y t t t' t .
p 1- 1- 1- 1 i- 1 I l
- 1 I i 1 I I 5 '- i--
,1 -- -
T-- - r-- -r- "-i-- --r-- r-- U 'l 1 i l I I i 1 c gg . . . ; , l i l i l l I i [ i I I g i 1 I 1 O LJ E
- 10 l
I i I I I I I I NOTACCEPTABLEl I I 8 I I I l 1 I I l 1 i
. _ _I _ _ ,..J _ _ .L _ _ __L.. _ _ l_ . .. . _I _ _ . . .! _ _ __L_.
1 I I I l- 1 I I I I I l- 1 1 I 5 , . . , . . . . I I I I I l- 1 I
,_ _I_ _ .l_ _ _ .L _ _ __L.. _ _ I_ _ ,..l.. . 1__ __L_.
I I I I I 'I I i i I i 1 1 1 I I 3.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 U-235ENRICMIENT(t/0)
.4 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 UNITS 2 & 3 FUEL MINIMUM BURNUP VS INITIAL.
ENRICHMENT FOR REGION 11 RACKS FIGURE 3.1-3
36
, s i i a i e i gg i i i l i i I
( ) i I I I I I
%./ '
- I g g
! I g ; , 7-~ , I I I i l l L t i ! I t 39 / i i l i l i I I I I I I I I l
. . ). _ _ _ .i _ .. _ _ i. . .l . . . . l. -- .
l. i I I I I I I I I E 25 ACCEPTABLEl ,I I I I 3 i , , , , 5
- I l I I I I I I o I I I I I I r-~ ~~7---'~- r-- -~ ~ 1 - - ' - - - --
- - T *- - - -
l- - -'
- p. I I I I I I I f20 ' ' ' ' ' ' '
g i i i I I I I co 1 1 1 1 I I I w , _ l. . .J .. l. J .._L_ . 1_ ___f.
., M 1 'l i I I I g 1 l i I I I l v 15 . ; ; ; l .
T I i i [ l l I I O I i i 1 i i I 3 r-- --1--- -r -
--i-- -- r-- --r-- --
l--- g i l i I ' ' i i i 1 NOTACCEPTABLEl I 10 g
- I i I I I i 1 L_ J_ .L. _J__,. _ f_ _ _ 1_ . l_
1 I I I I I I
~
I I I I I l I 5 8 I I I I I I l l I i 1 1
$_ r ,.p ,__,.,,,
- - t. . .. .
_ _ , 3_, 3 I I l l l 1 I I I i 1 I I q 4.0 2.5 3.0 3.5 4.0 4.5 5.0 5 "* U-235 ENRICINEN1 (t/0) l SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3
/
l UNIT 1 FUEL MINIMUM BURNUP VS INITIAL ENRICHMENT FOR REGION 11 RACKS FIGURE 3.14
CHECKERBOARD ALTERNATING ROW l cou m V a X X X X X X X X X<X X X X X X X X X X X X X X X ! l X X X X X X X X X X X X
% X X X X X X _
X X X X X X O O O 7 O O' O O O O
.e O O O ?
O O O O O O O O O O O O O O O O O O O O O O V E A N i, A N X Type 2 Fuel O Type 1 Fuel SAN ONOFRE NUCLEAR GENERATING STATION
- Units 2 & 3 0 - - - - ematv stor oe 'ocatioa rue' ero ^ae PATTERNS FOR REGION ll RACKS FIGURE 3.1 5
. CT ~
X X X X X X X X N X r X X r X X X TYPE 2 FUEL X X X X X X X X 1 O O O O RECON STATION
~
X X X X X X X X X X X X 'X X X X X X X X X X X X X X X X X X X X TYPE 2 FUEL X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X _ O O O O RECON STATION X X X X X X X X X X X X X X X
'O X X X X X X X
X X X- X X X X X X X X X X X X X X X TYPE 2 FUEL X X X X X X X X X X X X X X X X X X X X X X X X _ O O O O RECON STATION l 0 O O O O O O O TYPE 1 FUEL O O O O y ARRAY IN REGION !! l x l= TYPE 2 FUEL SAN ONOFRE l0 l = TYPE 1 FUEL NUCLEAR GENERATING STATION l Units 2 & 3 l l = EMPTY FUEL STORAGE PATTERNS FOR REGION ll RACKS RECONSTITUTION STATION FIGURE 3.1 -6
3.2 DECAY HEAT CALCULATIONS FOR THE SPENT FUEL POOL (BULK) O 3.2.1 SPENT FUEL POOL COOLING SYSTEM DESIGN Figure 3.2-1 illustrates the SFP cooling system outside the SFP. Figure 3.2-2 shows the pool cooling piping in the pool. Table 3.2-1 contains a summary of system operating parameters. The system is controlled manually from the main control board. Control room alarms for high fuel pool temperature, high and low liquid level in the fuel pool, and low fuel pool pump discharge pressure are provided to alert the operator to abnormal circumstances. A local alarm for low liquid level in the fuel pool is a3so provided. The SFP cooling system has two trains of cooling pumps, heat exchangers, and related piping. Spent fuel pool water is circulated by the cooling pumps taking suction near () the top of the pool and directing flow through the heat exchangers where the heat is transferred to the non-critical loop of the component cooling water (CCW) system described in UFSAR 9.2.2. From the outlet of the SFP heat exchangers the cooled water is discharged to the bottom of the SFP by a distribution header. The SFP cooling piping system will be modified by removal of the existing sparger lines on the discharge distribution header. All evaluations for the reracking of the SFP are based on thic modification. Normal operation of the SFP cooling system currently occurs with a maximum of 2-2/3 cores stored in the SFP and is proposed with a maximum of 6-1/4 cores stored in the SFP. This allows enough O TMFWOO57 3.2-1
room.for an additional full core offload. .Under normal operating conditions, including shutdown and refueling, one SFP cooling pump _and one SFP heat exchanger are in service to maintain a pool temperature of.140F or less. During a full core offload, two pumps and two heat exchangers are online and maintain a maximum pool temperature of 140F or less. Also, the shutdown cooling system,.shown in figure 3.2-3, may be used for SFP cooling by isolating the' normal pool cooling system and utilizing one low pressure safety injection pump and one shutdown cooling heat exchanger. This system lineup would be used to cool the SFP only during full core. offload. The clarity and purity of the water in the SFP is maintained by a purification system.. The purification system consists of a fuel. pool purification pump, filters, strainers, ion exchanger, and () surface skimmers. The purification flow normally is drawn directly from the fuel pool, except during refueling when a fraction of the purification flow is drawn through the surface skimmers. A strainer is provided in the pump suction purification line to remove debris from the line before the water is pumped through the filter and ion exchanger. Spent resins from the SFP ion exchangers are discharged to the solid radioactive waste system. The purification pump is stopped at 140F. The purification system sucticn piping is being modified by removing the distribution piping at the bottom of the pool. All evaluations for the reracking of the SFP are based on this modification. O TMFWOO57 3.2-2
The possibility of a siphon draining of the pool is precluded
- ("S through the design of the system and the use of administrative
- i. V controls. Leakage from the fuel-pool cooling system is detected I by a reduction in pool inventory and a leak sump high level alarm. Makeup to the SFP is from the Seismic Category I refueling water storage tank.
Alternate makeup water sources are also available from the nuclear service water, primary plant demineralized water via the reactor coolant.or boric acid recycle subsystem, and refueling water via the low pressure safety injection or containment spray or SFP cooling pumps. Makeup from the refueling water tank via the low pressure safety injection or containment spray pumps is available only if the entire reactor core is unloaded. O 3.2.2 DECAY HEAT ANALYSIS 3.2.2.1 Basis The SONGS 2E3 rsauto's r are rated at 3390 MW thermal. Each core contains 217 fuel assemblies. Thus, the average operating power per fuel assembly, Po, is 15.6 MW. l Unit i fuel will also be stored in the Unit 2 and 3 fuel pools. Unit 1 has 157 fuel assemblies and is rated at 1347 MW thermal. This corresponds to a Po of 8.6 MW/ assembly. O' TMFWOO57 3.2-3 l
The spent fuel decay heat loads are determined for two cases, based on the conditions delineated in the NRCs OT Position for Review and Acceptance of Spent Fuel Storage'and Handling Application, dated April 14, 1978(10), o Maximum normal heat load during normal (1/2 core) refueling e Maximum abnormal heat load during full core offload 3.2.2.2 Model Descrio,tLQD The NUREG-0800 Branch Technical Position (BTP) ASB 9-2(11),
" Residual Decay Energy for Light Water Reactors for Long Term
() Cooling" is utilized to compute the heat dissipation requirements in the pool. With the long term uncertainty factor, K, as specified in Standard Review Plan .(SRP) Section 9.1.3 (12) , the operating power, Po, is conservatively assumed to equal the rated power. This provides a margin of safety because the reactor may have operated at less than its rated power during much of the irradiation period for the fuel assemblies in the latest refueling batch. The computations and results reported here are based on the discharge (full core and half core offload) taking place when the inventory of fuel in the pool will be at its maximum, resulting in an upper bound on the decay heat rate. O TMFWOO57 3.2-4
i ' 'A review of the BTP ASB 9-2 notes that-the NRC methodology and assumptions are-valid up to 1.0 x 107 seconds (116 days). j l However, per SRP 9.1.3, for long term cooling in excess of 1.0 x i 107 seconds the decay heat calculation can be based on the BTP methodology when an uncertainty factor of 10% is used. -This uncertainty fact'or is used in the calculation for decay heat L ' loads beyond-1.0 x 107 seconds. The. normal offload scheme is as defined in' table'3.2-2. As required by SRP 9.1.3 III.1.h.i and 11(12), the normal' offload' L scheme includes the modeling of one refueling load after 150 hours decay plus one refueling load after 1-year decay. Additional decayed refueling loads sufficient to fill the'SFP to capacity are also modeled. The modeling of these additional refueling loads conservatively addresses the SRP 9.1.3 ( requirement that one refueling load after 400 days decay should also be' considered. The full core offload scheme is as defined in table 3.2-3. As required by SRP 9.1.3 III.1.h.iii(12), the full core offload scheme includes the modeling of one full core after 150 hours decay plus one refueling load after 36 days decay. Additional decayed refueling loads sufficient to fill the SFP to capacity are also modeled. The modeling of these additional refueling loads conservatively addresses the SRP 9.1.3 III.1.h.iv(12) 1 requirement that one refueling load after 400 days decay is also considered. O TMFWOO57 3.2-5
i r
- It should be noted that SRP 9.1.3 requires each refueling load shall be at equilibrium conditions. Branch Technical Position ASB 9-2 (Reference.11, Section B1) suggests ~ utilizing an operating history of 16,000 hours. The SONGS refueling cycles j i
vary and are factored into the SFP decay heat calculation as. such. The Unit 1 cycles are 500 affective full power days (EFPD) long and 52 spent fuel assemblies are discharged. The Unit 2 I cycle lengths and quantities discharged are (Unit 3 is similar): i End of Refueling Assemblies Calculated Operating Cvele Date Discharaed EFPD Time (Hours)' 1 Nov 1984 72 370 26,763 2 Jan 1986 88 270 15,979 3 Aug 1987- 108 402 19,385 From cycle 4 on, SCE has conservatively assumed 108 assemblies discharged (1/2 core).at 570 EFPD. Additionally, in the SRP 9.1.3 full core offload scheme assuming the full core placed in the SFP and 150 hours of decay, 50% of the' refueling batch would have been irradiated for no'more than 714 hours. (36 days) (24 hour / day) - (150 hours decay) = 714 hours Therefore, the required use of equilibrium conditions for this one batch is highly conservative. O TMFWOO57 3.2-6 ! l 4 _____m__ _ _ . _ ___ _ _ _ __ _ _ __ _ _ _ _ _ _ _ _ . _ _ . . _ .
n 3.2.2.3- Bulk Pool Temperature Results
,.O
' Tables 3.2-4 and 3.2-5 present the heat. rate per refueling batch EW and cumulative heat load in the SFP. The results are summarized below:
Normal. Refueling 25 MBTU/h
-(1/2 core. offload) 1355 fuel assemblies Full' Core Offload 51 MBTU/h 1572 fuel assemblies
. The existing SFP cooling system is adequate to maintain the SFP.
temperature below.140F for the normal refueling heat loads,
-assuming a single active failure required by SRP 9.1.3'III.1.d.-
A single active failure-for this system would be the failure of one cooling pump. Also in accordance with SRP 9.1.3 III.1.d, a single active failure need not be considered for the full. core offload. For both heat exchangers and pumps in operation for a full core offload, the SFP temperature will'be 156F assuming a conservative heat load of 53.4.MBTU/h. Therefore existing SFP cooling system is adequate to maintain pool temperature below boiling for full core offload heat loads. fb O TMFWOO57 3.2-7 I
- l. .
3.2.2.4 Snent Fuel Pool' Heat Load and'Coolina System Summary The spent fuel decay heat evaluations were-performed'in- l
-accordance with the method provided in the NRC BTP ASB'9-2(11)..
The evaluations conform to the following NRC guidance: y e The'"0T Position" which specifies that calculations 1 for the amount of thermal energy that.will.have to be removed by the SFP cooling system shall"be made in accordance with BTP APCSB 9-2(11) (now ASB 9-2). , e The BTP ASB 9-2(11) provides acceptable assumptions and' formulations.which may be used to calculate the residual decay energy release rate for. light-water-cooled reactors for long term cooling of the-reactor
.n O facility.
The original design specification for the SFP heat exchangers is described in table 3.2-6. It should be noted also that the heat. exchange. rate is based on a tube side inlet. temperature of 150F for conservatism.
.The maximum normal heat load for the SFP with the proposed new racks is 25 MBTU/h which occurs in the year 2002 (tables 3.2-2 and 3.2-3). This heat load can be transferred by the SFP cooling system to the CCW system using one pump (2200 gal / min) and two heat exchangers (1100 gal / min each) in order to maintain a pool temperature of 140F or less.
LO TMFWOO57 3.2-8
I; p. i i-
\
r'N .The actual flow rate with one pump and two heat exchangers
~
exceeds 2200 gal / min, and thus satisfies the heat exchange requirements. For this case, a CCW inlet temperature of 95F is
]
used.
- i l, The maximum abnormal heat load for the SFP with the' proposed new j racks is 51 MBTU/h'(25.5 MBTU/h per heat exchanger) which occurs in the year 2001-(tables 3.2-3 and 3.2-5). The calculated design temperature for CCW during a full core offload is 88F. For a conservative decay heat load of 53.4 MBTU/h and a CCW temperature of 88F,.the SFP bulk temperature will be 156F. This analysis bounds the decay heat load of 51 MBTU/h. Therefore, the SFP temperature can be maintained well below boiling with a full core offload using both cooling trains.
O For the maximum abnormal heat load (full core offload) the temperature of the pool water should be kept below boiling. The temperature is kept below boiling -(156F) as stated above. No single active failure needs to be considered for this abnormal case. Therefore, the existing SFP cooling system is adequate for the full core offload heat loads. , i O TMFWOO57 3.2-9
3.2.2.4.1 Spent Fuel Pool Cooling Alternative O The SFP cooling system is designed, procured and constructed to the requirements of Seismic Category I and Quality Class II. The SFP cooling system heat exchangers are cooled by the non-critical loop of the CCW system. The makeup systems are described in subsection 3.2.3. In the unlikely event of a loss of all normal cooling capability of the SFP cooling system, approximately 7.8 hours would elapse before the pool temperature would rise from 140F to 212F with a the maximum normal heat load of 25 MBTU/h. This allows time in which to effect repairs and restore cooling. Without cooling or makeup, with normal heat loads, the fuel () assemblies will be uncovered in 67 hours, assuming the Technical Specification water level of 23 feet above the fuel assemblies. This allows time to provide adequate makeup water to the pool. An alternate method of cooling during full core offload uses the shutdown cooling system, as described in the UFSAR, paragraph 9.1.3.3. Using this alternative, and a heat load of 51 MBTU/h, the maximum pool temperature would be 146F, assuming a CCW inlet design temperature of 95F. O TMFWOO57 3.2-10
l In'the unlikely event of a loss of SFP cooling system, y approximately 3.8 hours would elapse before the pool temperature
\
would' rise from 140F to 212F with the maximum abnormal heat load of 51 MBTU/h. This allows time to initiate the alternate cooling l 1 method dercribed in the following paragraph. 1 i l If available, the shutdown cooling system can be used as backup J cooling when the full core is removed from the reactor vessel. This is done by using the low pressure safety injection (LPSI) pumps to circulate water from the SFP through the shutdown cooling heat exchangers as shown in figure 3.2-3. The shutdown cooling system will be used as a backup to the SFP cooling system only if the full core is removed from the reactor vessel. Two spectacle flanges must be reversed to effect this mode of operation. Without cooling or makeup, with full core offload heat loads, the fuel assemblies will be uncovered in 32 hours, assuming the Technical Specification water level of 23 feet above the fuel assemblies. This allows time to provide adequate makeup water to the pool. TMFWOO57 3.2-11 - - - - _ - _ - - - _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ - -_. i
3.2.3 SPENT FUEL POOL MAKEUP
,r v t .
3.2.3.1' Normal Makeun Normal makeup to the SFP is from the cross-connected Seismic Category I refueling water storage tanks. The normal makeup path' is from the refueling water storage tanks via the SFP makeup pump to the pool (see figure 3.2-1). The SFP makeup capacity of 150 gal / min exceeds normal system leakage and evaporative losses. Maximum evaporative loss would be 105 gal / min in the unlikely event of bulk boiling with a full core offload heat load of 51 MBTU/h. The normal makeup system is the backup makeup system per the UFSAR table 3.2-1. 1 3.2.3.2 Additional Makeun
.r
( The following alternate sources are also available for additional makeup: A. Refueling water via the SFP cooling pumps. This system is the primary makeup system per the UFSAR table 3.2-1, and this system is not operable if the SFP cooling pumps are inoperable. B. Refueling water via the LPSI or containment spray pumps (only if the entire reactor core is unloaded) (spectacle flange changeover required). TMFWOO57 3.2-12
- l. ..
6 C. Primary plant demineralized water via the coolant or boric acid recycle subsystem. V[T D. Nuclear service water from any of eight hose connections near the pool. 3.2.3.3 Makeuo Water Ouantities The following tanks are available for makeup to the SFP: Tank Unit (al Nams capacity (callons) T005 2,3 Refueling Water 245,000
.T006 2,3 Refueling Water 245,000
, T1044 Common. Nuclear Service Water 25,000 T055 2 Primary Plant Makeup. Water 300,000
() T056 3 Primary Plant Makeup Water 300,000 The tanks listed are'available for both Units 2 and 3. The nuclear service water tank, which is common to both units, and the primary plant makeup water tanks are cross-tied between the~ units. Unit 3 has similar water supplies available. The capacities of these makeup water supplies will allow adequate time to take corrective actions. O TMFWOO57 3.2-13 L_____________________ _ _ . _ _
3.2.4' PURIFICATION SYSTEM EVALUATION
-q: ,
Q-- The evaluation ~of.the SFP' cooling system concludesLthat for normal refueling, the temperature of the SFP will remain below 140F. The SFP purification system is designed to operate below 140F. When the temperature is at or above 140F, the SFP purification system does not operate (see SONGS 2&3 UFSAR, paragraph 9.1.3.2). I' 3.2.5 COMPONENT COOLING WATER AND SALTWATER COOLING SYSTEMS EVALUATIONS Analyses were performed for the component cooling water (CCW) and saltwater cooling (SWC) systems to verify systems performance () with the' increased heat load associated with the SFP reracking
-project on SONGS 2&3. The normal offload and full core offload
' heat loads were considered. The results of these analyses indicate that the existing CCW and SWC systems are adequate to remove the increased heat load to the ultimate heat sink. No l hardware or system changes are necessary due to the SFP reracking.
i
'l O TMFWOO57 3.2-14
3.2.6 FUEL HANDLING BUILDING HVAC O d The existing design of the FHB-heating, ventilating, and air-conditioning (HVAC) system has been evaluated with respect'to the increased heat loads which result from the additional spent fuel storage and the existing HVAC design has been found to be 1 acceptable. The SFP design temperature of 140F is not exceeded during normal refueling heat load conditions; therefore, no design changes need to be made to the normal HVAC subsystem. For this analysis, 160F pool temperature was conservatively used for a full core offload. For this pool temperature, the normal ventilation subsystem has sufficient capacity to meet the original design basis of 104F, and no design changes are necessary. The building normal temperature range will not be exceeded during normal refueling or full core offload heat load conditions, using the normal HVAC subsystem for' ventilation.
.g THFWOO57 3.2-15
Table 3.2-1 PRINCIPAL PARAMETERS OF THE EXISTING FUEL POOL COOLING AND PURIFICATION SYSTEMS Soent Fuel Pool Coolina Puno Quantity 2 Design flow 2,000 gal / min Design head 46 feet Design temperature 200F Fuel Pool Purification Pumo Quantity 1 Design flow 150 gal / min Design Head 212 feet Design temperature 200F Spent Fuel Pool Heat Exchanaer Quantity 2 Tube side flowrate 2,000 gal / min Shell Side flowrate 3,150 gal / min Total design heat load (total 42.2 MBTU/h for two heat exhangers) Heat transfer conductivity (type 304SS) 8.84 BTU /h-ft-F O Tube surface area 1602 ft2 O TMFWOO57 3.2-16
)
j L /~ i I' \ ~" Table'3.2-2~ ANTICIPATED NORMAL OFFICAD SCHEDULE, Refueling Number of Unit 2' Decay Number of Unit 1 Decay Date Assemblies Time (vrs) Assemblies Time (vrs)- Nov 1984 72 17.58 -- - Jan 198G: 88 16.41 - - Aug 1987 .108 14.83 - - Jun 1989 108 13 75 -13.3 Jun 1991- 108 11' O - Jun 1993 108 9 52 9.3
.Jun 1995 108 7 0 -
Jun 1997 ~ 108 5 52 5.3 Jun 1999 108 3 - - Jun 2001 108 1 44 1.3 Jun 2002 108 150 hrs - - O I O TMFWOO57 3.2-17
7_______________-___-_-_-_____
. j ':
i Table 3.2'-3
-POSTUIATED FULL CORE OFFICAD SCHEDULE t
Refueling Number of. Unit 2 Decay Number of Unit 1 Decay Date Assemblies Time (vrs) Assemblies (a) Time (vrs) ' q Nov 1984- -72 16.58 - l Jan 1986 88 15.41 - - Aug 1987 108 13.83 - - Jun 1989 108 12 75 12.3 Jun 1991- 108 10 48 10.3 Jun 1993 108 '8 52 8.3' Jun 1995 108 6 52 6.3 Jun 1997- 108- 4 52- 4.3 Jun.1999 108 2 - - Jun 2001 108 36 days. 52 120' days'
, Jul 2001' 217 150 hrs - -
Unit 1: .331 assemblies stored Unit 2: 1241 assemblies stored 1572 spaces utilized-O ~
- a. To maximize the heat load for the full core offload condition, it is assumed that all Unit'l fuel assemblies are discharged to one unit.-
l TMFWOO57 3.2-18
g Table 3.2-4 ! ANTICIPATED NORMAL REFUELING HEAT LOADS Number of Number of i Refueling Unit 2 Heat Rate Unit 1 Heat Rate Cumulative j Date Assemblies (BTU /h) Assemblies (BTU /h) Heat Load 1984 72 .2.4588E5 -- -- 2.4588E5 1986 88 2.6071E5 --- --
'5.0695E5 1987 108 3.5847E5 -- --
8.6505ES. 1989 108 4.1741E5 75 1.7423E5- 1.4567E6 1991 108 4.3795E5 0 -- 1.8946E6 1993 108 4.5992E5 52 1.3304E5 2.4876E6 1995 108 4.8563E5 0 -- 2.9732E6 1997- 108 5.3012E5 52 1.5126E5 3.6546E6 1999 108 6.9022E5 0 -- 4.3448E6 2001 108 1.7952E6 44 3.2921E5 6.4693E6 2002 108 1.8195E7 -- -. 2.4664E7. O O TMFWOO57 3.2-19 - - _ _ - _ _ - _ _ _ _ - _ _ - l
() Table 3.2-5 POSTUIATED FULL- CORE:- OFFLOAD HEAT LOADS - Number of Number of-Refueling . Unit'2' Unit 1- Heat Rate Cumulative Date Assemblies Heat (b"FJRat
/h) ~ Assemblies ta) (BTU /h) Heat Load 1984 '72 2.4835E5 -- --
2.4835E5 1986 88 2.6441ES -- -- 5.1276E5' 1987 108 3.5994E5 -- -- 8.7270E5 1989 108 4.2756E5 75 1.7844E5' 1.4787E6 1991 108 4.4868E5 48 1.1984ES- 2.0472E6 1993 108 4.7194E5 52 1.3645E5 2.6556E6 1995 :108 5.0322E5 52 1.4477E5' 3.3036E6 1997' 108 5.8061E5 52 1.6247E5 4.0467E6 1999 108 9.5704E5 0 -- 5.0037E6 20011 108 8.6606E6 52 .1.1659E6 1.4830E7-2001 217 3.6426E7 --. --- 5.1256E7
.. a. To maximize the heat load for the full core offload condition, it is assumed that all Unit i fuel assemblies are discharged to one unit.
1 TMFWOO57- 3.2-20
I Table 3.2-6 SPENT FUEL POOL HEAT EXCHANGER DESIGN SPECIFICATION (PER HEAT EXCHANGER)' Original Normal Original Maximum Operation (*)- Refueling (b) . (Third core offload)- (Full core offload) Tube Side Flow: 2,000 gal / min 2,000 gal / min 1 Shell Side Flow 3,150 gal / min 3,150 gal / min Temperatures: Tube Side In. 150F 150F Tube Side Out 138F 129F Shell Side In 95F '95F Shell Side Out 103F 108F Total. Heat Load par Heat Exchanger 12.4 MBTU/h 21.1 MBTU/h
- a. Normal (Original) Operation includes operation of one pump and t one heat exchanger
- b. Maximum Refueling in(one train) cludes.o heat exchangers (two trains)peration of two pumps and two l
l TMFWOO57 3.2-21 E- _ - - - - - -
I
. SFPCOOUNG '
+ PUMP- - -
~ -
5FP HEAT ExcH waEn 3E 1
~
SFP HikT O UN REFUEUNG ' PUMP WATER - STOR E TANK' SPENT FUEL POOL (SFP) O SFP MAKEUP PUMP L 1 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SPENT FUEL POOL COOLING LOOP FIGURE 3.2-1 ; L o _ _ - - _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
y" a r 1 b.N '
. (M)%.
g " 0
'2 2
MM OE RT "6 FS L E Y O- NS O7 ' UG 2L P1 EIN D GL N T UE EL RO O-I NF AO - M
)).
UT M NO E T P T HC CP SF I D S
\G P SOB 0
'2 2
N 4 2, - mDz ozS$ Z p 53 SzmmD j m )sOz _ c=y u * " nO m 3O ommo c$ u
; EF ioo' 8 i
- z ' 3zO ymms 0 u .
1 1 5cmm u'"g llll,1
e .. 5
.R b d
LQ l g <g .g d l d h __ w$- l L1 h ty i O I glI
-l g __
i l i i k SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 PROPOSED UNIT 2 O- SPENT FUEL POOL COOLING PIPING SHEET 2 OF 2 FIGURE 3.2 2 j
6 4
?
e. SPECTACLE BLIND FLANGE - 8 ll 0 LPSI PUMP . 8 SPENT FUEL POOL (SFP) l SPECTACLE gD HEAT' BLIND FLANGE gg i
-1 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SHUTDOWN COOLING FOR SPENT FUEL POOL FIGURE 3.2 3
1. 3.3 THERMAL-HYDRAULIC ANALYSES FOR THE SPENT FUEL POOL []
%.J (IDCALIZED)
The purpose of the thermal-hydraulic analysis is to determine the maximum fuel clad temperatures which may occur as a result of using the spent fuel racks in the SONGS SFP. This evaluation is based on the C-E 16 x 16 fuel assemblies stored in the Units 2 and 3 racks, the results of which bound the case of the H 14 x 14 fuel from Unit 1 as stored in Units 2 and 3 racks (13). J.3.1 BASIS The following bases were used in the rack thermal-hydraulic analysis: () A. The nominal water level is conservatively assumed to be 22 feet above the top of the fuel storage racks, which is 23 feet-1 inch above the top of the fuel assembly. This water level also satisfies Technical Specification 3.9.11, which states "At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks." B. The maximum fuel assembly decay heat output is 51.6 BTU /sec per assembly following 7 days decay after shutdown (11). For conservatism, this value has been used for both Region I and Region II fuel assemblies. TMFWOO57 3.3-1 i _ _ _ _ . _ . . _ . _ _ _ . _ _ _ . _ . _ . _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ a
- c. The peak fuel rods are conservatively assumed to have 60%
,9
( j greater heat output than. average rods. 1 D. For normal operations, the maximum bulk pool temperature shall not e::ceed 140F. For conservatism, the' temperatures of the storage racks and the stored fuel are evaluated 3
)
assuming that the temperature of the water at the inlet to the storage cells is 150F during normal operation. E. Under postulated accident conditions, when no pool cooling systems'are operational, the maximum temperature at the inlet to the cells is assumed to be equal to the i saturation temperature at atmospheric pressure or 212F. 3.3.2 MODEL DESCRIPTION A conservative natural circulation calculation is employed to determine the thermal-hydraulic conditions within the spent fuel storage cells. The model used assumes that all downflow occurs in the peripheral' gap between the pool walls and the outermost storage cells and all lateral flow occurs in the space between the bottom of the racks and the bottom of the pool. The effect of flow area blockage in the region is conservatively accounted for (by modeling the rack supports larger than the actual blockage) and a multichannel formulation is used to determine the variation in axial flow velocities through the various storage cells. The hydraulic resistance of the storage cells and the i TMFWOO57 3.3-2 l
fuel assemblies is conservatively modeled by applying large uncertainty factors to loss coefficients obtained from various [')') sources since the high hydraulic resistance can reduce the circulation flow rate in the cells. The hydraulic resistance used in the computer code (17), is greater than the actual calculated value. Where necessary, the effect of Reynolds Number on the hydraulic resistance is considered, and the variation in momentum and elevation head pressure drops with fluid density is also determined. The solution is obtained by iteratively solving the conservation equations (mass, momentum and energy) for the natural circulation loops. The flow velocities and fluid temperatures that are obtained are then used to determine the fuel cladding temperatures. An elevation view of a typical model is sketched in figure 3.3-1 where the flow paths are indicated by arrows. l( ) Note that each cell shown in that sketch actually corresponds to a row of cells that is located at the same distance from the pool walls. This is more clearly shown in a plan view, figure 3.3-2. As shown in figure 3.3-2, the lateral flow area underneath the storage cells decreases as the distarice from the wall increases. This counteracts the decrease in the total lateral flow that occurs because of flow that branches up and flows into the cells. This is significant because the lateral flow velocity affects both the lateral pressure drop underneath the cells and O. U TMFWOO57 3.3-3
the turning losses that are_ experienced as the' flow branches up into the cells. These effects are considered in the natural. circulation analysis. 1 The most recently discharged'or " hottest" fuel assemblies are assumed to be located in various rows during different l l calculations in order to ensure that they may be placed anywhere within the pool without violating safety limits. In order to j simplify the calculations, each row of the model'must be composed of storage cells having a uniform decay heat level. This decay heat level may or may.not correspond to a specific batch of fuel, but the model.is constructed so that the total heat. input is correct. The " hottest" fuel assemblies are all assumed to be placed in a given row of the model in order to ensure'that conservatively accurate results are obtained for those assemblies. In fact, the most conservative analysis that can be performed is'to assume that all assemblies in the pool (or rows in the model) have the same maximum decay heat rate. This maximizes the total natural circulation flowrate which leads to conservatively large pressure drops in the downcomer and lateral flow regions which reduces the driving pressure drop across the-limiting storage locations. Table 3.3-1 presents the normal-operation results of the local rack thermal hydraulic analysis. Since the natural circulation velocity strongly affects the temperature rise'of the water and the heat transfer coefficient within a storage cell, the hydraulic resistance experienced by the flow is a significant parameter in the evaluation. In order O TMFWOO57 3.3-4
to minimize the resistance, the design of the inlet region of the f (3,) racks has been chosen to maximize this flow area. Each storage cell has one or more flow openings as shown in figure 3.3-3. The use of these large or multiple flow holes virtually eliminates the possibility that all flow into the inlet of a given cell can be blocked by debris or other foreign material that may get into the pool. In order to determine the impact of a partial blockage on the thermal-hydraulic conditions in the cells, an analysis is also performed for various postulated blockages (20, 40, 60, and 80% of all cells). Table 3.3-2 presents the results of this flow blockage analysis. 3.3.3 CLADDING TEMPERATURE The maximum cladding temperatures of the various conditions were calculated. They are 215.7F for normal operation, 269.8F for the postulated accident, and 233.lF for the postulated 80% flow blockage conditions. The postulated accident case is that when all SFP cooling is assumed to be inoperative, the water level ~is 22 feet above the top of the racks, and the water temperature at the surface is 212F. All calculated temperatures are lower than the maximum allowable fuel rod temperature criterion of 700F. O TMFWOO57 3.3-5
l l Table 3.3-1 I ' NORMAL OPERATION RESULTS' . LOCAL RACK THERMAL-HYDRAULIC ANALYSIS-
'c -
Temperature.of' Maximum
-Water at Inlet Maximum ~-
. Clad Surface Maximum
'to storaga cell- Cell Water Temperature Clad surface "
Temperature- Temperature (Typical Rod) (Peak' Rod)-
' 1'2'O ' 154.0 175.9 189.1 1 140 -171.7 192.9' 205.7 150 181.8 203.0 215.7 i
f All temperatures in degrees Fahrenheit. O O TMFWOO57 3.3-6
(... 4 / E
- 1. .
l3 ' t .p u '-
) i Table 3.3-2 2: .;
FIDW BLOCKAGE ANALYSIS RESULTS l
- l..
Maximum Cell Maximum Clad Surface Percent. Water Temperature Temperature' (Peak) Blockage (degrees Fahrenheit) (degrees Fahrenheit).
'O .181.8 215.7'
- 20. 182.1 216.1 2 '40 182.8 217.0 60- 185.1 220.1' 3- ,
80 195.2 233.1 Temperature of Water at, Inlet t to Storage Cell = 150F
\
l l O TMFWOO57 3.3-7
WATER SURFACE = 1 ( l F - ) A - WATER HEAD ABOVE RACKS . _ P euT
< , , , o < , , , , ,
7 I w / ~ ~h .,. , o -x o -> 3-,- o f l ! ROW
/
= 2 3 4 5 1 6 7 3 s#
o o A o o o o o a E n 8 h ~ 8 f'f'f'f'f'f'f'ff'1 - 1 % I s HEIGHT ABOVE FLOOR LATERAL FLOW REGION {_ g SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SPENT FUEL POOL NATURAL CIRCULATION MODEL (ELEVATION VIEW) { FIGURE 3.3-1
; ee ln .
I e
,8 .. eI[
i
- e , e < e -
. ~
~ 2OE
~
w .
; ~.
~
~
~
. ~
.. . -. ~
~
. . ~
~
~
~
. ~
~
. . , ^ ~
. .. ~
- R ~ V. .
- n. .:
O ~
. . . O ~
. L ~
F ~
. ~
.. E' ~
V ~
- . .... y O ~
~
* . B
- ... A ~
~
- ~
- T ~
H ~ G I
~
~
E ~ s.: O. . . :.
. H I
A
~
~
~
- E R
~
A . ~~ ~
. W ~
. . p: O L
. ~~ ~
F ~
... .. : L A
g ~~
- . . . R
~
~
.. : . R E
T a~- ~ ~
- p. . . E M
A L ~
. ..: : O ,
e~ ~ C ~
; N ~
W = O a ~~ x....:
,. *. D ~
. . .: o ~~ ~
...V . . '
~
e ~
. ~
2Og ~
. ~ .
. ~
- p. . . * . .
oOEzoOgE
~
~
~ nnzo3Qm
.. : . . * . . ., . . . , * . .*. . . * . ~
oOOr I>~j
... * ..~
.. ~
~
t .
. ... . .. .'.*. . * ,,A. y* .~ ,.
.. ..s>.
l
. * .**~~ . ~
M i m>Z o2O zm T 2 Corm >n : omZm3DH zo mH>HhZ C2amIEW
= mzH vCmr u oOr z 4C2>I i
o 2OCrg5z goomF _ur>Z < mT*
. Oc$ u.Y" l1llll.1:,l,ll;!
1 1 Reinforcement (Lifting) Block
] Assembly 3
~m
\+ l l \ l 1 l
l \ l l sh @w W# 3@ . i Rack I - b!ak[on ' s [kM O N %W rh O d/>s $ I
%?a Qf' VQ L
)
Base Plot Flow Areas Shaded SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 & 3 SPENT FUEL RACK INLET C FLOW AREA (PLAN VIEW) FIGURE 3.3-3 1201C/0309C
W k - 3.4. POTENTIAL FUEL AND RACK HANDLING ACCIDENTS L The method for moving the racks into and out of the SFP is discussed in paragraph 4.7.4.2. The methods and procedures utilized ensure that postulated. accidents do not result in a loss- , of cooling to either the SFP or the reactor, or result in a Kagg in the SFP pool exceeding 0.95. 1 L 3.4.1 RACK MISHANDLING i l During the reracking operation, it will be necessary to raise and I I maneuver the old racks out of the SFP in order to install the new spent fuel racks (see subsection 4.7.4). At no time will a rack containing any fuel (new or spent) be moved. The handling of these heavy loads will be accomplished by the use of a temporary construction crane and the cask handling crane. Both of these 1 cranes meet the intent of the design and operational requirements of Section 5.1.1 of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plant"(14). i
-)
The potential for mishandling of a rack during the reracking , i operation has been evaluated. If, during rack removal / L installation a rack is dropped, an inspection will be performed to determine whether there is any damage and identify any corrective action as necessary. At no time will the cask l handling crane'or the temporary construction crane carry a rack j l I 1 O TMFWOO57 3.4-1
.. J
directly over unprotected new or spent fuel. Procedures and r administrative controls governing the reracking operation will be r C]/ prepared and will ensure the safe handling of racks. Although spent fuel assemblies will be temporarily stored in the cask pool during the installation process, there will be a cover over the cask pool of sufficient strength and design to preclude an accidental drop onto the rack containing fuel. No heavy loads will be lifted over the cask pool area with fuel in the racks without the cover in place (see paragraph 4.7.4.5.). 3.4.2 TEMPORARY CONSTRUCTION CRANE DROP During the reracking operation, a temporary construction crane ( will be installed in the FHB as described in paragraph 4.7.4.3 of this report. If practical, the temporary crane will be installed as a single piece assembly. The installation will be done over the cask pool area with spent fuel and cover in place. If this is not feasible, the assembly of the temporary crane in the fuel buildings will be achieved such that no piece of the temporary crane will be carried over unprotected spent fuel during crane assembly. The cask pool cover will be designed for the temporary crane installation load drops (paragraph 4.7.4.5). Administrative controls will be used to ensure that the temporary crane and the cask crane are not operated over the cask pool l I TMFWOO57 3.4-2 l i 1
cover at the same time except during erection or disassembly of r' the temporary gantry crane. These administrative controls will preclude impact between the two cranes. l 3.4.3 LOSS OF POOL COOLING (STORAGE RACK DROP) ] Paragraph 4.7.4.4 of this report evaluates the consequences of a storage rack drop on the SFP liner. ! 3.
4.4 CONCLUSION
S During construction activities associated with the installation of high density spent fuel racks, procedures will be followed to () ensure that fuel and rack handling accidents do not occur. the administrative controls that are a part of the SCE heavy With loads control program, accidents that might affect criticality or SFP cooling will be precluded. The evaluation of other drops and their structural and radiological implications are discussed in subsection 4.6.5 and section 5.3 respectively. l O TMFWOO57 3.4-3 __________z________.____ _ _ _ _ _ _ _ _ .. )
3.5L TECHNICAL SPECIFICATION CHANGES O The proposed Technical Specification (15,16) changes are described below: A. Technical' Specification 5.6.l(b) will change the current 12.75-inch center-to-center rack storage, location spacing to 10.40-inch center-to-center spacing for Region I, and 8.85-inch center-to-center spacing for Region II. ! E. Existing Technical Specification 5.6.2 for dry storage of the first core in.the fuel pool in alternate rows and columns will be deleted. This. Technical Specification was only applicable for.the first core and the pool is filled with water which is maintained at'a' minimum level as () prescribed by Technical Specification 3.9.11, Water Level
- Storage Pool.
C. New Technical Specification 5.6.2.and accompanying figures 5.6-1, 5.6-2, 5.6-3, and 5.6-4 will define the fuel enrichment /burnup limits for storage of Units 1, 2, and 3 fuel in Region II of the high capacity spent fuel storage racks. This new Technical Specification will also define the conditions and storage patterns (checkerboard or alternating row) for which new or burned fuel, which does O TMFWOO57 3.5-1 l
f. i not meet the enrichment vs.-burnup criteria for unrestricted storage in Region II, may be stored in Region II. i i Lastly,,this new Technical. Specification.will define the conditions'(empty - alternating cells - empty) under which i a new/ burned' fuel reconstitution. station may be-h established in~ Region II. D. . Technical: Specification 5.6.4 will be revised to designate that no more than 1572 fuel assemblies may be stored inD the spent' fuel racks which is an increase of 772 from the-current limit ofE800. E. . Technical Specification 3.9.7 will be revised to list the following allowable lifts of heavy loads above stored (
. spent fuel:
- 1. Spent fuel pool gates shall not be carried at a height greater than 30 inches (elevation 36. feet 4-inches) over the fuel racks.
- 2. Testlequipment skid (4500 pounds) shall not be carried at a height greater than 72 inches (elevation ~39 feet 10 inches) over rack cells which contain Unit 2 or 3
. fuel assemblies or greater-than 30 feet 8 inches (elevation-64 feet 6 inches) over rack cells which contain Unit 1 fuel assemblies.
O TMFWOO57 3.5-2
r- 3. Installation or removal of the cask pool cover'over the cask' pool with fuel-in the cask pool.
- 4. The. lift of construction loads including the temporary gantry crane and the existing.and new fuel storage racks (including lifting equipment and rigging), above the cask pool with the cask pool cover in place and fuel in the cask pool. This includes temporary storage of these construction loads on the cask pool.
cover during construction. F. The basis for Specification.3.9.7 will'be revised to reflect the analysis for the heavy load drops associated with the revised Specification 3.9.7. O
<%3- TMFWOO57 3.5-3
s 3.6' REFERENCES O 1. . W.LE.' Ford,.III, CSRL-V, " Processed ENDF/B-V 227-Neutron-Group and'Pointwise Cross-Section Libraries ~for Criticality Safety, Reactor and Shielding Studies,"- ORNL/CSD/TM-160 (June'1982). . i i 2.. N. M.' George, etfal,~"AMPX: A Modular Code System:for; Generating Coupled Multi-group Neutron-Gamma Libraries from' ENDF/B," ORNL/TM-3706 (March 1976).
- 3. L. M. Petrie and N. F. Cross, " KENO'IV--An Improved. Monte Carlo Criticality Program,"~ORNL-4938 (November 1975).
- 4. M..N; Baldwin, " Critical Experiments Supporting Close
() Proximity Water Storage of Power Reactor Fuel," BAW-1484-7, July 1979.
- 5. J. T. Thomas, " Critical Three-Dimensional Arrays of U (93.2)
-- Metal Cylinders," Nuclear Science and Engineering, Volume 52, pages.350-359 (1973).
- 6. Harris, A. J., et al, "A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors," WACAP; 10106, June 1982.
)
l O TMFWOO57 3.6-1
1 f i t 7; Askew, J. R., Fayers,:F. J., and Kemshell, P. B., "A General Lt. Description of the Lattice Code WIMS," Journal of British' Nuclear Energy Society, 5, pages 564-584, 1966. ll 8.: England, T. R., " CINDER'- A One-Point-Depletion'and Fission Product Program," WAPD-TM-334, August 1962. 9 .' Melehan,-J. B., " Yankee' Core Evaluation ~ Program Final Report,"~WCAP-3017-6094, January 1971.
'r 10 . - Nuclear Regulatory Commission, Letter to All Power ReactorL
' Licensees,-from B. K. Grimes, April 14, 1978,c"OT Position for-Review and Acceptance of Spent Fuel Storage and Handling Applications," as amended by the NRC letter dated January 18,-_1979.
O 11. Nuclear Regulatory Commission,- Branch Technical Position ASB i
.l
'9-2, " Residual Decay' Energy for Light Water Reactors'for 1 i
.Long-Term Cooling",-Revision 2,. July 1981.
)
- 12. Nuclear Regulatory Commission, NUREG 0800, Standard Review l Plan 9.1.3, " Spent Fuel Pool Cooling and Cleanup System,"
Revision 1, July 1981. 1
)
- ]
I O TMFWOO57 3.6-2 1
- 13. San Onofre Nuclear Generating Station _ Unit 2 Facility Operating License NPF-10, Amendment 63, dated June 22, 1988, Docket No. 50-361; and San Onofre Nuclear Generating Station Unit 3 Facility Operating License NPF-15, Amendment 52, dated. June 22, 1988, Docket No. 50-362.
- 14. Nuclear Regulatory Commission, " Control of Heavy Loads at' Nuclear Power Plants", NUREG 0612, July 980.
- 15. San Onofre Nuclear Generating Station Unit 2 Facility Operating License NPF-10, Docket No.~50-361.
I
- 16. San Onofre Nuclear Generating Station Unit 3 Facility Operating License NPF-15, Docket No. 50-362.
- 17. " TRAM" - Thermal-Hydraulic Rack Analysis Model; Description and Verification, by J. C. Buker, WNEP-3530, May 1985.
- 18. NRC Safety Evaluation Report for Amendment No. 39 to Facility Operating License NPF-10 and Amendment No. 28 to Facility Operating License NPF-15 for SONGS 2&3 [ Enclosure 3 to December 2, 1985 Letter from George W. Knighton (NRC) to Kenneth P. Baskin (SCE) - Docket Nos. 50-361'and 50-362.
,[
l TMFWOO57 3.6-3 I
- 4. MECHANICAL. MATERIAL. AND STRUCTURAL CONSIDERATIONS
4.1 DESCRIPTION
OF STRUCTURE 4.
1.1 DESCRIPTION
OF FUEL HANDLING BUILDING The FHB is a conventional reinforced concrete structure containing the new and spent fuel handling, storage, and shipment facilities, fuel pool water cooling equipment, and decontamination area. The overall plan dimension of the structure is approximately 134 X 86 feet, with a maximum height of 110 feet. The structure is of heavy shear wall construction with a concrete-slab, steel-frame, composite-action roof system. Partial soil embedment of about 20 feet is present on three sides of the structure with no embedment on the fourth side. Typical e plans and sections of the FHB are shown in figures 3.8-34 through 3.8-37 of the UFSAR. The FHB has been designed as a Seismic Category I structure in accordance with the criteria, applicable codes, standards and specifications outlined in paragraph 3.8.4.2 of the UFSAR. The building exterior walls, floors, and interior partitions are designed to provide plant personnel with the necessary radiation shielding and protect the equipment inside from the effects of adverse environmental conditions including tornado, winds, temperature, external missiles, flooding, and earthquakes. TMFWOO57 4.1-1
i The SFPs.are 4- to 5-feet thick reinforced concrete walls, lined 1 (~s with stainless steel plates. The SFP is centrally located between the transfer pool and the cask handling pool. A stainless steel double liner plate system is installed along the interior surface of the SFP to ensure the leaktight integrity o f the pool, even in the presence of minor inherent concrete cracking of the SFP walls. The leak chase system is installed below the stainless steel liner plate system. l 4.
1.2 DESCRIPTION
OF NEW SPENT FUEL RACKS 4.1.2.1 Desian of New Soent Fuel Racks 4.1.2.1.1 Region I Racks The two Region I storage racks are composed of 12 x 13 individual storage cells made of stainless steel. See table 4.1-1 for additional rack data. These racks utilize a neutron absorbing i material, Boraflex, which is attached to each cell. The cells within a rack are interconnected by grid assemblies and stiffener clips to form an integral structure as shown in figures 3.1-1 and 4.1-1. Each rack is provided with multiple leveling pads (26 pads for each Region I rack) which contact the SFP floor or pool floor plates and are remotely adjustable from above, through the cells, at installation. The racks are neither anchored to the l l l i O TMFWOO57 4.1-2 I i
floor nor braced to'the pool walls. They also are not connected to each other. Also, the pool-floor bridge plates are not O', , attached to the pool floor. Figure 4.1-2 illustrates the basic. sections of the fuel rack assembly. They are the storage cell, the neutron absorber material, the Boraflex wrapper, the top and bottom grid assemblies, the base plate, and the leveling pad assembly. ' All . rack components are made from Type 304LN stainless steel unless otherwise noted (see' subsection 4.7.1). i i 4.1.2.1.1.1 Storace cell. The storage cell is fabricated from one sheet of 0.110-inch thick stainless steel. welded together at one corner. The cell material is stretcher-leveled to maintain ; dimensional stability and flatness. Each cell has an inner square dimension of 8.64 inches and is 189.50 inches long. The top end of each cell wall is angled outward to provide lead-in surfaces for fuel assembly insertion. 4.1.2.1.1.2 Neutron Absorber Material (Boraflex). Boraflex (neutron absorbing) consists of fine boron carbide particles distributed in a polymeric silicone encapsulant. Boraflex material is held in place on the side of the cell by the Boraflex wrapper (paragraph 4.1.2.1.1.3). Its length and width are designed to allow for both shrinkage and edge deterioration and still meet criticality requirements. Some cells have Boraflex on all four sides, some on three sides, and some on two sides. Boraflex is not required on Region I cell walls facing pool TMFWOO57 4.1-3
n-walls. Therefore, cells located in the interior of a Region I ! rack have Boraflex on all four sides. Periphery cells facing the j pool walls have Boraflex on the three sides not facing the wall. Periphery cells. facing adjacent racks have Boraflex on all four sides. Rack corner' cells which face two pool walls have Boraflex only on the two remaining sides.- Those corner cells adjacent to l another-rack and the pool wall require Boraflex on three sides. Corner cells adjacent to other racks in both directions have i Boraflex on all fourisides. 4.1.2.1.1.3 Boraflex Wranoer. The Boraflex wrapper positions the Boraflex on the side.of the cell and holds it in place (see figure 4.1-2). The wrapper is attached to the outside of the cell by spot welding the entire length of the wrapper via its side flanges. Manufacturing experience has shown that spot welding in this manner does not distort the wrapper such that the contained Boraflex is affected. The wrapper also~provides for venting of the Boraflex to the pool environment. The lateral clearance that the Boraflex has between the cell wall and the wrapper is designed to prevent pinching or binding of the Boraflex. This design precludes sagging or buckling inside the wrapper at any time during fabrication or in operation. 4.1.2.1.1.4 Grid Assemblies. The upper and lower grid 1 assemblies maintain the centerline to centerline spacing between , 1 the cell and provide the structural connections between the cells ! to form a fuel rack assembly. The lower grid assembly consists of box-beam members and side plates and is welded to'the base O k/ j TMFWOO57 4.1-4
plate. The bottom of the cell assembly is welded 1/2-inch off 7-~ the base plate to the lower grid assembly (giving a total fuel i A storage container length of 190.00 inches). The upper grid assembly consists of box-beam members and side plates. The cell assembly is welded to the upper grid assembly approximately 31 inches down from the top of the cell. 4.1.2.1.1.5 Stiffener Clios. Besides the grid assemblies 1 (paragraph 4.1.2.1.1.4), the Region I cell assemblies are also interconnected at their corners by use of cell-to-cell stiffener clips (see figure 3.1-1). The clips are of the same thickness (0.110-inch) as the cell material and are welded between adjacent cells at 10 locations along the length of each cell between the two grid assemblies. The purpose of the clips is to add structural stiffness to the racks. O 4.1.2.1.1.6 Base Plate. The base plate is a 1/2-inch thick plate with 4-inch diameter holes centered at each storage cell to allow coolant flow. At support locations additional flow holes are provided (see figure 3.3-3). The fuel assembly sits on the I base plate when it is stored (see figure 4.1-2). 4.1.2.1.1.7 Levelina Pad Assembly. The major components of the leveling pad assembly are the support block, the leveling pad, and the leveling screw (see figure 4.1-2). The support block is a 3-inch thick block welded to the bottom of the base plate. It ; has a threaded hole that centers on a base plate hole. The leveling screw rests inside the leveling pad and provides swivel O TMFWOO57 4.1-5 l
capability of the assembly in case of unlevel pool floor f"N conditions. The leveling screw threads into the support block b and the leveling pad sits on the pool floor or floor plates. In this manner, they transmit rack loads to the pool floor and l provide a sliding contact. The leveling screws are remotely adjustable from above, through the cells, base plate, and support ; I blocks. This permits the leveling adjustment of the rack. i 4.1.2.1.2 Region II Design The six Region II storage racks consist of stainless steel cells assembled in a checkerboard pattern, producing a honeycomb type structure as shown in figure 4.1-3. See table 4.1-1 for additional rack data. In other words, cells are located in every () other location and are welded together at the cell corners. results in "non-cell" storage locations, each one formed by one This outside wall of four checkerboard cells. Each cell is of the same basic design as described for Region I; i.e., the major components are the cell, the Boraflex (neutron absorbing) material, and the wrapper. The cells are welded to a base support assembly and to one another to form an integral structure without use of grids as used in Region I racks. This design is also provided with multiple leveling pads (33 leveling pads for each rack) which contact the SFP floor or pool floor plates (same type as in Region I) and are remotely adjustable from above, through the cells, at installation. The racks are neither O TMFWOO57 4.1-6
anchored to the floor nor braced to the pool walls or each f other. Also, the pool floor plates are not attached to the pool ^! floor. Figures 4.1-4 and 4.1-5 illustrate the basic sections of the Region II fuel rack assembly. They are the storage cell, the cover plate, the neutron absorber material (Boraflex), the Boraflex wrapper, the base plate, and the leveling pad assembly. All rack components are made from Type 304LN stainless steel unless otherwise noted (see subsection 4.7.1). 4.1.2.1.2.1 Storace Cell. The storage cell is faLTicated from one sheet of 0.110-inch thick stainless steel welded together at one corner. The cell material is stretcher-leveled to maintain dimensional stability and flatness. Each cell has an inner square dimension of 8.63 inches and is 190.00 inches long. (~}
%d 4.1.2.1.2.2 Cover Plate. The checkerboard cell configuration of a Region II rack assembly results in open "non-cell" locations around the periphery of the rack during rack fabrication. These final "non-cell" storage locations are formed with the cover plates. Each cover plate is one sheet of 0.110-inch thick stainless steel (same as the cells). Each cover plate is 190.00 inches long and is sized to close off the periphery rack openings. It is then welded to the adjacent cells (see figure 4.1-5). \
N- TMFWOO57 4.1-7
4.1.2.1.2.3- Neutron Absorber Material (Boraflex). Boraflex consists of fine boron carbide particles distributed in'a O polymeric silicone encapsulant. Boraflex material is held'in
. place on the side of the cell by'the Boraflex wrapper (paragraph
-4.1.2.1.2.4). Its length and width are designed to allow for both shrinkage and edge deterioration and still meet criticality requirements. Some cells have Boraflex!on all four sides, some on three sides, and some on two sides.- Boraflex is not required-on peripheral Region II cell walls. Therefore, cells located in the interior of a Region II-rack have Boraflex on all four s' ides. Periphery side cells have Boraflex on the three' rack interior sides. Rack corner cells.have Boraflex only on the two rack interior sides.
4.1.2.1.2.4 Boraflex Wranner. The Boraflex wrapper positions ( the Boraflex on the side of the cell and holds it in place (see figure 4.1-4). The wrapper is attached to the outside of the cell by spot welding the entire length of the wrapper via its side flanges. Manufacturing experience has shown that spot welding in this manner does not distort the wrapper such that the contained Boraflex'is affected.- The wrapper also provides for l venting of the Boraflex to the pool environment. The lateral l clearance that the Boraflex has between the cell wall and the wrapper is designed to prevent pinching or binding of the Boraflex. This design precludes sagging or buckling inside the wrapper at any time during fabrication or in operation. ! O TMFWOO57 4.1-8 l'
l 4.1.2.1.2.5 Base Plate. The base plate is a 1/2-inch thick plate with 4-inch diameter holes centered at each storage cell to allow coolant flow. Additional flow holes are provided at support locations (see figure 4.1-5). The fuel assembly sits on the base plate when it is stored (see figure 4.1-4). 4.1.2.1.2.6 Levelina Pad Assembly. The major components of the leveling pad assembly are the support block, the leveling pad, and the leveling screw (see figure 4.1-4). The support block is a 3-inch thick block welded to the bottom of the base plate. It has a threaded hole that centers on a base plate hole. The leveling screw rests inside the leveling pad and provides swivel capability of the assembly in case of unlevel pool floor conditions. The leveling screw threads into the support block and the leveling pad sits on the pool floor or floor plates. In this manner, they transmit rack loads to the pool floor and provide a sliding contact. The leveling screws are remotely adjustable from above, through the cells, base plate, and support blocks. This permits the leveling adjustment of the rack. 4.1.2.2 Fuel Handlina Southern California Edison fuel movement procedures require multiple verifications of any fuel movement plan. These procedures will be expanded to include certification of burnup as follows. At least 1 week prior to any movement of fuel from the reactor to Region I or transfer to Region II, fuel assembly TMFWOO57 4.1-9
burnups will be estimated with the best available methodology. (' This includes CECOR for Units 2 and 3, and INCORE3/ SCENIC for Unit 1 fuel. These best estimates of fuel assembly burnups will be available for all bundles which are to be moved and will prevent transfer to. Region II of any bundle which does not meet the burnup criteria. L, I 1 I
.1 O TMFWOO57 4.1-10 1'
l l
t' ,, i
,. *9 s y :
L ' Table 4.1-1
)
RACK DATA * (Each Unit) , Reaion I- Realon II Number of Storage 312 1260 Locations 1 Numbur of Rack Two 12 x 13 Six 14 x 15 : Arrays Center-to-Center 10.40 - 8.85 ' Spacing (inches) Cell Inside Width 8.64 8.63 j (inches) i i C-E 16 x 16 and/or O Type of T.:a1 C-E 16 x 16 and/or; H 14 x 14 E 14 x 14 Rack Assembly Outline ; Dimensions (inches) 126 x 136'x 196 125 x 134 x 196 1 Dry Weights (1bs) , Per Rack Assembly 52,103 35,286 l 1 I I I TMFWOO57 4.1-11 j
-r
,f -s ,s O ,fs
-s
- s
[ [',['
-s
- s
-s
-s
-s -s
-s -s
' ' : ' ; s, -s # s
- -s', ; s
, - % l
- : ', s - -s c<-s s ,,
-s
-s -s -s
-s -s
< -s -s
-s
-s
-s
-s , ,' s-s -s ,s
-s
-s.
< s -s -s -'
-s -s
( , 's -s
-s -s - s
-s -s -s
-s ,
) 5 l s
< -s
-s
-s
, f '- s - s- s ,fs -s,
- s ,
s -s -s, s - 1 s s, a s 3 <s- s - ,' ' , f '- s - - y' fp/ e*af/, i %s\ \, %,s s l i j l l [1 N K 'I ' Nhl l l l ik ! l h5I
, 1 i i
- ; 4
'N W :
I
; i ,
O ! '
, 0 l
1 i: }
" l
[ y ) N(N - i i u 'l-
\(%d -
g *
- - j,y, N h Figure shows 11 x 11 rack array.
SONGS Region I racks are 12 x 13 arrays. Internal support pads are omitted. Boraflex is not required on rack sides that face the pool wall. SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 FUEL STORAGE RACK (REGION 1) FIGURE 4.1-1
10.40 R E F. - CELL l ASSEMBLY r 0 1 j l I l l m !
! i l~ ~ 8.6 4 SOU ARE -
I. D. H l I 1
$55EMBLY ATOP GRID ASSEMBLY
\
r -r -- _ q j mE I'ms. i .. mar ~ -- l % , BORAFLEX . 19 0.00 R E F. l _. BORAFLEX l WRAPPER 3 I l
. L.
1 BOTTOM GR ID l ASSE MBL Y { , I l. l ' N l O i e t Y// ' 0 l _. '//////////A N BASE PL ATE 6.00 R E F. " SUPPORT BL OCK fr y ~ LEVELING PAD 1 h . .
\ "__
POOL LINER _.. .~ _x h h CONCRETE N PLATE Yp[6 SAN ONOFRE NUCLEAR GENERATING STATION NOTE: LINER BRIDGE PLATES Units 2 & 3 m NOT SHOWN REGION I RACK CROSS SECTION FIGURE 4.1-2
1 3 8
,1 l,,1 , . .
"N! j, s
I I i s 4
' !',y, 1 s :/
's! * ' '
i 2 I a 3 a a a a/ f/
, f
'N' ' *
%. i J "Q k s}/
o .% (d 4A
%~% _
U hg% #,d ?.
. ?.
- s. , ,
s', , .
,6 .
a Figure shows 12 x 14 rack array. j SONGS Region II racks are 14 x 15 l arrays. Boraflex is not required on the periphery cell walls. l SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 O REGION 11 FUEL STORAGE RACK i FIGURE 4.1-3
- 8.85 REF ----
l
.C CELL sN -
^- -
i, ASSEMBLY mw_ s~- , Y '
/ '
i POISO N ' M AT ER I AL _
. F ,? f ~~l M i
190.00 REF. / WRAPPER i M I i O O BASE PLATE i I j J
/ '
SUPPORT BLOCX q -- L EVELING P AD A ' r [4 i CONCRETE ' }j,N P00L PLATE LINER LEvrtmG 5.*REW isor to scAtt NOTE: LINER BRIDGE PLATES SAN ONOFRE NOT SHOWN NUCLEAR GENERATING STATION Units 2 & 3 REGION !! RACK Os CROSS SECTION FIGURE 4.1-4
i-l- (~g CELL-TO-CELL WELD . V - 8.63 scun E CELL CPENING
- TYP) CAL 7 -
CELL - - - ( ,
- I j
O O !
"^" "'^' % v j v 1 l
}
'! 8.95
~
RACK SUPPORTS LOCATION N Tr , l i i l O O O- vgu, v C) i
' s
\ l j' i COVER PLATE i
i J SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 ) REGION 11 RACK TOP VIEW FIGURE 4.1-5 ,
4.2 APPLICABLE CODES. STANDARDS. AND SPECIFICATIONS fm t - The design and fabrication of the spent fuel racks and the analysis of the SFP portion of the FHB have been performed in accordance with the applicable portions of the following NRC Regulatory Guides, Standard Review Plan Sections, and published standards: 4.2.1 FUEL HANDLING BUILDING - SPENT FUEL POOL ANALYSIS The analysis methods, criteria, codes and standards described in i subsections 3.8.4 and appendix 3A of the SCNGS 2&3 UFSAR were. used to analyze the FHB. The principal codes include: () ( American Institute of Steeel Construction (AISC), Manual of Steel Construction, 1969 Edition American Concrete Institute (ACI) 318-71, Building Code Requirements for Reinforced Concrete American Welding Society (AWS), AWS D1.1-72, Structural Welding Code TMFWOO57 4.2-1 _ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ . _ _ _ _ _ _ l
4.2.2 SPENT FUEL RACKS --DESIGN AND FABRICATION l L s_ Work related to the design.and fabrication.of the spent fuel racks will be in accordance with the applicable portions of the following codes and standards: ASME Section III Nuclear Power Plant Components 1986 Edition up to and including A-86 Addenda ASME Section III Code Data Reports and Code Symbol Stamping are not required. ASME Section IX Welding and Brazing Qualifications ASME Section V Nondestructive Examination AISC American Institute of Steel Construction Specification, Eighth Edition ANSI /ANS 8.17-84 Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors ANSI N16.1-83 Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors O TMFWOO57 4.2-2
p 1 ANSI N16.9-75 Validation of Calculational Methods'for 7^s7 - i s ,j Nuclear Criticality Safety .! ANSI N18.2-73 Nuclear Safety Criteria for Stationary Pressurized Water: Reactors ANSI N45.2.2-72 Packaging,. Shipping, Receiving, Storage and Handling of Items for Nuclear-Power Plants (During.the Construction Phase) ANSI N210-76 Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations l SSPC SP-10 Near White Metal Blast Clear.ing 10CFR21 Reporting of Defects and Noncompliance 10CFR50- Prevention of Criticality in Fuel Storage Appendix A-GDC and Handling No. 62 10CFR50 Quality Assurance Criteria for Nuclear ; Appendix B Power Plants and Fuel Reprocessing Plants 10CFR100 Reactor Site Criteria TMFWOO57 4.2-3
USNRC RG l'.13, Spent-Fuel Storage Facility Design Basis
,R,evL1, Dec-1975~
USNRC.RG 1.25, Assumptions Used for Evaluating the
~
.. March:1972 Potential Radiological Consequences'of a Fuel Handling Accident in the Fuel Handling and Storage Facility?for Boiling-o >
and Pressurized Water Reactorst r' i
.USNRC RG 1.26, Quality Group Classifications and-Rev'3, Feb 1976- Standards for' Water Steam and Radioactive-
. Waste Containing components of Nuclear.
Power Plant
- USNRC RG 1.29, Seismic Design Classification fReve3,'Sep 1978 USNRC RG 1.31, ' Control of Ferrite Content'in Stainless Rev 3, Apr 1978
, . Steel Weld Metal USNRC RG 1.33, Quality' Assurance Program Requirements Rev 2, Feb 1978 (Operation)
USNRC RG 1.61, Damping Values for Seismic Design of October 1973 Nuclear Power Plants USNRC RG 1.92, Combining Model Responses and Spatial Rev 1,.Feb 1976 Components in Seismic Response Analysis
,t O TMFWOO57 4.2-4 w__-__---__----_- __ .--_- __-- _
-t.
I l [{ USNRC RG 1.124,. Service Limits and Loading Combinations
.Rev 1, Jan 1978 for Class I Linear-Type Component Supports USNRC RG'3.41, Nuclear Criticality Safety in Operations Rev 2, April 1986 with Fissionable Materials at Fuels and Materials Facilities USNRC RG 1.75, Standard Format and Content of Safety Rev 3, Nov 1978 Analysis Reports for Nuclear Power Plants
- LWR Edition
.USNRC SRP'3.7 Seismic Design Rev 1, July 1981 USNRC SRP'3.8.4, Other Category I Structures, Appendix D, Rev 1, July 1981- Technical Position on Spent Fuel Racks USNRC SRP 9.1.2, Spent Fuel Storage Rev 3, July 1981 USNRC SRP 9.1.3, Spent Fuel Pool Cooling System Rev 1, July 1981 USNRC SRP 9.1.5, Overhead Heavy Load Handling System Rev 0, July 1981 TMFWOO57 4.2-5
U USNRC. Branch ASB'9.2' Residual Decay' Energy for Light' BTP ASB 9.2, Water Technical ~ Position-Reactors for Rev 2, July'1981 Long-Term Cooling (Attachment.to SRP 9.2.5, Ultimate Heat' Sink) USNRC Position OT Position for Review and. Acceptance'of
. Paper, . Spent Fuel Storage and Handling April 14, 1978- JApplicat' ions Rev Jan 18, 1979
( NUREG-0612, Control of Heavy Loads at Nuclear Power July 1980 Plants i IE Information Fuel Binding Caused by Fuel. Rack Notice 83-29 Deformation IE Information Gaps in Neutron-Absorbing Material in Notice 87-43 High-Density Spent Fuel Storage Racks. O TMFWOO57 4.2-6
=--__-___-_________--_________
4.3 SEISMIC INPUTS (- NJ The objective of the seismic analysis for the spent fuel racks is to determine the structural responses resulting from the I simultaneous application of three orthogonal seismic excitations. The method of analysis employed is the time history method. Seismic floor response spectra for the SFP floor have been developed using the methods described in subsection 3.7.1 of the SONGS 2&3 UFSAR. The free field ground design input response. spectra and the corresponding time-history is as given in the SONGS 2&3 UFSAR, figures 3.7-1 and 3.7-2, respectively. The parameters of the original lumped mass model of the FHB were adjusted to reflect the increased mass corresponding to the new f i_),/ high density spent fuel storage racks completely filled with Units 2 and 3' spent fuel assemblies. The resulting instructure response spectra at the pool floor are shown in figures 4.3-1 through 4.3-6. These spectra were then used to generate statistically independent 80-second duration time history records, one for each of the three orthogonal directions (figures 4.5-2 through 4.5-7). Each pair of time histories has a correlation coefficient of less than 0.1. The response spectrum from each corresponding time history record is also plotted on the above referenced figures to show their enveloping of the floor response spectra. Since the spent fuel racks have no j i O TMFWOO57 4.3-1 1 _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ J
C "r . I-
' connection with the" pool-walls or with each other, the_ pool floor _
]
-. time histories are_used as: input.to the dynamic analysis of-the.
racks, as-described in paragraph 4.5.2.2.1. 1 O O TMFWOO57 4.3-2
l O I f i I
'f I
l i
' O
/ A
/ v J )/(
T
'f
^
o
)
N
.h Z
m C O m Z Hm m o v-
- J ~m{m
~ \
\
, ~M<
- m s
- mg x
- ja N
- $$I N m
\
c l
\
l
\
l \ m io s at N " O
)G( NOITARELECCA SVN ONOi.W3 N0013VW D3N3WV1INO SIV nu!ts Z 9 C 3
SONDS 3 YNO C d037 801101N ds~OOB HOWiZONIV1 NS )NHC dOW #% QVNidiND 083 Sd301 d100W3 t'C L
d
~
O /^\ L) -
. I I
f I i
/
I .. (
,, 0
/ ^
4 M CL
\ U i >-
O 2 W D I' O
$ w WW Z 0
C,m Z E, w s 4 _mg _w W
\-
~
$ N E!S \
_gs N
#3 \
_l s I \ l \
\
~
w o v n N o* (D) N0llV831300V SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 i SONGS 2 AND 3 FUEL BUILDING POOL \ FLOOR HORIZONTAL EW (NE368 C4) FOR 4% DAMPING DBE SPECTRA FIGURE 4.3-2
i l d O f w
\J i
/
, s i
/
^
' (A I- >
O (N z O Z$d ss N x - t 2$b \
~h \
~$$
- E $g \
CI
-Qw \
l \ l '
\
(D) NOI.I.Vhl373DOV SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3
' SONGS 2 AND 3 FUEL BUILDING POOL FLOOR VERTICAL (NE360 C4)
FOR 4% DAMPING DBE SPECTRA FIGURE 4.3 3
i
~
C$ O f' ,- ' N_)y I l I I I
\
y .
- O
/ e-
.. I ^ l M CL
\ O v
\
\ O Z ;
W \ D O 4 w
-( O
- W9 t/) W s
-sE W x
-g$
-M
_ \
\
_8R C lf)
\
_esW \ E \
- \
l - g I l
\ ),
3 p . . . . ,. (0) NOl1VB373DOV SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 , SONGS 2 AND 3 FUEL BUILDING POOL U) FLOOR HORIZONTAL NS (NE391 C4) { FOR 2% DAMPING OBE SPECTRA l FIGURE 4.3-4 1
Nw
'/ - \
1 s 0 r 1
- )
- S
, P C
(
/ N Y
C E [ - U Q E_ A R R F
- TE s CP -
i EO 1 iSE PL / iE V / i S N / N E iOY g / PR ' i SO ET - RS iRHI OE l O LI M i FT
- - . - - . - . - - - - - - . - ~ - .
' 1 5' 4 3 2 1 o.
= I OV zo_l= xLaJOOC L L
a gzoZOm2m DgEa$> doz zCo m>!7m"yl c. E >W 3 n iOz0m n >5 u mC - sac $ 1 t m 0m 5iO$yI 3 mG zm$= mOa g EszO men mS sC2m = .Y
. g'~%g O N) t l
1 I I I .
', O
. ; w
/ ^
- / {
- ' o
- v g
O
% Z Y
O . i LJ '
.b v b E
_GE .
-E9 ww .
-wg w
\
\ :
-[w p
_ \ ! _mg \ . 1m \
-mi \
88 en l \ l \ l \ I (o) NOI.l.YB373DOV SAN ONOFRE , NUCLEAR GENERATING STATION Units 2 & 3 t SONGS 2 AND 3 FUEL BUILDING POOL FLOOR VERTICAL (NE396 C4) l FOR 2% DAMPlNG OBE SPECTRA FIGURE 4.3-6
- - _ - - - - - - - - - - - - - - - )
i ., n F {' !
- 4.4 IDADS AND IDAD COMBINATIONS 4.4.1Il SPENT) FUEL' POOL'
~
4 . 4 .1.1 - Loads The:following loads-were considered'in the.SFP analysis:.
- A'. Structural; Dead Load (D)
The dead loads include theLweight of the structure,.the .
-vertical.and horizontal' hydrostatic loads'-in.the.SFP,'the weight.of.the new spent fuel racks l(with a full complement
-of spent' fuel), and a 50 lbs/ft2 uniform load on all floor-
~ slabs to' account for miscellaneous equipment, piping, and electrical, raceway. loads.
B. . Live Load (L) f
- The live loads include-the weight of the cask (250 kips located in the cask storage pool)'and the weight.of the railroad car / cask (394 kips located in the' railroad unloading area). ' Vertical and lateral hydrostatic loads are considered as live loads in the fuel transfer and cask storage pools since these pools do not require water at
, all times. i l l TMFWOO57- 4.4-1 l
r W 1 x. 4 overall; floor occupancyllive loads are very small in-
;p comparison'to_the structural weight and.their effects are s
negligible when considering-the seismic response /of the entire' structure. Additionally, the SFP basematf(the area-of concern) has no floor 1 occupancy live loads. Therefore, floor occupancy live'. loads were not considered in the evaluation. C. Normal' operating Thermal Loads (To) Normal thermal' loads (temperature differences and temperature gradients) are produced due-to the temperature
' distribution through the concrete wall /basemat and~
stainless steel liner plate system.during normal operating or shutdown conditions. These' thermal loading conditions are applied to the walls and basemat.of all three'(cask,
)
storage, and transfer)' pools. Temperature differences J represent a uniform temperature change above an initial unstressed temperature _(this thermal load causes onlyL membrane forces), while. temperature gradients _ represent a linear variation of temperature across:the thickness'of a wall cnr basemat (this thermal load causes only bending 1' moments). 1 The following temperature conditions are used in the ! conservative structural evaluation: b O. TMFWOO57 4.4-2 1 1 F
1 Pool Water 150F
/~St Outside Ambient Air 40F
'Q Outside: Soil 40F I Inside Ambient Air 65F Initial' Installation 65F (Zero-Thermal Stress) j D. Abnormal Thermal Load (Ta)
Abnormal thermal loads (temperature differences and temperature gradient)'are produced due to the temperature distribution through the wall /basemat and liner: plate during a full core off load (including the effects'of concrete. gamma heating). These temporary thermal loads are evaluated for the concrete walls, basemat, and liner
'l ) plate system of the three pools under the following two situations:
- 1. Spent fuel pool with water; transfer and cask' pool empty
- 2. All three pools with water Temperature differences and gradients are determined based i
on water temperature of 180F (a conservative abnormal ! maximum heat load of 162F and gamma heating of 18F). l 1 \ l Thermal loads will also be reevaluated based on a water i 1 l temperature of 212F. l 1 (2) TmF.00s, 4.4-2
U .
.. E. Design: Basis Earthquake!(DBE) Load-(E')
1.r' l'N Three-directional seismic loads ~are. applied-based.on.the.
, site specific free" field DBE response spectrum shown in-
-the'UFSAR figure 3.7-1. This spectrum is applicable to both horizontal components. lThe vertical-motion spectrum has the same shape as the horizontal, but is two-thirds-times the horizontal. The resultant seismic load distributions used in the' evaluation were established by_a three-component square root of the sum of the squares (SRSS) combination technique considering seismic excitation in all three coordinate directions acting.
concurrently. F. Operating Basis Earthquake (OBE) ' Load (E) ' The free' field OBE spectrum acceleration values:are one-half the DBE values. Three directional seismic; loads are applied and combined'as discussed for the DBE case - above. 4.'4.1.2 Load Combinations In the SFP analysis, the following load combinations, from the SONGS.2&3 UFSAR, paragraph 3.8.4.3.2 were considered. Load i combination 8, tornado loading is not applicable to the area of ! the structure being evaluated. O TMFWOO57 4.4-4
I "; l o
, " Load Case -f !
1
]/" , l No . -
h <
- 1. LNormal. Case-1.4D +11.7L
'2. Severe Environmental' Case ~
1.25D + 1.25L + 1.25E '
- 3. Severe Environmental Case 9
1.25D + 1.25L + 1.25E + 1.0To
- 4. Abnormal / Severe' Environmental Case 71.0D + 1.0L + 1.25E + 1.0Ta-() 5. Abnormal / Severe Environmental Case 1.0D + 1.25E + 1.0Ta-
- 6. ' Abnormal / Extreme Environmental' Case 1.OD + 1.0L + 1.0E' + 1.0To
- 7. Abnormal / Extreme Environmental Case 1.0D + 1.0L + 1.0E' + 1.0Ta
- 8. Abnormal Case (Not applicable) 1.0D + 1.0L + 1.0Wt + 1.0To + 1.25 Ho I
O -TMFWOO571 4.4-5 _ _ _ _ . _ _ -_ _ . _ __ _ _ . _ _ _ . _ _ . _ _ _ ______..___m.___ _ . _ _ i i iimW
.p 1
+ , . p INote:'.For; load-cases.4, 5,.6, and 7,l pipe. rupture l'oad.and .
, miscellaneousimissile ' loads 'are excluded because they: are .not'
~J
. applicable:for.the area being evaluated.
,r s
'4.4.2l. . SPENT FUEL' RACKS' 4.4'.'2.1 Loads
~
The following loads were considered in'the rack design': Dead Load' ('D) '= Dead weight-induced stresses (including M fuel.' assembly weight). Live Load. .(L) = ' Uniform loading over a limited area ofr the top of the rack.of'300.lbs/ft 2. Fuel Drop- (Fd) = Force caused by theLaccidental' drop of
. Accident' the heaviest' load'from the maximum
- Load-- possible height.'
Spent Fuel '(Pf) = Upward force on the racks. caused by'. i Handling Machine _ postulated stuck fuel assembly'(6000
. (SFHM) Uplift pounds).
Load Seismic (E) = Operating Basis Earthquake l~ Loads O TMFWOO57 4.4-6 _ _ _ . _ _ _ _ - _ _____.___-2__m._ _m_.___._..m____-_.________-.____u_____ _ . _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
() (E') = Design Basis Earthquake s Thermal (To) = Differential temperature induced loads Loads (normal condition). L"
~
(Ta) = Differential temperature induced loads (abnormal design condition). I_ 4.4.2.2 Load Combinations Each component operating condition has been evaluated for the applicable loading combinations shown in table 4.4-1. O O TMFWOO57 4.4-7
l
/i Table 4.4-1 LOADS AND LOAD COMBINATIONS FOR SPENT FUEL RACKS Load Combination Acceotance Limit' a-1.7 (D + L) NF 3340 of ASME Code Section III 1.3 (D + L + To) 1.7 (D + L + E) 1.3 (D + L + E + To) 1.3-(D + L.+ E + Ta) 1.3 (D + L + To + Pf) 1.1 (D + L + Ta + E')
'1. The abbreviations in the table above are those used in Section 3.8.4 of the Standard Review Plan (SRP) where each term is defined except for T Fd and Pg. The term T is
.fft )r definedhereasthehighest$e,mperatureassociatedvikhthe postulated abnormal design conditions. The term Fd is the force' caused by the accidental drop of the heaviest load from the maximum possible height, and Pf is the upward force on the racks caused by a postulated stuck fuel assembly.
- 2. The provisions of NF-3231.1 of ASME Section III, Division I, shall be amended by the requirements of Paragraph c.2.3 and 4 of Regulatory Guide 1.124, entitled " Design Limits and Load Combinations for Class A Linear-Type Component Supports."
- 3. For the faulted load combination, thermal loads will be neglected when they are secondary and self limiting in nature and the material is ductile.
TMFWOO57 4.4-8
l .0
'4 . 5 DESIGN AND ANALYSIS PROCEDURES 3g' - 4.5.1 ANALYSIS PROCEDURES FOR THE SPENT FUEL POOL.
4.5.1.1 LSnent Fuel Pool Structure Finite Element Analysis Increasing the SFP storage:chpacity will structurally affect only.
< the basemat and walls in and around the SFP. The finite element-model used for-analyzing the FHB is refined (element size) about the.SFP to assure accurate results in areas of concern. The.
remainder of the structure (above the. pool deck) is modeled in sufficientLdetail to account for'its static / dynamic effects on. the. fuel. pool. area. The finite element model was analyzed using the Bechtel Structural Analysis Program (BSAP) (25) . This is a general
. purpose computer program which uses the direct stiffness approach to perform linear elastic analyses of three-dimensional (3-D) structural models.
The' finite element model of'the FHB used for this evaluation is consistent with the original lumped mass model (UFSAR figure
- 3.7-22). The building is modeled in.3-D using " plate", " beam",
" brick", and " bound" (boundary) elements at and below the pool
" deck, and is represented with a stick model for the portions above the pool deck. The stick model portion of the building is adequate since it is modeled with sufficient detail to account for its stiffness and mass and it is distant from the area of TMFWOO57 4.5-1
___1_ _ - - _ _ - - -
concern (SFP basemat). This results in a finite element model with a fine mesh in the spent fuel area which yields accurate results in the area to be evaluated. Figure 4.5-1 provides an isometric view of the model. The static and dynamic analysis models are identical except for the boundary-elements (soil springs) used for each model'.
. Add.itionally, the dyncmic nadel hus cpring cle:acnta cadcd to represent the hydrodynamic loads of the oscillating water.
The springs representing the oscillating water are modeled based upon AEC Report TID-7024, Chapter 6 " Dynamic Pressure on Fluid Containers"(1) . The six soil boundary elements (three transnational and three () rotational) are attached to the basemat master node located he the center of gravity of the basemat. The soil boundary _ elements are based on the FHB soil stiffness parameters listed in the SONGS 2&3 UFSAR, table 3.7-6. Masses are lumped to the appropriate nodes of the model and a free-vibration modal analysis is performed in which enough modes are extracted to achieve 100% participation of the mass. The results of the free-vibration modal analysis are then used to perform OBE and DBE response spectra analyses. The damping values used in the analyses are shown below. The values listed are in agreement with the GONGS 2&3 UFSAR, table 3.7-3. Q TMFWOO57 4.5-2 i
l' 1 1 ( Concrete Steel Soil Water OBE 3%- 3% 8% 0.5% DBE 6% 5% 10% 0.5%. l The concrete section' evaluations were performed us'ing OPTCON which is a module of the-BSAP. computer program. For the purposes of this; evaluation, OPTCON is used_to evaluate existing! concrete H sections withLknown dimensions and areas.of steel.- All b-evaluations'are performed using the strength; design method of ACI. 318-71.; The' moments and membrane forces due.to thermal loads as
- 7. - determined in BSAP are for elastic (non-cracked) sections.
Moments due to thermal gradients are self relieving'in' nature.and OPTCON' accounts for the reduced 1 moments of the cracked section
-using an iterative approach which takes into consideration equilibrium;and compatibility equations for the cracked section as allowed by the applicable design code.
OPTCON.also accounts for a' hot liner plate on one side of the concrete section'. A hot liner plate induces. additional moment into the concrete section. The effects of yielding of the' liner plate.(if it occurs) on concrete is also considered.
.TMFWOO57 4.5-3
l' 4.5.1.2 Liner and Anchorace Analysis em The liner plate and its anchorage were evaluated for in-plane L horizontal loads caused by spent fuel rack. friction loads during a seismic event. The total horizontal load associated with the I racks for each appropriate load combination was uniformly applied to the liner plate floor system for the purpose of evaluating the horizontal effects. The. floor liner plate was evaluated for transverse (vertical) loads from the spent fuel rack support pads and impact loads from postulated load drops. The liner plate system for the proposed condition was also
. ,, evaluated for thermal effects with the parameters of the original 1 \
(_,/ analysis (see paragraph 4.6.1.3.B for results). 4.5.1.3 Foundation Stability and Soil Bearina The SONGS 2&3 UFSAR, table 3.8-19, lists the maximum foundation bearing pressure (worst case) as 21 kips /ft 2. The corresponding allowable bearing pressure is 44 kips /ft 2. The soil bearing was evaluated by comparison to existing margins (see paragraph 4.6.1.3.C for results). ( TMFWOO57 4.5-4
W
":s
- 4. 5. 2 ' DESIGN AND ANALYSIS PROCEDURES FOR SPENT FUEL STORAGE RACKS s
&O The purpose ~of this~ subsection is to demonstrate the adequacy of the spent fuel rack design.to store C-E 16 x 16 and H 14 x'14 fuel assemblies under normal and accident loading conditions.
The seismic analysis, which produces the governing loading
' conditions, produces results which are dividcd.into two types (stresses and displacements).
The stresses were checked against the design limits to ensure the structural adequacy of the design. The horizontal displacement i results were. checked against the rack clearances to determine whether the racks' collide with another rack or the pool wall. The vertical displacements of the support pads, due to rack [} rocking and lift-off, were used to show that the racks do not-overturn. The results showed that the high density spent fuel-racks are structurally adequate to resist the postulated stress combinations shown in table 4.4-1, and the racks have margin against overturning and against rack collision. 4.5.2.1 Analysis Overview The spent fuel storage racks are seismic Category I equipment,
- therefore, they are required to remain functional during and after a DBE(2). Since the racks are free standing (neither l-1 L TMFWOO57 4.5-5 l l
u_u________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _
l , anchored to the pool floor, to the pool wall, nor structurally interconnected) and the fuel is free to move inside the cell.
~
within the limits of the clearance between-the fuel and cell, a nonlinear seismic analysis was performed. To ensure the limiting values have been obtained with a nonlinear analysis, numerous conditions with different combinations of parameters were analyzed in order to identify the combination of parameters which produces the limiting conditions. The parameters within the pool configuration which affect the response of the fuel racks are: e Damping e Impact between.the fuel assembly and cell e Fluid coupling between the fuel racks and pool wall and
;O between fuel racks e Coefficient of friction between the fuel rack support points and pool floor e Structural characteristics of the fuel racks (Region I, Region II) e Fuel loading within the fuel rack With the exception of damping and the impact properties between the fuel and cell, these parameters were varied for the 12 cases summarized in table 4.5-1 to obtain the bounding values of fuel i
O TMFWOO57 4.5-6 l
rack responses (rigid body motion, structural deformation, loads, ( stresses, etc.) which were used to demonstrate the structural adequacy of the fuel racks. The analysis was performed on a 3-D ; nonlinear dynamic finite element model of multiple racks which was subjected to the simultaneous application of three statistically independent, orthogonal acceleration time histories at the pool floor. The pool floor acceleration time history data developed for the DBE are shown in figures 4.5-2 through 4.5-7. The steps in the seismic analysis were: A. Develop 3-D nonlinear dynamic finite element models of the fuel rack modules consisting cf beam, mass, dynamic gap, and friction elements. B. Perform time history analyses on the nonlinear dynamic models for the bounding cases using the dynamic capabilities of the nonlinear modal superposition solution technique of the Westinghouse Electric. Computer AHalysis (WECAN) Code (3,4), C. Compute the stresses in the fuel rack at the critical structural locations using the loads from the previous step. These stresses were checked against the design limits given in table 4.4-1 to ensure the adequacy of the design. i
- f%
O TMFWOO57 4.5-7
4.5.2.2 Seismic Model
.V b
Since the fuel assembly is not structurally connected to the cell
. wall and the fuel racks are free standing, the seismic model must have the capability to address a wide variety of rigid body motions:
e The fuel moving in the clearance between the fuel and cell with subsequent impact on the cell wall (" fuel rattling condition"); e The rack sliding on the pool floor (" sliding condition") ; e One or more support pads momentarily losing contact with _ the pool floor (" tipping condition") ;
'\_/
e Rack rocking onto one support pad and pivoting about that pad-(" torsional rotating condition"); e The rack may also experience a combination of sliding and tipping conditions. These mechanical nonlinearities in the fuel rack dynamic responses required a strong emphasis on the modeling of both the linear and nonlinear elements. TMFWOO57 4.5-8
To' determine the dynamic response of a system with multiple gg s nonlinearities, as described above, the analysis must be conducted on a model which realistically represents the dynamic characteristics.. That is:. the model must have~ realistic linear stiffness and mass properties as well as the correct number and
-geometric. location'of impact points, a realistic value of11mpact stiffness, and' sufficient Dynamic Degrees of Freedom (DDOF) in order to develop-the higher modes of response which are associated with impact forces. -The linear and nonlinear elements used in the model must have the capabilities described in the following paragraphs.
The linear. elements represent both the fuel and rack'stiffnesses-as well as the geometric properties (support pad' spacing, fuel grids locations, etc.) with sufficient DDOF to capture the dynamic response of the combined fuel and rack system. More
)
specifically the number and location of fuel to cell contact points were accurately modeled in order to provide the correct impact force magnitude and location so that the correct modes of response of the fuel and cell were produced. The fuel rack height, mass distribution, and support pad spacing were modeled to provide a representation of the rack rigid body stability characteristics. The nonlinear elements which model the impacting between the fuel assembly grids and cell included the impact stiffness, impact damping, and gap size. The nonlinear elements that model the support pads which may lift off and impact the floor, or may O i TMFWOO57 4.5-9 _ - _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ - _ _ - - _ _ _ b
J 1 I slide relative'to the pool liner,. included'the. impact' stiffness, ,[ impact ~ damping, and-Coulomb' friction to appropriately simulate pg .this interface between the fuel. rack and pool floor. 4.5.2.2'1 .' Model Description. The-time history analysis was performed on a 3-D nonlinear finite element model which represents multiple racks in the pool. The following sections describe <in detail the structural model, the nonlinear single rack model, the nonlinear partial fuel loading model, and.the nonlinear multiple rack model. To compile-the finite element model, the properties:of.the. linear. elements and-nonlinear' elements must be calculated. The properties.of the linear. elements are obt'a'ined from a 3-D finiEe-(O ,/ - element structural modeloof the fuel rack. The linear properties,; referred'to as.the effective-structural properties, are calculated from the struct' ural-model and used as the basis of the-nonlinear model. -The; nonlinear elements, with properties-such as impact stiffness, gap, impact damping, and friction coefficient, are added.to the' effective structural prcperties to produce the nonlinear model. The details of the structural model, effective structural properties, and the nonlinear model are: presented in the following paragraphs.
- O TMFWOO57 4.5-10
= _ _ - _ _ = - _ - - _ _ _ - _ _ - _ _ _
i 1 E I i 4.5.2.2.2 Structural Model l G l The structural model is a 3-D finite element model of the fuel j rack. Since the spent fuel racks are a two region design, and the racks for each' region are structurally different, two different structural models are used in the analysis as shown in figure 4.5-8 for Region I and Region II. The fuel rack is composed of the following structural components: support pad assembly, base plate, cells, and cell to cell connections. These components are attached-in a manner which produces an overall rack structure. The structural input properties of each component are discussed in the following paragraphs. A () The support pad assembly is composed of a leveling pad, leveling screw, and support block. The structural components in the assembly, leveling screw and support block, are modeled as beams with area and inertia values obtained from cross sectional properties. l The base plate assembly in the structural model has the effective stiffness values of the base plate, support blocks, and cells or bottom grid welded to the base plate. The base plate is modeled as effective beams which connect the support pads and the cells. Since the cells (Region II) or bottom grid (Region I) are welded to the base plate, the inertia and area of the base plate O TMFWOO57 4.5-11 i
p effective beams are based-upon the cross section which includes a portion of the base plate and the bottom grid (Region I) {J "l' portion of the cell wall (Region II). or a The cells are square (0.110-inch wall thickness) tubular sections which have 0.020-inch thick wrapper plates on the four sides to ) support the Boraflex sheets. The cells are modeled as beams with i area and inertia. based upon the cross section of the 0.110-inch l thick cell wall and only one wrapper plate on the tensile side of the cell. The use of one wrapper on the tensile side of the cell is based upon the post buckled condition which produces structurally ineffective members of the wrapper plates which are loaded in compression. It is noted that the dynamic analysis is based upon section properties which include the wrapper plate on-the tensile side, but the stress analysis is based upon section () properties which do'not include the wrapper plate, thereby producing a. conservative analysis. The cell to cell connections are the members which. form the structural connection between the cells and produce the overall shear connection of the rack assembly. In Region I the connection is produced by stiffener clips which are welded 1 between cells as shown in figure 3.1-1, and in Region II the i connection is produced by cell to cell welds at the cell corners as shown in figure 4.1-5. This connection is modeled with effective beams which connect between cells. The properties of j i these beams are obtained from a finite element unalysis of the cell wall and the stiffener clip or weld. The finite element TMFWOO57 4.5-12 l
modeloused to determine the equivalent properties of the' Region
/\
b II cell-to-cell connection beam is'shown.in figure'4.5-20. This model' captures the local cell wall flexibility at the weld. It' consists of 1000 constant-strain plane-stress elements. Because of symmetry only half of two.of the side panels of one.of the cells is modeled. The weld connection is at the center of this model. Although the actual side panels of the cell are 90' apart, in the model the two side panels-are in the same plane. This.modeling technique is justified'because the.two sides deform independently. The displaced cell wall geometry (magnification = 200) ' due to cell-to-cell shear transfer is also 'shown in figure 4.5-20. Thefproperties of the effective structural model are obtained from the results of the structural model. The effective model,. () shown in figure 4.5-9 for Regions I-and II,'is composed of-elements which represents the cell assembly, cell to cell connection, and support pad / base plate. The properties of these elements are calculated from the. load and displacement results of the s'tructural model, as discussed in the following paragraphs. l 4 A finite element model of the effective structural model is compiled and run, and the mode shapes, including the higher cell modes, of the effective model are compared with those of the structural model to ensure that the effective structural model is an adequate. dynamic representation of the structural model and O TMFWOO57 4.5-13
has a' sufficient number of modes to produce the higher mode
'ir~)N s_ . response due to fuel impact loads. Details of the modes and frequencies are presented in Section 1.5 of Reference 17.
The properties of the cell assembly 1,n the effective structural model are the same as those in the structural model. i i The properties of the cell to cell connection in the. effective model are rotational stiffnesses (K0 ) in units of in-lb/ rad. These values are the ratio of bending moments reacted by the cell i to cell connection members divided by the angular rotation of the connection at the cell from the structural model. At each l elevation of cell to cell connection, this calculation is j performed at each cell location and averaged for the total cross j
-s section to obtain the effective rotational stiffness at that \ssl elevation. The details of this calculation are presented in Section 1.1 of Reference 17. The cell to cell connections, which produce the overall shear connection of the rack, are placed at locations along the length of the cells to produce the required frequency of the rack. As shown in figure 4.5-9, there are 10 cell to cell connections (welded clips) plus one top grid connection for the Region I effective structural model, and there are 17 cell to cell connections (welds) for the Region II effective structural model. The value of the rotational stiffness between the bottom of the cell and base plate is calculated by the same method as the cell to cell rotational stiffness values. It is included in the stiffness matrix I 1
t element, [K), between the bottom of the cell and base plate. TMFWOO57 4.5-14 ]
g The effective support pad / base plate properties are calculated in two stages. The overall base plate rotational stiffness is calculated from the structural model, and then the combination of rigid beams with vertical spring elements on the ends is'used to represent the rotational stiffness while providing the. effective-vertical stiffness and geometric spacing of the corner support pads. The overall base plate rotational stiffness is calculated by. dividing the rotation of the rack base by the total moment
~~
applied to the. base. The rotation of the rack base is calculated by. dividing the vertical displacement at each cell location on the base by the distance from that point to the rack centerline,.
'and calculating the average for all the cell locations in the rack base. This accounts for both the support pad deformation and base plate deformation at the locations between support. pads.
The vertical stiffness of the effective corner pads is cblculated. by equating the rotational stiffness of the vertical spring / rigid beam' assembly to that of the base plate rotational stiffness (Kg). Using four support pads, one at each corner, and a spacing of (L) inches between support pads, the effective vertical stiffness per corner pad is Ky = Kg/L 2. The details of this I calculation are presented in Section 1.4 of Reference 17. ' O TMFWOO57 4.5-15 _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ - _ . _ _ _ _ _ _ _ - . __ .1
1 l
.l
.4.5.2.2.3 ' Nonlinear.Model, Single Rack The nonlinear model, shown in figure 4.5-10, is a 3-D model composed of effective properties from the structural model, to- l account for the= rack structure, with additional elements to l account.for the. fuel assembly, fuel to cell gap, fuel
' hydrodynamic mass, support pad' boundary' conditions of aufree standing rack, and the hydrodynamic' mass between the fuel rack and the pool wall or between fuel racks. A better illustration of the elements of a 2-D view of the 3-D model is shown in figures 4.5-11 and 4.5-12 for Regions'I and-II respectively.
The effective components / properties from1the structural model are: () e The cell assembly represented by 3-D beam elements with effective stiffness properties and mass density;- e Cell to cell connections represented by rotary stiffness elements in both horizontal axes; e Cell to base connection represented'by a stiffness matrix element with rotary stiffness in both horizontal axes, lateral stiffness in both horizontal axes, and rigid connection in the vertical axis; e Support pad corner spacing represented by rigid 3-D beams elements connecting the center and the four corners; O TMFWOO57 4.5-16
l f'[o e The. effective support pad / base plate vertical stiffness represented by a 3-D friction element which has the nonlinear capability to slide on the horizontal plane or lose contact in the vertical direction (support pad lift-off) and impact upon contact. The fuel assembly is represented by a combination of 3-D beam-elements and rotational stiffness elements. Th'e beam elements-have properties of inertia, area, and mass density. The area and inertia values are based upon the fuel rod cladding and fuel skeleton cross section. The rotational stiffness elements have rotational stiffness about both horizontal axes to account for 4 the connection of the fuel rods at the fuel grid locations. The connection between the fuel base and fuel rack base plate is () modeled with a 3-D friction element which has the nonlinear capability to allow the fuel assembly to lift-off the base plate and produce impact forces. The dynamic properties of this fuel assembly model were qualified by comparing mode shape and frequency values from the finite element model with values supplied by the fuel vendor. The fuel to cell gaps are modeled by 3-D dynamic elements with gaps, impact stiffness, and impact damping values to represent the gap between the fuel and cell. The fuel assembly impact stiffness and damping values for the C-E 16 x 16, and E 14 x 14 fuel assemblies were used in the analysis. The maximum value of the gap between the fuel and cell is conservatively used in the TMFWOO57 4.5-17
I y l . model. This is-done by using the tolerances to produce the minimum fuel grid dimension and maximum cell opening. The gap is { input as a concentric gap where the fuel is initially in the j center of the cell and can impact any of the four sides of the cell. The impact stiffness between the fuel and cell is based upon the series combination of the fuel grid impact stiffness and the local stiffness of the cell wall. The fuel grid impact ! stiffness is supplied by the fuel vendor, and the local cell wall stiffness is calculated from the displacements of the cell wall, l' modeled as a flat plate, due to loads applied at the corners of the fuel grid. Details of this calculation are presented in Section 2.0 of Reference 17. The impact damping of the fuel grid is based upon values supplied by the fuel vendor. There are 12 gap elements located along the length of the fuel assembly'to represent 11 intermediate grids and one top fitting. f\
\
The hydrodynamic mass between the fuel assembly and cell is modeled by a 3-D mass matrix element in both horizontal directions. Since the fuel assembly is an open array of rods, the proximity effect of the cell wall on the rods only affects the edge row of rods and does n a produce a significant hydrodynamic mass on the fuel assembly. Thus, the flow around the rods is the condition which produces the governing hydrodynamic mass. The calculation of the hydrodynamic mass due to flow around the fuel rods is based upon the kinetic energy of the fluid (5). The hydrodynamic mass elements are modeled at 12 locations along the length of the fuel with values proportional to the effective fuel length at each location. O TMFWOO57 4.5-18
The hydrodynamic mass between the fuel rack and the pool wall, or O between fuel racks in a multiple rack model, is modeled by 3-D mass matrix elements in both horizontal directions. The hydrodynamic mass elements are modeled at 12 locations along the length of the cell with values proportional to the effective cell length at each location. The values of hydrodynamic mass for the fuel rack are based upon potential flow theory and are calculated by evaluating the effects of the gap between the rack and the pool wall or between racks by using the method outlined in the paper by R. J. Fritz(6). A more detailed discussion on hydrodynamic mass is presented in paragraph 4.5.2.2.7 Fluid coupling. 4.5.2.2.4 Nonlinear Model, Partial Fuel Loading Models O Nonlinear models for partial fuel loading were compiled for three configurations; quadrant fuel loading, four-row fuel loading, and empty rack. These models were used to address specific types of motion. The quadrant fuel loading model, shown in figure 4.5-13, was used to address the fuel loading eccentricity which will shift the center of gravity toward one corner pad and cause the rack to rock onto one support pad and produce torsional rotation motion. The four-row fuel loading model, shown in figure 4.5-14, was used to address the fuel loading on one side of the rack which will cause the rack to rock up on one side with tipping motion. The empty rack model was used to calculate the motion of an empty rack. The displacement results of these three models TMFWOO57 4.5-19
and.the full rack model were used to determine which two racks produce the maximum relative motion, and also justify the placement of fuel-in the rack;in any configuration; The quadrant loading and four row loading models are compiled -
' with the.same support base configuration as the single-rack model. In the center of the model is the cell' structure which
, has the stiffness properties of the empty cells but the total structural. mass of'the rack.- At the center of gravity'of the fuel loading is.the cell structure which has the stiffness
- properties'of the cells'with fuel but no cell structural mass..
The empty cells structure is coupled in the horizontal' directions with the cells with fuel to provide the. correct stiffness of the-rack. Coincident with the-loaded cells structure is the fuel assemblies model.. The finite element details in the cells O structure and. fuel assemblies model are the same as those used in the single rack model, but with stiffness and mass properties
~
equivalent to the number of cells or fuel assemblies in the partial loading condition. The hydrodynamic mass' elements which represent the rack to pool. wall hydrodynamic mass are connected to the empty cells assembly in the center of the rack model. 4.5.2.2.5 Nonlinear Model, Multiple Rack Model Since the relative motion between racks is needed to address-
. rack-to-rack interaction, a nonlinear model of multiple racks was compiled. The finite element details in each rack of this model are the same as those presented for the single rack model. There O
.TMFWOO57 4.5-20
-_.--____mu_____ ____.___u_---__--________2--_ __ _.-...-.._-.aL.___.__m.._ . - _ - . -
-are two different multipl'e rack, full / full and empty / full,.for j,
\ss each region:as shown in figures 4.5-15 through 4.5-18. The dynamic degrees'of freedom for these models are 190' full / full and 129 empty / full for Region I Land 180 full / full'and'123 empty / full for Region ~II. The full / full model was.used to produce the 1
. maximum' absolute displacement of the racks in order to address l
l the relative motion between the rack and pool wall. To address the relative-motion between racks, the combination of full and partial fuel loaded racks which produce the maximum' relative-
- displacement must be used. The results of the rack'models-for full fuel loading,' partial fuel loading in quadrant, partial fuel loading in four. rows, and empty rack, presented in Subsection 4.1.3 of Reference 17, were used to show that the maximum-relative displacement was produced by the combination of a fully loaded rack and an empty rack.
O The plan view of the multiple rack model, figure 4.5-19, shows the-position of two rack models with solid lines and the position of two " effective" rack models with dotted lines. The effective racks are not actual models, but are shown there because they represent the effect of adjacent racks on-the two actual racks. Thus the two actual racks respond in a manner which accounts for the effect of adjacent racks in both N-S and E-W directions, and can be referred to as a four-rack model. Details of the four-rack model are presented in Section 5.0 of Reference 17. The value of the rack to rack hydrodynamic mass is based upon the initial spacing between racks. ( TMFWOO57 4.5-21
L In addition the plan view, figure 4.5-19, shows the hydrodynamic ! i
/~ elements between the racks and the pool wall. Since the seismic C')
model has multiple fuel racks, the hydrodynamic mass value for each rack is based upon the appropriate gap between the rack and j i the pool wall. Therefore, the seismic analysis accounts for the different gap sizes. The hydrodynamic mass in the E-W direction for rack 1 is based upon the initial distance between rack 1 and l j the east wall of the SFP which is 11.10 inches for the Region I { model and 5.28 inches for the Region II model. The E-W - ~ hydrodynamic mass for rack 2 is based upon the initial distance between rack 2 and the west wall of the pool which is 54.9 inches for Region I and 48.0 inches for Region II. Since the distance between the racks and pool wall on the north end and south end are approximately equal, the hydrodynamic mass () for the N-S direction on both racks is based upon the initial. distance between the racks and pool wall on the north wall and south wall, and averaged over the four racks in the N-S direction. A more detailed discussion on hydrodynamic mass is presented in paragraph 4.5.2.2.7 Fluid Coupling, and the details concerning the averaging technique for the N-S direction are presented in Section 3.2 of Reference 17. 4.5.2.2.6 Damping There are two types of damping present in the dynamic response of a nonlinear structure such as the fuel rack (structural damping and impact damping). TMFWOO57 4.5-22
.The structural damping values used for the seismic analysis of fuel rack structures are 2% for OBE and 4% for-DBE. These values are in accordance with Regulatory Guide 1.61(7) for' welded steel structures and the SONGS 2&3 UFSAR section 3A.
The damping which-occurs during impact is produced by the small amount of local plastic deformation which takes place.- The two types of locations in the rack which. have impact are .the fuel' grid to cell wall and the support pad to the pool floor. The
" damping value used for the fuel grid impact, which was supplied by the fuel vendor, is 4.4%. The damping value used for.the support pad impact is based upon steel on steel impact. The range of coefficient of restitution for. steel on: steel in Reference 18 is 0.5 to 0.8. Thus, conservatively using a value O of 0.85 coefficient of. restitution produces an effective impact damping of 4% for the support pads.
4.5.2.2.7 Fluid Coupling The effect of water on the dynamic response of a. submerged structure is significant, and must be included in the modeling. If one body of mass (mi) vibrates adjacent to another body of mass (m2), and both bodies are submerged in'a frictionless fluid medium, then Newton's equations of motion for.the two bodies have the form: O TMFWOO57 4.5-23 i e-____________-_-____--_-___-_. ._ __ __ _ j
(my + M11) 51-M12 2 = applied forces on mass al l
-M 21 = applied forces on mass m2 1+ (m2 + M22) 2 E,E2 i
denote absolute accelerations of mass mi and m2, respectively. Fluid coupling coefficients, M11, M12, M21 and M22, depend on the shape of the two bodies, their relative disposition, etc. Fritz(6) gives data for Mij for various body shapes and arrangements. It is noted that the above equations indicate that the effect of the fluid is to add a certain amount of mass to'the body (M11 to body 1), and an external force which is proportional to the acceleration of the adjacent body (mass m2). Thus, the acceleration of the one body affects the force field on another. This force is a strong function of the interbody gap, reaching large values for very small gaps. This inertial coupling is called fluid coupling. It has an important effect in rack dynamics. The motion of the rack as well as the lateral motion of a fuel assembly inside the storage location will encounter this effect. The hydrodynamic coupling between the fuel and cell and between the rack and pool wall is represented in the models by mass matrix elements. The hydrodynamic mass between the fuel assembly and the cell walls is based upon the fuel rod array size and cell O TMFWOO57 4.5-24
- - ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __
inside dimensions using the technique of potential flow and
. kinetic energy. The hydrodynamic mass is calculated by equating the kinetic energy of the hydrodynamic mass with the kinetic i energy of the fluid flowing around the fuel rods. The concept of kinetic energy of the hydrodynamic mass is discussed in a paper by D. F. DeSanto(5), and the details of this calculation are presented in Section 3.3 of Reference 17.
The hydrodynamic inass between the racks and between the racks and pool wall was calculated by evaluating the effects of the gaps between the racks and the pool wall using the method based upon potential flow theory outlined in the paper by R. J. Fritz(6), To account for the flow in three dimensions, the hydrodynamic mass for flow in the horizontal direction around the racks and for flow in the vertical direction up over the top of the racks and down below the bottom of the racks are calculated independently, and combined to produce an overall hydrodynamic mass value. The details of this calculation are presented in Section 3.1 of Reference 17. The effects of rack to rack gaps and rack to pool wall gaps are included in the analysis. 4.5.2.2.8 Friction Coefficient Since the fuel racks are free standing, the frictional resistance in the interface between the support pads and pool floor is the only horizontal constraint. Thus the value of friction O TMFWOO57 4.5-25
coefficient must be accurately represented in a manner to in ( ,) conservatively calculate the displacements of the fuel rack. The values used in the analysis are based upon tests performed by Rabinowicz(8). According to Rabinowicz, the results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of friction coefficient to be 0.503 with a standard deviation of 0.125. The upper and lower bounds (based on two standard deviations) are 0.753 and 0.253 respectively. In order to address the upper and lower bounds of friction, two separate analyses are performed on the rack models with friction coefficient values of 0.8 and 0.2. The results of these analyses showed that different dynamic behavior was produced for each coefficient, thus producing the bounding response. However, in order to address a combination of behavior characteristics for an intermediate friction coefficient, an analysis was performed using the mean friction coefficient (0.5), and the results presented in Section 4.1.2 of Reference 17 show that rack displacement for the 0.5 coefficient of friction was bounded at all times by the maximum response of the 0.2 and 0.8 cases. 4.5.2.3 Time History Evaluation The seismic analysis of a free standing fuel rack is a time history analysis performed on a 3-D nonlinear finite element model subjected to the simultaneous input of three statistically independent acceleration time histories at the pool floor elevation. The analysis was performed on the Westinghouse O 4.5-26 TMFWOO57
Electric Computer Analysis (WECAN) Code (3, 4) using the dynamic
/~')I ' capabilities of the nonlinear modal superposition method.
WECAN is a general purpose finite element code which has been reviewed and approved by the NRC. The general reviews of the overall code by the NRC are: e Documents submitted to the NRC for review are: Westinghouse Report WCAP-8252(3) Westinghouse Report WCAP-8929(4) e NRC Review at Pittsburgh, PA on October 1-5, 1984 by Messrs. P. Sears and P. Milano, Reference Docket No. 99900404/84-03 () In addition to the general reviews, the application'of the nonlinear modal superposition method in the WECAN Code for spent fuel rack seismic analysis has been specifically reviewed by the NRC during the licensing review for the following fuel rack dockets. o Duke Power Co., Oconee Unit No. 1&2, Docket No. 50-269, 50-270, 50-287, during 1980. e Duke Power Co., McGuire Unit No. 1&2, Docket No 50-369, 50-370 during 1984. O TMFWOO57 4.5-27
i L< o Consumers Power Co., Palisades Plant, Docket No. 50-255, during 1986. The following is a list of rerack projects in which the seismic analysis was performed on the H WECAN Code. Utility Site Name Arkansas Power and Light Arkansas l'& 2 Carolina Power and Light Shearon Harris 1, 2, 3, &4 Carolina Power and Light H. B. Robinson Consumers Power Palisades Duke Power Oconee 1, 2&3 McGuire 1 & 2 Florida Power and Light Turkey Pt.-3 Georgia Power A. W. Vogtle 1 () Gulf States Utilities River Bend 1 Northeast Utilities Millstone 1 & 3' Philadelphia Elec. Co. Peach Bottom 2 & 3 Public Service of New Hampshire Seabrook Tennessee Valley Authority Bellefonte 1 & 2 Texas Utilities Comanche Peak 1 & 2 The nonlinear modal superposition method was developed to analyze nonlinear structural' dynamics problems involving impact between components and Coulomb friction. The finite element method is used to express the equations of motions with the nonlinearities represented by a pseudo force vector. The details of the O TMFWOO57 4.5-28
)
i _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ __ ._ ._ J
7__-_---_ nonlinear.. modal superposition method used in WECAN are. presented ( .below. References 19 and 20 provide additional information and;
~ application of the nonlinear modal superposition method.
The natural frequencies and mode shapes for the nonlinear structure are obtained by' reduced modal analysis or full modal analysis.- These frequencies and mode shapes represent the reference state of the nonlinear structure. During the time history analysis, as the nonlinear behavior comes into action, l the true frequencies and mode shapes. change. The effect of the
. variation of the true frequencies and mode shapes from the original ones is represented by pseudo forces on the right-hand side of-the equation of motion.
The generalized equation of motion of a structure is: [M](E) +-[Cnl]($) + [Knl](X) = (F) (1) Where, [M] is a mass matrix. [Cnl] is a nonlinear damping matrix, dependent upon velocity and displacement. [Knl] is a nonlinear stiffness matrix, dependent upon displacement (X},{X),(5) and (F) are displacement, velocity, acceleration and applied force vector. Let [Cnl] = [C] + [C] O TMFWOO57 4.5-29
s 1 (~% . and () [Knl) = [K) + [5] (2). where [C] and [K) are the damping and stiffness matrices representing the reference state of'the structure. The [6] and [5] are the damping and stiffness matrices dependent on ve'locity and displacement. . Equations (2) are substituted in equation (1). This gives: [M](E) + [C](X) + [K](X) = (F) - (Fnl) (3) Where, the pseudo force vector is defined by
-(Fnl) " [E]($) + [E](X) (4)
As the nonlinear properties are normally restricted to a small k portion of the structure, only a small number of finite elements-with nonlinear properties are used to model the structure. So the matrices [6] and [5] are sparse, and the cost of calculation of the pseudo force is small'if computations are performed at element level. The homogeneous, undamped equation of motion
~
representing the reference state of the structure is: [M](N) + [K)(X) = (0) (5) Let [ w] and [4] be the natural frequency and normalized mode shape matrix associated with equation (5). The following transformation, (X) = [$)(q) (6) O TMFWOO57 4.5-30
- l yo is substituted'in' equation (3) pre-multiplied by [$]T, employing-the orthogonality relations expressed by
[$)T[M][$) ='[I]. [$)T[C)[$) = [2C w] I and [93T[K][$) = [w 23 the resulting modal equations become 2 (6) + [2 C w .] {4) ~+ [w 3(9) " (Q) - (Onl) (7) where, G= percentage of the critical damping for the jth mode.
-{Q).= [$]T(F) = generalized applied force vector.
(Onl) " [$3T (ynl) = generalized pseudo force vector Arrays (q), (q) and (y) are the modal displacement, velocity and-acceleration vector respectively. The generalized pseudo force vector is a function of displacement and velocity. For a given time step, it can be approximated by' Taylor series-as follows: (Qn1)lt " (Qn1) (t - T)k (8) k=0 kl dtk ,T T<t<T+ T k=0,1,2....... O TMFWOO57 4.5-31
i Equation (7) represents a set of uncoupled equations. These equations are' integrated analytically to, eliminate numerical ; damping or frequency distortion during integration. Equation (8) represents the extrapolation'of generalized pseudo force vector j j by Taylor series. The number of terms that can be included in L the Taylor series is determined by the continuity of (Qnl) and-its time derivatives. The more the number of terms, the larger the allowable integration time step. The' extrapolation is done in the modal space, which is of relatively small' size. Each additional term of the series will require small additional storage in computer core. For the most practical applications, it suffices to include only the first two terms of the Taylor series. For a given time step, modal equations of motion are integrated-() analytically. associated with the nonlinear elements are calculated. .This Then the displacement and velocities of the nodes information is used to calculate the generalized pseudo force vector and its time derivatives. Then.the modal equations are integrated for the next time step. The nonlinear model was run for the bounding cases listed in
' table 4.5-1 which account for the variation of parameters such as friction coefficient (0.2 and 0.8), Region I and Region II rack structure, and fuel loading. The results from these runs include the fuel to cell impact loads, support pad loads, fuel rack structure internal loads and moments, support pad lift-off, fuel rack sliding and structural displacements. Since the seismic O TMFWOO57 4.5-32
l 7, analysis was conducted on a multiple rack model, the relative
+
(~s) / (where relative is the displacement of one rack with respect to an adjacent rack) displacements between racks at bo'.h the bottom and top of the racks as well as the absolute (where absolute is the displacement of a rack with respect to the pool floor) displacements at the bottom and top of the racks were obtained. The values of these results were searched through the 80 seconds duration of the time history to obtain the maximum values. The maximum values of the loads and moments were used in the stress analysis, and the displacement results were used to show that significant separation margin against collision with an adjacent rack or the pool wall remains and that there is ample margin against overturn.
,g A number of conservatism have been incorporated in the analysis
- I
\ ' and are listed below.
A. All fuel assemblies are treated as if they respond in phase which results in the maximum rack response. B. Friction coefficients of 0.8 maximum and 0.2 minimum are used in the analysis. C. A low value of 4.4% is used for the fuel assembly grid impact damping. (D V-TMFWOO57 4.5-33
L: D., Hydrodynamic mass is based upon constant gaps. As the gap
.f g decreases the! hydrodynamic mass restoring force increases;
'~
; but since the analysis is based,upon constant gaps, the displacements which'close the gaps are conservative because the restoring: forces. increase.
The seismic model, which uses.four effective' support pads
~
E.. to represent the 26 (Region I) and 33 (Region'II) actual support pads, rocked onto one support pad and produced'
' rotational motion.' Since the actual rack has multiple interior support points and will not lift off onto one support point,.the support pads in contact will resist the rotational motion.
- 7. No friction.is used in support pad ball joints to. resist
( rotation when the rack rocks'onto one. pad.
}
G. Gaps between fuel and cell were maximized and produce the maximum impact forces. H. The magnitude of the separation between the racks and between a rack and the pool wall, which is maintained during the seismic event, shows the conservatism in the overall design. In addition to the conservatism that have been incorporated its the analysis, several uncertainties have been addressed: l O 'TMFWOO57 4.5-34
;7 r A. Weld. performance vs.-fuel rack' seismic' response n
l [' , k 's OT B.- . Hydrodynamic mass distribution'in.the N-S direction
- c. -Applied coefficient of friction-l-
These items are addressed in the following paragraphs.
'To address the uncertainty of.the structural performace of'the.
fuel rack welded structural connections,,in lieu of load-deflection tests, a sensitivity study was performed. This study., presented.in Section 1.2 of Reference 17, concluded.that the seismic model is 'aut accurate representation of the fuel rack l' dynamic characteristics, since the primary frequency of the rack was reduced by 4% producing a 2% increase in rackidisplacement due to a conservative reduction of 10% of the welded connections. A second uncertainty involves the use of an average-distribution of the hydrodynamic mass of.the racks in the N-S direction. A study of the effect of applying an average hydrodynamic mass to each-rack versus lumping the hydrodynamic mass on the racks closer to the pool walls was performed to evaluate this uncertainty. Section 3.2 of Reference 17 discusses the details of the evaluation.- It was concluded from the results of the study that the average hydrodynamic mass distribution,.used as
'the basis of the N-S hydrodynamic mass in-the seismic analysis, produces rack displacements greater than the lumped hydrodynamic
' mass distribution method. Since the lumped hydrodynamic mass O TMFWOO57 4.5-35
_ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ = _ _ _ _ _ _ _ - ._
; distribution is the:more refined method, the use of the average hydrodynamic mass distribution in the seismic analysis produces
, conservative displacements, and the lumped hydrodynamic mass- -
q I
. distribution' produces 5% less rack displacement.
i JAnother uncertainty _that has been addressed is the uncertainty of I the actual coefficient of friction at the support pad.to pool. i floor interface. This situation has been addressed by a study.
-which evaluates the response of the racks for several different coefficients of friction. Subsection 4.1.2 of Reference 17
. presents. additional information about the study performed. The results of the study showed that using the upper and lower bound values of 0.8 and 0.2 for the coefficient of friction captures the maximum response'of the racks. Actual values for the ,
( coefficient of friction between the two extremes-produce less
)
rack response, and the mean value of coefficient of friction of 0.5 produces 22% less rack response than the 0.8 and 0.2 values. To' summarize the. uncertainties, it was shown that their individual effect is small (will not increase seismic displacements'more than 2%), and their combined effect is negligible since two of the uncertainties decrease displacements and one increases displacements. I i O TMFWOO57 4.5-36
'% J 1
.)
Table 4.5-1 LISTING OF SEISMIC ANALYSIS BOUNDING CASES RACK TYPE FUEL LOADING FRICTION COEFFICIENT Region I Partial, Quadrant 0.2 Region I Partial, Quadrant 0.8 Region'I Partial, Four Rows 0.2 Region I. Partial, Four Rows 0.8 Region I Full / Full 'O.2 Region I Full / Full 0.8 Region I- Empty / Full 0.2-Region I Empty / Full 0.8 Region II Full / Full 0.2 Region'II Full / Full 0.0 Region II Empty / Full 0.2 0 Region-II Empty / Full 0.8 i TMFWOO57 4.5-37
.c ..
O
% L g
5
%h g%
saga n 9 %> s:egpy
' g E
5 a
,gp e
4 gyN
.#g,,gSas o N
~
% N e e i
NUCLEAR EN RATING STATION O ISOMETRIC VIEW OF SONGS 2 AND 3 FUEL HANDLING BUILDING FIGURE 4.5 1
V SONGS 2&3 FUEL BLDG. TIME HIST. , HOR. NS DBE (NE368 C4) FROM 0 TO 40 SEC, 1.00 l 0.75-
, 0.50-3 0.25- ,
$ 0.00 *'
' I
' i
$ @ d I '
j i ! l a i f ' l
} .
I U -0.25 - i i
- 0.50 -
O' -0.75 -
-1.00 , , , f 0 10 20 30 40 TIME (SECONDS) i SAN ONOFRE O NUCLEAR GENERATING STATION UNITS 2 & 3 ACCELERATION TIME HISTORY N-S DBE O TO 40 SEC.
FIGURE 4.5-2
f% SONGS 2&3 FUEL BLDG. TIME HIST HOR. NS DBE ~(NE368 C4) FROM 40 TO 80 SEC. 1.00 0.75 - L 0.50-E 0.25-l 1I I t k O.00 '
! < I .I b !! Y
$ f f l l
! j l
i I f fl l ]'l Ilf 7l l
' y) ~ 0.25 - lI N
-0.50 -
f -0.75 -
-1.00 , , ,
40 50 60 .70 80 TIME (SECONDS) SAN ONOFRE NUCLEAR GENERATING STATION O' UNITS 2 & 3 ACCELERATION TIME HISTORY N-S DBE 40 TO 80 SEC. FIGURE 4.5-3
SONGS 2&3 FUEL BLDG. TIME HIST. HOR. EW DBE (NE386 C4) FROM 0 TO 40 SEC. i 1.00 0.75-0.50-O ' O.25-5 ;i / i
.I 5 0.00 , ; t
- a n J g 'ry '
l t l i l1 : ! : l a i 1 , i d -0.25 - W '
-0.50 - ,
-0.75 -
-1.00 , , ,
0 10 20 30 40 TIME (SECONDS) l SAN ONOFRE l NUCLEAR GENERATING STATION UNITS 2 & 3 k ACCELERATION TIME HISTORY E-W DBE O TO 40 SEC. FIGURE 4.5-4
1 7! SONGS 2&3 FUEL BLDG. TIME HIST HOR. EW DBE (NE386 C4) FROM 40 TO 80 SEC. { 1.00 0.75-0.50- ' s A I i 8 ' vi 0.25- ,j 2 p, gj l , y 0.00 h L !, t t. dIll g* d I, i, l I 1 l
! A h[. ,M k 1I g i j
i l 11
- ,(y ;
]
i d -0.25 - l ri I W ' )
-0.50 - i i
-0.75 -
-1.00 , , ,
40 50 60 70 80 TIME (SECONDS) SAN ONOFRE NUCLEAR GENERATING STATION O' UNITS 2 & 3 ACCELERATION TIME HISTORY E-W DBE 40 TO 80 SEC. FIGURE 4.5-5
1 l , i (D
.V SONGS 2&3 FUEL BLDG, TIME HIST.
I VERT. DBE (NE360 C4) FROM 0 TO 40 SEC. 1.00 0.75-0.50-O 0.25- i f 1 O I l 0.00 ' I i i 1 f ' 5 7 g I t , i 1 . y ; ; d s I l l U -0.25-W
-0.50 -
-0.75 -
-1.00 0 10 0 $o 49 TIME (SECONDS)
SAN ONOFRE NUCLEAR GENERATING STATION 1' O ' UNITS 2 & 3 ACCELERATION TIME HISTORY VERT. DBE O TO 40 SEC. FIGURE 4.5-6
e ( SONGS 2&3 FUEL BLOG. TIME HIST VERT. DBE (NE360 C4) FROM 40 TO 80 SEC. 1.00 0.75-0.50-v 0.25-E O 00 f ' ' I N'- 1 -M ' 1 iOl ebl hL l I[ l Fl M7 f
' I f 4 hi / f l4 l
U -0.25 - W
-0.50 -
-0.75 -
-1.00 , , ,
40 50 60 70 go i TIME (SECONDS) i SAN ONOFRE-NUCLEAR GENERATING STATION O UNITS 2 & 3 ACCELERATION TIME HISTORY VERT. DBE 40 TO 80 SEC. FIGURE 4.5-7 , 1
REGION I REGION 11 s _ { m 6_Irrlir'O 7
- lIf_n z ;;=:::=r ;!--
Top Grid 4.j ; J ; . , ; ,,; - j _. '2;%. _ gj . - ! Cell to Cell Connection - --
, ~l ly I ' !
; e- , ,
Cell f A l < 1 i r 1 - l
-1 i <
-.1 i<
-1 i <
1 i<
-J l <-
B
/ ase Plate --
Support Pad -
\jI le l
l
' SAN ONDFRE NUCLEAR GENERATING STATION O UNITS 2 & 3 STRUCTURAL MODELS REGION I & II FIGURE 4.5-8 l
l
REGION I REGION li
~!htrit::E- _ -
E...c C M",e era asstus '] :r asseus -
- \ , ;
\ l l
! /j I
# #j i i,
'I! !I
. 1
- -% i !!
l
- g-C)
.\J IVIEUlsI'* IVIIRiss'" g
- t. L . . . . . . .
t, - f h I i SAN ONOFRE
'O1 ~
NUCLEAR GENERATING STATION UNITS 2 & 3 EFFECTIVE STRUCTURAL MODELS REGIONS I & II FIGURE 4.5-9 i - _ - _ _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - - _ _ _ _ . - - - _ _ _ _ _ _ _ _ _ _ l
3 /? . 1 N [s N s
'[/l /
s- - _- \ se z N , , s, ':<1 A A- 4 $ N N) s '/4
\v se yg#/!
O N ' s, '/4 xe A\ x O "=" M M s='" 3-D NONL N SE SMIC MODEL E 45 10 1201C/0309C
1 [ l o NONL INE AA SE15Ulc . WCDEL. SINGLE m ACK
' ra o rue or 3-o .oegu wvDACOVN AWIC W AS$,
~
.['T - NE T1 b CEL ' ASSE WBLY q ry g._ ASSEMBLY - -
7 CD 6+ d
' ,"d ruEL GA!:) -
FrN 55
' Wa, Lp a w s
' 6~.jp L
aD r 6W.J MAS FUEL 'N
# y"*.J-N ,L.5 J * '
/ r ,
rVEL 70 CELL" GAP ELE ME NT. ' NW # r , as I L.p a 6 u. W ^ ry n, 4a u a /
- u. f W ^ r L.p a 6y ,s
. d"'D.- ..
e.,. ./ - U"
'l tJ : -
. 6Ob: -
l 5";b. - e , /
...y/ ,
wunc ELEMENT D
/
/
[N3s .- mm,--s 4 e f/ / /////// . SAN ONOFRE (~'s NUCLEAR GENERATING STATION
-Q ' UNITS 2 & 3 NONLINEAR SEISMIC MODEL (2-D VIEW OF 3-D MODEL)
REGION I FIGURE 4.5-11
NONL INE AR SE ISWlC WDDEL, SINGLE Q Acx 42-0 v2s or 3.g gong NE TI [ELNcx
~~
3 CELL ASSEWBLYq \ rVEL ASSEWBLY] gy - wWas ! 9 Man em. - ,
+ .W .,.
# ,W e, Lp a w s A UEL A '
N ,Lp a A , , b,N f [ br
- u. f.
e ,
,, Lp a a %
6~b / a&Etim- ' Ly s u,% u. Nr A e , a W 2b d Ly s u,%
/
e~m. /
- W, j
W.
,,y 6"b "
M WO / d. L N ELEWENT ,
, , /
A A
' '# ~
!EEWYNT f///////////// .
i SAN ONOFRE NUCLEAR GENERATING STATION (^'/ s, NONLINEAR SEISMIC MODEL (2-D VIEW OF 3-0 MODEL) REGION II FIGURE 4.5-12
i EMPTY CELLS CELLS WITH FUEL FUEL
@/
~
t
~
h N e S b 2ee
/ e
\ ~_
g 2
~ -
SAN ON0FRE NUCLEAR GENERATING STATION { UNITS 2 & 3
\ NONLINEAR SEISMIC MODEL PARTIAL FUEL, QUADRANT LOADING REGION I FIGURE 4.5-13
_ - - _ - _ __ - i
1 I l CELLS WITH FUEL FUEL EMPTY CELLS
'N m. 2 7
s s @m 1 e d3
~ @
U@ s g E .
)
~
y e s o v^ N 1 u SAN ONOFRE l NUCLEAR GENERATING STATION ! l
~
UNITS 2 & 3 1
- i. NONLINEAR SEISMIC MODEL 1 PARTIAL FUEL, 4 ROW LOADING REGION I FIGURE 4.5-14
__1___- - _ - - - _ _
O
""
- M.t^3f# # r.T hd&' =
r
/- :-> 13 67 ::
/
/ .,
N d"k ' d2 K
,' , ~ ,
/
/ :-: ' /<3 ad ,', ~ ,
O r, ' :..:
/
/ :..: ~
T"$ $? - :..:
/
/ :: ~
1"s ,
$? - :..: /
/ :-: ~
Ts .
^ sh -
/ s
/ :..: ~
19 s? -
- ..: /
/ :: -
1"$ .
?? -
- .c /
/ :..: ~
T@ .
$? -
- ..: /
/ :-: ~
13 ._',
- st -
- ..: /
/ :_: ~
13 -
- s? -
- ..: y O/ :.c ~
1s er -
- ..: /
/ y '
/ ~.. /
/
,/ Co
.a, to Ca <O /
//////////////////////
R ACK 2 FULL R AC K t FULL SAN ONOFRE NUCLEAR GENERATING STATION O UNITS 2 & 3 MULTIPLE RACK MODEL FULL / FULL REGION I FIGURE 4.5-15
O
""*'faV#b "#h&"'l""*' o /v /
- -: @9 -:-: /
.. . 2. , .,
/
/ :..: % , $2 - :..: /
/ :-: '
$? - :.c /
/ :: s 37 - :..: /
./
/
- ..: ~
s 32 - :..: /
~
$? - :..: /
/ :: -
$7 -
- e /
/ :.: - - $7 -
- /
/ :-: -
s-h c:c :: /
/ :..: -
- $? -
- /
O// :: ~ s? -
- /
/
'a j
on..
/
ao _ j/
/ CD G) LD G3
//////////////////////
2t%' "r'Ji' l SAN ONOFRE NUCLEAR GENERATING STATION O UNITS 2 & 3 MULTIPLE RACK MODEL EMPTY / FULL REGION I FIGURE 4.5-16
O "u'""'?,.M i;9 ,=.? wJn'"ua l
..e ' - :
/ & &
e/
/ :-: '
9 -
- -: / p'
/ :.c '
- -> ,/
e
/
/
- ~ .
^
e -
- -: y -
- .c ~
9 -
- .e /
/ x ~
x 9 -
- /
/ x ~
9d -
- -: /
/ . e b @t.d. .
L.t. ..
/ :* - d.2 rL%.
. x.
~
/
. c 12 - ., . e
/ .~, ! d -
d / E G Gd O/ x .
~
- -: /
\ /a /
/ @ '
o , GJ LD O /
/
///////////////////////
"stt "E"t l
4 SAN ONOFRE O NUCLEAR GENERATING STATION
- UNITS 2 & 3 MULTIPLE RACK MODEL FULL / FULL REGION II FIGURE 4.5-17
C-/ :: N
"""%WJb W53&i"""
c Q
' /'
x 9 -
- /
- c <
9 -
- e /
- c ' ,
9 -
- -: /
- c '
9 -
- .e /
~
'g
. 9 -
- -: /
/
x $? -
- .e f
, b tb d
/ $,c. b E. d I.-n',
b + Dd
/
/
k ~
/
V A @$d . d l.l ..
/
/.
y j OI/% .. b, D [D JJ.. & ,/
/-
/////////////////////// Et%! "#&'
SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 & 3 O MULTIPLE RACK MODEL EMPTY / FULL REGION II FIGURE 4.5-18 A__-___--____________- . _ .
MULTIPLE PACK POOL K) DEL (PLANVIEW) 7
\ -
p z l L " l I l I I l M H $ r i I I I
's N l I y
' / 'x N I
i '
/
\ f N l f l
/
N N ' 4N
/
s/ 4 f f y
/ \ / N
/ ..N / _ _
s
/ Mg N/ 4 s O '
RACK T0 RACK X-DIRECTION RACK T0 RACK i I X-DIRECTION y M H i _ _ Y-DIR __ _k _ M 4_ X o RACK 2 RACK 1 _ Y SAN ONOFRE NUCLEAR GENERATING STATION
'( UNITS 2 & 3 MULTIPLE RACK MODEL (PLAN VIEW)
REGION II FIGURE 4.5-19 . i
- l
~
- q) Displaced i Geometry- e 1
, ... ... .. ... . . . ... I
.. . . - ..- ... ... . . . . 1I I I I I I I l l -
q .. ... .. Y
.. ... ... . . . h. g..
( Weld 1l l l l l l l l l - 1 I CELL WALL r
/ i FINITE ELEMENT MODEL 1
// l Portion of Cell l
/ >
r 1 Wall Modeled
)
rdi
- w. .
\ l 1
)
p
/ '
Kb SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3
- CELL TO CELL WELD FINITE ELEMENT MODEL R E GIO N 11 FIGURE 4.5-20 l i
i _ _ _ - _ _ - - - - - _ - _ - - - - - - - - - - - - - - - _ - - - - -
y , 4 J4. 6 STRUCTURAL ACCEPTANCE CRITERIA j jb '. h' " 4.6.1
~
STRUCTURAL ACCEPTANCE CRITERIA FOR. SPENT FUEL POOL-r STRUCTURE:
'4.6.1.1 Critaria b,
The stresses / strains resulting from the loading combinations described in subsection 4.4.1 satisfy the following acceptance criteria: e spent Fuel Pool Concrf?.e'. Structure The' design stress limits described in. paragraph'3.8.4.5 of the SONGS 2&3 UFSAR were used for the evaluation of the SFP. reinforced' concrete structural' components. 4.6.1.2~ Material Properties The following naterial properties were used in the' analysis.of the SFP structure: A. Concrete Young's modulus Ec = 530,000 kips /ft2 Poisson's ratio vc = 0.17 ' f'c = 4 kips /in2 thermal expansion coeff. c = 0.0000055 O TMFWOO57 4.6-1 x.________________-_-______-___-_-
() B. Reinforcing Steel
-Young's modulus Es = 4,176,000 kips /ft2 q Poisson's ratio vs = 0.27 fy = 60,000 psi modulus of elasticity = 29,000,000 psi modular ratio = 8 C. Liner Plate System Reliner Plate: ASTM A240, Type 304L fy = 25,000 psi Liner Plate: ASTM A240, Type 304 fy = 30,000 psi Anchorage Steel: ASTM A36 fy = 36,000 psi O
4.6.1.3 Results The results are presented in this section for the design evaluation of the SFP basemat, walls, and liner plate system. A comparison between the current evaluation and the UFSAR values is presented for the pool basemat and walls. A. Pool Floors and Walls The concrete section evaluatio".s were subdivided as follows due to different concrete section dimensions. O TMFWOO57 4.6-2
- 1. North and south SFP walls
()' 2.- 3. East SFP wall West SFP wall L
- 4. SFP basemat The loading cases 6 and 7, which have large temperature gradients, have the most significant effect on concrete
, compressive stresses for both mat and wall locations. The utilization factors for the maximum stress are presented in table 4.6-1. The largest utilization factor for concrete section is 87.2% (at least a 12% margin remains against the section allowable), which resulted from loading case 6. The utilization factor is defined as the percentage of resistance of the reinforced concrete section that has been utilized relative to the zero () curvature line. A utilization factor of 100% indicates that the section is fully utilized by the design load. B. Liner and Anchorage The existing liner plate system is evaluated for the new spent fuel rack induced loads and for two postulated load
- drops. Both local and overall effects on the liner plate system are evaluated. The ACI and AISC codes were used for the liner plate evaluation.
O TMFWOO57 4.6-3 L__ _ _ - _ _ ___-_ -_-_- _ __ _ ____________-_____-__--_______- -________ _ _ ____
Local effects due to the' rack vertical loads are etaluated based on the worst case single support pad loads"for the O spent fuel racks-for,both Region I and Region II.- The evaluation shows that the single support pad loads are acceptable-except when applisd directly over'or; immediately adjacent to the leak chase-channels. The support pads over or adjacent to the leak chase channels will.be.provided with load spreading or bridging plates to
~
assure the. actual' concrete bearing is within the ACI. allowables. overall rack horizontal loads due to friction were evaluated for both Regions I and II racks in the SFP. The evaluation shows that the liner plate system can withstand the loads imposed by the new spent fuel racks without any required liner plate system modifications. Actual horizontal rack loads 5,132 kips Reliner plate capacity 20,500 kips original liner plate capacity 19,330 kips. Anchorage system capacity 8,240 kips The liner plate and anchorage system capacities are based on AISC material allowables increased by a factor of 1.6-for DBE conditions as allowed by UFSAR paragraph 3.8.4.5 [e.g.: tension = 1.6 (0.6 Fy )). O TMFWOO57 4.6-4
The original analysis for thermal effects on'the SFP liner plate was based on conservative parameters with a pool temperature of 220F and an initial unstressed liner plate temperature of 60F resulting in a temperature differential of 160F. The actual governing maximum pool design basis temperature is 212F and the initial unstressed liner plate temperature is 65F (resulting in a temperature differential of 147F), which creates a less severe condition than that originally analyzed. Therefore, the results of the original analysis remain valid, i.e., the liner plate will not buckle due to design thermal conditions. Additionally, load drop evaluations were performed for lifted items to determine their effect on the liner plate. These items were the SFP gate, a test equipment load, and an empty spent fuel rack. The Region I type rack governs over Region II type rack because it is much heavier. A minimum water depth of 40 feet was assumed. The evaluation was performed in accordance with BC-TOP-9A. The analysis indicates that the liner plate could be perforated due to a SFP gate drop (maximum concrete penetration about 1-1/2 inches deep by 31 square inches) or an empty spent fuel rack drop (maximum concrete penetration about 5-3/4 inches deep by 63 square inches). The test equipment drop is enveloped by an empty spent fuel rack drop. O TMFWOO57 4.6-5
- {'
The immediate loss of water due to liner plate perforation is;approximately 11-gallons which is the quantity ofiwater to fill the applicable leak chases.- The worst case rate of discharge to the leak detection sump is less than 49 gallons per minute. The 49 gal / min rate is' based on the maximum flow achievable-(with maximum pool head and-unrestricted: flow into the leak' chase channel) through the leak chase discharge line (3/4-inch pipe)'into'the detection sump. The worst case concrete penetration is approximately 7% of the basemat thickness; therefore, the water will be maintained within the pool / leak chase system. ..The maximum loss of water within the pool due to the postulated - liner plate perforation - (49 gal / min).- is - well'within the makeup water supply of 150 gal / min-(SONGS 2&3 UFSAR, table 9.1-2), thereby assuring that the Technical Specification water level within the pool (s) can be maintained. i This evaluation shows that no modifications to the liner plate or building are required to accommodate the new spent fuel racks being added to the SFP. 1 l-C. Foundation Stability and Soil Bearing As shown in the SONGS 2&3 UFSAR, table 3.8-19, the worst case foundation bearing pressure for the fuel handling building is 21 kips /ft2 (seismic loading case). The O TMFWOO57 4.6-6 ________n______._____ _ . _ _ _ __
+ ,
4 allowable bearing pressure for this'casalis 44 kips /ft 2 , The allowable is more than twice as"large as the UFSAR V)
,/'
actual. Since the total change in mass due to the addition'of the high density spent fuel racks.is less than-74 and dynamic characteristics are slightly impacted (frequency shifts less than 13%, see table'446-3), the resulting increased bearing pressure for.the. current evaluation will remain'we11'within the allowable.. D. UFSAR Comparisons The table 4.6-3 comparison shows that the finite element model (previously discussed in paragraph 4.5.1.1) shown in figures 4.6-1 through.4.6-3 behaves very similar to the lumped parameter model shown in'the SONGS 2&3 UFSAR figure () 3.7-22. The dynamic characteristics of-the current finite element model-(FEM) are compared with those stated in UFSAR table 3.7-10. The two dominant modes for each direction are compared. The model (lumped vs. FEM) comparisons show that the FEM accurately depicts the FHB and the differences in the modal frequencies (FEM lower than lumped model) are as expected (more mass - lower frequency). The verified FEM was used to compute the resulting stresses from the new spent fuel racks and additional stored fuel for the structural evaluation of the FHB. O TMFWOO57 4.6-7 (
The governing results for this evaluation are compared against the governing UFSAR values given in the SONGS 2&3 O UFSAR table 3.8-10. The table 4.6-2 comparison shows that the current evaluation of the SFP basemat and walls results in reduced section moments and membrane forces due to the refined analysis and design techniques discussed previously even though the loads due to the spent fuel racks increased. The SONGS 2&3 UFSAR, table 3.8-10, lists a value of 224 kips /ft as the maximum computed basemat shear which corresponds to an allowable value of 304 kips /ft, resulting in a 26% margin. The shear value listed is for a cantilever portion of the basemat away from the SFP area and would not be affected (in shear) by the added mass within the pool which is bounded by heavy shear walls. However, the basemat shear within the pool boundary (pcol floor) would be affected and the maximum computed value is 83 kips /ft which corresponds with an allowable of 119 kips /ft, resulting in a minimum 30% margin. Table 4.6-2 shows that the current evaluation results in increased margins (differences between code allowables versus actuals). For example, the governing basemat load interactions per the UFSAR are at approximately 98% of capacity (a 2% margin), whereas the corresponding value for the current evaluation is approximately 87% capacity (a 13% margin). TMFWOO57 4.6-8
)
m. i3 E. Comparison of Floor Response Spectra Evaluations were performed to' determine the' effects of the high~ density spent fuel racks on the floor response spectra (FRS).of.the FHB. The evaluation covered-both'OBE and DBE responses with comparisons being made for 2%. damping at three representative elevations. (basemat:' 17 feet-6 inches, pool deck:- '63 feet-6 inches, and roof:~ 114; feet-0 inches). All comparisons were made using unwidened spectral' data. The computer model,. input time-history record, computational techniques, soil data, composite modal damping;and spectra combining methods were all consistent with the methodologies employed and data used in the original UFSAR analyses-(UFSAR figure 3.7-22). O The increase in total mass from the addition ~of high density spent fuel racks-(about 7% for the vertical-direction and about 3% for the horizontal-direction) produced a maximum frequency shift of 3.2% for the dominant vertical structural mode. The-frequency shift for'the combined horizontal and vertical modes were less. From a comparison of the 2% damped instructure floor response spectra for both the original UFSAR analysis'and the current BSAP analysis, performed to incorporate the effects of the high density spent fuel racks, it was seen that the gross response of both cases are very similar O TMFWOO57 4.6-9 l
---___ _ -__ A - _ _ - _ _ _ .
with the differences being very minor and insignificant. Figures 4.6-4 through 4.6-9 show the comparisons for the DBE responses at the basemat and the top of the structure l (these are typical comparisons). A comparison of the peak responses for both analyses showed that in most cases the origina1' analysis exceeded t the current analysis by values ranging up to 15%. Of the eighteen sets of spectral output being evaluated (two horizontal and one vertical response for both DBE and OBE inputs at three separate elevations: 3 x 2 x 3 = 18), only five cases resulted in the current spectral peaks being greater'than the original computed values. Four of the five instances are associated with the DBE (vertical spectra at all three elevations and east-west horizontal at the basemat) and one with the OBE (north-south horizontal at the basemat). The magnitude of the four exceedances associated with the DBE are 3% or less while the one OBE exceedance is 10%. A comparison of the responses at the zero period acceleration (ZPA) level showed that the two analyses are essentially the same for all cases compared except that the current analysis exceeded the original analysis by less than 9% for the DBE vertical response (no difference in OBE vertical response) for all three elevations empared. TMFW0057 4.6-10
p
.Since'all.of-the .'above. comparisons.were'made'with the- d
- unwidened-(raw); spectra,-it is n'oted that much of the differences would_ effectively. disappear ~inu the process of .l
. 1 widening, enveloping and. smoothing the'FRS curves.
L e
'4. 6. 2 : LSTRUCTURAL ACCEPTANCE CRITERIA FOR SPENT' FUEL STORAGE' RACKS -
4.6.2.1 Criteria ~
-. There . are 'two sets of criteria: to: be satisfied b'y the racks: .
l A. Kinematic. Criterion-O
'This criterion seeks to' ensure that the rack is a physically stable structure. The SONGS racks are evaluated for margin against overturning'and also for rack
. displacement to ensure that rack to rack and rack to pool _
wall impact does not occur. P O TMFWOO57 4.6-11 ___._______m__.__________________
'a i .
[- 4 .B. ' Stress' Limits
!1 Tl The stress limits of the.ASME Code, Section III~,
Subsection NF,1 1986 Edition up to and' including A-86 Addenda are used since this code provides the most y appropriate and consistent? set of111mits for various
-stress types and-various' loading conditions.
4.6.2.2 . Stress' Limits for Specified Conditions The. structural analysis of the SONGS racks.is a-limit type. analysis per ASME Code, Section III,-Subsection NF,. paragraph-
'NF-3340(9). Per the-requirements of the SRP (NUREG-0800)f, Section 3.8.4, Appendix D.the normal plus upset loads"are multiplied by'l.7 and the faulted loads are multiplied by 1.1 to.
1 getithe-" factored". loads. These factored loads are then compared
'to the;" limit'.' load . (also called the lower bound collapse -load) to obtain the Margin to Allowable. The limit load is calculated using the yield stress:from ASME Code,lSection~III, Appendix.A at 150F and assuming elastic perfectly plastic material (no strain I
hardening). Welds are proportioned to meet the requirements of p. NF-3342. 2 (e) (4) . The Load Combination and Acceptance Limit table for limit analysis from the SRP (NUREG-0800) Section 3.8.4, Appendix D, Table 1.is shown in table 4.4-1 and repeated below. O TMFWOO57 4.6-12 l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ = _ _ _ _ _ _ _ _ _ _ _ _ ____ _______ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ Limit Analysis Load' Combination- Accentance: Limit _
- 1. - 1.7 (D+L) -NF.-3340 of.ASME
- 2. .1'.3 (D+L+To) Code Section III
- 3. 1.7 (D+IAE)
- 4. 1.3 (D+L+E+To)
-5.- 1.3 (D+L+E+Ta)
- 6. 1.3 (D+L+To+Pf)
- 7. 1.1-(D+L+Ta+E')
Abbreviations are defined'in paragraph 4.4.2.1. Note.that SRP 3.8.4 Appendix D lists XVII 4000 rather than NF 3340 for the acceptance limit'. However, Appendix XVII has now been incorporated'into Subsection NF and what was XVII 4000 is now NF c' 3340. Margins to Allowable shown in tables 4.6-4 and-4.6-5 are for the
~
. limiting load combinations 3, 5, anf'7. 1The Margin.to Allowable (MA) is calculated, as shown in-equation form below, by comparing ;
the acceptance limit with the applied load.- The acceptance limit is the limit' load of the structural component (see NF 3340 of ASME Code Section III),.and the applied load is the factored load obtained from the load combinations specified above. Since the-acceptance limit.is the same for all seven load combinations, it
~
is possible to meet the requirements of the load combinations by addressing three limiting combinations. lLoad combination 7 is
~ limiting because-it is the only combination involving the DBE O TMFWOO57 4.6-13
l condition. Load combinations 3 and 5 are the limiting combinations of combinations 1 through 6 as shown in the G(~N following paragraphs. Load combination 3 [1.7(D+L+E)] envelops load combination 1
'[1.7(D+L)].
Load combination 5 [1.3(D+L+E+Ta)] envelops' load combinations 2 [1.3(D+L+To)] and 4 [1.3(D+L+E+To)]. Since stresses caused by the stuck fuel assembly load condition.(P f) are much lower than stresses caused by the'OBE (E), load combination 5 also envelops load combination 6 [1. 3 (D+L+To+Pf) ] . The two columns on tables 4.6-4 and 4.6-5 are labeled OBE and DBE. The column labeled OBE is the MA for either load
/"N combination 3 [1.7 (D+L+E) ] or_ load combination 5 [1.3(D+L+E+Ta)s b whichever is the more limiting condition. Except where indicated on the tables, load combination 3 [1.7(D+L+E)] is more limiting.
The column labeled DBE is the MA for load combination 7 [1.1(D+L+Ta+E')]. The MA shown in tables 4.6-4 and 4.6-5 is defined as MA = Allowable Load - 1 = Limit Load 1 Applied Load Factored Load Specifically, for the two reported conditions: tO
\- / TMFWOO57 4.6-14
- _ _ = _ _ _ _ - _ - _ _
- k. where load combination 3 [1.7(D+L+E)] is limiting -
MA = Limit Load - 1 OBE 1.7(D+L+E). where load combination 5 [1.3(D+L+E+Ta)] is limiting - MA = Limit Load -1 OBE 1.3(D+L+E+Ta) DBE i MA = Limit Load -1 DBE 1.1(D+L+Ta+E') 4.6.2.3 Results for Rack Analysis Tables 4.6-4 and 4.6-5 show the minimum MA for the various components and welds on the SONGS racks.- The adequate margin in each case shows that the racks meet the structural requirements of the ASME Code. In addition, the impact loads on the fuel assemblies due to the interaction with the rack during a seismic event have been determined. The maximum calculated seismic impact load at a O TMFWOO57 4.6-15 lI L___-____-__-_________
I spacer grid location is 2142 pounds, which is less than the allowable spacer grid strength for the more limiting C-E 16 x 16 fuel assemblies (24), 4.6.3 SPENT FUEL HANDLING MACHINE (SFHM) UPLIFT ANALYSIS i An analysis was performed to demonstrate that a rack can withstand an uplift load of 6000 pounds produced by a jammed fuel assembly. Using worst geometry assumptions, the stresses resulting from this load were calculated and compared to the acceptance limits. This loading condition was determined not to be a governing condition and is covered by the results reported
- in tables 4.6-4 and 4.6-5 for the limiting loading combinations.
In addition, since the gross stresses remained within the elastic regime, there is no change of rack cell geometry of a magnitude sufficient to cause the criticality acceptance criterion to b'e violated. 4.6.4 FUEL ASSEMBLY DROP ACCIDENT ANALYSIS 4.6.4.1 Statement of Problem 4.6.4.1.1 Drop Cases Two cases were considered for the accidental drop of a fuel assembly onto or into the racks. These were: TMFWOO57 4.6-16
l I A. Westinghouse 14 x 14 standard fuel assembly with control rods, total dry weight of 1260 pounds, dropped from a conservative height of 24.9 feet above the pool floor, B. Combustion Engineering 16 X 16 fuel assembly with control rods, total dry weight of 1540 pounds, dropped from a conservative height of 21.7 feet above the pool floor. 4.6.4.1.2 Drop Orientations i Three orientations of drop were considered. These were: A. Drop of an assembly onto the top of the racks with the assembly in a vertical position, O B. Drop of an assembly onto the top.of the racks with the assembly in an inclined position, and C. Drop of a fuel assembly through an empty cell to the bottom of the pool. 4.6.4.1.3 Acceptance Criteria The acceptance criteria used were: A. Fuel criticality does not occur, and B. Perforation of the pool liner does not occur. O TMFWOO57 4.6-17
' .N 4.6.4.2 Model Definition 4.6.4.2.1- Assumptions for Energy Dissipation a
i For evaluation of the cases defined above, the following general ; assumptions for_ energy dissipation were made: A. The fuel assembly falls-freely ~in an' infinite pool of static water with hydrodynamic drag being considered,- B. No energy is dissipated in the rack structure during the drop, C. The pool liner and floor flexibilities are neglected. The only flexibilities considered are those component parts at
)
the bottom (impact) end of the' fuel assembly, D. No energy is dissipated in the fuel rods, and E. the kinetic energy of the fuel assembly is totally converted into strain energy of the assembly structure. 4.6.4.2.2 Assumptions for Drag Determination The mathematical model used to evaluate the impact velocity was based on the following conservative assumptions: O TMFWOO57 4.6-18 i 1 1 i
i A. The drag coefficient used was that for a flat plate normal l to the flow direction. The minimum value found in the literature (10), which represents flow at high Reynolds numbers, was used and was taken as a constant for the entire drop event. This is conservative since'the drag coefficient decreases as the flow velocity increases. B. The minimum frontal area of the fuel assembly was used to determine the drag force. Work-energy relationships were used to solve the nonlinear kinetic equation relating fuel assembly weight, buoyancy, velocity, and drag. The finite difference technique was used to integrate the drag force term. The increment size was varied to determine that the solution had converged properly. O 4.6.4.3 Droo Analysis Results 4.6.4.3.1 Satisfaction of Criticality Criterion Criticality calculations show that with 2000 ppm Boron in the fuel pool water (the normal condition is a minimum of 2350 ppm), fuel criticality does not occur. Thus, for the fuel drop accident the presence of the Boron ensures that the criticality criterion is satisfied for all cases. O TMFWOO57 4.6-19
4.6.4.3.2 Satisfaction of Pool Liner Integrity Criterion es U Drop orientations A and B are considered together since the same philosophy covers both cases. For these cases, either the fuel assembly will remain on top of the racks after impact or will impact the pool floor with a lower velocity than the drop through I case since part of the potential energy will be absorbed by the initial impact with the top of the rack. Therefore, perforation of the pool liner is enveloped by the case of drop of a fuel assembly through a cell. Each of the three cases (see paragraph 4.6.4.1.2) was evaluated to determine the velocity of impact with the pool liner. In each
~
case the structure at the lower end of the assembly, i.e., bottom nozzle, guide tubes, etc., had enough strain energy capacity to () absorb the drop kinetic energy. When consideration was given to the " footprint" of the dropped assembly, the stresses imposed on the pool liner were determined to be 43% of the ASME Code Allowable Limit for Faulted Conditions. The pool liner will therefore not be perforated for any of the drop accidents. 4.6.5 OTHER EQUIPMENT DROP ANALYSIS There were two types of analyses performed for drops onto the SFP racks (see figure 4.6-10). The first analysis postulated a drop of the SFP gate. The second analysis examined a drop of test equipment load. These analyses are discussed below. A V TMFWOO57 4.6-20
f A. Spent Fuel Pool Gate' Drop Load Analysis ( Evaluation of a pool gate drop was based on a drop height of.30 inches above the top of.the rack. This height is, administratively controlled by permitting a maximum vertical clearance of 10 inches between the bottom of'the-gate and the bottom. ledge of the gate opening until gate; is laterally' moved clear of the SFP. The dimensions of the-gate'are'41.0 x 343.5 x 0.75 inches. The weight of the gate is 4500 pounds.
.The' maximum penetration occurs for the case.when:the-gate impacts'the rack at 45*. The resulting penetration depth-i.s 21.2 inches. This results in potential fuel. damage in
)' six cells.
The amount of penetration was determined from
" conservation of energy"; i.e., the energy absorbed through plastic deformation:of the rack was equated to the change in potential energy of the gate. Because all the deformation was assumed to occur in the rack and drag due !
I to water was ignored, a conservative upper bound on the ) a' penetration was determined. i Energy absorbed by the rack was based on a " knife-edge" 1 1 penetration of the cell wall. Because the gate thickness I I
-4 is only 0.75 inches, the force to initiate this type of O TMFWOO57 4.6-21
= _ ___ _ ___ _ _ . - - - - _ _ _ _ _ - .
t
.ti E-Penetration:is significantly less than any'other mode of- q penetration. The absorbed energy-was calculated by
\ conservatively assuming. perfectly plastic deformation at a. {
l
- load which results in'a. shear-stressiequal'to 57% of the minimum compressive yield strength of-the.' cell wall.. The i
. .. I
. impact location;was selected such that the maximum number
~
of fuel. assemblies was.affected. i i These calculations were done for.a Region II-rack., This y region is limiting because it.has only one' cell walli ' between adjacent storage locations, whereas Region I has .
.two cell walls between adjacent storage locations.
A drop analysis was alsol performed for the SFP' gate impacting'the. pool floor liner. The analysis assumed a
,Q drop height of 50 feet 6 inches (from 4 feet 6. inches. .j V above the, pool-deck) and evaluated the consequences on the pool floor liner system and the basemat. concrete. The
~
results of that analysis are bounded by the. rack drop-l results presented in paragraph 4.7.4.4. The radiological consequences of this drop are presented in subsection ; i 5.3.6. .j B. Test Equipment Load Drop Analysis I. The test equipment load drop analysis assumes the drop of a 4500 pound piece of equipment from a height of 47 feet above the pool floor (administrative 1y controlled to be 1 , l l O _ 005, 4. .
o ' f
~ ~
~
foot above the operating floor (elevation 64 feet 6 inch'es)). -The test equipment' consists of;a 4-foot by [} 6-foot base'with a 200-inch long vertical H-beam attached to'the base at one of the 4-foot edges. Additional equipment is attached to both the base and the H-beam. The height of the water in the pool during the drop accident is l assumed to be 40 feet. .'Therefore, the equipment will fall 7 feet and then enter the water. -Then:
.the equipment will' fall through the water until it impacts the tops of the racks which are at 16 feet 4 inches above the. pool floor. . Some conservative drag calculations were made for this piece'of equipment. These resulted inLthe equipment impacting the top of:the racks with a velocity lof approximately 206'in/s.' The kinetic energy of the.
equipment is.then converted into strain energy.in the. rack-structure. calculations were made to determine the load required to compressLa fuel. rack cell. As-the cell is displacedfthe load on the cell will rise. - At'some point the sides.of the cell will locally buckle.. This does not result in collapse of the cell but.only means that the cell walls function at'a reduced effective-width. Beyond this point the load will still rise but at a slower rate than before the cell-walls locally buckle. When the yield point of the cell is reached then the local-buckling increases rapidly and the load the cell can withstand-decreases to a l lower value. Then the cell load decreases very slowly as O TMFWOO57 4.6-23 _o____1_____ . _ _ _ _ _ _ _ _ _ _ _ __ _ __ l
the cell is compressed. The penetration into the rack top 7-s if the equipment base.is conservatively assumed to be at
! an angle with the horizontal of.45' when it impacts the rack was calculated. The maximum penetration in this case is approximately 16 inches.
The top of the Unit 1 fuel assembly is approximately 51.5-inches below the top of the rack. Therefore, this drop will not result in damage to Unit 1 fuel assemblies. The top of the Units 2 and 3 fuel assemblies is approximately 13.2 inches below the top of the rack and this drop would result in damage to 14 Units 2 and 3 fuel assemblies. An additional analysis was made to determine the maximum drop
' weight under..which no fuel assembly damage results. It was determined that for a drop height of 72 inches above the top of the rack the test equipment will' impact the top
~ of the rack with a velocity of 177 in/s. The penetration into the rack top if the equipment base is conservatively assumed to be at an angle with the horizontal of 45' when it impacts the rack was calculated. The maximum penetration in this case is 13.0 inches. The top of the Units 2 and 3 fuel assembly is approximately 13.2 inches below the top of the rack. Given the drop penetration of 13 inches no fuel damage occurs. O TMFWOO57 4.6-24
.g ,
.q
, .g These.' calculations were done for'a. Region'II. rack. -Since this type of rack has-only one-cell wall between' adjacent L
storage [ locations'and the Region..I rack has two cell walls h between adjacent storage locations', the Region II rack'ist
.the limiting casa..
Administrative controls will'be. implemented to provide assurance'that the: radiological consequences of these: drops are acceptable. The administrative ~ controls are presented in subsection 5.3.5. 9 The drop'of the test equipment onto the SFP floor'was-investigated. It has been shown that the results of.such a postulated event would be' bounded by the rack drop'. analysis presented in paragraph 4.7.'4.4. .O 9 4.6.6 RACK- DISPLACEMENTS - From the nonlinear time history analysis, the maximum Region I.
-rack displacement (absolute displacement) was determined to be 1.80 inches in the' east-west direction and 1.50 inches in the north-south direction. For the Region II racks, the maximum displacement was determined to be 1.39 inches in the east-west direction and 1.44 inches in the north south direction. The remaining rack.to pool wall gap is calculated by taking the nominal initia1Lclearance between the rack and the pool wall and 1- then subtracting the installation tolerance (0.25 inches),
O TMFWOO57 4.6-25
fabrication tolerance, total thermal growth of one rack (0.10 inches), and the seismic displacement of the rack. The minimum
)
remaining rack to pool wall gap is determined to be 1.90 inches and is based on the nominal initial rack to wall gap of 3.75' inches in the north-south direction for a Region I rack and the seismic displacement of 1.50 inches'(table 4.6-6). It is noted that the maximum displacement of 1.80 inches for Region I east-west direction does not produce the minimum pool wall gap
~
because of the large (11.10 inches) nominal initial clearance.
/
The most limiting relative displacement between racks as determined from the time history results is 1.39 inches. (A larger relative displacement of 1.73. inches occurs in the east-west direction between the two Region I racks. However, the gap between these racks is large and this case is not limiting.) () Using the appropriate nominal initial clearance between racks of. 3.44 inches and then subtracting the installation tolerance (0.25 inches), fabrication tolerance, thermal growth of two racks (0.20 inches total due to 0.10 inches per rack), pool construction tolerances (which may reduce rack to rack gaps a maximum of 0.34 inches), and the seismic relative displacement between racks (1.39 inches), the remaining rack to rack gap is determined to be 1.26 inches (table 4.6-6). From these results it is concluded that the racks are spaced with sufficient clearance so that rack to rack and rack to pool wall impact does not occur. O TMFWOO57 4.6-26
- - _ - _ _ - - - _ - _ _ _ _ - - - _ _ - _ _ - _ - - _ - - i
Also extracted from the time history results is the maximum support pad. vertical displacement (lift-off) . - The_ maximum. support pad lift-off is found to be 0.23 inches. For this magnitude of pad lift-off the. factor of safety against rack overturning 'is determined to be greater than 48 which satisfies the requirements of Section 3.8.5.II.5 of the SRP. 4.6.7- RACK IhCATION VERIFICATION The applicable plant procedures which govern activities after a seismic ~ event will be revised to include a requirement to perform a walkdown of the SFP to check the rack configuration. This walkdown will be performed after confirmation of an OBE event. O TMFWOO57 4.6-27
l . Table 4.6-1 i CURRENT EVALUATION RESULTS FOR THE SPENT FUEL POOL WALLS AND BASEMAT Utilization Factor *,(%) North and South' Walls:
. Horizontal Reinforcement 38.0 Vertical Reinforcement 29.4 East Wall:
Horizontal Reinforcement 18.1 Vertical Reinforcement 40.4 West Wall: Horizontal Reinforcement 20.9 Vertical Reinforcement 65.1 Basemat: North-South Reinforcement 50.5 East-West Reinforcement 87.2 O
- The Utilization Factor is defined as the percentage of resistance of the reinforced concrete section that has-been utilized relative to the zero curvature line.
TMFWOO57 4.6-28
r
. Table;4.6 COMPARISON OF GOVERNING RESULTS FOR p THE ORIGINAL DESIGN VERSUS THE CURRENT EVALUATION FOR THE SPENT FUEL POOL-
. Max Axial Flexural Flexural Location Governing Load Load Load in Spent- Load Pu Mu Mu(Max) .Mu/Mu(Max)
Fuel Combination (kips) (K-ft/ft) (K-ft/ft) Pool (a) (b) (c) (d) 7-Foot Thick UFSAR 7 -527 2604 2660 0.98 Basemat in Pool Area . CURRENT 6 102 1546 1767 0.87
.(E-W- EVALUATION Reinf) 4-Foot Thick 'UFSAR 7 -404 445 947 0.47 (N or S)
Spent O -Fuel Pool U Wall ' CURRENT 7- -239 195 665 0.29 (Vert EVALUATION Reinf) 5-Foot Thick -UFSAR 7 0 666 674 0'.99 (West) Spent-Fuel Pool Wall . CURRENT 6 202 191 266 0.72 (Vert EVALUATION Reinf)
- a. The UFSAR values are from UFSAR table 3.8-10. The current evaluations are the maximum values obtained and not necessarily at the previous locations.
b.. Refer to UFSAR paragraph 3.8.4.3.2.A
- c. Sign convention for Pu: Compression (-), Tension (+)
- d. Maximttm flexural interaction capacity (Mu(Max)) given the axial load shown (Pu).
TMFWOO57 4.6-29 l C_ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ I
l f_ Table 4.6-3 COMPARISON OF MODAL CHARACTERISTICS FOR THE LUMPED s PARAMETER MODEL VERSUS THE CURRENT EVALUATION Frequency (Hz) Participation (%) Critical Damping (%) Direction UFSAR Current UFSAR Current UFSAR Current East-West 2.38 2.53 48.6 54.9 9.79 9.72 North-South 2.58 1.67 41.9 34.5 9.87 9.87 Vertical 2.58 2.67 10.0* 25.4* 9.87 9.87 Vertical 2.99 2.89 83.9* 71.5* 9.98 9.93 North-South 5.53 4.84 37.4 36.2 9.94 9.21 East-West 5.85 5.16 38.1 38.8 9.87 9.77 UFSAR values in the above table are obtained from UFSAR table 3.7-10.
- The summation of participation of vertical masses for the UFSAR and the current evaluation are approximately the same. The frepencies for the two modes are closer spaced for the current evaluation which accounts for the participation shift.
ll qgg TMFwooS7 4.6-30
Table 4.6-4 MINIMUM MARGIN TO ALLOWABLE REGION I OBE(a) DBE(D) Support Pada 0.89 0.40 Cells 0.78 0.32 Grids 2.42 1.38 Cell to Cell Clips 1.49(c) 2.46 Welds Cell to Grid 0.46 0.21 Cell to Clip 1.67 0.55 Grid to Grid 4.04 1.90 Grid to Base Plate 1.70 0.71 Cell Seam 1.22 0.30 Cell to Wrapper 0.63 0.41
- a. Load Combination 3 [1.7(D+L+E)] unless otherwise specified.
- b. Load Combination 7 [1.1(D+L+Ta+E')).
- c. Load Combination 5 [1.3(D+L+E+Ta)3-TMFWOO57 4.6-31
- f. ..
2r~R-L) ' Table 4.6-5 i MINIMUM MARGIN TO ALIOWABLE q REGION II OBE(a) DBE(D) Support Pads 0.84 'O.35 Cells 0.74 0.27 Welds Cell to Base Plate 1.54 0.56-Cell'to Cell 0.68(c) 0.80 Cell Seam 0.85(c) 0.27 Cell to Wrapper 1 0.61 0.40 O
'n. . Load Combination 3'[1.7(D+L+E)] unless otherwise specified.
- b. ' Load Combination.7'[1.1(D+L+Ta+E')].
- c. Load Combination 5 [1.3(D+L+E+Ta)3-O TMFWOO57 4.6-32 m_.______ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _
) Table 4.6-6 RACK GAP SPACING RESULTS (Sheet 1 of 2)
Relative Gpp Item Absolute (inches)(Gpp ai (inches) (b) Initial Nominal Gap 3.75 3.44 Installation / Tolerance 0.25 0.25 Fabrication Tolerance (c) _o_ _o-Pool Construction Tolerance (d) --- 0.34 Calculated Thermal Growth 0.10 0.20 Reduced Gap 3.40 2.65 Calculated Seismic Displacement 1152 1x12 Remaining Gap (e) 1.90 1.26
- a. Rack to Wall
- b. Rack to Rack
- c. The summation of the dimensions of the widest sections of the racks (the base plate dimension) in the north-south and the east-west directions will be maintained less than 510.36 inches and 267.34 inches respectively (Region II).
Application of these values as maximums eliminates the need to include fabrication tolerances in the limiting gap calculation. Any slight variations in these values will be accounted for in the existing non-limiting gap locations, and will not affect the remaining gap dimensions.
- d. The pool construction tolerance was equally divided between the rack to rack gaps and not the rack to wall gap in the
' north-south directions. l In the non-limiting east-west direction, the pool construction tolerance was included in the rack to wall gaps and/or the rack to rack gap. O TMFWOO57 4.G-33
L 1 Table 4.6-6 RACK GAP SPACING RESULTS (Sheet 2 of 2)' l a. It 'should be noted that the reported calculated seismic
. displacements of the racks are the maximum values of rack displacement at the top.or bottom of the rack. The minimum reduced gap'is based on the rack base plate dimension which is the widest section of a rack. Therefore, at the top of the rack where the initial gap is larger, the remaining gap will be greater than that calculated by the above described method. (In other words, the plane of the rack wall will be held within the base plate dimensions including fabrication /
assembly tolerances.)
;-~ ,
V l TMFWOO57 4.6-34
Tc B l l-l l, l I i
'l o
I u ^t 1 l' I < .) l
/" r r- ~ == _ _ _
'=* wr+D,)7
-g/E -- _ . i _
pf47=- )lL24@i&jk;cc,&_ 5 4 . p @ 1:9 I 9 l# - =f"? h ,
-g ,$ $[f ik '
'h .-
~
-df ~h . .J ,
I
- J .-
t 7: fil , ; (( hy
)
O W [h-- ) 12 r= 4
;ikt ii
; g
==1
_5 __ i
- i. , i S Ah j h! [-, hk, h - --[
= = --
)===-
W==_= .- ,g= &- =.-- - ll:" H~;;? - -. l.l
,=- .-
SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SONGS 2 AND 3 FREE VIBRATION 0- ANALYSIS (ALL STANDARD FUEL IN SFP) ELEVATION LOOKING NORTH MODE NO. 4 FIGURE 4.6-1
^
0 l l I I , I ' I I I I i l I
! ' /
r ! ! , i g -4 -
/
M iL.~" E 1_ ~_lAl IY h [y'p '#j y-- K 9E I MbbbiNl :
, % ._ __ _ i
_ IM l'L Vi , ,1,,[ ,,-- y r I.
.l I[h M.u-wqp hN U . :.D-Ep-ps gl l h,,4;, ? '- - -1"L---J Ji i'
_ilt l
;, ggl _la . p p .,_ ..ut.
lL t i ' pi ,, ( ,fM-H 1- - 2 l S ( {'. ca L L h < -1 i i 6M4.Ciflg7J -NU'-- M3 IA T - f. . Liv m. PTh ll ka (, .Q
~
e 4f
! j_ _ __
l - I E 9 _T ___ 0dif' ru ,- ,-- s}_ _ m z , g = e- - Tf7* $s q/T~ - ? T' MM 1
'l i , y- ,
, 7-
.. y;W q sjt.:- a d.--_J-d4 - -
(' __ pg~,2
/I pl4_-.4 9
/M305=P-f- . - - - -f\ _._M(. M iL d_.h 1::: o i / T ---
> ni 2d / '~ "' !
e W-nyn-
'r
- U-h=f' G;p. --'jfl*
l N=:"" ~ _
%' 'l3
\! ;;
lf;' g Id% iV k=.{pr?._
~*- *=&%Yff.ri&2ff_-iWf_=_,
SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SONGS 2 AND 3 FREE VfBRATION ANALYSIS (ALL STANDARD FUEL IN O SFP) ELEVATION LOOKING EAST MODE NO. 5 FIGURE 4.6-2
i y
- I I :1 4
g I I \ 1
\.
- 5. l -
'I \
l - I I I
-L.
/ mo ._) > > ,
A - _ . _4 _;,. .
- . _!__ _ _, )
_ha=;: ,ejy:pr -- - ~- .. _ _ ,rt : izwn - - - di pr_--q' ", - - mm J,f,=
-s , _
- y. --.
~
N,;.d s y-. r-O w M- /r: ==% Q il ___gd g .a_ _p_4_ p+L.J ._ u J. .p. j. ._. aj _ij _. # $ -tt81 :3.Lx:m
.._uS: _u F_4_%pp -h _
l(J _.-___4'! -
- li L . i .
r i Lg f 0i U ._l5! N_ ._ J J_ _UJ. p E
'IW 'Il I(
~
1p1,g$gp*- 1 I. . ,= _ .g 4/e s. ire.=p
= = = .- ga c g y_ g- (l-
_ . J- -# ! _ -fP b
._ ,1. ;iJ .__
i; r 3- ~"# g! i , r' _1f- _ -it
.- =
.w.=. ._--y ,=_
= a
,k- -
%.gv9'
,gaW8E\ .sM.- h%}f t L lj =_Jy c I g g.~ *= _ i - - _ m __ _ g
- <_--_:_ _ . . _ _ L_ .w: )f Wr-m g 8 8".' "" ==
E2 = = ==gCC 2.b~=E?P.?==fEC_is.lC_- ._V_ __v . .= - i=E2.u. _ .O SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SONGS 2 AND 3 FREE VIBRATION " -[ ANALYSIS (ALL STANDARD FUEL IN SFP) ELEVATION LOOKING EAST MODE NO. 6 FIGURE 4.6-3 0 j n .0 / / / \ ~ { f' N Y m / in J> I O D i Z r ?l - 0 v - 7,' O O M Q E w -. 1 ~ -Ec , ) ~ _ao (2 5 1 1 I [ l j ~ f O O Q CO N O LO t O N - C * (0) N011V831300V SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 lO , SONGS 2 AND 3 FUEL BUILDING RAW FLOOR SPECTRA EL.1T 6" DBE HORIZONTAL (N S) FOR 2% DAMPING FIGURE 4.6-4 l- ~ .O / " I \ p'~) l 'O. ( 1 ( i ) { I /' W m / (n > O j ' Z 4 % 8 . LJ d (A -g } Q 'N - .A y _ -m ; . Z5W L -Es -DO - l' I I I " o O O O N c 0 v' n N e C - (D) N011V831300V SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 O SONGS 2 AND 3 FUEL BUILDING RAW FLOOR SPECTRA EL.17' 6" DBE HORIZONTAL (E W) FOR 2% DAMPING FIGURE 4.6 5 ii i 0 _' 1 x _ ^ _ k ' m ' 1 _ S' ) _ S _ \ 3 D _ N _ b i O _ s C _ \ \ E i l S - / ( / D r O I R E P 1 _ I T N =_ I RE AR I - SR - u I FU ,_ I UC - - _ 1 - w 1 - 1 . 0 9 8 7 6 5 4 3 2 j O.0 1 1 ^Ov Z O ='<Wh oO4 m>Z o2O =m 2C i>" omzm3MEo N>4 o2 CsE m IP= mO2 " >EuEmFap=92O m>4 ro , g m S M > m r a ' e; omm s <m3 a m@ nY a>5E* 3OcDm' a'e* ,i : !l,! l O i - I ,y / .'u} I ( / > * < w W m ,' I tn A> 1 O 9/ Z -s- O -= w - tn m% v bm O O M E L.! ) D. ./') %. / g -- Qy - l2Do5 I I l i _ l = l t I " O in v to N - o c) to h- to to t n N - O - (D) N01.1.VB31300Y ! SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SONGS 2 AND 3 FUEL BUILDING RAW IQ FLOOR SPECTRA EL.114' 0" DBE Q HORIZONTAL (N S) FOR 2% DAMPING FIGURE 4.6-7 i I ~. O l " I' I ( . f- < w P 'm V (f) 1 - ? O =. P' 2 - O g A $ ~ u- - Sb O w , W 0.. 5 - M . f - <x e 1 e5 00 ) _ l l 1 - l l I I , O LO + n N' - o c) cc h to in e n N - o (D) N011Vhl31300V SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SONGS 2 AND 3 FUEL BUILDING RAW FLOOR SPECTRA EL.114' 0" DBE HORIZONTAL (E W) FOR 2% DAMPING FIGURE 4.6-8 i l ' yy v 1 ( ,I a f / . i n ./ m / m .. .:::- o .0 z ' O _ / o W w M~ __ m w v \ O s 9 \ " W i C ~ l t < (V !E ; - zee i _2g D 4 1 i 'l i l i i - o c) co h- to to + n N - C O, (0) N01.LVB373OOY SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 p SONGS 2 AND 3 FUEL BUILDING RAW V FLOOR SPECTRA EL.114' 0" DBE VERTICAL FOR 2% DAMPING-FIGURE 4.6-9 C o,. ~ n - , SPENT FUEL'; POOL (UNIT-: 2)' q q(f I r , n.. T l u ik. ;:':= . 7- c: o ..- ! y *.a e.. TEST ' EQUIPMENT. DROP HEIGHT OVER UNIT .1',- FUEL .) \- i
- y; i (a '.8/
=- . [o ~9 199 O TEST EQUIPMENT ! DROP HEIGHT OVER - UNIT & 3 FUEL , GATE DROP ' D r* .5., :j i o HEIGHT ' N. .) ., . s . .: -1 - =g 4 tT'$'? o __ 's' TOP OF- # ,f.'.# RACK. l l .' s- .d, '. - ! e.4,,.,. ..... . . _ _p : i SECTION VIEW i LOOKING SOUTH l SAN ONOFRE NUCLEAR GENERATING STATION l Units 2 & 3 l v- DROP HEIGHTS QVER RACKS i l' FIGURE 4.6-10 ) L -- - - _ - - - - :J 4.7 MATERIALS. OUALITY CONTROL, AND SPECIAL CONSTRUCTION TECHNIOUES 4.7.1 CONSTRUCTION MATERIALS Construction materials for spent fuel racks conform to the requirements of ASME Boiler and Pressure Vessel Code, Section III, Subsection NF. All the materials used in the construction are compatible with the SFP environment and do not contaminate the fuel assemblies or the SFP water. The racks are constructed from Type 304LN stainless steel except the leveling screws which are SA-564 Type 630 stainless steel. Welds for the rack fabrication will be visually examined. Westinghouse has used visual examination (in lieu of liquid (} penetrant) for all rack orders. Only under special conditions and only for a very small percentage of the welds was dye penetrant ever previously used. The basis for the use of visual examination is (ASME Code) Section NF-5230, which allows visual examination for welds with a throat thickness of less than 1 inch. Welds with a throat thickness greater than 1 inch are examined in accordance with that code section. The visual examination is performed immediately after the weld is made. The inspection is recorded on shop routings for record-keeping purposes. The visual examination for the particular weld i size (less than 1 inch) and type (single pass) gives assurance comparable to a dye penetrant examination. O TMFWOO57 4.7-1 L L Neither section III or V of the ASME Code require use of magnification for inspection. It is E' policy to use the unaidedi eye:for normal visual inspection. -If.however, any indication is observed,-evaluation is made using 3-5X magnification. . t 4.7.2 NEUTRON ABSORBER MATERIAL q The neutron absorbing material, Boraflex, used'in the SONGS spent fuel rack construction is manufactured by Brand Industrial Services, Inc., and fabricated to safety-related quality assurance criteria of 10CFR50, Appendix B. Boraflex.is-a silicone based polymer containing fine. particles of boron carbide Boraflex contains a minimum B10 .in a homogeneous,-stable matrix. (_) areal density of 0.026 gm/cm2 for Region I racks and.0.016 gm/cm 2 for Region II racks. Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments, and to verify its structural integrity and suitability as a neutron absorbing material.(11,12) Tests were performed at the University of Michigan exposing Boraflex to 1.03 x 1011 rads gamma radiation with a substantial concurrent neutron flux in borated water.- These tests indicate that Boraflex maintains its neutron attenuation capabilities before and after being subjected to an l environment of borated water and 1.03 x 1011 rads gamma radiation and a total neutron fluence on the order of 1020 n/cm2,(12,13) TMFWOO57- 4.7-2 l ~'i Additionally, further testa (11) have recently been conducted and (G preliminary results indicate that some shrinkage (a maximum of about 2%) can occur in Boraflex, and that this shrinkage is complete at approximately 1 x 1010 rads gamma. Long term borated water soak tests at high temperatures were also conducted.(14) It was shown that Boraflex withstands a borated water immersion of 240F for 260 days without visible distortion or softening. Boraflex maintains its functional performance characteristics and shows no-evidence of swelling or loss of ability to maintain a uniform distribution of boron carbide. The actual tests verify that Boraflex maintains a long-term material stability and mechanical integrity, and can be safely () utilized for neutron absorption in spent fuel storage racks. Boraflex is the neutron absorbing material that will be used in the new SONGS 2&3 spent fuel racks. This material assures a-shutdown margin of 5% with no boron in the SFP water. Three plants (Point Beach, Prairie Island, and Quad Cities) have reported the results of their Boraflex surveillance. Of these three, the Boraflex material used at Point Beach Nuclear Power ; Plant has received the highest accumulated dose. This Boraflex j material has been in use for a total of 5 years, and some of the I) R/ i, TMFWOO57 4.7-3 1 l Boraflex panels have received a 20-year equivalent radiation dose due to the spent fuel management techniques used at Point Beach. The examination of the 2-inch by 2-inch sample coupons at Point Beach (which had a maximum exposure of 1.6 x 1010 rads gamma) showed that the coupons had experienced changes in physical I characteristics such as color, size, hardness, and brittleness, with some sample thinning. However, the nuclear characteristics of the samples had not' experienced any unexpected changes, and I the neutron absorbing properties of the samples met the acceptance criteria for maintaining .the 5% Ak/k shutdown margin. Point Beach also examined two full size (150 inches long by 8 1 ( inches wide) Boraflex' panels, which had a maximum exposure of J about 1 x 1010 rads gamma. These panels had a far' lesser m ount j of physical changes than the 2-inch by 2-inch sample coupons. Thus, the examination of the Point Beach coupons and Boraflex panels. indicates that, while some physical changes in Boraflex may occur with accelerated radiation expocure, the Boraflex will retain its neutron absorbing characteristics. Prairie. Island has also examined two large (8 inches by 12 inches) Boraflex coupons (22). One of the coupons (which had a 6-month exposure) had an appearance similar to the as-manufactured Boraflex. The other coupon (which had a 12-month exposure) had some slight physical changes similar to that experienced by the l Boraflex panels at Point Beach. The Boraflex panels in the Quad Cities racks (which had an exposure of approximately 1 x 109 rads gamma) were examined by a O TMFWOO57 4.7-4 - - _ _ _ _ - _ _ _ - - - _ _ - - ________=_____ -________-_ _ _ ___ _ _ - - _ _ neutron surveillance technique. Gaps were noted in the Boraflex panels, and review of the size and number of gaps was performed. This review indicated that the gaps were attributed to a rack design and fabrication process which did not allow the Boraflex to shrink without cracking. The Quad Cities racks were designed to hold. smaller-BWR fuel. The fabrication process required the Boraflex material to be glued to the stainless steel fuel rack walls. Also, the Boraflex remained tightly clamped during service. This did not allow for the-predicted shrinkage'of Boraflex wnich resulted in gaps developing. Less than half of-the Boraflex panels at Quad Cities had gaps, varying in' length up to a-maximum of 4 inches, and were located at various locations along the' height of the panels. A Kegg analysis of the Quad Cities SFP demonstrated tha';,these gaps did not cause Quad Cities to' exceed its 0.95 limit on Keff. The SONGS racks are' designed to hold the larger PWR fuel assemblies. The Boraflex sheets are not glued or clamped in-place, but instead are supported by the Boraflex wrapper (see figures 4.1-2 and 4.1-4 and paragraphs 4.1.2.1.1.3 and 4.1.2.1.2.4). The SONGS spent fuel racks design requires consideration for edge deterioration and 3% shrinkage. The installation of the material in a stretchect or restrained condition will not be permitted. l (The installation of torn or cracked sheets of Boraflex also will not be permitted.) Southern California Edison has a full-time O TMFWOO57 4.7-5 _ _ _ _ _ _ _ _ _ _ - _ \ r__. _ _ . _ _ _ _ _ . _ _ - . _ . . _ _ _ . _ _ _ _ _ . _ _ _ - - l Quality Assurance representative assigned to the fabricator's q I
- ,r- shop to ensure that the-specification requirements are adhered to
.t < . - by the fabricator. L In conclusion,.the SONGS rack'dosign features and fabrication ~ L procedures ensure that any shrinkage.of the Boraflex sheets q during in service irradiation.will not result in potential l criticality problems for the stored fuel by: 1 e Using oversized Boraflex sheets to provide for edge deterioration and shrinkage. e Allowing for Boraflex shrinkage with no restraint mechanism (no adhesive allowed). e Implementing administrative controls to provide _ assurance ) that the gamma ray exposure is limited to that exposure previously demonstrated to.cause no significant material deterioration (1 x 1011 rads). e Providing for, installation of Boraflex in a non-stretched condition, with no tears or cracks. e Maintaining a full-time SCE Quality Assurance representative during fabrication in addition to the Boraflex manufacturer's and rack fabricator's own Quality Assurance staff to ensure specification requirements are met. O TMFWOO57 4.7-6 _.ma---_____-m _____________.2_.- _ _ _ _ _ _ . - . _ _ _ _ _ - _ _ _ . - Further,.the experience at Point Beach indicates that some physical changes may occur in Boraflex, but that the Boraflex will retain its neutron attenuation properties. Both testing of Boraflex and the experience at Quad Cities indicate that some shrinkage in Boraflex may occur, but that this shrinkage is limited.to maximum of about 2% of the length of the Boraflex. This resulted in some gaps in the Quad cities Boraflex panels because the racks did not permit the Boraflex to shrink without cracking. In any case, due to the small size and.the random orientation of the gaps,-the calculated Keff of the Quad Cities SFP did not exceed the 0.95 limit. Since there are differences in the installation process of the Boraflex used at Quad Cities and SONGS, the anomalies experienced at Quad Cities are not expected at SONGS. O 4.7.3 QUALITY ASSURANCE The design, procurement, fabrication, and installation of the new high density spent fuel storage racks comply with the pertinent Quality Assurance requirements of Appendix B to 10CFR50. In addition, the Quality Assurance Program of the Westinghouse Nuclear Components Division (WNCD), manufacturer of the racks, also conforms to the requirements of the American Society of . Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, O TMFWOO57 4.7-7 i Section III and VIII (Subsection NCA-4000) , MIL-I-45208, /~N- MIL-Q-9858A, and RDT F2-2. The Quality Assurance Program at WNCD Q is implemented through the Westinghouse Water Reactors Division Quality Assurance ~ Plan as described in WCAP 8370(15), 4.7.4 SPECIAL CONSTRUCTION CONSIDERATIONS The reracking of the SFPs at SONGS 2&3 (see figure 4.7-1, Fuel Handling Building Unit 2) will be accomplished in the " wet" condition (spent fuel stored in pool during reracking) subsequent to cycle 5 refueling. At that' time there will be approximately 480 spent fuel assemblies per unit being stored in each SFP. Special considerations will be employed in the construction planning, equipment / tools development, sequencing of activities, j) and administrative procedures and controls to ensure work is performed consistent with ALARA considerations and as safely as practically achievable. i 4.7.4.1 Removal / Installation SecuenciDS The rack removal / installation program will utilize the storage of approximately 180 fuel assemblies in the cask handling pool (adjacent to the SFP) during the reracking process (see figure 4.7-5). A H Region II rack will be placed in the lower portion of the cask handling pool for the fuel storage. Pool cooling and purification for the cask pool will be maintained at the UFSAR (subsection 9.1.3) requirements throughout the period in which O'. '~ TMFWOO57 4.7-8 1 spent fuel resides within the cask handling pool'(see paragraph [ 4.7.4.5). A cover.will be placed directly over the entire cask The ~ ' pool to protect the spent fuel from postulated load drops. . cover will be designed and analyzed in accordance with-Appendix A of NUREG 0612 for the governing appl' i cable construction load drops. The storage of approximately 180 fuel assemblies in the cask handling pool during the raracking process will greatly enhance the overall safety, margins and ALARA program throughout the operation by: A. Reducing the number of required fuel shuffling operations-and the number of fuel assemblies shuffled each time; () B. Allowing greater horizontal distance between fuel assemblies and work areas in the pool throughout-most of-the raracking program; C. Providing more spent fuel storage locations for greater flexibility in isolating spent' fuel away from areas which would require the use of divers;- D. Lowering the probability of having an accident in the SFP involving spent fuel assemblies (by reducing the number of assemblies available and increasing the distance between fuel assemblies and work areas / safe load paths). O TMFWOO57 4.7-9 I l The present' configuration of spent fuel racks at SONGS 2&3 / consists of 15 racks (10 - 8 x 8 arrays and 5 - 4 x 8 arrays)- arranged as shown in figure 4.7-2 for Unit 2 (Unit 3 is mirror image of Unit 2). The final configuration of H high density spent fuel racks is shown in figure 4.7-3 which is comprised of two Region _I racks (racks 1 and 2) and six Region II racks (Unit 2 orientation will be used throughout this section in both descriptions and figures, for simplicity in presentation, with the understanding that they also apply to Unit 3 in an opposite hand orientation). The rack removal / installation sequencing approach for SONGS 2&3 has been developed to best meet the following criteria: A. Maintain a safe load path for heavy loads at all times and meet provisions contained in paragraph 4.7.4.6; f')l B. Provide at least three empty rows of cells (no spent fuel contained within) adjacent to any work area within the SFP; C. Provide maximum flexibility in spent fuel storage locations; D. Minimize required amount of spent fuel shuffling; and i O TMFWOO57 4.7-10 I 'ib j
- d. '
) i E. Allow for the use of divers'for those activities which l () .cannot be reasonably handled remotely. will"be in place. A diving plan i H. The rarack sequencing for SONGS 2&3 is comprised of four general j l steps starting.from.the original condition.shown.in figure 4.7-4 l depicting 480 fuel assemblies (shown in red) residing in the SFP..-The number of fuel assemblies used in describing the various operations in the raracking effort are conservative estimates of actual conditions to be encountered during.the r implementation process. Therefore, the numbers are approximate in nature and are given to establish the numerical magnitude of -the assemblies involved at each stage. Southern California Edison's intent is to follow the sequencing- ) described below. In the event that it is not possible to explicitly'iollow the sequencing, alternative approaches utilized h will best comply.with the five criteria listed above. The first step in the process (see figure 4.7-5) consists of:. . placing a H Region II rack in the lower portion of the cask L handling pool; placing approximately 180 fuel assemblies into the H rack; placing the protective cover completely over the cask pool; relocating the remaining 300 fuel assemblies to the north end of the SFP (as shown) ; installing the temporary gantry crane (for use about the pool area); removing the six most southerly racks (10-15 per figure 4.7-2) from the SFP; and removing the existing piping and supports from the vacated area. O TMFWOO57 4.7-11 L 1 l i f~~') Step one of the sequencing plan provides ample separation between V the spent fuel and the work area (about 8-1/2 feet), by having one row of empty racks (racks 7-9) plus one row of empty cells.in l' between. The movement of the racks, piping, support material, and all other items entering / leaving the pool will be controlled to stay within established safe load paths as shown on figure 4.7-9. Step two in the sequencing (figures 4.7-3 and 4.7-6) starts with the installation of H Region II rack 7. Rack 7 is placed in the pool at this time to serve two purposes; minimize the consequences of a load drop (see paragraph 4.7.4.4), and provide additional flexibility for the temporary placement of fuel. Next, H rack 8 is installed in the pool and approximately 132 spent fuel assemblies are placed in it, then existing racks 7-9 are removed, followed by the removal of the piping and supports from the vacated area, which completes this step. Sequencing step three (figure 4.7-7) consists of installing H rack 6; placing all fuel assemblies into H racks 6 end 8; removing the remaining racks (1-6), and then removing the remaining piping and supports from the pool floor. q 1 l O TMFWOO57 4.7-12 J l The initial step four tasks *(figure 4.7-8) consists of installing-H racks 1-5 (racks 1-4 for Unit 3 only); removing the temporary ] } u gantry crane; removing.the protective cover from over the cask handling pool; and'atuffling the approximately 180 fuel-i assemblies stored in the cask pool back into the SFP. The' final step four tasks.for Unit 2 consists of removal of the fuel storage rack from the cask pool (using the cask handling crane). For Unit 3 the final step four tasks' consist of reinstallation of t!e temporary crane, transfer of the cask pool rack to the~SFP (permanent rack 5),'and removal of the temporary, crane. 4.7.'4.2 Safe Load Paths 5() Safe load paths are used throughout the project. Figure 4.7-9 depicts the basic load paths within the FHBs at the pool deck level. Material will enter / leave the pool deck floor (elevation
- 63. feet-6 inches) via the access hatch (location 1 in-figure 4.7-9) which opens directly to the access bay at grade (elevation 30 feet-0 inch). The existing cask handling crane will be used to-move material into/out of and about the southern portion of the building (cask pool, cask washdown area, and access hatch area). Material entering the building for placement in the SFP will follow the arrows from point 1 to point 2 where they will be placed on the cask handling pool cover for transfer to the temporary gantry crane. The temporary gantry crane can move over the pools in a north-south direction from the wall at the cask O TMFWOO57 4.7-13
l washdown area to the fuel transfer pool. The temporary gantry ] crane will hoist material off the cask pool cover and move it to { point 3 (H rack 7 location, see figure 4.7-3) where it will be lowered into the pool to a depth of within 1 foot of the top of rack 7. The material will then be moved northward until it clears rack 7, at point 4 (rack 5 position in figure 4.7-3), where it will be lowered to within 2 feet of the pool bottom (to assure adequate clearance over items residing on the pool floor). It will then proceed along the east side of the SFP until it is at the northern-most location. It will then be moved to its final location (point 5) and lowered into place on the pool floor. Material leaving the pool will be moved in the reverse order of that presented above, with the load being placed on the cask pool The path from .( ) cover for transfer to the cask handling crane. point 2 to point 6, shown in figure 4.7-9, will be used for the temporary storage of items entering / leaving the pool as required. Material leaving the pool will normally be taken from the cask pool cover (by the cask handling crane) to the cask washdovn area (point 7) for further decon/ packaging prior to leaving the building. From point 7 the material will be moved by the cask handling crane to the access hatch (point 1) and lowered to grade. Au described above and previously in the rerack sequencing (paragraph 4.7.4.1), at no time during the reracking process is any heavy load (greater than 2000 pounds, present Technical TMFWOO57 4.7-14 Specification limit) planned to be lifted over unprotected spent fuel except-during installation of the cask pool cover which will [ be done in a manner that will preclude an accidental drop of the cover into the cask pool'(see paragraph-4.7.4.5). Heavy loads 'will be lifted over the fuel stored in the cask handling pool, but only after the protective cover has been installed. Heavy load lift heights will'be limited to 12 inches above floor / cover surfaces at the pool deck level'and 24 inches above the pool floor, except at the designated lifting points for entering / leaving the pool and during transition from floor to cask pool cover. The entire area serviced by the cask handling crane has no safe shutdown equipment located within it or.beneath it, and therefore, no area restrictions will be required for movement of material within this portion of the building. '( ) 4.7.4.3 Temoorary construction Gantry Crane The reracking process at SONGS 2&3 will require the use of'a temporary gantry crane (see figure 4.7-10 for a schematic representation) for handling loads within the SFP area. The temporary gantry crane will be designed, tested, and installed specifically for use during the reracking process. The maximum anticipated lift is estimated at 27 tons; however, the crane will be an upgraded commercial class and rated at 35 tons. It will also meet the following criteria: A. The requirements of CMAA-70 and chapter 2-1 of ANSI B30.2, i O TMFWOO57 4.7-15 p. h l B. All' load bear 1ng members will have. minimum factors of safety of: ( o Three against' minimum tensile yield strength, 1 l o Five against average ultimate strength, l C. Applicable SRP load combinations including DBE,(SONGS 2&3 l . equivalent of SSE) with load on hook, D. Equipped with dual ~ holding brakes on both main and auxiliary hooks rated at 150% of the motor's' maximum rated ~ torque value, E. -Equipped with gantry and trolley brakes rated at 150%'of the motor's maximum rated torque value, and ] Hooks (main and auxiliary) load tested at twice their -{ ) F. rated capacity and then nondestructive examination (NDE) tested to ensure material integrity is maintained. The testing requirements for the temporary gantry crane are presented in paragraph 4.7.4.6 which addresses the control of Heavy Loads Evaluation. 4.7.4.4 Postulated Construction Load Droos The SFP area at SONGS 2&3 has been evaluated for potential construction load drops during the reracking process. The governing load drop analyzed was a H Region I rack weighing ] TMFWOO57 4.7-16 approximately 50,000 pounds being dropped 50 feet-6 inches to the (,) SFP floor. It was assumed that the rack would impact on one corner support foot with the rack's center of gravity aligned vertically directly over the impacting foot. The analyzed drop is conservative due to the planned rerack sequencing and safe load path conditions. The actual postulated worst case drop would be a E Region II rack weighing about 31,000 pounds (vs. 50,000 pounds analyzed) from a height of 48 feet-6 inches (vs. 50 feet-6 inches analyzed). The planned rerack sequencing and safe load path programs have successfully reduced the worst case drop conditions and the probability for such an incident occurring. This has been accomplished by restricting lift height conditions and by placing ,-. E rack 7 in the pool first (see figure 7.4-3) ; and then requiring \_) all other heavy loads to enter / leave the pool directly over it (see paragraph 4.7.4.2, Safe Loads Paths). In order to prevent a dropped rack from tipping and damaging spent fuel stored in an adjacent rack, a restraint or protective device will be used when lowering the racks into or out of the pool. The H Region I type racks are now reduced to a maximum drop height of about 22 feet over the unprotected portion of the pool floor. The 22 feet maximum drop height also applies to all other items entering / . i leaving the pool with the exceptions of; first six racks being removed (maximum rack weight of about 27,000 pounds), piping and supports associated with the first six racks removed, and the placement of H Region II rack 7 (actual worst case postulated drop). s v TMFWOO57 4.7-17 i l The analysis of the Region I rack drop (conservative condition) j - ) ) indicates that the following could occur: l e The, stainless steel liner plate (3/16 inch thick) and the reliner plate (1/8 inch thick) would be penetrated,
- e. The concrete basemat (pool floor) would be penetrated about 5-3/4 inches which is about 7% of its thickness, )
e Leakage from the pool would be confined to the leak chase system, o The immediate water loss would be about 11 gallons (maximum capacity of a leak chase channel), q 4 \_ e The maximum flow rate from the pool (into the leak chase system) would be limited to approximately 49 gal / min. The existing SFP makeup water supply is 150 gal / min; therefore, the Technical Specification water level will be maintained. 4.7.4.5 Use of Cask Pool and Cask Pool Cover The cask handling pool will be used to store fuel assemblies during the construction phase of the reracking effort. This will minimize fuel assembly movements. The existing 4-inch cooling line supplies 370 gal / min to the cask pool. Analysis shows that this flow rate is capable of removing 6.7 MBTU/h from the cask 7-TMFWOO57 4.7-18 / pool by discharging the water to the SFP via the gate opening. Under these conditions, the cask pool temperature will be less than 140F, and the SFP temperature will be approximately 107F, assuming the CCW design temperature of 95F. A maximum decay heat production of 6.7 MBTU/h will be allowed; further, any combination of fuel assemblies may be used provided it has a minimum of 70-day decay. To provide further assurance of adequate mixing, a temporary cooling line will be added to discharge cooling water at the cask pool bottom. The E seismic analysis bounds the use of the Region II storage rack in the cask pool area (figure 4.7-14). The cask handling pool, which is located adjacent to the SFP (see figure 4.7-2) for each unit at SONGS 2&3, will be covered during the reracking program. The cover is required to perform two basic functions, protect the spent fuel being stored in the cask pool and provide a working platform for the transfer of loads L a deen the cask handling crane and the temporary gantry crane. Because spent fuel is being stored in the cask pool the design and installation of the cover will preclude the possibility of the cover being dropped into the cask pool during its installation and removal operations. The cask pool cover is planned to be assembled prior to its placement over the cask pool. It will be moved into position (possibly with the use of rollers) as one complete piece with spent fuel in the cask pool (figure 4.7-11) while maintaining a O TMFWOO57 4.7-19 , L l i I minimum distance above the deck (see figure 4.7-12). The cover ( ,) will then be bolted in-place. The removal.of the cover would be ) accomplished in a similar manner. The cover will extend beyond the pool edges in the north-south direction (see figure 4.7-11) adding' additional safety margins to precluding the possibility of it entering the cask pool during installation and removal. The cask pool cover will be designed / analyzed in accordance with the requirements of Appendix A of NUREG 0612 for impact loads resulting from postulated construction load drops. It will also be designed to provide a suitable working surface (laydown area) for the transfer of loads between the cask handling crane and the temporary gantry crane. 4.7.4.6 Control Of Heavy Loads Evaluation 7s The hoisting of all heavy loads within the FHBs will be accomplished by either the existing cask handling crane (125-ton rated capacity main hook and 10-ton rated capacity auxiliary hook) or the temporary construction gantry crane (35-ton rated capacity main hook and 5-ton or less rated capacity auxiliary hook). Figure 4.7-13 shows the relationships between the cask handling crane, the temporary gantry crane, and the spent fuel handling machine. The provisions of the heavy loads program for reracking SONGS 2&3 are presented below in relation to the specific guidelines given in NUREG 0612(16) Subsections 5.1.1 and 5.1.2, option 3. p -\ TMFWOO57 4.7-20 , y _ _ . - . - - - . - - _ . _ - . _ _ ~ - - _ - _ - - _ _ - _ _ - _ . _ - - _ _ - - _ . - - _ _ - - _ ) : n < i. The general 1' requirements'(seven' items. listed in'Secticd 5.i.1-of-NUREG'0612)'for-SONGS 2&3 reracking heavy loads program consist of the.following. 'A.= Safe' Load Paths-1 Safe' load paths for the handling of all. heavy' loads 1within the FHBs will be maintained in accordance with. thel 1 guidelines of;NUREG 0612 Subsection 5;1.1 (1), except.the marking of'the floor. The' entire area;of the building serviced by the cask handling crane is available'for-the , movement of heavy loads (see paragraph 4.7.4.2).and ] thereforeffloor markings are not: required. The only'other area involved-is within the wetLSFP which cannot be marked. l B. Procedures Procedures will be developed'for the handling of a'll' heavy loads within.the FHB for the reracking program. The developed procedures will comply with the guidelines of NUREG 0612 Subsection 5.1.1 (2). C. Crane Operators l '. L . All crane operators for the cask handling crane and the temporary gantry crane will be trained, qualified, and conduct themselves in accordance with Chapter 2-3 of ANSI O TMFWOO57 4.7-21 , i J ___-___--___-_.___.____..__.-_m__ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ t B30.2-19761while handling heavy loads during the reracking ( of SONGS 2&3.--Therefore, the guidelines of NUREG'0612 Subsection 5.1.1-'(3) will be complied with. D. SpecialLLifting Devices-
- j. -
All special lifting devices to be employed.in.the handling of heavy loads within the FHBs at SONGS 2&3 during the ~ reracking program will satisfy the guidelines of' ANSI .N14.6-1978' as amended by NUREG 0612 Subsection 5.1.1-(4).. Therefore,'the guidelines of NUREG 0612 Subsection 5.1.1 (4):will be complied with. E. . Lifting DevicesLThat Are'Not Specially Designed , - 7s \s All: slings used in handling heavy loads with the FHB during the reracking program will comply with the guidelines of NUREG 0612 Subsection 5.1.1 (5). 1 i F. Cranes The testing, inspection, and maintenance of the existing cask handling crane will be performed in compliance with the existing NUREG 0612 program for SONGS 2&3. I The in-place inspection and maintenance of the temporary gantry crane during the reracking program will comply with the guidelines of NUREG 0612 Subsection 5.1.1 (6). O TMFWOO57. 4.7-22 , t ~-z_________.-_m.-___.m-__m__. ._.___-___m____ _ _ .__ t yL ./" . i t: \~si The. testing of.the temporary gantry crane will be accomplished inLthe following manner:
- 1. operational tests per ANSI B30.2-1976 Section 2-2.2.1 will be performed at the factory prior to' shipment, and/or at the' site prior to final' installation,'and' l in-place (over pools). prior to initial-uce;.
, J l
- 2. Rated load test per ANSI B30.2-1976 Section 2-2.2.2' 'f i
(at 1.25 x 35. tons) will be performed at the factory ! prior to' shipment and/or at!the site: prior to final installation; Also, a modified load test of the hoist and. trolley 3. 3-- (at 1.25 times the maximum anticipated load of approximately 27 tons) will be performed in-place .(over covered cask pool only) prior to initial use. The above stated program for.the testing of the. temporary gantry crane meets the intent of NUREG 0612 without requiring testing of the crane directly over a SFP containing spent' fuel. Therefore, all of the guidelines for inspection and maintenance and the intent of the guidelines for testing from NUREG 0612 Subsection 5.1.1 (6)'will be complied with for the temporary gantry crane. l' l-/~). 'l. j TMFWOO57 4.7-23 - l l-W_:__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . / Also, all of the applicable requirements of the existing SONGS 2&3 NUREG 0612 program will be complied with for the existing cask handling crane. G. Crane Design The design of the existing cask handling crane is qualified under the established NUREG 0612 heavy loads program for SONGS 2&3. The design for the temporary gantry crane will be in accordance with the applicable criteria of Chapter 2-1 of ANSI B30.2-1976, and CMAA-70 and therefore, will comply with the guidelines of NUREG 0612 Subsection 5.1.1 (7). The specific requirements (five items listed in Section 5.1.2 (3) O of NUREG 0612) for SONGS 2&3 reracking heavy loads program will consist of the following. A. No " hot" spent fuel (as defined in Section 2 of NUREG 0612) will be stored in a SFP during the reracking of said pool. Therefore, the guidelines of NUREG 0612 Subsection 5.1.2 (a) are not applicable. B. This item has two specific requirements, namely, that no movement is made within 25 feet of " hot" fuel and that movements of loads within 25 feet of decayed fuel has the shift supervisor's approval. Since no " hot" spent fuel will be stored in a pool during its reracking the O 4.7-24 TMFWOO57 K t bs guidelines of NUREG 0612 Subse'ction 5.1.'2-(b) which apply I,)' to." hot" fuel are not applicable. Sufficient fuel decay; . of-approximately 70. days (as shown in Figure 2.1-1 of 'i NUREG,0617)-will have taken place prior'to any heavy' load being moved over or within the SFP-or within 25: feet- _ (horizontal) of any spent; fuel (complies with NUREG guideline). .In order to meet the second requirement.of this guideline, administrative controls.will be-established through procedure application and scheduled work activities..to govern the movement of heavy loads-within'25 feet (horizontal) of spent fuel. The associated procedures and work activities will have.the approval of the shift supervisor. prior to their' implementation (complies ~with the intent of NUREG guidelines). .Therefore, the intent of the applicable _ guidelines of-7-~ s_/ NUREG 0612 Subsection 5.1.2 (b) will.be complied with. C. The areas covered (at the pool deck elevation and below) ' by the' cask handling crane and the temporary; gantry crane do not include any' equipment associated with redundant or alternate safe shutdown paths;.therefore,z the requirement - for mechanical stops or electrical interlocks is not applicable. Postulated load drops will be' analyzed to demonstrate that they.will not cause damage that could result in criticality, cause leakage that could uncover the fuel, or cause loss of safe shutdown equipment. Therefore, the applicable guidelines of NUREG 0612 Subsection 5.1.2 (c) will be complied with. O TMFWOO57 4.7-25 l e _____m__________.-m__am._ _ _ _ . _ _ _ _ _ _ - _ _ _ _ _ . _ - _ _ _ _ . _ . _ . _ _ _ - . _m_. . _ _ - _ _ _ _ _ t; / r (K)< D. Administrative controls will be established to limit the maximum lift height of heavy loads within the FHBs. 'Within theLSFP the maximum lift height'will be 24 inches i above the pool floor (except when entering / leaving the pool as described in paragraph 4.7.4.2). The 24-inch lift -height in the. pool is'to assure adequate clearance over items existing cn1 the pool floor (piping, supports, etc.). .The allowable lift height When moving heavy' loads-about the' slab at elevation 63 feet-6' inches (pool deck) and.over the cask ~ pool cover will'be limited to 12. inches except at lifting points. The heaviest anticipated load' will be about. 30 tons - (temporary gantry crane) which ~is considerably,less than a fuel cask (subject of this ,_ section of the'NUREG). Therefore, the intent ~of the q (_/ guidelines in NUREG 0612 Subsection 5.1.2 (d) will be met. E. Postulated load drops along the specified safe load paths will be analyzed in accordance with the guidelines of Appendix A of NUREG 0612. Therefore, the guidelines of NUREG 0612 Subsection 5.1.2 (e) will be complied with -along the designated safe load paths. u nV TMFWOO57. 4.7-26 , ,' ' 1 j j 'b w;wiy . . p. < ' a ., x - . ,s .., t o (. *'kh e, / , ,a. k ,p. ( 3 f ( iS M' #..,.h[ '[.k A?? , :. .- 1.F11 '?j - I h[ I \,0.mpc,5 n-q, y g ;g' . s ' n-1.a.;.m,f$i,pd'{ i ~-,:jf1 - ~ ,. . #: $ ![ 4 . b: .; j - ( / L 'i' . !.sJfjl1[i d.h..i $l ;; .m n , ,: y /' ~ j9 ,;. . y m .- uaey
- v. m>. @m, o.si 5 F-%4,.,
1 I ' I+ ,' h ,b j 's g , I , f* .,'$,sjf a 1 - e / y ;-E'p 4;%q%g s:q / I.[-) 5 f,. f((,.. - I J r > J iM . @ t 5.. y;0 y;h/djikjikh) 4 g / #[ . WPil,(.s; a. Ff 'Q;*,f ki i .e <4 . w%$JD n.. . c , s. p ,.M - t r.x.-a .. m Aq .. h.a , ., is s_a?39cg~m e qr 4 @4I ld'.$ IQ F( ' AQ l f,;h n l l["}**fR ( mlh. a.m&m]a{ w* k - g ' ' 4 g r J. gpy 9 9 g. _' D i 'f-llr7 .e + - ., y fj ~ - (@.
- c. , dj a
,. m s je - ,qh 3 d* a Nq .(( 9 ' 'l;I N, 474* . . p . "+ . [, .- ;, n. n! \ 5*_ , -l 'q' , j gj ex A Z "P y%9777; ' ' A 4 0 i" daO~%w gg;w WLwM Vy < , k., DW ? 1 m' 'i. p w,. .u,w pp aw cu , W $ maww)3+gl.?i k+Of {j h;?I M_'f we .v i u t f T N . wl]: m .
- 'kk lpdll ,
g: ^ p, Tins Y, O( Al y ' ,f if.c $:];,W ?. M;, , < @ NZ y n 7 M , * ~ ,. -f g 3 g C - g
- .p u.
l -l ,.. $<; IP ~.. E i: i
- J[
-x Of y 1 Y' y [; gM 7$ NJ:q{ ? Kn9 W g "i g ;p- f;;:; >; $,U%- Q ?) R ll gi my .lf. , A^*:f'i;? I l}{V ' '-g ging'=1'. .q gjgtt /' 1 < + Q ,ygl .Mg: gqg< . ,jd% %;; ,, ' i. - , ' 4:g$. ma ana NNMlb # .- < L '4 l l , 1 n M 4 m O 4 O U A i l .l \ I ! <g - l i _v, N .m ?:(J x. a < w , , , ,r N, ,I I u + [ -2 a @p ;hmg F-] " p w* eI +*g +- +4% _;. 2 cr _Ju = E ~ _.Je __1.c m' .o. y 4 c .- L yg ,m.q . = m.! l j o J, , ..- a. , = . k , js"mZ >m W -_$d. -, a M.,Y
- h.- - a ,
an ' ' ~ 4 i ,, r 1 M1. a <+t,- _ , , , M .} ' m Cg l ,3 9 r, .- , ) / t - c v + +'; m.., v .i -m o --- Y f .g,__ "%b 5 h,f.c
- a S ~
( ** T ,apDE , ' . . . .%g & # &' - r,- * ~ . . . d-{ 4.4 ., r- .. ~. .I m.
- x. . ..,_, ,
. i .. ., .T T +5s% x .- 3ki $$.d.$.$.JJT p, . o a4\ -I h- n' j m > , ,,,e, ,, l,. 1T T -H +-[-2& jm @'--J*,l J Nh , '.4 ,.4%.j,'j. d %3 ..H ; g ra h m T l ) ,.g,.t, >_ e 4 . . 4- ; g m~ p=1 g l'. '_', '- 4 . m-- g j __L m ' 2 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 O ' SPENT FUEL POOL (UNIT 2) ORIGINAL CONDITION FIGURE 4.7 2 7 _ 1 1 ( t 1 f 5 A O 4 O O _ j 9 f J- _ L-' _ h W- , \ I I W;<>,,,,, c ~ r { so g m, u,, e_ . , , ,w . $c C ( ' J r - r M .f lll'l- ' , %' E! r ,,,, - h ' l 'N f ' ' ' ffNE uM, , ,' , . - a ' kW Q . W::: WN, . -f ,' l , r
- ; . : . :: .+
) I c : ; ;;. . ui__. .., ,,l **.n. 1 &~ill. & rs:C:}av, * ., l l ,' ' g i w. s'>.' ,l'*0I$ Y~
- u. . . . .
w .,, l**. ' '* ,*1'S,_h, v, - C ',< .'.', , ';; ' l ,' . ' '.']p p- - i i \ I t b L-- SAN ONOFRE NUCLEAR GENERATING STATION 1 Units 2 & 3 SPENT FUEL POOL \ (UNIT 2) FINAL CONFIGURATION FIGURE 4.7 3 k ;m"y M' 1 j - . w ;g aa y - .w 7: ?} ; 1 mi vn . : + N a e 4 - w = _' d ' m- ) W% _ T' ~ 2 'r'. l nl2 G IT s G wK .s NN m : I W' %J .] C (U N ng EL v. .~. a UO .' ~x rf QO w 2 EP S
- w+0 x
.2 / 2 W L =t - ,e KE " ' CU _W.*~zG ;c . S K AF a C s' A R R ,m&n:M" L : ,w g ET i a. G L -;* g N RNE %ini e rg Taw " g a I T S I P ;W m X E S 9 s N ? A %*E I i i C_ S 4 ,* N %Qwj - O ' I , . : , T a A .D^5:.- ,.M f ' 4 + Q ^f C O L n e .5 .q .] . E m G m#'C . "4 c ^ A T R O ., S ~
- m. m. . L E
f, . m- w; U R,M{:.~ 3; W . l Q aj F ?. h' ,.?A e-0 i%
+ m 4 g E N g g R I f' U C f i f ?nK g G N ^ c': f .i 4N{Q 31 O@g g I F E _-_ g sg U ~ c# i _ _ n _ s ._ Q : x;y _ E J.- i y g S ; t K $5@ta.h ~ _n_ ,, '. _. w a - ~ & ,, _ . , C . m A s _- A P m,eM - m E R - E $ D c, c, . 1 + y' u rK R A 5Mnn% '__ e _ - . aC D R Y mg%o E Ewge~.@,__ .___ W, _.mRA T D s :: _ .~ . L A C e-- - M:5.wg;s A E _ .w O O V _. P S k nu&r% -
- ~
O _ -- _: S O R NLg;r;w.Z.:._~_ A P ~ - r _ _k g - - C R F m<. O p %. d ; _ ~ __ N I E V S T R 8g - - R P 94uz.%. ~ 4 S E I O R 5 1 O P - > L E - P $ws.cW $i c: ~ h fui,. pr n A - - B M V O 0 0 # 0 1 U S b[z&n$ C ~ ne,O ~~ '~ * ~~ ~.~4 " ~-~. - *k E S S A C E V I 3 G K N C I S D N A kW:g:u* nn y- ~ ~ ~ ~ ~ ~- - y L E T C N A I A R G N r- ~ ~ E t ~ U m r~ .r- , ; MSE ^ I F T E P 5 1 . {s' 0 O R N l P ,' 8 R w'3 4 1 P E E E r . E E T V V AO ;\ y n7l j, C C O h+ama2vtg e A L A L CM OE M E k)w%. hum;an.g k viW*~~ gb u% P P L R R ll\\ \l>))Iii \li \)\\ i O f G N I , 3 f C C, m^$]g sg N ; s.: ^p , g E a. ( U - w Q i E - S . OK 3 1 t. p C _ A - = E M' 7 R d E , , .+.~ + R * . .,'Q . v - .~ D E N n t S E % u' ., 5 O M P h%'y- % O .rt" gt-R k P ka;A., i 8 M;ff 3: & g i. i 7 .e ;M DnW S .e: K a:e C gs:q'n e A R W L L m, e qn j_. A T O S N mJ;z- n u: n; I m 'p;g; u .s G N I C N ' . g . :,.-- E 1 ^ -. U Q ~ E ' S ' K C ~N A , R , l
- i. z K M
;iE' n q0 t 8 R q , ry.. kit fhi f ey . D w,;y,ke tg c, E 7.%g . g ;U$ S %! Q : O D , .,j P %hW; s- _ O ?_.%;;~.. % ,O +Q R E ja .,@}
- ,- Q E
P YgR . .. _l &. , <aui .L- qLf ? N 4
- ,l .
'5.r a K Mn9. "a +3 ::s:?'- k 3 C A R i W_ L @n.? ,R9 y L A T N^ ' ?,l{g _ S N _h I m9,;c khs _ t. y l('?Y I ,$n - y [ p:e.. Ahs j j,_n 4; 7 * {. w
- f 6 P o,. "$: #
7, t"_ _ N I D' I a - m , ,. C 1 N g e E ; z' 1 ?.pa zsN U &, R$'; - r T* g Q ' y E " '~ s. . y S ^' $ n' g ,' [. ,C g K - ., hn n' g i q C w';;;;w h 5 ? ? A _~ _T l S 6 d R w s,5~iS .m + E *= p ,S - R r c16 g y D g E y 3 S : t O ra r P M O R _ P - 5 1 S K C A R w u W L n;;e p% = g- = L z A , i 3 f. {, Vm n . " , +, v%p : ,. H '~ T ) i. t S N j- I m f e N , r~3 ) LJ ' .T.ij.' *:. - ") k? - . 1- ,4 i A. I 1 ~. ., N.Nx-[ X l ,. M N 4 \n- ; ?- h ) " , ,. y l 3 9:$ 3 b S. i r _ , ins- - e ny , ,m ,o 4 m 31 m, a l n- v 4 - j. ,; ~ , o r o. r 's e n- ~2~ 3, I n n a hl l I I-- ; r i d _ _. .... 3 Pl I i-1F- __ 1 --- - 1 r- 1 r 1 r ' -e r- - - 5 5 5 Xl l i ' 6- - i I i e _m - i g , I .c yf,,;, - - . ..$ .T.I* r a., l "* 'Y O\ A'- X= DESIGNATED HEAVY LOAD LIFT POINTS WITHIN SPENT FUEL POOL i l f i SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 SAFE LOAD PATHS O- FUEL HANDLING BUILDING 1 (UNIT 2) l FIGURE 4.7-9 - T , . t * @. I gg y v. i w ' *E 32'- 6" g' *1:d /. d. $. EL .9 2'-II" ' El 9t'-5,, 1 -; ,, , 3 l l W ~ ~ < ' 7 - AUX. MONORAIL l l I a h. 'R15'- 9" 'L I 24'-6" 1 C-C OF RAFL jL . f.- E L .63'- 6" L.4- 2'- 3" i _-l' git / ' I (TYp)' i J' l ll =. i i i O i s- ~ ' lllf l ' d' . ff,k %l.l.j s ffE .E L 17 '- 6 " ~- ' ' %yp _ _ _ SECTION - SPENT FUEL POOL (UNIT 2)' SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 TEMPORARY CONSTRUCTION GANTRY CRANE FIGURE 4.7-10 L, [ Dg 1 h x)'.' , , i , 1 l.glI l 1 i l,' 1 oot covcx , l g I l ,' e i l i I l' i i l l i- i a ji l- 1 l 878 I I edi. A lI l- -CASK Pool l l5l g l ,, l. I w^us ,1i ll,
- j Tc g G Wg 1
I I IOI ll, l. l@ i,l8 l i 1., l ,4 l li !fl ' l@l l I lpll ' I I lEl C< I4 l l lll lli i i lh ! l.I I -ll l l :l.1l(t ' i l I l- l l ilp - %lI ki. lo L_________ . I _ _ _ _ _t- i. PLAN (CASK POOL COVER) . SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 PLAN (CASK POOL COVER) ~ FIGURE 4.7-11 _ --..__.u._ . _m --.._mm_-. -._ - - . _ _ . - _ _ _ . l 1 - --CASM MANDL/MG CR4HE NOOK l \ POOL COVER .$,; COVER UPPORTBEAMk . e.2.: )" .. . ..I *4". ,". 'k..s ., ?' . . o. ..
- i n l
CASK P O O L. v,y S E C Tlok/ @ t/ SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CASK POOL COVER INSTALLATION FIGURE 4.7-12 P,h)N w' ^m ."- - f}5 ' UaIwMSfW .n W _ ' j: - f .j " ,N A m_p ? bR ', = S $ gNCfM e C ~ .l + <y' q .
- . .a_
%f % Ms,g# wWWg$ m g _ + ;. gx M;yw ~ n d~ 7 , q . ~ E - . f~ hy:! .Y c *wW ~ ~ - .- mp /n&@ e q: ,. f%" , 0,4 $ id~ u%M:q4 's .s , g& p AS +m~ ; x . haps. t n m"" , b %gg myypwp ' f . Cn Y w u+ o M R ~ 's + . -A i 7: . D R YC 2,Q ~ KH S S EA .O RE AAR PTN3;""e**h,' CWA s- ~ 4 C ,M EAR N Am fLr . n 1 = '2~ . sc2 , ' _ T(;G-C L ,by 1 - ; _ ~m % .w C ~ . + . y4 : > ,wunO',- mr ;i N @j ' bg R Kt Ev SoV n AoO 3 C p C.e - N 'm n - 77 ~ i a'
- c
NT N ie . m. g I L nG UIN> D , 'f D N s nN(Um :g^ = % ism 7_ Pe 7yc 6 g3,- - 3.S ,a. 1 L , sn?Aj% ;2 ? % .% SP5H E . a ~ '- gf[(" - . A j a;e - ;gQ'j=; ,. w' , wx 1- , , ]6.m}5 H 5 nL - i - = . c. ~
- ,j 's i' G Lm%
Me. . r~ J p" S M , 9 ,. G 7 m. + 7'- - g- w i-v_ .-_ - . yw r a e - e & ;_' g(1_<.wVM"b a 4 ,h[v
- .~ -
5 }f :. mp'u:hyg [ rC gG F Ffpf ;1 -(: w r - ' - e,r \ n e r-c : {;, . 5 g- t lIl1l i\l'!ll 4:.' : y , _ A,&ys' i:x3w3sn<. j* s.$.g W ha <. g3 am %[& s - ~ wo 2 pLJ ;Y:+1Qd,fn ' ~ h.. :g5 s < m e,"~-@V.,~ag'*i"i'33;yAnqn:,ill L ' ~ y- [ 2 2 - . #o- -'x ii a. OXImwa' E l'T.# r N ;y : t - A , 3Q>Ad;- f:n A s a.NmT gD~:,c<. qM, wgew_~Y~g u n
- a ,. -
; c mWR,.n'g. Ce fup ,<' l;M ' s 1.s#- G. - . 0 ty ~ ^ yf w# ' ~ i* ;.ra ,r ' WJ am",- I g
- gNc
- .e.pM;gg w
W D a.Ng . ' ,% g%- h5 ;;% a;y^ c~ OLm t wn.; n: ? ; n[4' ws m4.MAw c;w- ^?+ ^ ,- ; L gN> "e@" . % c. J wpaN i - x .sm An ? ' f L .m r ' '%Wm #wg' ~ . .w= _^ -W~v s #:m., , .- % .V <aA - eN;. _ s . Y / _ , v=~ - ,5. f' _ W'1 f . Y. _ '- ,.b h*2 < t-h # , O .s . A - ,M,M':'.(& 6 R - R. A. &r., % $ D D2+ G h' ar
- a fRuewwaw/
- w ;. ~ KH SSA S& jfMe M sG Lx 4.k't h P N AAE R Ag:;e I ~ 2 yh E R t - CW A . Cni%- _ .$A2O- 2? : .T _ .%@~i ; nsW, e.c R@%-
- a. O [c.' T :
,m, J. s s* .;u - 9/ . *Q p w '_N t dm .i 2(Y- . pM, uy - ;L R , ..g:K L E 8 s,S i-O V mg S $ x-c AOO ,! C.PCawQE#t s l ^ si Y w RmMM A R R,S Oh@ O SF ~ 9 - ;.P ..c Q, MC ETC - 9 2lm,E A.u uf ;_ *= m g r~ e y E^ N ~/t: - ;=s Y c. %:;c 'C '- 3 P _ , w AN~a:=y@e.y% & ' R %&m ?C+m,zN gw~?.%m Se : ( _ ,f(s 1 g ' ,a ,; ,H u yA g ;> h'r S n_ . ~.,.' M u T .6 f~ M % s.- . . = . _a'~ . w3.nT - s.X _ L _ #, Rm s- 'j j-e- - ~ - ,:c3' . . 7 sk5Yn - 1 (Eya yngu;Qw\,'n 4=.#.i. A ~ wn ; ; k .
- 2*
- pJ v
e ~ nht .A ~4 Q:l#<'-T_i ' M' 4s - j 1 $ Wfy , y,s-k#gp g~. F ghruy ya- s.w? .a. ~ - 5 . ,xk . %' - c _ rV , :=; ..- A;%?r$ Ni2 i r.' ~ ikeLtwh j r y !g . s.<<. w _ m- 1 l l l SPENT FUEL POOL ~ 7, (UNIT 2) N.]7 1 N l t J, SPENT CASK HANDLING ,. POOL 2 FUEL Yf e i , 4S O (J w - r-POOL J REGION II "! e ( STORAGE $ 0 RACK I U_ ' p r I I I I I I e 6 "8 l l l
- b a
l2 4.82 _ ._9. 5 9 14 4 _ PLAN VIEW l SAN ONOFRE i NUCLEAR GENERATING STATION Units 2 & 3 , ( NOTE: ALL DIMENSIONS TEMPORARY CASK POOL - V] SHOWN IN INCHES STORAGE RACK SHEET 1 OF 2 FIGURE 4.7-14 a . SPENT FUEL.: POOL yy (UNIT 2)- p
- L___________! U l_______________._J 9.59_ _
...b ... ,* k. ,f, ,. 4:i' *s ' TOP OF- CASK '. POOL COVER ' OPERATING FLOOR i i N = :4 '. ' R
- r. n,
5 . 4,, -g 4y m y- POOt. GATE . . .I OPENING - 8, ,. f,.~. s .j(T3 g 2'- 0" . :. g i. ,4 . ' *A* . 3, ,,- ? .'.s~;. s .<,g .. .a . ,f,t: i- i .' ' - M-SECTION VIEW LOOKING SOUTH SAN ONOFRE NUCLEAR GENERATING STATION l, Units 2 & 3 .I i
- 5.1 -
'\~ STOF, AGE RACK SHEET 2 OF 2 FIGURE 4.7-14 __z __i_ _ _ ___2 _ . _ _ _ _ _ _ _ _ __ _ .) 4.8' bORAFLEX TESTING AND INSERVICE SURVEILLANCE 1 4.8.1 PROGRAM INTENT The inservice surveillance program for the Boraf1'ex neutron absorber material ~is~ designed to monitor the performance of the Boraflex over time in the SFP environment. The program will' provide' data on Boraflex behavior without affecting the function of the Boraflex fabricated into the spent' fuel storage racks.. The neutron absorber' rack design includes a Boraflex verification-view-hole in the cell or wrapper wall so'that the presence of. Boraflex material may be visually confirmed at any time. Upon completion of' rack. fabrication, such an inspection is performed. This visual inspection, coupled with the E Quality Assurance program controls (confirmed by SCE inspectors) and the use of- ,{ } 1 qualified Boraflex neutron absorbing material, satisfies an initial-verification test to assure that the proper quantity and placement of material was achieved during fabrication of the racks. This precludes the necessity for onsite neutron radioassay measurements (blackness testing). 4.
- . TEMPORARY CASK POOL l
8.2 DESCRIPTION
OF SPECIMENS The-Boraflex coupons used in the surveillance program will be representative of the material used within Region I and Region II locations. They will be of the same composition, produced by the O TMFWOO57 4.8-1 1
same method, and certified to the same criteria as the production lot Boraflex. The sample coupons will be of the same thickness
~( )as the-Boraflex used within'the' storage system. Each Boraflex specimen will be encased in a stainless steel jacket of an identical alloy to'that used inLthe' storage system, formed so as to encase the Boraflex material and fix it in a position similar to that designed into the storage system. 'The jacket will be mechanically closed without welding'in such a manner as to retain its form throughout the use period yet allow rapid and easy opening without contributing mechanical damage to the Boraflex specimen contained within. There is at least one coupon specimen from each production lot of Boraflex. The specimens will be located in the SFP such that they will receive a representative exposure of gamma radiation.
) The specimen locations will be adjacent to designated storage cells stich that the straps used to suspend the specimens can be -removed at any time, providing access to a particular specimen. I 4.8.3- SPECIMEN EVALUATION i The coupon evaluation at SONGS 2&3 will be based on EPRIs
" Appendix C - Guidelines for a Standard Boraflex Coupon Surveillance Program" dated December 1988(23). Basically, istandard, special, and archive coupons will be obtained from all i lots of Boraflex used in the racks. Standard coupons will be examined and returned to the pool according to a schedule i TMFWOO57 4.8-2 ; 1
relating to refueling outages. At least one special coupon will 1 be removed approximately every 5 years for more thorough examination and will not be returned to the pool. Archives will be maintained outside the pool for comparison to both standard and special coupons. Scheduled evaluations of the Boraflex coupons will include at least the following examinations: A. A visual inspection of the general overall appearance of the coupon will be made. Any qualitative changes in surfaces, outline and general appearance will be recorded. Length, width, and thickness measurements will be made to B. determine if any changes from pre-exposed dimensions have occurred. ll C. The coupon will be weighed. The weight will be recorded and any change noted. D. Shore A hardness measurements will be taken to provide e rough estimate of the gamma exposure and to detect any gross softening of the Boraflex. E. Rudioassay of the surface of some of the coupons for beta and gamma radiation will be performed to provide an indication of the extent of water permeation into the coupons. O TMFWOO57 4.8-3
j F. Neutron attenuation measurements will be made on special coupons to verify the uniformity of B 10 loading at several locations on the coupon. G. Specific gravity and coupon volume by immersion will be determined for special coupons. As discussed in subsection 4.7.2 irradiation tests (21,22) have been previously performed to test the stability and structural integrity of Boraflex in boric acid solution under irradiation. These tests have concluded that there is no evidence of deterioration of the suitability of the Boraflex material through a cumulative irradiation in excess of 1 x 1011 roentgens gamma radiation. As more data on the service life performance of f Boraflex becomes available in the nuclear industry in the coming years through both experimentation and operating experience, SCE will evaluate this information and will modify the surveillance program as determined warranted and justified. To assure.that the gamma ray exposure to the SONGS Boraflex sheets is limited to that exposure previously demonstrated to cause no significant material deterioration, i.e., to less than 1 x 1011 roentgens, administrative controls will be instituted. These controls will direct that to the extent practicable, a fuel assembly will be discharged from the reactor core into the rack location in which it is expected to remain. Furthermore, an exposure history for each rack location will be maintained to O TMFWOO57 4.8-4
.i IL
; .3 ; include:' (1) the residence time if greater than 90 days, and j 1j
'(2) the age if less than 1 year from time of reactor discharge, for each fuel assembly having resided therein. .. l 1 'O -i lO TMFWOO57 4.8-5
i '
, .14.9 REFERENCES
L 1'. United States AEC Report TID-7024, " Nuclear Reactors and earthquakes",' August 1963. 4 2. USNRC Regulatory Guide 1.29," Seismic Design Classification," Rev.3,'1978.
- 3. 'WECAN " Documentation of Selected Westinghouse Structural Analysis Computer Codes," WCAP-8252.
- 4. WECAN -~" Benchmark Problem Solution Employed for.
Verification of the WECAN Computer Program," WCAP-8929.
'5. D. F. DeSanto, "Added Mass and' Hydrodynamic Damping of Perforated Plates. Vibrating in Water", ASME' Journal gf.'Pr.gssure Vessel Technoloav, May 1981.
- 6. ~R. J. Fritz, "The Effects of Liquids on the Dynamic Motions-of Immersed. Solids," Journal of Engineering'for Industry,-
Trans'. of the ASME,' February 1972,.pp 167-172.
- 7. USNRC Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants," Rev. 1, 1973.
.g . -8. " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility", Prof. Ernest Rabinowicz, MIT, a report for Boston Edison Company, 1976.O TMFWOO57 4.9-1
!5 ti ( :9. ASME' Boiler &-Pressure Vessel Code, Section III', Subsection-f NP (1986 Editit up'to and including A-86 Addenda). !. 10. Arthur P.-Fraas;and M. Necati Ozisik, Heat Exchanaer Desian, John Wiley & Sons,: Inc., New York,'1965,:p. 295.
- 11. J. S. Anderson, "Boraflex Neutron. Shielding Material - '
Product 1 Performance' Data," Brand Industries, Inc., Report' l
'748-30-2 (August 1981).
- 12. . J. S. Anderson, " Irradiation Study of Boraflex Neutron-Shielding-Materials," Brand Industries, Inc., Report 748-10-1 (August 1981).
'13.: J.-S. Anderson,f" Irradiation Study of Boraflex Neutron Absorber-Interim Test Data," Brand Industries,-Inc., Report No. NS-1-050f(Interim), Rev. 1, 11/25/87.4 a- 14. J. S. Anderson, "A' Final Report on the Effects of.High Temperature' Borated water Exposure'on BISCO Boraflex Neutron Absorbing Materials," Brand Industries, Inc., Report 748-21-1 (August 1978).
- 15. WCAP 8370, The Westinghouse Electric Corporation Quality Assurance Plan, Revision 9, Amendment 1, February 1978.
O TMFWOO57 4.9-2
i 1 r
. 16. ) . Nuclear Regulatory Commission, " Control of Heavy Loads at, . _;ii Nuclear Powerplants",fNUREG 0612, July;1980.
- 17. H.LE.-. Flanders, " San.Onofre Units 2'& 3' Fuel Rack Seismic Analysis Methods and Parameters", WNEP 8901, Westinghouse Electric Corporation,LNuclear Components Division, Pensacola, Florida, January.1989.-
- 18. .A. Higdon and W._B. Stiles, " Engineering Mechanics, Vector Edition", Prentic-Hall, Inc., Engelwood Cliffs, N.J., 1962,
- p. 639.
- 19. V. N. Shah,-G. J. Bohm, and A. N. Nahavandi, " Modal Superposition Method for Computationally Economical
' Nonlinear Structural Analysis",-ASME Journal of-Pressure() LVessel_ Technology, Vol. 101, May 1979, pp. 134-141.
- 20. 'V. N. Shah'and C. B..Gilmore, " Dynamic Analysis of a Structure with coulomb. Friction", Submitted for Presentation at'the 1982 ASEE Pressure Vessel Piping Conference, Orlando, Florida, June 27-July 2,.1982.
- 21. Bisco Products, Inc. Technical Report No. NS-1-050 (Interim), " Irradiation Study of Boraflex Neutron Absorber, Inerim Test Data", June 25, 1987.
I i O TMFWOO57 4.9-3 _ _ _ ____.----------------_--_ - - _ _ _ . )
- 22. R. W. Lambert of Electric Power Research Institute, May 26, 1987 memorandum to Attendees of Boraflex Review Meeting at the EPRI Workshop of May 20, 1987, EPRI-RP-2813-4.
- 23. Electric Power Research Institute, EPRI NP-6159, "An Assessment of Bor3 flex Performance in Spent-Nuclear-Fuel Storage Racks", December 1988.
- 24. December 27, 1978 Proprietary submittal to the NRC, J. G.
Haynes (SCE) to R. L. Baer (NRC) letter Dc.cket Nos. 50-361 and 50-362.
- 25. Bechtel Structural Analysis Program (BSAP), CE800, Version F1-54.
O O TMFWOO57 4.9-4
5.- COST / BENEFIT. RADIOIDGICAL AND ENVIRONMENTAL ASSESSMENT 5.1 COST / BENEFIT AND ENVIRONMENTAL ASSESSMENT The cost / benefit' analysis.and environmental assessment for the proposed;reracking.of SONGS 2&3 are presented in the following. sections. 5.1.1 NEED FOR INCREASED STORAGE CAPACITY A. Currently, ths licensed: capacity of the Units 2 and-3 pools is 800 storage cells each. Table 5.1-1 provides the.
. anticipated-refueling schedule for SONGS Unit 2 (Unit 3 schedule is similar with refueling dates approximately 1 year later) .- and the expected number of fuel assemblies that will be transferred into each of the SFPs for SONGS 2&3 at each refueling. Also shown in table 5.1-1 are spent fuel transfers from SONGS Unit 1. Unit 1-fuel..is currently being stored at Units 2 and 3 to facilitate Unit-1 continued operation.B. With the present licensed capacity of the Units 2 and 3 pools it will be impossible to' conduct a full refueling beyond the cycle 7 outages in 1993 and 1994, respectively. Loss of full core reserve will occur in 1991 and 1992, respectively.
-TMFWOO'57' 5.1-1 4 ) --:-_-_-__-_______u___ - - - - - - _ _ _ _ _ _ _ _ _- -.f C. The SONGS 2&3 SFPs are expected to contain approximately
/,N.() 480 spent fuel assemblies each at the time of reracking, during cycle 5 operation. D. Based on the proposed increased storage capacity of 1572 spent fuel elements, the estimated date when the SFPs would be filled is as provided in table 5.1-1. Full core reserve would be retained through cycle 11 which begins in the year 2001 for Unit 2 and 2002 for Unit 3. E. Adoption of this proposed spent fuel storage expansion would not necessarily extend the time period that spent fuel assemblies would be stored onsite. Spent fuel will be sent offsite for final disposition ender existing contract with the U.S. Department of Energy (DOE) pursuant () to the Nuclear Waste Policy Act of 1982. Although the contract specifies start of spent fuel acceptance in 1998, the Federal facility is not expected to be available prior
' to 2003.5.1.2 ALTERNATIVES Southern California Edison has considered and evaluated various alternatives to the proposed increase in spent fuel storage capacity at SONGS 2&3. The storage of spent fuel at SONGS is an interim solution until a Federal repository becomes available to receive spent fuel from SONGS 2&3, currently expected to be in the year 2003 in accordance with the Nuclear Waste Policy Act of
- i. \.
TMFWOO57 5.1-2 ! l w _ _ __ - _
1
'1982 (Public Law 97-425). Furthermore, commercial reprocessing of spent fuel has not_ developed as had been anticipated at the time these' facilities were' designed. The NRC issued in 1979 a final " Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Reactor Power Reactor Fuel"(1) which analyzed alternatives for the handling and interim storage of spent fuel as well as the possible restriction or termination of the generation of spent fuel. One of the fuel storage alternatives considered in detail in the Statement is the expansion of onsite fuel storage capacity by modification of existing SFPs which is the proposed approach for SONGS. However, since there are variations in storage designs and limitations, for example as a result of spent fuel already stored in the pools, the Statement recommentis that' an evaluation of other alternatives be done on a case-by-case basis.O Southern California Edison has identified alternative spent fuel storage methods currently utilized by the nuclear industry and evaluated their acceptability for use at SONGS. These methods include: A. High density racks in existing SFPs B. Fuel rod consolidation in existing racks C. Construction of a new SFP at SONGS The following additional alternatives are currently not utilized by the nuclear industry but were evaluated by SCE for their applicability and feasibility for SONGS: O TMFWOO57 5.1-3
() D. E. Shipment of spent' fuel to a-reprocessing facility Shipment of spent fuel to a Federal or commercial' storage / disposal facility F. Shipment of spent fuel to another reactor facility I C. Reduced generation of spent' fuel H. Shutdown after current capacity exhausted These alternatives were evaluated with respect to their environmental impact, time needed to'become operational, and the overall cost. Each alternative is addressed below. A. High Density Racks in Existing Spent Fuel Pools The use of high density storage racks in the existing SFPs D
'(_,b has been selected by SCE as the approach to increase the storage capacity in each of the two SFPs from the 800 to 1572. Of the approaches currently being utilized in the nuclear industry, this approach has been determined to have the smallest overall impact and to meet the time schedule need for the expanded capacity. This approach is described in this report.B. Fuel Rod Consolidation in Existing Racks l' The existing racks were not designed for consolidated fuel loads; therefore, this alternative is not feasible. j TMFWOOS7 5.1-4 _ _ - - - _ - - - - - - - - ---- - -a
C.- Construction of a New Spent Fuel Storage Pool at SONGS p)-
\- ' Additional storage capacity'could be developed by one or -l more new and independent spent fuel storage. facilities in a separate building on the SONGS site or offsite, either similar to the' existing pools or as.a dry storage installation using a vault, metal cask, dry well or concrete cask concept. Federal law precludes construction of additional fuel storage facilities on a military reservation. The SONGS site is located on a military reservation. While such' systems are technically feasible, SCE has determined they would'be far more costly than the use of high density racks in the existing pools. These I alternatives are not suitable for the SONGS site due to severe space limitations in general and offsite dose l'~) .(_j limitations at the plant boundary for the dry storage lconcept. 'In addition, a new, site specific design and construction of such facilities, including equipment for the transfer of spent fuel, would be required. It is not likely that such.an effort could be completed in time to meet the need for additional. storage capacity. The same considerations, with the possible exception of space limitations, would apply for an offsite construction. However, the associated site specific evaluations and I licensing aspects would be compounded. Furthermore, such construction would not utilize the existing expansion capabilities of the existing pools and thus would be. 1 O TMFWOO57 5.1-5 h l --=:_-___-____-_ 1
m .c {,; ;,. s li s wasteful. -Southern California Edison;has concluded that~ the construction'of a new spent fuel. storage ' facility is; : o-not a reasonable? approach for SONGS 2&3.. D.- Shipment of Spent Fuel to a Reprocessing Facilityf 1-No commercial or: Federal' facilities are presently in operation in the United States for the reprocessing:of. spenti~ fuel from commercial reactor facilities ~. Furthermore, there is no prospect for such facility in'thei foreseeable future. While such; facilities exist-outside the United States, use of these facilities for the reprocessing ofl fuel.from U.'s. facilities is not permitted'
.by' law. This' approach,.therefore, is not a viable solution.'O E. 1 Shipment of Spent Fuel to a. Federal or Commercial Storage / Disposal Facility Shipment of spent fuel to a permanent Federal fuel storage disposal' facility is the preferred alternative to increasing and maintaining adequate spent fuel storage capacity at SONGS 2&3. At the present time, there are no existing Federal or commercial disposal facilities available in the United States. . Southern California Edison has made contractual , i arrangements with the DOE whereby spent nuclear fuel O .TMFWOO57 5.1-6 l ii _____..____.__._..___________.__...i____ _
. and/or.high' level nuclear waste will-be accepted-and j disposed;of:by the U.S.' DOE-when such-facilities will be available. While plans are being formulated bysthe department lfor the construction;of a spent _ fuel repository ~in accordance'with:the-Nuclear Waste Policy Act, this
~ facility is-not' expected to be available to accept spent fuel any earlier than the year 2003.- ' Southern. California ~ Edison'has an existing' contract with the General'ElectricLCompany Morris Facility in Morris, Illinois but only for the storage of 88 additional Unit 1 Lspent fuel assemblies and no storage of Units-2 Land 3- ' spent fuel. ~ .In summary,,sh'ipment of spent fuelLfrom SONGS 2&3 to a- ~ ~ Federal orlcommerical storage or disposal facility'is not a viable approach at this time..L F '. - Shipment of1 Spent Fuel to Another Reactor, Facility The National Waste, Policy Act and 10CFR53 clearly places the responsibility.for interim storage of spent fuel with each owner or operator of a nuclear power plant. Shipment of SONGS.2&3 spent fuel to another reactor facility could potentially provide-a.short term relief to the storage capacity. problem'at SONGS. At present, SCE has~no license
'to transship fuel to another facility. Transshipment of spent fuel to another facility will not result in any O TMFWOO57 ~5.1-7
+I m ,
. additional net long term storage capacity. Permanent'_transfereof; SONGS 2&3, spent fuel to other facilities would
~onlyecompoundLproblems at those/ facilities, for example, 6
'potentially greater environmental impact than~ associated j jwith thefproposed increased storage at SONGS 2&3.- At-the
'l presenti SONGS Unit 1 fuel is being stored'in the_ Units 2 l and-3 SFP to facilitate Unit 1 continued operation.- 1 l +Based on~the~above considerations it-is concluded that the' transshipment?of SONGS spentJfuel to another' reactor facility is not a viable solution. l G. Reduced' Generation of' Spent. Fuel' Improved usage of fuel in the reactor of SONGS Units 2&3 () and/or operation;at a reduced power level would extend'the life of the' fuel-in the reactors. In'thefcase of extended-burnup of fuel assemblies'the fuel' cycle would be extended and fewer offloads would take place. However, the
~current, storage capacity for 800: assemblies would still be exhausted. prior to availability of a federal' repository. Operation at reduced power would not make
-effective use of'available resources, thus causing economic penalties. This is not a viable approach.E ['2 TMFWOO57 5.1-8
-~ .in m__-_-_ .____E__-. - _ . _ _ _ . _ _ _ _ _ _ . - _ _ _ _ _ _ - . _ . _ . . - . . . _ _ _ _ _ _ _ _ . . - _ _H. ' Shutdown After_ Current Capacity Exhausted , f /y 1 iX_/f ) Table 5.1-l'provides the anticipated spent fuel' storage- < needs for SONGS Unit 2 (Unit 3 schedule is similar with refueling dates approximately 1 year later). If the SFP storage capacity for each unit were to remain at 800, the storage capacity would become exhausted in. 1993 and 1994 assuming _no further action were takeni and
~each unit would have to be shut down at the end of the following cycle. This stop in operations would result in no further generation of spent fuel, thereby eliminating _ the need for increased spent fuel storage: capacity. The impactiof terminating the generation of spent fuel by
.O . (_/ - ceasing the' operation of existing nuclear power plants
- -(i.e., ceasing. generation of electric power) when their SFPs'become filled was evaluated in the final NRC Generic Environmental Impact Statement (1) ;and found to be undesirable.
L Southern California Edison calculated estimates for costs ! of replacement power based on the most recent rate of return. The assumption was made that SONGS 2&3 could be 1 operated without maintaining full core reserve capacity. Thus, Unit 2 and Unit 3 cycle 7.in 1993 and 1994, respectively, would be the last refueling possible with the existing storage capacity. Tables 5.1-2 and 5.1-3 O TMFWOO57 5.1-9 __=____-__-_
l '
' indicate.the average yearly fuel cost increases'for SONGS -( -2&3"after~ shutdown in 1995 and 1996. Plant shutdown would-place-an unacceptable financial burden on the residents within SCEs service area and cannot be justified.Shutdown of SONGS 2&3 after exhaustion.of current storage capacity is not a viable solution. In summary, based on its evaluation of the currently existing. methods, SCE has selected the use of high density. racks in the existing SFPs. Fuel rod consolidation, potentially will be proposed'in the future for additional spent fuel storage capacity. Construction of new storage capacity has not-been selected because of Federal law, time, and cost restraints. 5.1.3. TOTAL RERACKING. COST The total combined cost associated with the proposed Units 2 and 3 reracking modifications is estimated to be approximately
$50,000,000. This amount includes all costs for design and fabrication of the new spent fuel racks. The total combined cost above also includes all costs for engineering, installation, and support costs at the site, and all overhead costs, including allowance for funds used during construction.The prime reason SCE wishes to rerack during cycle 5 operation is to minimize the risks and dollar costs associated with removal of old racks and SFP hardware and new rack installation with a heavy TMFWOO57 5.1-10
L l ~ j '. Jspent fuel load in-the SFP. At the beginning of cycle.5 ?" () approximately 480'of.800' rack 1ocations will be. filled. No other U.S. nuclear power plant has reracked with 480 rack-locations.
~filled. If delayed until cycle 6, an additional 108 fuel assemblies will be in the racks, increasing the fuel inventory up to approximately 588. Reracking during cycle 5 will provide for " the following: A. An improved exclusion zone between stored fuel and construction work areas. This margin is important for meeting ALARA goals. Although SCE will_ conduct underwater removal of old racks and hardware remotely as reasonable and practical, it is likely that divers will be used for the difficult: tasks. () B. Significantly reduced spent fuel shuffling'from the old' racks to the new racks assuring that transshipment between' Unit 2 and Unit 3 will-not-be needed. C. More margin from lifts near spent fuel.when removing old racks. If reracking were to be delayed until cycle 6 operation, an additional factor of transshipping between SONGS 2&3 would be involved. Any proposed fuel transfer between Units 2 and 3 has several unknowns at this time: O TMFWOO57 5.1-11
'A. Our' current GE-IF-300 seven-element cask is not g(( ji necessarily available to us since we have not entered any <L long-term lease or purchase agreement with the cask vendor. J B. The GE-IF-300 cask is not compatible with our Units 2 and
- 3. fuel without some limited hardware and license
[ modifications. C. If.we use another cask, other than IF-300, then we will require a' licensing submittal (amendment) to validate the acceptability of another cask's use in Units 2 and 3. The additional costs associated with transshipment would be approximately.$1,000,000 for startup and $10,000 to $15,000 per () spent fuel. assembly. These costs are based on the recent experience with transshipment from Unit 1 to Units 2 and 3. J 5.1.4 RESOURCES COMMITTED Reracking of the spent fuel pools will not result in any irreversible, irretrievable commitments of water, land, and air resources. .The land area now used for the SFPs will be used more efficiently by safely increasing the density of fuel storage. l 1 O TMFWOO57 5.1-12
jj The materials used'for new rack fabrication are discussed.in F [~} subsection 4.7.1. These materials are not expected to I '^ significantly foreclose alternatives available with respect to 1 any other licensing action designed to alleviate shortage of
]l
. spent fuel storage capacity. '5.1.5 THERMAL IMPACT 0;; THE ENVIRONMENT Section 3.2 considered the following: the additional heat load and the anticipated maximum temperature of water in the SFP that would result from the proposed expansion; the resulting increase in evaporation rates; the additional heat load on component i and/or plant cooling water systems; and whether there will be any significant increase in the amount of heat released to the environment. The maximum SFP heat load increases from 40.3 MBTU/h to 51 MBTU/h, an increase of 10.7 MBTU (in the worst
- O)
~case). The thermal power of the plant is 3390 MW, and the electrical power is 1100 MW, which means the heat rejection to the environment is 2280 MW. 2280 x 10 6W x 3.412 BTU /h-W = 7.78 x 109 BTU /h = 7780 MBTU/h 10.7 . o,0014 . o,14g 7780 The proposed increase in storage capacity will result in an insignificant impact on the environment. O TMFWOO57 5.1-13
O s e ' ae lt . at i rna l- ar , d b) eed e a m3 ytet n o
. e( .1ao l r a. o l d f sl 2081997755364208 so 7665512895645677 'yl s n ts t f o Aop 124578902356789 . 11111111 lo o o - 3 l eosi .n e & an t pet n r 2 tI a ia ws o s o m3lc oi c T i .b o h t xdml sg l i one n l n rase ti u U p sg ok f S 1 p2aa nc r D a r a n o E t ) t0o er i f E i s2 si7t re a N n e( en2s tU ar t n 1 3 Uid E - le a e8 sg i 0 G fbG . deh8 en a 2 A omr - - 5020202O50000 .ht g om si 'm R ea - 7 5 5 5' 7 gt pu s 8 O rsh - n tn T)U est - i oai rs k 1 us c 1 S bas lthn
- 2 m i e ti pa a r- 4 1 L . u D ud a s, r e
. ET N . fe,m b 5 UI e g4 e uP y o FN rr0r oF t t e U a0 e S yi c l T -hh2e n ts O b N- 2 tc h agin s _O a T E PS SG t isd i snt lnle wii
.d t r ea liid i etb eN nee aerh csah si pg t EO Uia a TS lr - l rit i x ai A fba 2888888888888888 i ap mech d M omh 7800000000000000 m xd - I . ec 11111111111111 i sen rede n X rss . se a ohah o O esi il ftot i R BAD sll ,. l t P m i bidddhfh a P u mweeecfc r A N n z si oi i 3 tssul e'e s ri u hwew .h p x i antis r e n eetneoe -
) ) )U1cruolcl e
'( -3 ( 6 ( i ic c s e 4679135791357913 t l t t t yl y n t 8888999990000011 .i ooaclc e a 9999999990000000 nn1nnc u c D 1111111112222222 wU o gf g i o teeln n l htirr i ni . sanaaetit .t o hU gaaayn n ry st ssartrte i ert i eii r e e ei s l baseeeeeeeeeeeee s h r r o p r pl e e muunnnnnnnnnnnnn u enguuuuuuuuuuuuu 2etrrto m oos d bp oir f vauJJJJJJJJJJJJJ tutMM tlta xsuspe e oJA i sa R N nshEEiaoaah UAtGGSLwLcT 12 34 5 6 O m 79*
l. I i L/~~l Table 5.1-2
'%) 'ANNUAL POWER REPLACEMENT COSTS ATTRIBUTED TO SONGS UNIT 2 j Months Yearly Power Present Value Cumulative Present Out of- Replacement Cost 1988 Dollars Value 1988 Dollars Year Service ($1000) ($1000)(U ($1000) 1988 0 N/A N/A N/A 1989 0 N/A N/A N/A 1990 0 N/A N/A N/A 1991 0 N/A N/A N/A 1992 0 N/A N/A N/A 1993 0 N/A N/A N/A 1994 0 N/A N/A N/A 1995 6 144200 58200 58200 1996 12 308306 111178 169378 1997 12 334506 107702 277080 1998 12 363752 104570 381650 1999 12 393608 101029 482679 200G 12 419808 96209 578888 2001 12 453319 92758 671646-
\' 2002 12 488049 89165- 760811 2003 12 510593 83289 844100 2004 12 542277. 78980 923080 2005 12 567258 73766 996846 2006 12 603207 70036 1066882 2007 12 '623923 64680 1131562- '2008 12 670230 62036 1193598 2009 12 703742 58159 1251757 2010 12 738898 54522 1306279 2011 12 775883 51117 1357396 2012 12 814695 47923 1405319 2013 10 712760 37435 1442754
- 1. Cost of Capital equal to 12%.
O TMFWOO57 5.1-15 _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ - . _ _ _ _ - _ _ - _ _ - _ - - - _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ .__-_-______-___-__----__--_-___--________---_a
1y Table 5.1-3 L] ANNUAL POWER REPLACEMENT COSTS ATTRIBUTED TO SONGS UNIT 3 1 l J Months Yearly Power Present Value Cumulative Present Year Out of Replacement Cost 1988 Dollars Value 1988 Dollars Service ($1000) ( $1000) ") ($1000) ) 1988 0 N/A N/A N/A 1989 0 N/A N/A N/A 1990 0 N/A N/A N/A 1991 0 N/A N/A N/A 1992 0 N/A N/A N/A 1993 0 N/A N/A N/A 1994 0 N/A N/A N/A 1995 0 N/A N/A N/A 1996 6 160208 57773 57773 1997 12 330870 106531 164304 1998 12 375365 108023 272327 1999 12 389910 100080 372407 2000 12 430500 98659 471066
/~} 2001 12 452640 92619 563685 -('~' / 2002 12 496920 90785 654470 2003 12 513525 83767 738237 2004 12 548580 79898 818135 2005 12 574410 74696 892831 2006 12 605160 70263 963094 2007 12 635295 65859 1028953 2008 12 667890 61820 1090773 2009 12 701285 57956 1148729 2010 12 736340 54333 1203062 2011 12 773178 50939 1254001 2012 12 811800 47753 1301754 2013 10 710325 37307 1339061 l
- 1. Cost of capital equal to 12%.
1 TMFWOO57 5.1-16 l l
5.2 RADIOLOGICAL EVALUATION 5.2.1 SOLID RADWASTE Currently, contaminated resins are generated by the SFP purification system. The SFP purification system will not require upgrading as a result of reracking. No significant increase in volume of solid radioactive wastes is expected due to operation with the new racks. It is estimated that a minimal amount of additional contaminated resins will be generated by the SFP cleanup system during reracking. The currently calculated and predicted isotopic inventories of the SFP resin are presented in table 5.2-1. The SFP purification pump activity was calculated by assuming its o) l internal volume is filled with refueling cavity water at the start of the refueling operation. The fuel pool filter activity inventory, or loading, was calculated by assuming all of the
" CRUD" present in the refueling cavity at the start of refueling operations is deposited within the filter, and then decayed for the duration of the refueling outage. The SFP ion exchanger activity inventory, or loading, was calculated by assuming all the radioactive nuclides in the refueling cavity at the start of the refueling operations, except noble gases and tritium, are deposited within the ion exchanger, and then decayed for the duration of the refueling outage. Since noble gases are not deposited in the filter and ion exchanger, refueling water O TMFWOO57 5.2-1P B i i activities'at the-start of-the refueling. outage-are used;to yield' p ., l maximum component inventories. JActual. filter.and ion. exchanger
-decontamination factors.are not;used, which yields ~ conservative results.m 5.2.2 GASEOUS, RELEASES The dose rate to offsite individuals. caused by rele'ases'from the s FHB is the. subject of the'UFSAR subsections,11.3.3 and 12.4.2.- Since tritium is the only airborne effluent and is not affected by the SFP raracking' modifications, Lit will continue to be released from the FHB to the environment through the continuous 'k
-exhaust vent stack at the rate.of approximately 585 curies' par.3-weeks of refueling operations, and at the-rate of approximately 320 curies per year-of normal' plant operations (see UFSAR table 11.3-9) which yields maximum offsite' dose rates on the order of l'x'10-4 rem /yr, which is negligible.. _l Table 5.'2-2 presents a listing of. typical radionuclides identified by gamma spectroscopic' analysis of FHB air samples.
=Because the quantities of airborne'nuclides which exist-in the-FHB are low, the actual contribution from airborne sources to pffsite dose rates is negligible.i O TMFWOO57 5.2-2
5.2.3 PERSONNEL EXPOSURE r~% Southern California Edison has evaluated the factors that contribute to the personnel exposure during normal and refueling operations at SONGS 2&3. Dose rates at onsite locations from the SFP are determined by taking into consideration reracking modifications. Because the refueling frequency does not change, the only result of the reracking modifications which could potentially impact dose rate analysis is the ability to store more fuel. The sources for radiation exposure are: (1) SFP water contamination, (2) fuel handling airborne activity, (3) SFP purification system components, (4) stored fuel, and (5) fuel fragments. O Since 23 feet of water shields personnel from stored fuel the exposure due to the stored fuel is negligible, on the order of 1 x 10-6 mrem /h. An increase in the amount of stored fuel due to reracking has an insignificant impact on the dose rate analysis. Dose rates were calculated along the perimeter of the SFP, and at the center of the pool's surface. The dose rates in the vicinity of the SFP during a refueling outage, due to the SFP water activity are presented in table 5.2-3. O TMFWOO57 5.2-3
Reracking modifications including storage of an increased number
. of spent fuel assemblies do not affect the refueling cavity }isotopic profiles; therefore, the SFP purification system components: (1) source strength spectrums used to determine doses'do not, change; (2) design basis dose rates do not change; and (3) resin replacement and backflush frequencies do not change. Furthermore, the ion exchanger resin replacement and the filter backflush processes are remotely operated and personnel are not exposed. Doses due to the radionuclides in the SFP were not calculated in the original design calculations so no comparison can be made between existing and new values. For doses due to FHB airborne activity, the existing values are ( ) given in the SONGS 2&3 UFSAR, table 12.4-9. A comparison of these values with the new calculated values shows that the annual doses for normal operation and for refueling outages will not increase and will remain well within the guidelines given in 10CFR20. It is not anticipated that the Health Physics program for SONGS 2&3 will need to be modified for this increase in storage capability. I O TMFWOO57 5.2-4
5.2.4: RADIATION PROTECTION DURING RERACKING ACTIVITIES' 5.2.4.1 General'Descrio' tion of Protective Measures-5.2.4.1'.1L Introduction' The raracking project for SONGS SFPs'is being planned in. accordance with SCEs commitment to its as low ~as is'_reascnably-
' achievable (AIARA) policy. The combination of' routine radiochemical analyses, radiological surveys', and' extensive contamination control experience will provide adequate technical ' bases for implementing comprehensive radiological controls and ensuring personnel doses are maintained AIARA. Notwithstanding, the Health Physics organization has performed a' radiological characterization of the SFPs. This formal program of sampling, sample' analysis and evaluation, and direct measurements.in each of the three SFPs'will provide an extra measure of assurance toward; personnel safety and control of any waste materials which are generated. ,The Unit 1 SFP has been characterized because spent fuel bundles from Unit 1 will be transshipped to the Units 2 and 3 pools prior to the reracking.The results of the characterization effort will be used to make appropriate decisions concerning contamination control techniques, decontamination equipment and effective waste processing. The physical and radiological characteristics of SFP O TMFWOO57 5.2-5 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . . - - _ _ . - - )
f. l i 1 CRUD will affect'both procedures and equipment used for the - decontamination of equipment'and waste material being removed j from the SFP. f The use of remote methods will be evaluated for.each aspect of' the work. Divers will be used only when remote and other methods have been determined by both Health Physics and Construction to be impractical as determined by pre-established criteria. Construction will' develop each diving plan and Health Physics shall concur with it prior to each diving evolution. Health Physics and Construction maintain approval authority for each diving evolution. A parallel effort is being applied to planning for the use of divers should remote technology be impractical. Not planning for this possibility would violate AIARA principles. 5'.2.4.1.2 Spent Fuel Pool. Characterization A. Purpose The purpose of this' effort is to provide sufficient technical data concerning the radiological environment within each SFP to ensure adequate planning for providing surveys, establishing radiological controls and for waste processing, packaging, decontamination, shipping, and burial, as necessary. i O TMFWOO57 5.2-6
m - . r I 's e
, g L B. . ObjectivesJ l.Specific' objectives: include:the following:. p.
- 1. Identify radioactive nuclide types.and' concentrations and dose rate profiles for typical.in-poolicomponents.-
2.. Determine contaminants-and' concentrations [i..'SFP' water.
- 3. . Determine' sediment characteristics.-
- 4. ~ Provide preliminary 10CFR61 waste characteristics and classification analyses.
C. Application of Information The:information.obtained will be applied to-the g () appropriate portion of the Health 1 Physics program. lit is not anticipat'd e that'the Health Physics program will need
'to'be modified for this work.D. THealth Physics' Instrumentation The~information derived from the' characterization.will verify the calibration requirements for the selection of underwater survey instrumentation, airborne radioactivity survey monitor types and techniques, and the choice and implementation-of contamination control equipment. O TMFWOO57 5.2-7
1 f
.. A temporary monitoring' system will be utilized to monitor ,ry .Q gamma radiation levels in the SFP areas. Continuous air monitors will be used, as necessary, to support' specific activities'that may produce airborne radioactivity. Periodic radiation and contamination surveys will be j conducted, as necessary. I E. .' Dosimetry i All personnel in the work area will be required to wear i personnel monitoring equipment. At a minimum this will consist of a thermoluminescent dosimeter (TLD) and self-reading pocket dosimeter. Additional personnel monitoring equipment,'such as extremity or multiple whole body badges (plus remote readout equipment on divers) will be utilized as required. The calibration of dosimetry devices will be appropriate for the environment to be encountered. This calibration will be performed in accordance-with existing or revised station dosimetry procedures. l O TMFWOO57 5.2-8 l l
i i
'5.2.4.1.3- ALARA Activities '(f A. Preplanning Efforts 'Preplanning. efforts to maintain personnel doses ALARA involve each major discipline on the Project Team. Examples of'ALARA motivated preplanning efforts include, but are not limited to.the following: ^
- 1. Planning for all underwater work is being. approached' from the philosophy that the use of remote methods will be evaluated for each aspect of the work.
Prudent judgement dictates parallel planning'for the use cf a diver. Construction techniques, equipment, and work sequencing are being chosen accordingly. O 2. Construction sequencing shall include' provisions for using divers should the primary remote technique prove unusable.
- 3. The positioning of spent fuel to reduce personnel ,
exposure has been integrated into all planning.
- 4. Both preconstruction and construction decontamination efforts are planned to minimize the potential for fuel fragment and hot particle exposure and to reduce general area dose rates.
- 5. Appropriate training for complicated and/or high exposure and high risk tasks will be provided prior to task performance.
TMFWOO57 5.2-9
_ . _ = - - _ _ _ _ _ _ _ - - _ _ _ ___________ _ __ _ __ - - -- --. ._ _ y-U
- 6. Work evolut' ions will be controlled and monitored by Health Physics personnel at all times.
-) 'l B. ALARA Job Reviews i l The ALARA job reviews, both pre- and post-job, will be addressed in the Health Physics work control plan and performed in accordance with SONGS procedures.
"Tailboard" sessions, both pre- and post-evolution, will l be mandatory for each diver entry in the pool.and will likewise be addressed in the plan.C. Key Performance Indicators
.(~h Formally established key performance indicators (KPIs) %./will be tracked by the Health Physics organization with-the results and identified trends reported to management on a routine basis. These KPIs include, but are not limited to the following:
- 1. Diver entries into SFP
- 2. Personnel contaminations
- 3. Person-rem (especially diver)
- 4. Fuel handling building contamination levels and total FHB area (in units of ft 2) contaminated at that level
- 5. Gross microcuries/ milliliter in SFP water
- 6. Dose rate at SFP water's surface (poolside)
- 7. Dose to " immersion TLD(s)"
O TMFWOO57 5.2-10
i 5.2.4.1.4 Training Personnel shall-be trained and qualified in accordance with SONGS procedures prior to performing work in SFP areas. Appropriate task-specific training will be provided for high dose rate tasks. Diver's training shall include instruction in_nonitoring their equipment (diver's suit, etc.) and the work area with an I underwater probe. 5.2.4.1.5 Health Physics Work Controls A formal Health Physics work control plan shall be approved by i the Health Physics Manager. This plan will delineate each major radiological control element being applied and is in addition to
- SONGS procedures.
Specific work controls are intended to be used for diving operations (see paragraph 5.2.4.3). 5.2.4.1.6 Surveys l The types and frequency of radiological surveys will be identified in the Health Physics work control plan. All radiological surveys will be conducted in accordance with station procedures. TMFWOO57 5.2-11
a. Diving specific surveys are'to be performed'in accordance with
~ - ..t the established SONGS diving procedure.. A 3-D. survey of the:in-pool work area will be done periodically.to support. diving: operations.- The extent and frequency.of supplemental surveys. during the dive duration'will be established in the Health Physics work control plan. 5.2.4.1.7. Contamination Control n The major-elements of the. contamination' control program for.this project will be. delineated in the Health Physics work control
-plan. Normal Zone 3 requirements (fuel fragment / hot particle) are. adequate for all above water activities. These are contained in established SONGS procedures. . The station radiation protection staff will closely monitor and' control all aspects of the work to ensure personnel exposures are . maintained AIARA.5.2.4.1.8 Decontamination Efforts ongoing decontamination efforts are planned in the pool and during pool-side operations. Specific decontamination needs will be determined through periodic surveys delineated either on the Radiation Exposure Permit (REP) or in the Health Physics work control plan. Any diver entry shall be preceded by vacuuming of the work area. The diver will normally vacuum his way into and out from the work site. O TMFWOO57 5.2-12
l LAll equipment and'other items removed from'the water will be' appropriately decontaminated. Tools'and equipmers will be removed from the SFPs in accordance with station procedures. Additional decontamination, if required, preparation, and: packaging for shipment offsite will be done in designated prepared areas. 5.2.4.1.9 Solid Radwaste Materials Material and equipment used for reracking, and any waste hardware or solid materials generated, will be decontaminated and/or packaged, transported and disposed of in accordance with applicable Federal and State regulations including 10CFR71 and !
- 49CFR173. The 10CFR61 waste characterization analyses will be ]performed to determine burial classification and packaging f i requirements. ] J 5.2.4.2 Anticipated ExDosures Durina Rerackina
.i These estimates are made based on the anticipated construction plan, including fuel transfers, the use of remote tools, and divers, and the onsite decontamination, cleanup, and packaging of the old storage racks. Also, current pool radioactivity levels, were used in calculating these exposures. The types of activities and projected person-rem estimates are summarized in table 5.2-4. Column one values are derived using the anticipated construction approach. The estimates in column two are O TMFWOO57 5.2-13 I
. predicated on the approach that remote efforts prove impractical f' -and diving operations are required forfin-pool work. The: total Lb}' combined occupational ~ exposures for the Units 2'and 3 reracking
~ .is estimated to be approximately 47 person-rem. This value'is an-estimate of the-expected. accumulative dose and does not. represent the maximum or upper bound conditions.5.2.4.3- Exnosure Controls Durina Divina Operation The'use of remote methods will be evaluated for each aspect of
.the work. Divers will be used only when remote and other methods have been determined by both Health Physics and Construction to be impractical as determined by pre-established criteria.Construction will develop each diving plan and Health Physics () shall concur with it prior to each diving evolution. Physics and Construction maintain approval' authority'for each Health diving evolution. A parallel effort is being applied to planning for this possibility. Major activities in this preparatory effort include: the incorporation of diving operations " lessons learned" from those utilities which reracked.with divers, including revision of the " Radiological Controls for Nuclear Diving" procedure, the formal SFP characterization effort, and appropriate diver-specific actions in areas such as ALARA prejob planning, training, contamination control, surveys, and management oversight. O TMFWOO57 5.2-14
The: Health Physics work control plan will implement SONGS Health ( ), Physics Management's hhilosophy that HealthiPhysics exercises I
. primary: authority over all radiological aspects of diving:L: operations that potentially affect the health and safety of.the. diver (s). This will include, at a minimum, diving decision-making criteria and a checklist, final dive approval by Health. Physics, and stop work criteria. Specific work controls will be established for diving operations in accordance.with the station's procedure, " Radiological Controls and Guidance for Diving Operations" and in accordance with USNRC IE'Information Notices: No. 84-61 - Overexposure of Diver in Pressurized Water Reactor (PWR) Refueling Cavity, dated LO (s/ ' August 8, 1984. No. 82-31 - Overexposure of Diver During Work in Fuel Storage Pool, dated July 28, 1982. Examples of appropriate diving operations work controls include, but are not limited to the following: A. Installing a highly visible, physical underwater barrier when practicable to delineate the work zone. The diver may be confined to a suspended cage or underwater work platform. Other controls which may be used in conjunction with a cage or platform, are the use of colored underwater O TMFWOO57 5.2-15
{i ' M[ r
,,f vsv ' lights and/or weighted plastic; streamers, netting or. ropes: - >c i- ~ KL // suspended from a rigid frame to define.the bounds of' diver.movement.- B.- Maintaining visual surveillance of. divers during diving-operations.
- l C.: Monitoring ~ diver exposure continually during high !
potential exposure and/or high risk activities. I D. Establishing radiological hold-points, as appropriate,-in complicated or lengthy diving work evolutions. i E. Computing. stay times for each diver prior to each' dive. F. Maintaining close communication and cooperation with the diver by radio. G. Monitoring the diver's movements from the surface and-quickly passing stop work instructions by radio should access-controls be' compromised. Certain diving ~ operations controls that specifically address discrete radioactive particles (DRPs) include, but are not
-limited to the following:i j vn TMFWOO57 5.2-16
A. Protection factors for typical diving suit materials will k__ be determined to ensure divers are afforded' adequate protection. B. A diver will normally vacuum'his way to and from' work area, and the work area itself prior to beginning work,_to reduce dose contribution-from crud. The following ALARA diving operation controls are indicative of those that will be applied: A. "Tailboard" sessions, both pre- and post-evolution, will be mandatory for each diver entry in the pool and will' likewise be addressed in the Health Physics work control s plan. B. The positioning of spent fuel to reduce personnel exposure-will be integrated into all planning. As many of.the newer spent fuel assemblies as practical will be stored in the cask pool area. C. Pre-construction and construction decontamination efforts are planned to minimize the potential for fuel fragment and hot particle exposure and to reduce general area dose , I rates. O TMFWOO57 5.2-17
l 1e L' D. Establishment of gross microcuries/ milliliter in SFP
. water, dose rate at.SFP water's surface (poolside),entries into SFP water and person-rem, especially diver as KPIs on'which identified trends will be reported to management on a routine basis. E. . Appropriate training for complicated and/or high exposure and high risk tasks will be provided prior to task performance. Personnel involved in diving operations shall be trained and qualified in accordance with SONGS procedures prior to performing work in SFP areas. Appropriate additional dive-specific training will be provided. Diver's training shall include instruction in gs monitoring their equipment (diver's suit, etc.) and the work area
\~ / with an underwater probe.The types and frequencies of radiological surveys will be identified in the Health Physics work centrol plan. All' radiological surveys will be conducted in accordance with Station procedures. Diving specific surveys are to be performed in accordance with-the established SONGS diving procedure. A 3-D survey of the ; in-pool work area will be done periodically to support diving
. operations. The extent and frequency of supplemental surveys during the dive duration will be established in the Health Physics work control plan.O TMFWOOS7 5.2-18
+Each diver will.be equipped with a calibrated alarming dosimeter and other personnel monitoring dosimeters as required by the REP. As e minimum,.the alarming dosimeter will have a remote readout monitored by Health Physics personnel. The non-remote readout' dosimeters will be checked periodically. The types of construction activities'and projected person-rem estimates are summarized in table 5.2-4. The information in column two in table 5.2-4 is predicated on the assumption, recognized to be very conservative, that the planned remote work evolutions prove to be impractical.- That is, all underwater work hasz to be done by divers. Because of the proposed distance to be maintained between the diver and the nearest spent fuel bundles, there will be negligible contribution to diver dose from the O spent fuel. Therefore, the submersion dose due to current radioisotope concentrations in the SFP water is used for the purpose of estimating diver person-rem. This is a conservative estimate since no decontamination measures are applied. 5.2.5 DISPOSAL OF RACK AND OTHER MATERIALS The 15 spent fuel storage rack modules that will be removed from each SFP weigh between 13,705 and 27,058 pounds each. The total l weight of these racks is approximately 170 tons and the racks occupy a total uncompacted volume of approximately 15,000 ft 3. l They will be adequately decontaminated, packaged, and shipped to a licensed radioactive waste processing facility. I TMFWOO57 5.2-19 i
O Shipping containers will be used for shipment offsite that meet-V ' Department of Transportation (DOT) the requirements of the U.S. regulations pertaining to radioactive waste shipments, including limitations with respect to the waste surface dose and radionuclides activity distribution. Shipping containers will be certified to meet all the requirements for a strong tight package. The maximum weight of a loaded shipping container will be in accordance with the American Association of State Highway and Transportation Officials (AASHTO). Trucks and drivers used for rack and waste transportation will have all permits and qualifications required by the Federal DOT and DOT for each State through which the truck will pass. At the waste processing facility, the racks and other materials will be decontaminated to the maximum extent reasonable. Remaining portions of the racks, and other materials, and contaminated waste generated from decontamination will be buried at a licensed radioactive waste burial site. In preparing non-decontaminable waste for shipment and subsequent burial, volume reduction methodologies will be considered. 5.2.6 RADIOLOGICAL IMPACT ON THE ENVIRONMENT The radiological impact of the SFP reracking project on the environment is negligible. The major source of potential l radioactive releases from the FHB is the reactor coolant system O TMFWOO57 5.2-20 \ l ___--_____-_ _ _ - _ _
which enters the SFP during refueling. The reracking project i does not change the way refueling operatic 3s are conducted. 1 The increase in normal radioactive effluent has been evaluated and has been found to be insignificant. The increase in offsite dose rates under normal conditions has also been analyzed and found to be negligible because the evaporation of tritium from the SFP water remains unchanged. Offsite doses due to SFP boiling were not analyzed in the original design so no comparison can be made. However, offsite' doses due to this abnormal condition are well within 10C7R20 guidelines. The offsite doses caused by a gate drop accident increase due to the increased number of fuel assemblies which may be potentially damaged. The consequences of this event are maintained well within 10CFR100 limits (see subsection 5.3.6), by taking credit for the FHB () emergency HVAC filtration operation. O TMFWOO57 5.2-21 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ __ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _d
i Table 5.2-1 L NORMAL AND MAXIMUM ISOTOPIC INVENTORIES OF THE-L SPENT FUEL POOL PURIFICATION SYSTEM ION EXCHANGER (VCi) U . Existing- Reracking Existing Roracking
]!aclide Normal (*3 Normal (*) garimum(*) Maximum (*)H-3 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Cr 4.4E+5 3.1E+5 4.4E+5 .3.1E+5 Mn-54 9.8E+4 7.0E+4 9.8E+4 7.0E+4 Co-58 4.5E+6 3.3E+6 4.5E+6 3.3E+6 Fe-59 7.6E+5 1.9E+5 2.6E+5 1.9E+5 Co 6.5E+5 4.7E+5 6.5E+5 4.7E+5 Br-84 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Kr-85m 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Kr-85 0.0E+0 0.0E+0 0.0E+0 0.0E+0' Kr-87 0.0E+0 0.0E+0 0.0E+0- 0.0E+0 Kr-88 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Rb-88 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Sr-89 9.1E+4 6.6E+4 1.6E+6 1.1E+6 Sr-90 3.3E+3 2.4E+3 1.0E+5 7.5E+4 Y-90 9.9E+1 7.0E+1 5.2E+3 3.7E+3 Sr-91 9.4E-7 6.7E-7 6.9E-6 5.1E-6 Y-91m 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Y-91 5.3E+6 3.8E+6 3.2E+6 5.5E+6 Er-95 1.7E+4 1.2E+4 1.7E+6 1.3E+6 - Mo-99 3.1E+6 2.2E+6 9.6E+6 7.0E+6 Ru-103 1.1E+4 7.9E+3 1.2E+6 8.9E+5 Ru-106 3.2E+3 2.3E+3 9.0E+4 6.7E+4 Te-129 0.0E+0 0.0E+0 0.0E+0 0.0E+0 I-131 2.3E+7 1.7E+7 2.3E+8 1.7E+8 Xe-131m- 0.0E+0 0.0E+0 0.0E+0 0.0E+0 I-132 1.5E-40 1.1E-40 1.8E-39 1.3E-39 Te-132 3.3E+5 2.3E+5 4.7E+6 3.5E+6 I-133 6.8E+2 4.9E+2 7.3E+3 5.4E+3 Xe-133 0.0E+0 0.0E+0 0.0E+0 0.0E+0 I-134 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Cs-134 7.3E+6 5.3E+6 2.7E+7 2.0E+7 1 I-135 2.8E-9 2.1E-9 3.0E-8 2.3E-8 : Xe-135 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Xe-135m 0.0E+0 0.0E+0 0.0E+0 0.0E+0 Cs-136 1.7E+6 1.2E+6 2.1E+6 1.5E+6 Cs-137 5.4E+6 3.*E+6 1.1E+8 8.3E+7 Xe-138 0.0E+0 0.vE+0 0.0E+0 0.0E+0 Ba-140 3.1E+4 2.2E+4 1.0E+6 7.2E+5 La-140 8.4E+1 5.8E+1 4.4E+3 3.3E+3 Ce-144 1.0E+4 7.2E+3 1.3E+6 9.8E+5 Pr-143 7.3E+3 5.2E+3 9.3E+5 7.0E+5
- a. Calculated values based on definition of normal )
and maximum found in SONGS 2&3 UFSAR Section 11.1. O TMFWOO57 5.2-22 _-___________________._._.____m._. _ _ _ _ _ _ _ _ _ _ .._._ _ _
, a.:s Table 5.2-2
. f).. ~FUEL MANDLIFG BUILDING REFUELING OPZRATION AIRBORNE RADIOACTIVITY CONCENTRATIONS'(Shtet 1 of 2), 4
- L ' Actual'(*)
I )Iuclide 1 ci/uniti H-3 ND (b) Na-24 ND '. E-40 < 2.422E-10 Ar-41 ND. Cr-51 < 5.662E-11 Mn-54 ND Mn-56 ND Co-57 < 2.301E-12 l- C0-58 < 2.020E-11
.Fe-59 < 2.427E-11 C0-60 ND l Ni-65 ND En-65 < 2.735E-11 Br-82 < 1.266E-11 Br-84 ND- ;Kb-85m < 5.911E-12 Kr-85m .ND Kr-85 < 3.307E-9 . O Kr-87 Kr-88 Rb-88 < 1.997E-10 < 7.511E-11' ND Sr-89 ND Sr-90 ND Y-90 ND Y-91m- <-6.011E-10 Sr-91 ND Y-91 < 5.015E-9 -Sr-92 ND -Nb-95 < 1.345E-11 Zr-95 ND Zr-97 < 2.027E-10 Tc-S9m ND Mo-99 < 7.501E-11 Ru-103 < 6.866E-12 Ru-106 ND Cd-109 -
< 8.524E-11 Ag-110m < 1.213E-11 Te-129 ND I-130 < 2.101E-11 Xe-131m < 3.387E-10
- a. These values are indicative of refueling conditions.
- b. ND - Not Detected O TMFWOO57 5.2-23
, Table.5.2-2 FUEL HANDLING BUILDING' REFUELING OPERATION AIRBORNE RADIOACTIVITY CONCENTRATIONS (Sheet 2 of.2)Actual (a) Nuclide ( ci/ unit) I-131 < 5.5835-)2 I-132 ' ND (b . i Te-132 < 5.837E-12 Ba-133 ND I-133 < 1^.252E-11 Xe-133; < 7.070E . Cs-134 ND I-134 ND I-135- ND Xe-135 < 7.033E-12 Xe-135m ND Cs-136 ND Cs-137 < 1.667E-11 Xe-138 ND Ba-140- ND La-140 ND
.Ce-141 < 4.302E-12 Pr-143- ND Co-144 < 3.695E-11 W-187 < 3.334E-11 Pb-212 < 1.822E-11 Np-239 < 4.446E-11
- a. These values are indicative of refueling conditions.
- b. ND - Not Detected O TMFWOO57 5.2-24
l 4 j-Table 5.2-3 DOSE RATES IN THE VICINITY OF THE SPENT FUEL POOL (arem/h) Normal (a) SFP Maximum (a) SFP. Actual Plant Location Conditions Conditions Data Along the SFP 4.04 to 7.94 38.1 to 75.0- 2 to 5 perimeter At surface of- '17.6 166 2'to 4-the SFP' The results-shown in this table are-based uponisource
. strength spectrums determined 4 days after reactor. shutdown, when peak Xenon-133 activities occur. Since the primary . contributor to these doserates is Xenon-133, dose rates at any other time will be lover.O
- a. As' defined in UFSAR section 11.1.
O TMFWOO57 5.2-25 I
i Table 5.2-4 ESTIMATED RADIATION DOSE 3 FOR CONSTRUCTION ACTIVITIES (Sheet 1 of 2) Column 1 Column 2 Anticipated Diver Project Only Approach Approach Item Description (rem) (remi-Install / remove cask pool rack 0.24 0.24 Shuffle fuel 0.57 0.57 Install / remove cask pool cover. 0.16 0.16 Install / test / remove work platforms and crane 0.53 0.53 i Remove old racks e Preparation of work site 0.38 0.42 e Tie-down bolt removal (top of rack) 0.59 0.95 e Installation of lifting fixture (top of rack) 0.35 0.34 e Monitor the lifting and rigging activities 0.88 0.93 e Hydrolaze/ vacuum rack 1.40 1.40 e vacuum path / work area 0.00 1.67 Remove support structure and piping e Unbolt support beams 0.82 0.82
.[/ \_ -s} e Recove and cut support beams e Cut and remove purification piping 0.76 0.88 0.76 0.88 e cut and remove purification supports 0.94 0.94 e Cut and remove sparger piping 4.19 4.19 e Cut and remove sparger supports 4.76 4.76 e Cut'and remove miscellaneous piping 0.68 0.68 e Cut and remove miscellaneous items 0.09 0.09 e vacuum path / work area 1.88 1.88 Install new racks e Install floor plates 0.28 0.28 e Release rigging 0.11 0.11 e Elevation survey 0.06 0.06 e Install, level and torque new rack 1.40 1.58 e Remove lift fixture from new rack. 0.12 0.18 Drag test new racks 0.07 0.07 Prepare racks and other radwaste for offsite shipment 4.00 4.00 Miscellaneous (including mobilization and demobilization) 1.22 1.34 TOTAL FOR UNIT 2 27.36 29.83 TOTALS FOR BOTH UNITS
- 46.51 50.71
- Assumes Unit 3 requires 30% less person-rem due to learning curve efficiency increase.
TMFWOO57 5.2-26 L L
Table 5.2-4 ,. ESTIMATED RADIATION DOSES FOR CONSTRUCTION ACTIVITIES I (Sheet 2 of 2) Assumptions:
- 1. Column 1 values are based on the anticipated construction approach where the in-pool work is accomplished by a combination of diving operations and remote technology.
- 2. Column 2 values are based on the contingency approach where all remote efforts prove to be impractical'and diving operations are required for all in-pool work.
- 3. The current measured radionuclides concentrations in the SFP water remain constant throughout the work.
4.. All values are estimates of expected cumulative dose and do not represent maximum or upper bound conditions. O TMFWOO57 5.2-27
5.3 ACCIDENT EVALUATION O 5.3.1 SPENT FUEL HANDLING ACCIDENTS I 5.3.1.1 Fuel Assembly Dron Accident The design basis accident for a dropped fuel assembly is discussed in the SONGS 2&3 UFSAR, paragraph 15.7.3.4. The installation of the new spent fuel racks will not have any impact on the previous analysis as presented in the UFSAR. The structural evaluation of the fuel assemblies will not be impacted since the fuel is not being changed. The fuel assemblies are stored within the spent fuel rack at the bottom of the SFP. The limiting fuel handling accident, at the SONGS 2&3 h facilities, is a maximum vertical drop of a vertically-oriented SONGS 2&3 assembly to an end-on initial impact on the SFP floor. This limiting fuel handling accident encompasses an accident for a SONGS 1 fuel assembly at the SONGS 2&3 SFPs(2), The maximum possible drop distance for a fuel assembly, in the SFP, is 254 inches. The worst case horizontal impact results from the maximum vertical drop followed by the fuel assembly rotating to the horizontal position and striking a protruding structure. Interlocks on the spent fuel handling machine ensure that the fuel assemblies are moved at a minimum safe water level (see UFSAR subsection 9.1.4.) e TMFWOO57 5.3-1
The worst case fuel assembly drop discussed above causes the
'\s,) failure of fuel rods in four rows parallel to one assembly face l or a total of 60 fuel rods. This is the largest number of fuel rods which could fail as a result of the worst case fuel drop.The FHB ventilation system was evaluated assuming the failure of all 236 fuel rods in one assembly to demonstrate that the system
-is capable of handling a release of that severity.The resultant radioactive release would escape from the SFP and be exhausted from the FHB over a period of approximately 2 hours. This evaluation is described in section 15.7 of the UFSAR(3) and shows that post-accident doses are within the limits of 10CFR100. Based on the above discussion, the radiological consequences of'a postulated spent fuel handling accident will remain the same as discussed in the SONGS 2&3 UFSAR. The proposed SFP modifications will not increase the radiological consequences of fuel handling accidents previously evaluated in paragraph 15.7.3.4.6 of the SONGS 2&3 UFSAR. Specifically, doses to the control room personnel and offsite personnel will not be increased. Fuel handling accidents for Unit 1 fuel are enveloped by fuel handling accidents involving Units 2 and 3 fuel (2), O TMFWOO57 5.3-2
L 5.3.1.2- Cask Dron Analysis yy V 5.3.1.2.1 Cask Handling As discussed in SONGS 2&3 UFSAR, subsection 9.1.4, the cask-handling crane is prohibited from traveling over the SFP or any q unprotected safety-related equipment. Thus, an accident resulting from-dropping a cask or other major load into the SFP is not credible. Furthermore,.the potential drop of a spent fuel cask is limited to less than an equivalent 30-foot drop onto a flat, essentially unyielding, horizontal t. surface. This evaluation is found in the SONGS 2&3 UFSAR, paragraph 15.7.3.5 and the assumptions have not changed. Therefore, the conclusion remains valid. 5.3.1.2.2 Radiological Consequences As stated in paragraph 5.3.1.2.1 there are no radiological consequences-for the cask handling accident. The radiological consequences for a dropped-fuel assembly are found in the SONGS 2&3 UFSAR, paragraph 15.7.3.4.6, and are still valid. 5.3.1.2.3 overhead Cranes The spent fuel cask crane is not capable of traveling over or into the vicinity of the SFP, except as described in the SONGS 2&3 UFSAR, subsection 9.1.4. A complete cask crane component
- O I
TMFWOO57 5.3-3 I
_-_ -s s b i [" description, cask handling' description, and cask crane' design f,
'i evaluation ~are provided in SONGS 2&3 UFSAR, _ paragraphs 9.1.4.1,1 9.1.4.2, ~and 9.1.4.3 and will not be affected as a result of the raracking.The temporary crane is addressed in subsection--3.4.2. 5.3.'1.3b: Abnormal Location of a-Fuel Assembly Accidents can be postulated which would increase reactivity. (i.e., misloading an assembly with a burnup and enrichment
- combination outside of the acceptable area, or dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975' is applied. This states that one is not required to assume.two unlikely, independent, concurrent events to ensure protection.
against a criticality accident. Thus, for accident conditions, the presence _of soluble boron in the storage pool water can be assumed as a realistic. initial condition since not assuming its presence would be a second unlikely event. The abnormal location of a fresh unirradiated fuel assembly of
-4.1 w/o enrichment could, in the absence of soluble neutron absorbing material, result in exceeding ~the design reactivity limitation (Kegg'of'O.95). This could occur if the assembly were to be either. positioned outside and adjacent to a storage rack module or inadvertently loaded into a Region II storage cell, with the latter condition producing the larger positive O 5.3-4 TMFWOO57reactivity increment. ' Soluble neutron absorbing. material, () however, is present in the SFP water (for which credit is permitted under;these conditions) and would' maintain the reactivity substantially less.than.the design limitation. The. presence of.approximately 2000 ppm (actual minimum of-2350 ppm) boron in the pool water will decrease reactivity by about 30% Ak. . Therefore, for postulated accidents,.should there be a reactivity' increase, Kegg would be less than or equal tol0.95 due to the effect of the dissolved boron. 5.3.1.4 Fuel Handlina Accidents Durina' Construction
.The probability of a fuel handling accident occurring during construction is not significantly higher than during normal' fuel handling operations, because even though a significant number of fuel handling operations'will be performed during construction the same equipment and procedures will be used. Because the fuel enrichment and burnups of the fuel assemblies stored in the SFP at the time of construction will be less than or equal to those assumed in the UFSAR, and because the assemblies stored in the . pool will be decayed for at least 70 days prior to the onset of moving cycle 4 spent fuel assemblies into the cask pool, the consequences of a fuel handling accident during construction are enveloped by the fuel handling accident discussed in the UFSAR which assumes failure of 60 fuel pins. The offsite doses which result from this scenario are well within the limits imposed by 10CFR100. Therefore, the consequences of a fuel handling O TMFWOO57 5.3-5 Iu.-______.___._.______-_____._.__m. _ . . _ _ _ . . . _ _
accident will not:be increased by the proposed modification. A LD drop onto spent fuel temporarily stored in the cask pool is
.V precluded by;the cask pool' cover.{ 5.3.2 FUEL DECAY Technical Specification 3.9.3(4,5) requires a decay time of 72 l hours for irradiated fuel prior to any movement of fuel into the pool. This allows sufficient time for short lived isotopes to decay. This decay time is consistent with the UFSAR a'ccident analysis. As a result, the radiological consequence of a fuel assembly drop is less than the limits of 10CFR100. 5.3.3 LOADS OVER SPENT FUEL Administrative procedures and Technical Specification 3.9.7(4,5) limit the maxim'um weight of loads that may be transported over spent fuel. 5.3.4 IDSS OF SPENT FUEL POOL COOLING FLOW Loss of SFP cooling flow has been evaluated in paragraph 9.1.3.3 of the UFSAR. The probability of this type of accident occurring is not affected by the installation of high-density spent fuel
. racks because the existing SFP cooling system pumps and heat exchangers can be used without modification. It is possible that the pumps may be modified to improve operating characteristics, TMFWOO57 5.3-6however, the consequences of the loss of cooling would not be significantly increased due to the increased mt- er of spent fuel I elements stored. An analysis of a pool boiling. accident has been performed using 1 the highest expected fuel enrichments and burnups. This analysis shows that without taking the FHB charcoal and HEPA filters into account, the offsite dose will be less than 1 x 10-3 rem which is well within the limits defined in 10CFR100. Therefore, the radiological consequences of this accident are not significantly increased by the proposed modification. 5.3.5 RADIOLOGICAL EVALUATION OF TEST EQUIPMENT DROP ONTO THE
~RACFS A drop of the test equipment skid from a height of 47 feet above the pool floor onto racks containing Units 2 and 3 fuel assemblies would produce unacceptable radiological consequences. Therefore, in order to assure that excessive radiological consequences do not occur due to a test equipment skid drop, administrative controls will be in place. These controls will consist of the following: A. The height above the pool floor that the skid may be carried over rack cells which contain Unit 1 fuel assemblies shall be limited to 47 feet (elevation 64 feet-6 inches). e TMFWOO57 5.3-7
i
~
- !' j.;;
When'it is lowered, it shall'be lowered over empty. racks
~B.
. () or rack' cells containing Unit 1 fuel assemblies only.C. The maximum height that.the skid will-travel horizontally. over the racks shall be 72 inches. A drop fromLthis. height will not-damage Units 2 and 3; fuel assemblies. Withithese controls.in place, it will ensure that the fuel assemblies are not damaged, since the depth of penetration.will not impact the racks'at the level where the Unit i fuel assemblies are located. By this-means the radiological consequencessfor test' equipment drop will be bounded-by the-results of'the_ radiological consequences'for a gate drop. 5.3.6 RADIOLOGICAL EVALUATION OF GATE DROP ONTO THE RACKS O A calculation ~was performed to evaluate the radiological consequences of a gate drop onto racks containing fuel assemblies. The postulated drop would damage six fuel' assemblies in the racks. The results of the calculation show that, with credit taken for the FHB filters,'the thyroid doses would be less than 67 rem at the exclusion area boundary which is well within the 10CFR100 limits. All other doses would remain significantly lower than that level. L O TMFWOO57 5.3-8 I-
5.3.7 SHIELDING EVALUATION
'(A calculation which is based on a maximum possible loading in the SFP with only 3 days decay of fuel gives the following dose rates (from fuel only) in occupied areas adjacent to fuel pool walls in Unit 2. West Wall: <0.2 mrem /h East Wall: <2.0 mrem /h With 30 days decay of fuel the above dose rates adjacent to the east wall drop to the following: East Wall: <0.25 mrem /h To minimize these dose rates, fuel decayed less than 30 days will () not be stored adjacent to the east wall or appropriate Health Physics controls will be implemented, unless more realistic analyses or measurements prove that such administrative controls are not required. 5.3.8 SEISMIC EVENTS The probability of occurrence of a seismic event is unaffected by the installation of the high-dedsity storage racks. The consequences of a DBE havo been analyzed and the results confirms i that the racks will maintain the required structural integrity and continue to provide safe storage for spent fuel. The spent i fuel racks are designed to Seismic Category I requirements as l TMFWOO57 5.3-9 l
defined by USNRC Regulatory Guide 1.29, Revision 3. The individual storage cells, individual racks, and the layout of the racks in the SFP comply in all respects with published NRC requirements including, but not necessarily limited to, seismic design criteria and guidance provided in SRP Section 9.1.2 " Spent Fuel Storage", SRP Section 3.8.4 "Other Category I Structures, Appendix D, Technical Position on Spent Fuel Racks", SRP Section 3.7.1 " Seismic Analysis", and NRC Position Paper "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications". O O TMFWOO57 5.3-10
- - _ _ - - _ _ _ _ _ - _ - - _ _ _ _ _ _ = _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ ~5.4 REFERENCES
i
- 1. NUREG 0575, " Final Environmental Impact Statement on Handling..
and Storage of Spent Light Water Power Reactor Fuel", Vol. 1-3.USNRC August 1979.
- 2. . March 11,.1988 letter from M.'O. Medford (SCE) to Document Control. Desk' (NRC) ; Subject. . Docket Nos. 50-361 and 50-362, San Onofre Units 2&3.(TAC Nos. 66970 and 66971).
- 3. -San Onofre Nuclear Generating Station Units 2 and 3, Updated Final' Safety Analysis Report, Docket No. 50-361 and 50-362.
.4. San Onofre Nuclear Generating' Station Unit 2 Facility Operating License NPF-10, Docket.No. 50-361.O
- 5. San Onofre Nuclear Generating Station Unit 3 Facility Operating License NPF-15, Docket No. 50-362.
l TMFWOO57 5.4-1 _ _ _ _ _ _ - - ____- . 1
- 6. RESPONSE TO REOUESTS FROM NRC RADIATION PROTECTION BRANCH DURING THE JUNE 3. 1988 MEETING IN ROCKVILLE. MARYIAND
\
'l RPB NO. la Sources in the Spent Fuel Pool Water Provide a description of fission and corrosion product sources in the spent fuel pool (SFP) water from: (a) introduction of primary coolant into SFP water, (b) movement of fuel from the core into the pool, and (c) defective fuel stored in the pool.Include a listing of the radionuclides and their concentrations (expressed in mci /ml) expected during normal operations and refueling. The radionuclides of interest should include Co-60, h Co-58, Cs-134, and Cs-137. RPB NO. la Response Typical radionuclides identified by gamma spectroscopic analysis of liquid SFP samples are presented in table 6-1. Radionuclides present in the Unit 2 fuel pool should be considered to be representative of non-outage concentrations and Unit 3 radionuclides concentrations are representative of outage concentrations. O TMFWOO57 6-1
L, i RPB NO. Ib
-O Airborne Radioactive Sources ~ =Provide a: description of radioactive materials that may.become' airborne as a result of failed fuel and evaporation ~(e.g. Kr-85, l and H-3,'respectively). The radionuclides description should include calculated or measured concentrations expected during-normal operations and during refuelings. RPB NO..Ib Response Typical airborne radionuclides concentrations in the FHB during
'both outage and non-outage. conditions based on measurements are presented in table 6-2. Note that the outage and non-outage values are the same.RPB NO. Ic Miscellaneous Sources of Exposure Address the effects of more frequent replacement of demineralized filters on cumulative dose equivalent if this is a factor that results from the modification. k TMFWOO57 6-2
RPB NO. Ic Response 10 v There are no specific radiation exposures to personnel during the changeout of the SFP cooling and cleanup system (SFPCC) ion exchangers and. filters. The design of both SONGS 2&3 is such that these filters are backflushed to the crud tank. Likewise, the resin is discharged to spent resin tanks for batch processing. These activities are work evolutions which.do not , accumulate additional dose at SONGS. RPB NO. 2 A. Provide a. description of the. dose rate at the surface of the pool water from the assemblies, control rods, burnable poison rods or any miscellaneous materials'that may be () stored in the pool. Additionally, provide the dose rate from the individual fuel assemblies as they are being placed into the fuel racks. Information_ relevant to the depth of water shielding the fuel assemblies as they are being transferred into the racks should be specified. If the depth of water shielding over a fuel assembly while it is being transferred to a spent fuel rack is less than 10 feet, or the dose rate 3 feet above the SFP water is greater than 5 mR/hr above ambient radiation levels, then submit a Technical Specification specifying the minimum-depth of water shielding over the fuel assembly as it is O TMFWOO57 6-3 l
r being transferred to the fuel rack and the measures that 1 will be taken_to assure that this minimum depth will not be degraded. B. Address.the dose rate changes at the sides of the pool
. concrete shield walls, where occupied areas are adjacent to these walls, as a result of the modification.-Increasing the capacity of the pool may cause spent fuel assemblies to be relocated close to the concrete walls of the pool, resulting in an increase of radiation levels in occupied areas. Please evaluate this potential problem. RPB NO. 2 Response A. The contribution of the spent fuel essemblies to the dose rate at the surface of the pool is negligible. For the maximum pool loading and only 3 days decay of all spent fuel, the dose rate from the fuel is on the order of 1 x 10-6 mrem /h (calculated) at the surface. Attenuation occurs in 23 feet of pool water above the fuel assemblies. An increase in dose rate in occupied areas has not been observed from individual fuel assemblies as they are being placed in the fuel racks. An amendment to the plant Technical Specifications is not needed because this is addressed and approved by the Safety Evaluation Report, dated June 22, 1988 which allows the minimum depth of shielding water (8 feet - 5 inches) and the maximum dose TMFWOO57 6-4 I 1 l
rate due to the fuel assembly (5 mrem /h). The SFP water
/) level will not be lowered below the minimum level required Vfor Unit 1 fuel handling activities nor below the Technical Specification limit for SONGS 2&3 spent fuel storage. The existing SFP low level alarm has been' reset' in order to allow for this lower SFP water level. The maximum elevation of the fuel assembly is determined by limit switches on the SFHM'and by administrative controls. The water level in the SFP will be reduced from its normal level (27 feet - 6 inches above the spent fuel) to-a level of 24 feet - 6 inches above the spent fuel to allow the handling of Unit 1 fuel casks without placing the cask crane block in the water. B. A calculation based on the maximum possible loading in the O
.() SFP with only 3 days decay of fuel shows the following dose rates (from fuel only) in occupied areas adjacent to fuel pool walls (Unit 2):West Wall: <0.2 mrem /h East Wall: <2.0 mrem /h
)With 90 days decay of fuel, the dose rates are reduced to: West Wall: <3 x 10-3 mrem /h East Wall: <4 x 10-2 mrem /h j 1 1. O TMFWOO57 6-5 l I i i l
18 RPB NO. 3 LO Dose Rates from Spent Fuel Pool Water l-. L Provide information on the dose rates at the surface of SFP water resulting from radioactivity in the water. . Include:- (1). dose rate levels in occupied areas and along the edges 1and center for the pool and on the fuel handling crane;. (2) effects of crud buildup; and (3) based on refueling water activity, the dose rates before,.during, and after refueling. RPB NO.' 3 Response Typical gamma dose rates at the surface of the SFP water, near. the center of the pool, vary from 2 mrem /h during nonrefueling activities up'to 5 mrem /h during refueling.. The value of 5 (f mrem /h is due to radionuclides in the water, the-spent fuel, and radioactive contamination on the fuel handling bridge. The estimated dose rate from radionuclides in the water and from the spent fuel is 2 to 4 mrem /h during refueling. Typical gamma dose rates at the surface of the SFP water, along the edges of the pool, are 2 to 3 mrem /h with small areas reading up to 5 mrem /h due to localized deposits of radioactive material. Close to the concrete sides of the SFP at each elevation of the building, outside the FHB, dose rates are as follows: On the 45-foot elevation, dose rates are typically <0.2 mrem /h gamma O TMFWOO57 6-6
radiation in accessible areas. On the 30-foot elevation, dose rates are typically 2 to 3 mrem /h in accessible areas. Dose ( rates in some rooms may be up to 5 arem/h due to the storage of radioactive materials in the room. Dose rates in the fuel pool heat exchanger rooms are typically 10 mram/h when the system is in service. Dose rates in accessible areas on the 15-foot elevation vary from <0.2 to 2.0 arem/h gamma radiation. RPB NO. 4 Done Rates from Airborne Isotopes Based on the source terms, provide the dose rates from submersion and dose commitments from exposure to the concentration of Kr-85 and H-3. O RPB NO. 4 Response Using the source terms from the UFSAR design bases, the calculated dose rates from airborne radionuclides are 0.43 mrem /h for normal operation and 1.25 mrem /h during refueling operations. These dose rates are primarily from tritium with negligible contribution from noble gases. Tritium concentration 1 (calculated) is 8.3 x 10-7 mci /cm3 (normal) and 2.4 x 10-6 mci /cm3 (refueling). O TMFWOO57 6-7
l
'RPB NO. 5' YDose Assessment from Modification Procedures A. Discuss the manner in which occupational exposure will'be kept as low as reasonably achievable (AIARA) during modification. Include the need for and the manner in which cleaning of the crud on-the SFP walls will be performed to reduce exposure rates in the SFP area. B. Discuss vacuum cleaning of SFP floors if divers are used and the distribution of existirig spent fuel stored in racks to allow maximum water shielding to reduce dose rates to divers. C, Describe plans for cleanup.of the SFP water to minimize radioactive contamination'and to ensure fuel pool clarity l and underwater. lighting acceptance criteria to help ensure good visibility. D. Discuss underwater radiation surveys that will be made befora any diving operation, These surveys should be performed before or after any fuel movements or the movement of any irradiated components stored in the pool. 1 l i
\' l O _s, e_. l l
/" , [MM -
,, , , , s g,, ,9,~ , 4 ,, m , StateLyour~ intent to-equip each diver-with a calibrated' 6 1 E.. ' [l alarming dosimeter and personnel monitoring dosimeters, i which should be. checked periodically to ensure that qq U m prescribed dose limits are not being exceeded.- F. Discuss any preplanning of work by divers as. required. l G. Discuss.your provision for surveillance'and monitoring of af ~the SFP; work area by Health Physics personnel during the: modification.
' fu ..RPB No. 5 Response
~A. The radiation protection aspects of tho'SFP modification are the responsibility of.the Station Health' Physics Manager, who is assisted by his staff, with the support of. the Corporate Health Physicist and his staff. Periodic radiation and contamination surveys will be conducted in work areas as necessary.- Portable area radiation monitors will also be ~u sed. Where there is a potential for significant airborne radionuclides concentrations, continuous air samplers will be used in addition to periodic grab sampling. Personnel working-in. radioactively contaminated areas will wear protective clothing and when required by work area conditions respiratory protective equipment, as required by the applicable Radiation Exposure Permit (REP). Personnel monitoring equipment is assigned to and worn by all O TMFWOO57 6-9
i L personnel in the work area. At a. minimum, this equipment D will consist'of a TLD and self-raading pocket dosimeter. [O Additional personnel monitoring equipment,.such as extremity badges, will be utilized as required. Radiation exposure from radioactive particles will.be controlled through compliance with the station program for l 1 conducting work in areas where discrete radioactive particles are known or suspected to exist. The refueling water level will be maintained in the SFP to keep exposure ALARA. Should crud buildup ever become significant on the SFP walls around the pool edge, it could easily be removed. A (,) B. The diver.will normally vacuum his way to and from the work area, and work area itself prior to beginning work and after work is completed. Planning will be employed with spent fuel bundles shuffled accordingly, to ensure the distribution of existing spent fuel bundles will provide adequate water shielding. , i i C. The SFP cleanup system will be used to control radioactive contamination and to ensure fuel pool clarity. l O TMFWOO57 6-10 ( I
Health Physics personnel assigned to the raracking project will'have..the a'uthority to stop work based on. inadequate lighting and/or pool clarity.- D. Radiation surveys will be performed in accordance with the established station procedure that controls nuclear diving operations, entitled " Radiological Controls for Nuclear-Diving". : Additional survey requirements can be identified' in the Health Physics work control plan, approved by the Radiation Protection Manager, written specifically for this work. E. Each diver will be equipped with a calibrated alarming dosimeter and other personnel monitoring dosimeters as required by the REP. As a minimum, the alarming dosimeter will have a remote readout monitored by Health Physics personnel. The nonremote readout dosimeters will be checked periodically. F. Preplanning of work by divers will be controlled by procedure entitled " Radiological Controls for Nuclear Diving". Supplemental preplanning requirements will be identified in the Health Physics work control plan, approved by the Radiation Protection Manager, written specifically for this work. O TMFWOO57 6-11
l G. Health Physics surveillance and monitoring of the SFP work area will be provided commensurate with the work activities being performed. A Health Physics Technician to Worker ratio will be identified in the Health Physics work control plan for reracking. This ratio will ensure adequate coverage. Additionally, personnel will be periodically surveyed for radioactive particles. T Personnel exiting the work area will be required to I proceed to a whole body contamination monitor to check for radioactive contamination. \ Specific surveillance and monitoring provisions will be : established in the REP and the Health Physics work control I plan, approved by the Radiation Protection Manager, written specifically for this work. O RPB NO. 6 Provide an estimate of the total man-rem to be received by personnel occupying the SFP area based on all operations in that area including those resulting 1om RPB Nos. 2, 3, and 5 above. Describe the impact of the spent fuel storage modification on these estimates. O TMFWOO57 6-12
RPB No. 6 Response O The Health Physics work control plan for the entire sequence of operations is currently being developed. When completed, this work plan will define, plan, and schedule radiological protection activities. Each major work activity, including contingencies, has an estimated radiation dose associated with it (see table 5.2-4). O l l l l 9 O TMFWOO57 6-13
q
)l i 7 v Table 6-1 1 RADIONUCLIDES IN.THE SPENT FUEL POOL WATER l}
. Unit 2 Unit 3 Radionuclides (uci/ml) (UCi/ml) ')1 Mn-54 3.5E-6 -1.2E-5 Co-57 9.9E-6 1.7E-5 Co-58 4.3E-4 5.3E-3 Co-60 4.7E-5 7.4E-5 Zn-65 Not Present 1.9E-6 s Nb-95 Not Precent 5.9E-5 Zr-95 Not Present' 7.3E-5 Ru-103 Not Present 3.3E-5 Sb-124 3'.2E-6 1.5E-5 Sb-125 4.1E-5 4.0E-5 Cs-134 1.4E-3 1.9E-4 I-135 8.4E-6 Not Identified Cs-137 1.6E-3 5.3E-4 Nb-95 Not Present 5.9E-5 Zr-95 Not Present 7.3E-5 Ru-103 Not Present 3.3E-5 i l I l () TMFWOO57 6-14 l
r TOblo 6-2 AIRBORNE RADIONUCLIDES IN THE FUEL HANDLING BUILDING Radionuclides . _ uCi/ml Na-24 ND K-40 <2.422E-10 Ar-41 ND Cr-51 <5.662E-11 Mn-54 ND Mn-56 ND Co-57 <2.301E-12 Co-58 <2.020E-11 Co-60 ND Fe-59 <2.427E-11 Ni-65 ND Zn-65 <2.735E-11' Br-82 <1.266E-11 Kb-85m <5.911E-12 Kr-85 <3.307E-09 Kr-87 <1.997E-10 Kr-88 <7.511E-11 Sr-91 ND Sr-92 ND Y-91 <5.015E-09
. Y-91m <6.011E-10 Nb-95 <1.345E-11 Zr-95 ND Zr-97 <2.027E-10 Mo-99 <7.501E-11 14' Tc-99m ND Ru-103 <6.866E-12 Cd-109 <8.524E-11 Ag-110m <1.213E-11 I-130 <2.101E-11 I-131 <5.583E-12 I-132 ND I-133 <1.252E-11 I-134 ND I-135 ND Xe-131m <3.387E-10
-Xe-133 <7.707E-12 ,Xe-135 <7.033E-12 Te-132 <5.837E-12 Ba-133 ND Cs-134 ND Cs-137 <1.667E-11 Ba-140 ND La-140 ND Ce-141 <4.302E-12 ' Ce-144 <3.695E-11 W-187 <3.334E-11 Pb-212 <1.822E-11 Np-239 <4.446E-11 ND = Not detected
.V TMFWOO57 6-15 ,I _ _ _ _ - _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ . _ _ _}}