ML20127A091
| ML20127A091 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/31/1985 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20127A081 | List: |
| References | |
| NUDOCS 8508050401 | |
| Download: ML20127A091 (119) | |
Text
{{#Wiki_filter:7, . PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 SPENT FUEL STORAGE CAPACITY MODIFICATION SAFETY ANALYSIS REPORT DOCKET NOS. 50-277 AND 50-278 MAY 1985 - REV. O JULY 1985 - REV. 1 3 A k 7 6 -
.e TABLE OF CONTENTS Pace
1.0 INTRODUCTION
1-1 1.1 Current Status 1-1 1.2 Summary of Report 1-1 1.3 Conclusions 1-2 1.4 References 1-2 2.0
SUMMARY
OF EXISTING RACK DESIGN 2-1 3.0 NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS 3-1 3.1 Neutron Multiplication Factor 3-1 3.1.1 Normal Storage 3-2 3.1.2 Postulated Accidents 3-4 3.1.3 Calculation Methods 3-5 3.1.4 Criticality Analysis Results 3-7 3.1.5 Acceptance Criteria for Criticality 3-8 3.2 Decay Heat Calculations for the Spent Fuel Pool 3-8 3.3 Local Thermal-Hydraulic Analyses for the Spent Fuel Pool 3-9 3.3.1 Criteria 3-9 3.3.2 Key Assumptions 3-10 3.3.3 Analytical Methods and Calculations 3-11 3.4 Potential Fuel and Rack Handling Accidents 3-13 3.4.1 Rack Module Mishandling 3-14 3.4.2 Flow Blockage Analysis 3-14 3.4.3 Accident Condition 3-15 3.5 Technical Specifications 3-16 3.6 References 3-16 4.0 MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS 4-1 4.1 Description of Structure 4-1 4.1.1 Description of Spent Fuel Pool Structure 41 4.1.2 Description of New Spent Fuel Racks 42 i Pev. 1
I l l TABLE OF CONTENTS (CONTINUED) ! Pace 4.2 Applicable Codes, Standards, and Specifications 4-5 4.3 Seismic and Impact Loads 4-8 4.4 Loads and Load Combinations 4-10 ! 4.4.1 Spent Fuel Rack 4-10 4.4.2 Spent Fuel Pool Structure 4-10 4.5 Design and Analysis Procedures 4-11 I 4.5.1 Design and Analysis Procedures for Spent l Fuel Storage Racks 4-11 l 4.5.2 Design and Analysis Procedures for Spent l Fuel Pool Structure 4-14 l l i 4.6 Structural Acceptance Criteria 4-20 l
- 4.6.1 Structural Acceptance Criteria for Spent l Fuel Storage Racks 4-20 4.6.2 Fuel Handling Crane Uplift Analysis 4-21 4.6.3 Fuel Assembly Drop Accident Analysis 4-21 4.6.4 Fuel Rack Sliding and Overturning Analysis 4-22 4.6.5 Fuel Assembly / Module Impact Evaluation 4-23 4.6.6 Structural Acceptance Criteria for Spent Fuel Pool Structure 4-23 4.6.7 Spent Fuel Pool Serviceability Checks 4-25 4.6.8 Pool Slab / Fuel Rack Interface Loads 4-26 4.6.9 Evaluation of Spent Fuel Pool Slab for
- l Fuel Handling Accident 4-27 4.6.10 Spent Fuel Pool Liner Plate Evaluation 4-27 4.7 Materials, Quality Control, and Special Construction Techniques 4-28 4.7.1 Construction Materials 4-28 4.7.2 Neutron Absorbing Material 4-28 4.7.3 Quality Assurance 4-29 4.7.4 Construction Techniques 4-30 11 Rev. 1 e
A - - - -
r TABLE OF CONTENTS (CONTINUED) Page 4.8 Testing and In-Service Surveillance 4-32 4.8 1 Initial Verification 4-32
- 4. 8.> 2 Tericdic Verification / Surveillance 4-32 7
4.9 References 4-34 5.0 COST /8ENEFIT ASSESSMENT 5-1 5.1 Cost / Benefit Assessment 5-1 5.1.1 Need for Increased Storage Capacity 5-1 5.1.2 Estimated Costs 5-2 5.1.3 Consideration of Alternatives 5-2 5.1.4 Resources Committed 5-3 5.1.5 Thermal Impact on the Environment 5-4 5.2 Radiological Evaluation 5-4 5.2.1 Solid Radwaste 5-4 5.2.2 Gaseous Radwaste 5-5 5.2.3 Personnel Exposure 5-5 5.2.4 Radiation Protection During Re-Rack Activities 5-8 5.2.5 Summary of Re-Rack Operation 5-10 5.2.6 Re-Racking Exposure Estimate 5-14 5.3 Accident Evaluation 5-15 5.3.1 Spent Fuel Handling Accidents 5-15 l 5.3.2 Conclusions 5-15 111 Rev. 1
.f
f I
; LIST OF TABLES Table Title Pace 2-1 Quantity, sizes anu weights of Existing Spent Fuel Rack Modules PBAPS Units 2 and 3 2-3 l
l 3-1 Parameters for BWR Fuel Assemblies to be stored ! in the Peach Bottom Racks 3-18 3-2 Benchmark Criticality Experiments 3-19 l 3-3 Sumary of Cooling System Analysis Results 3-20 4-1 Rack Module Data (Per Unit) 4-36 4-2 Storage Rack Loads and Load Combinations 4-37 4-3 Spent Fuel Pool Governing Design Load Combinations 4-38 4-4 Slab Design Load Sumary 4-39 4-5 Sumary of Design Stresses and Minimum Margins of Safety 4-40 4-6 Drywell Shield Wall Original Design Sumary 4-41 4-7 Maximum Allowable fuel Rack / Pool Floor Interface Loads 4-42 4-8 Pool Floor Loads 4-43 5-1 Existing Spent Fuel Storage Capacity Spikv 5-2 Spent fuel Storage Capacity After Rerack -18 5-3 Identification of Region Materials for QADM00G Model of Spent Fuel Pool 5-19 5-4 Estimated Doses During Reracking 5-20 1 9 iv Rev. 1 e
I LIST OF FIGURES t Fiqure Title Page 2-1 PBAPS Unit 2 Existing Spent Fuel Pool Layout 2-4 2-2 PBAPS Unit 3 Existing Spent Fuel Pool Layout 2-5 3-1 Spent Fuel Pool Natural Circulation Model 3-21 (Elevation View) 3-2 Spent Fuel Pool Natural Circulation Model 3-22 (Plan View) r 3-3 Spent Fuel Rack Inlet Flow Area 3-23 [ (Plan View) i 4-1 Unit 2 Spent Fuel Pool Plan View 4-44 i 4-2 Unit 2 Spent Fuel Pool Cross Section 4-45 , 4-3 Unit 2 Spent Fuel Pool Isometric View 4-46 f 4-4 Spent Fuel Pool Storage Rack Arrangement Unit 2 4-47 4-5 Spent Fuel Pool Storage Rack' Arrangement Unit 3 4-48 4-6 Fuel Storage Rack Assembly Details 4-49 4-7 Fuel Rack Plan View 4-50 4-8 StructuralModel(QuarterRack) 4-51 4-9 3-0 Nonlinear Seismic Model 4-52 i 4-10 Section of 3-D Nonlinear Seismic Model 4-53 l
,4-11 Reactor Building - Plans and Sections of Structure 4-54 l l
4-12 Reactor Euilding - Finite Element Model 4-55 ! 4-13 FEM Undeformed Shape - Entire Structure 4-56 ; 4-14 Unit 2 Typical Section Looking North or South 4-57 i
. 4-15 Critical Section at Slab / Wall Joint 4-58 V
e r
LIST OF FIGURES Fiqure Title Pace 5-1 PBAPS Spent Fuel Pool QADMOD-G Model (SectionViewEast-West) 5-21 5-2 PBAPS Spent Fuel Pool QADMOD-G Model (SectionViewNorth-South) 5-22 5-3 PBAPS Spent Fuel Pool QADMOD-G Model (PlanView) 5-23 - i l t I i 1 I I i s i L l . 1 l vi i ! , r
1.0 INTRODUCTION
1.1 CURRENT STATUS Philadelphia Electric Company (PECo) is currently pursuing the design and manufacture of new spent fuel storage racks to be placed into the spent fuel pools of Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3. The purpose of these new racks is to increase the amount of spent fuel that can be stored in the existing spent fuel pools. The racks are designed so that they can store spent fuel assemblies in a high density array. This Safety Analysis has been prepared to support PECO's request for NRC review and approval of new spent fuel racks in accordance with PBAPS Units 2 and 3 Facility Operating Licenses DPR-44 and OPR-56(1). There are two spent fuel pools at PBAPS; one for each nuclear unit. The existing racks in each of these pools have 2608 total storage cells. In the 1987-88 time frame, these units will lose their full-core discharge reserve storagecapacity(764 assemblies);andinthe 1991-1992 time frame, they will no longer have the capacity to store fuel discharges from the operating units. l Therefore, to ensure that sufficient capacity continues to exist at PBAPS to store discharged fuel assemblies, PEco plans to replace the existing storage l racks with new spent fuel storage racks whose design will allow for more dense l storage of spent fuel, thus enabling the existing pools to store more fuel in the same space as occupied by the current racks. 1.2
SUMMARY
OF REPORT 1 This Safety Analysis Report follows the guidance of the NRC position paper entitled, "0T 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(2). Sections 3.0 through 5.0 of this report are consistent with the section/ subsection format and content of the NRC position paper, Sections !!! through V. 1-1 Rev. 1 A
This report contains the nuclear, thermal-hydraulic, mechanical, material, structural, and radiological design criteria to which the new racks are designed. The nuclear and thermal-hydraulic aspects of the report (Section 3.0) address the neutron multiplication factor, considering normal storage and handling of spent fuel as well as postulated accidents with respect to criticality and the ability of the spent fuel pool cooling system to maintain sufficient cooling. Temporary fuel storage considerations during rack removal and installation are also addressed. Mechanical, material,andstructuralaspects(Section4.0)involvethe capability of the fuel assemblies, storage racks, and spent fuel pool structure to withstand effects of natural phenomena and other service loading conditions. The environmental aspects of the report (Section 5.0) concern the thermal and radiological release from the facility under normal and accident conditions. This section also addresses the occupational radiation exposures, generation of radioactive waste, need for expansion, commitment of material and nonmaterial resources, and a cost-benefit assessment.
1.3 CONCLUSION
S On the basis of the design requirements presented in this report, operating experience with high density fuel storage, and material referenced in this report, it is concluded that the proposed modification of tne PBAPS Units 2 and 3 spent fuel storage facilities will continue to provide safe spent fuel storage, and that the modification is consistent with the facility design and operatingcriteriaasprovidedinthePSAPSUpdatedFSAR[3]andOperating Licenses.
1.4 REFERENCES
- 1. PBAPS Units 2 and 3, Facility Operating Licenses OPR-44 and OPR-56, Docket Nos. 50-277 and 50 278.
12 Pov. 1 _t _ _ _ _ _ _ _ _ _ _ _ __
1
- 2. Nuclear Regulatory Comission, Letter to all Power Reactor Licensees, from B. K. Grimes, April 14, 1978, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", as amended by the NRC letter dated January 18, 1979.
- 3. PBAPS Units 2 and 3, Updated Final Safety Analysis Report, Docket Nos. 50-277 and 50-279.
9 I I l 8 11 Rev. I t_' '
2.0
SUMMARY
OF EXISTING RACK DESIGN PBAPS Units 2 and 3 spent fuel pools each contain 26 freestanding spent fuel assembly storage modules having a capacity to store 2608 spent fuel assemblies. The quantity, size and weight of each of the existing rack modules is provided in Table 2-1. Figure 2-1 and 2-2 illustrates the current spent fuel rack module layout in PBAPS Units 2 and 3, respectively. The rack modules consist of an array of storage cavities having nominal center-to-center spacing of 7.0 inches; each storage cavity can accommodate one fuel assembly. The fuel assembly storage cavities are structurally connected to form 26 freestanding spent fuel assembly storage modules per spent fuel pool. The existing high density spent fuel racks are free-standing, all-anodized aluminum construction. Each rack consists of six basic components:
- a. top grid casting
- b. bottom grid casting
- c. poison can assemblies
- d. side plates
- e. corner angle clips
- f. adjustable foot assemblies Each component is anodized separately. The top and bottom grids are machined to maintain nominal fuel spacing of 7.00 inches center to center within the rack and a spacing of 9.75 inches between centers of cavities in adjacent .
racks. Poison cans nest in pockets which are cast in every other cavity opening of the grids. This arrangement ensures that no structural loads will be imposed on the poison cans. The poison cans consist of two concentric square tubes with four Boral plates located in the annular gap. The Boral is positioned so it overlaps the fuel pellet stack length in the fuel assemblies by 1 inch at the top and at the bottom. The outer can is formed into the inner can at each end and seal welded to isolate the Boral from the spent fuel pool (SFP) water. Each can is pressure and vacuum leak tested. The grid 2-1 Rev. 1
I structures are bolted and riveted together during fabrication by four corner angles and four side shear panels. Leveling screws are located at the rack corners to allow adjustment for variations in pool floor level of up to
+ 1 inch. To maintain a flat, uniform contact area, the bearing pad at the bottom of each leveling screw is free to pivot.
I 2-2 Rev. 1 I l
TABLE 2-1 QUANTITY, SIZES AND WEIGHTS OF EXISTING SPENT FUEL RACK MODULES PBAPS UNITS 2 AND 3 Number Quantity Rack Size of Cavities Weight Per Rack Total Weight 1 8x8 64 8,704 8,704 3 8 x 10 240 10,880 32,640 7 8 x 12 672 13,056 91,392 14 9 x 12 1,512 14,688 205,632 _1, 10 x 12 120 16,320 16,320 26 2,608 354,688 2-3 Fev. 1 A
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3.0 NUCLEAR AND THERMAL HYDRAULIC CONSIDERATIONS The nuclear criticality design considerations and analysis methods used for PBAPS are the same as those used by Westinghouse for several other nuclear power plants which the NRC has previously reviewed and found acceptable. Those plants are listed below: Utility Plant Fuel Type Arkansas Power and Light Arkansas 1 & 2 PWR Carolina Power and Light Shearon Harris 1, 2, 3 & 4 PWR & BWR Carolina Power and Light H. B. Robinson PWR Duke Power Oconee 1, 2 & 3 PWR Duke Power McGuire 1 & 2 PWR Florida Power and Light Turkey Point 3 PWR Gulf States Utilities River Bend 1 PWR Public Service of New Hampshire Seabrook PWR 3.1 NEUTRON MULTIPLICATION FACTOR Criticality of fuel assemblies in the spent fuel storage rack is prevented by the design of the rack. This is done by maintaining a minimum separation between assemblies and using a fixed neutron absorbing material between adjacent fuel assemblies. The design basis for preventing criticality is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (kerr) of the fuel assembly array will be less than0.95asrecommendedinANSI/ANS-57.2-1983(1) and in NRC "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978, and revised January 18,1979(2), 3-1 Rev. I 1 _ - - ____
The codes, standards, and regulations (or pertinent sections thereof) used to meet the above design basis are listed below: USNRC Regulatory Guide 1.13 " Spent Fuel Storage Facility Design Basis," Proposed Rev. 2, Dec. 1981. USNRC Standard Review Plan, NUREG-0800, Section 9.1.2." Spent Fuel Storage," Rev. 3 July 1981. l USNRC Branch Technical Position CPB 9.1-1, " Criticality in Fuel Storage Facilities." l USNRC Guidance, "0T Position for Review and Acceptance of Spent Fue.1 Storage and Handling Applications," April 14, 1978, and modification dated January 18, 1979. l ANSI /ANS-8.1-1983, " Nuclear Criticality Safety In Operations with Fissionable Materials Outside Reactors." ANSI /ANS-57.2-1983, " Design Requirements for LWR Spent Fuel Storage Facilities at Nuclear Power Plants." ANSI /ANS-52.1-1983, " Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants." ANSI N16.9-1975, " Validation of Calculational Methods For Nuclear Criticality Safety." l 3.1.1 Normal Storace
- a. The fuel assembly contains the highest enrichment authorized without any control rods or any noncontained burnable poison and is at its most reactive point in life. The GE 7 x 7 fuel assembly is more reactive than the GE 8 x 8 or 8 x 8(R) fuel assembly. Table 3-1 lists the fyel parameters used in the analysis.
3-2 Rev. I t i- o .. ..
The assembly is conservatively modeled with water replacing the assembly grid volume and no U-234 or U-236 in the fuel pellet.
- b. The moderator is demineralized water at the temperature within the design limits of the pool which yields the largest reactivity. A conservative value of 1.0 gm/cm3 is used for the density of water. There is no dissolved boron present in the water.
- c. The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the design. The nominal case calculation is infinite in lateral and axial extent. Huwever, neutron absorber plates are not necessary on the periphery of the modular array 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 cells with neutron absorber material is a conservative assumption.
- d. Mechanical uncertainties and biases due to mechanical l tolerances during fabrication are treated by either using
" worst case" conditions or by performing sensitivity studies a'nd obtaining appropriate values. The items included in the analysis are:
Neutron absorber pocket thickness Cell ID Center-to-center spacing Cell bowing The calculated method uncertainty and bias is discussed in Section 3.1.3. 3-3 Rev. 1 G .
l
- e. Credit is taken for the neutron absorption in full length structural materials and in solid materials added specifically for neutron absorption. A minimum poison loading is assumed in the poison plates and B 4 C particle self shielding is included as a bias in the reactivity calculation.
3.1.2 Postulated Accidents The criticality analysis includes postulated accidents so that the double contingency of ANSI /ANS 57.2-1983(1) is met and that the effective neutron multiplication factor (Keff) is less than or equal to 0.95 under all conditions. l Accident conditions that will not result in an increase in Keff of the rack 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 assembly has approximately thirteen 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. They include a dropped or misplaced fuel assembly outside the periphery of the racks. The rack neutron absorber loading is designed such that the nominal rack Keff will be approximately 0.90 since there is no soluble boron in the spent fuel pool. This will allow a 0.05 delta-K margin to the design limit for uncertainties. biases, and accident conditions. Thus, for postulated accidents, Keff is less than or equal to 0.95. The " optimum moderation" accident is not a problem in spent fuel storage racks because presence of poison plates removes the conditions necessary for
" optimum moderation". The Keff continually decreases as moderator density decreases from 1.0 gm/cm3 to 0.0 gm/cm3 in poison rack designs.
3-4 Rev. I n' _ _ _ _
I The PBAPS Updated FSAR evaluates the potential for a cask drop over the spent fuel pool. The rerack program will not alter the results of that evaluation I and will not increase the likelihood of a cask drop affecting storage rack criticality. The PBAPS Updated FSAR evaluates the potential for tornado generated missiles hitting the spent fuel pool,and protection of the spent fuel from such missiles. The rerack program will not alter the results of that evaluation and will not increase the likelihood of a tornado generated missile affecting the storage rack criticality. The results of earthquake loads on the deformation and position of the racks is described in Section 4.6. The racks are designed to withstand these loads and maintain a relative fuel position resulting in a Keff less than or equal to 0.95. 3.1.3 Calculation Methods The calculation method and cross-section values are verified by comparison with critical experiment data for assemblies similar to those for which the racks are designed. These benchmarking data are sufficiently diverse to establish that the method bias and 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 (3,4) for cross-section generation and KENO IV(5) for reactivity determination. The 218 energy group cross-section library (3) that is the common starting point for all cross-sections used for the benchmarks and the storage rack is generated from ENBF/B-IV data. The NITAWL program (4) includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is performed by the XSDRNPM program (4) 3-5 Rev. 1 D
which is a one-dimensional SN transport theory code. These multigroup cross-section sets are then used as input to KENO IV(5) which is a three dimensional Monte Carlo theory program designed for reactivity calculations. A set of 27 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 (Boral, steel, water) that simulate LWR fuel shipping and storage conditions (6,7) to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials (8) (Plexiglas, steel and air) that demonstrate the wide range of applicability of the method. Table 3-2 summarizes these experiments. The average Keff of the benchmarks is 0.9998 which demonstrates that there is no bias associated with the method. The standard deviation of the Keff values is 0.0014 Ak. The 95/95 one sided tolerance limit factor for 27 values is 2.26. Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0032 Ak. The fabrication tolerances for the racks allow for the nominal center-to-center spacing to be randomly reduced for individual cells. This change which results in an increase in Keff, is included in the calculation of the worst case Keff. The effect of the tolerances on material thicknesses also results in an increase in Keff which will be treated conservatively as a bias. The neutron absorbing material particle radius is used in the analysis to determine the neutron absorbing material self-shielding reactivity bias for the neutron absorbing plates. The particle size distribution is based on data supplied by the manufacturer. The final result of the analysis is that the criticality design criterion is met when the calculated effective multiplication factor, plus the total uncertainty (TV) and any biases, is less than 0.95. 3-6 Rev. 1 l
3.1.4 Criticality Analysis Results The spent fuel storage rack is described in Section 4.0. The minimum B10 loading in We poison plates is 0.021 gm 8 10/cm2, For normal operation and using the method in the above sections, the Keff for the rack is. determined in the following manner. Keff = Kworst + Bmethod + Bp art + [(ksnominal)2
+ (ksmech)2 + (ksmethod)2]1/2 where:
Kworst = worst case KEN 0 Keff which includes asymmetric fuel assembly position, material and mechanical. Bmethod = method bias determined from benchmark critical comparisons. Bp art = bias to account for the poison particle self-shielding. ksnominal = 95/95 uncertainty in the nominal case KENO Keff. ksmethod = 95/95 uncertainty in the method bias, ksmech = 95/95 uncertainty to account for thickness, spacing and bowing tolerances. The final Keff from this analysis for all analyzed conditions is less than 0.95, including all uncertainties at a 95/95 probability / confidence level. The nominal Keff calculated in the criticality analysis for normal storage is 0.9198 with a 95 percent probability /95 percent confidence level uncertainty of +0.0045. The maximum Keff calculated for normal storage or postulated accident conditions is 0.9357 including uncertainties at a 95/95 probability / confidence level. Thus, all acceptance criteria for criticality are met. 3-7 Rev. 1 l
1 3.1.5 Acceptance Criteria for Criticality The neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions. Methods for initial and long-term verification of poison material stability l and mechanical integrity are discussed in Section 4.8. 3.2 DECAY HEAT CALCULATIONS FOR THE SPENT FUEL POOL The Spent Fuel Pool Cooling System design is described in the PBAPS Updated FSAR(9), Section 10.5. The heat load resulting from the presence of 3819 spent fuel assemblies is within the capabilities of the existing cooling system to maintain pool bulk water temperature at or below a design basis temperature of 1500 F. The analysis of the Spent Fuel Pool Cooling System capacity is based on the first assembly in each discharge being unloading to the spent fuel pool 120 hours after reactor shutdown. Fuel assemblies in normal discharges are conservatively assumed to have an exposure of 40,000 MWD /MTU. THe fuel decay energy release rates are evaluated in accordance with the NRC Branch Technical Position APCSB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling". For analysis in the normal condition, the spent fuel racks are considered to be filled with spent fuel discharged on 18 month refueling schedules. The spent fuel pool cooling system is in operation with the design cooling water flow rate and temperature on the tube side of the SFP heat exchangers and the design flow rate of pool water to the shell side of each heat exchanger at the calculated bulk pool temperature. For this normal refueling condition, pool outlet temperature can be maintained below the design temperature of 1500F with only two of the three heat exchangers in service. If only one heat exchanger is in service, 1500F will be reached in approximately 7.3 days, assuming an initial pool temperature of 1100F. 3-8 Rev. 1 a_
l For analysis in the abnormal condition, all but 764 cavities (one full core) of the spent fuel racks are filled with spent fuel discharged according to the anticipated 18 month refueling schedule. With all other cooling system conditions the same as stated above, the spent fuel pool temperatures are I calculated after a full core is discharged to fill the remaining spaces. For this limiting full core offload, all three heat exchangers must be in service ! to maintain pool outlet temperature below 1500F. With one and two in service, the time to reach 1500F is approximately 7.3 days and 26.9 days, respectively assuming an initial temperature of 1100F. With only one exchanger in service, the time to reach the boiling temperature of 2120F is 27.5 days after loading has begun, assuming an initial pool temperature of 1100F. The maximum spent fuel pool heat load occurs after the full core discharge condition described above. In this case the Residual Heat Removal System is available if necessary, to supplement the Spent Fuel Pool Cooling System to maintain temperatures below 1500 F. This heat load is well within the capability of the Residual Heat Removal System. A sumary of the cooling system analysis is presented in Table 3-3. The results of a total loss of spent fuel pool cooling are presented in Section 3.4.3 assuming the normal water level is maintained. 3.3 LOCAL THERMAL-HYDRAULIC ANALYSIS FOR THE SPENT FUEL P00L' The purpose of 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 Peach Bottom spent fuel pools. 3.3.1 Criteria The rack design must allow adequate cooling by natural circulation and by flow provided by the spent fuel pool cooling system. 3-9 Rev. I t
- i. _ _ _ _ , . - . - - . . - - . --
The criteria used to determine the acceptability of the rack design from a thermal-hydraulic viewpoint is summarized as follows: I
- a. The coolant must remain subcooled at all points within the pool when the cooling system is operational.
- b. When the cooling system is postulated to be inoperable, the temperature of the fuel cladding must be sufficiently low that no structural failures would occur and that no safety concerns would exist.
3.3.2 Key Assumptions
- a. The nominal spent fuel pool water level is approximately 24 feet above the top of the fuel storage racks.
- b. The maximum fuel assembly decay heat output is 11.39 Btu /sec per assembly (based on Ref. 10) following 312 hours decay time (based on averaging decay time to first and last fuel assembly discharge) after shutdown.
- c. Bulk thermal-hydraulic analysis shows that the maximum pool temperature will not exceed 1500F when the spent fuel pool cooling system is in operation. However, for conservatism, the temperatures of the storage racks and the stored fuel are evaluated assuming that the temperature of the water at the inlet to the storage cells is 1500F during normal operation.
- d. When the pool cooling system is not in operation, the maximum temperature at the inlet to the cells is assumed to be equal to the saturation temperature of 2120F at the top of the pool.
3-10 Rev. 1 A
a 3.3.3 Analytical Methods and Calculations j A 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 and a multi-channel formulation is used to determine the variation in axial flow 4 velocities through the various storage cells. The hydraulic resistance of the t storage cells and the fuel assemblies is conservatively modeled by applying l large uncertainty factors to loss coefficients obtained from various sources. Where necessary, the effect of Reynolds Number on the hydraulic resistance is considered, and the variation in momentum and elevation head pressure drops I 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 3-1 where the flow paths are indicated by arrows. Note that each storage location shown in the sketch actually corresponds to a row j of storage locations that is located at the same distance from the pool walls.
- This is more clearly shown in a plan view, Figure 3-2.
For modeling purposes the lateral flow area underneath the storage cells is decreased as the distance from the wall increases. This counteracts the l 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 the i turning losses that are experienced as the flow branches up into the cells. i These effects are considered in the natural circulation analysis. i i ! 3-11 Rev. 1 l ' t- -- - . _ . _ - _ _ _ - ~ ---- _ _.-_ _ _ _ - - - - _.. - - - _ - _ - _ . _ . - _ - -
i Fuel assemblies from the most recently discharged batch (" hottest" fuel assemblies) are assumed to be located in various rows during different calculations in order to ensure that they may be placed anywhere within the pool without violating safety limits. In order to 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. This is the approach that has been used to perform the analysis for the Peach Bottom spent fuel storage racks. 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 to minimize the resistance, the design of the inlet region of the racks has been chosen to maximize this flow area. Each storage location has one large flow opening as shown in Figure 3-3. The use of these large flow holes virtually eliminates the possibility that all flow into the inlet of a given storage location 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 periormed for various assumed blockages. 3-12 Rev. 1 I
3.3.3.1 Normal Operation Results Basis: a. Cooling System Operational (i.e., temperature of water at inlet to storage racks maintained at 1500F or below)
- b. 312 hours (average decay time for discharged batch) after shutdown-Decay Heat = 11.39 Btu /second/ assembly
- c. Uniform decay heat loading in pool - No credit for icwer actual heat input
- d. The peak rod is assumed to have 60 percent more heat output than average rod
- e. 5 mil thick crud layer on all rods f.. All storage locations filled with spent fuel resulting from normal refueling
- g. Minimum heat transfer coefficient in the storage cells is based on laminar forced flow. The analysis conservatively does not take credit for laminar or turbulent free or forced turbulent convection.
Results of the analysis show that no boiling occurs at any point within the storage racks when the normal cooling system is in operation or whenever pool temperature is maintained at or below 1500F. 3.4 POTENTIAL FUEL AND RACK HANDLING ACCIDENTS The method for moving the racks into and out of the spent fuel pool is briefly mentioned in Section 4.7. The methods utilized ensure that postulated accidents do not result in a loss of cooling to either the spent fuel pool or the reactor, or result in a k rre in the spent fuel pool exceeding 0.95. 3-13 Rev. 1
3.4.1 Rack Module Mishandling The potential for mishandling of rack modules during the rerack operation will be precluded through the use of heavy load handling procedures, load paths, and installation procedures. At no time will the cask handling crane carry a rack module over top of stored spent fuel. 3.4.2 Flow Blockage Analysis The effects of a postulated flow blockage accident were calculated. Basis: a. Cooling system operational (1 e. temperature of water at inlet to storage racks maintained at 1500F). 1
- b. 312 hours (average decay time for discharged batch) after shutdown decay heat = 11.39 Btu /sec/assy.
- c. Uniform decay heat loading in pool - no credit for lower actual heat.
- d. The peak rod is assumed to have 60 percent more heat output than average rod.
- e. 5 mil thick crud layer on all rods.
- f. All storage locations filled. '
- g. Minimum heat transfer coefficient in the storage cells is based on laminar forced flow. The analysis conservatively does not take credit for laminar or turbulent free or forced turbulent convection.
3-14 Rev. 1 1
e Results of the analysis show that should up to 90% flow blockage occur. there would still be no local or bulk boiling inside the cells. Because of the large flow openings that are used in the storage racks, it is very improbable that a complete blockage could occur. 3.4.3 Accident Conditions Under postulated accident conditions where all spent fuel pool cooling is assumed to be inoperative the spent fuel is cooled by allowing the pool water to boil to remove the decay heat from the spent fuel pool while maintaining pool water level. Although it is highly unlikely that a complete loss of cooling capability would occur, since the Residual Heat Removal System can be connected to the spent fuel pool, the racks are analyzed for this condition. With the Spent Fuel Pool Cooling System not in service, and a pool outlet temperature of approximately 1000F before beginning to load the pool with a full core offload 120 hours after shutdown, the time for the pool to reach the boiling temperature of 2120F is approximately 82 hours after loading has begun. i Basis: a. The temperature of water at the spent fuel racks' inlet is assumed to be 2120F which corresponds to the saturation temperature at the top of the pool.
- b. A spent fuel pool water level of approximately 24 feet above the top of the racks is maintained by providing makeup water to the pool from the installed makeup system, or other means.
- c. The assemblies that are evaluated are initially put into the pool at 312 hours after shutdown.
- d. The peak rods are assumed to have 60 percent greater heat output than average rods.
- e. All storage locations are filled and all downflow occurs in the peripheral gap.
3-15 Rev 1 e t
_ _ , , -. a , _---,a _.- a - _.m . - Results of this analysis show that due to the effects of natural circulation, the fuel cladding temperatures are maintained sufficiently low to preclude structural failures. The maximum calculated fuel cladding temperature is 2540F.
! Boiling of the water within the storage locations does not occur since the ,
calculated maximum water temperature of 2250F is below the saturaticn temperature at the top of the racks of approximately 2400F. 3.5 TECHNICAL SPECIFICATIONS The racks have been designed to ensure subcriticality based on the existing PBAPS Technical Specifications 5.5., B and D which requires theek rt of the spent fuel storage pool to be less than or equal to 0.95. and an average fuel assembly loading not to exceed 17.3 grams U-235 per axial centimeter of total active fuel height of the assembly. The spent fuel pool thermal load analysis is based on thermal loads resulting i from the reactor being subcritical for at least 120 hours prior to movement of fuel assemblies from the reactor vessel to the spent fuel pool. One hundred twenty hours is the minimum time required to prepare for fuel transfer from the reactor to the spent fuel pool. This spent fuel decay time limits the decay heat load to within the limits analyzed. The installation and use of new spent fuel storage racks does not necessitate the revision of any existing PBAPS Technical Specifications.
3.6 REFERENCES
- 1. ANSI /ANS-57.2-1983. " Design Requirements for LWR Spent Fuel Storage Facilities at Nuclear Power Plants."
- 2. USNRC Guidance, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978 and modification dated January 18, 1979.
3-16 Rev. 1
- 3. W. E. Ford, III, et al., "A 218-Group Neutron Cross-Section Library in the AMPX Master Interface Format For Criticality Safet/ Studies,"
ORNL/CSB/TM-4 (July 1976). 4 N. M. George, et al. "AMPX: A Modular Code System for Generating Coupled Multi-group Neutron-Gama Libraries from ENDF/B," ORNL/TM-3706 (March 1976).
- 5. L. M. Petrie and N. F. Cross, " KENO IV--An Improved Monte Carlo Criticality Program," ORNL-4938 (November 1975).
- 6. S. R. Bierman, et al., " Critical Separation Between Subcritical Clusters of 2.35 wt percent 2350 Enriched U02 Rods in Water with Fixed Neutron Poisons." Battelle Pacific Northwest Laboratories PNL-2438 (October 1977).
- 7. S. R. Bierman. et al. " Critical Separation Between Subcritical Clusters of 4.29 wt percent 235U Enriched UO2 Rods in Water with Fixed Neutron Poisons." Battelle Pacific Northwest Laboratories PNL-2615 (March 1978).
- 8. J. T. Thomas, " Critical Three-0imensional Arrays of U (93.2) --
Metal Cylinders," Nuclear Science and Engineering, Volume 52. Pages 350-359 (1973).
- 9. Philadelphia Electric Company " Peach Bottom Atomic Power Station Units 2 and 3. Final Safty Analysis Report (FSAR)", 1972 as amended by UFSAR Docket 50-277 and 50-278.
- 10. U.S. Nuclear Regulatory Comission (USNRC) Branch Technical Position APCSB 9.2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling."
3-17 Rev. 1 A
r l t TABLE 3-1 PARAMETERS FOR BWR FUEL ASSEMBLIES TO BE STORED IN THE PEACH BOTTOM RACKS l l FUEL BUNDLE 8 x 8R 8x8 7x7 Enrichment in w/o U-235 3.5 w/o 3.5 w/o 3.5 w/o Lattice Pitch 0.640" 0.640" 0.738" Number of Fuel Rods / Assembly 62 63 49 Number of Water Rods / Assembly 2 1 0 Fuel Rod Pellet 0.0. 0.410" 0.416" 0.487" fuel Rod Pellet Theoretical Density 95% 95% 95% Fuel Rod Clad 0.D. , 0.483" 0.493" 0.563" Fuel Rod Clad Thickness 0.032" 0.034" 0.032" Fuel Rod Clad Material Zr-2 Zr-2 Zr-2 l Water Rod 0.D. 0.591" 0.591" --- Water Rod Tube Wall Thickness 0.030" 0.030" --- Water Rod Material Zr-2 Zr-2 --- FUEL CHANNEL l Material Zr-4 Zr-4 Zr-4 Thickness 0.113" 0.113" 0.113" Outside Square Dimension 5.84" 5.84" 5.83" l l 3-18 Rev. I 1 .
.I
TABLE 3-2 BENCH CRITICAL EXPERIMENTS [6,7,8] General Enrichment Separating Characterizing Description w/o U235 Reflector Material Separation (cm) K,rr
- 1. 002 rod lattice 2.35 water water 11.92 1.004 1 .004
- 2. 002 rod lattice 2.35 water water 8.39 0.993 1 .004
- 3. 002 rod lattice 2.35 water water 6.39 1.005 1 .004
- 4. 002 rod lattice 2.35 water stainless steel 4.46 0.994 1 .004
- 5. U02 rod lattice 2.35 water stainless steel 10.44 1.005 1 .004
- 6. UO2 rod lattice 2.35 water stainless steel 11.47 0.992 + .004
- 7. U02 rod lattice 2.35 water stainless steel 7.76 0.992 + .004
- 8. U02 rod lattice 2.35 water stainless steel 7.42 1.004 1 .004
- 9. U02 rod lattice 2.35 water boral 6.34 1.005 1 .004
, 10. 002 rod lattice 2.35 water boral 9.03 0.992 1 .004 A 11. U02 rod lattice 2.35 water boral 5.05 1.001 1 .004
- 12. U02 rod lattice 4.29 water water 10.64 0.999 1 .005
- 13. U02 rod lattice 4.29 water stainless steel 9.76 0.999 ! .005
- 14. U02 rod lattice 4.29 water stainless steel 8.08 0.998 1 .006
- 15. UO2 rod lattice 4.29 water boral 6.72 0.998 ! .005
- 16. U metal cylinders 93.2 bare air 15.43 0.988 1 .003
- 17. U metal cylinders 93.2 paraffin air 23.84 1.006 1 .005
- 18. U metal cylinders 93.2 bare air 19.97 1.005 ! .003
- 19. U metal cylinders 93.2 paraffin air 36.47 1.001 ! .004
- 20. U metal cylinders 93.2 bare air 13.74 1.005 t .003
- 21. U metal cylinders 93.2 paraffin air 23.48 1.005 1 .004
- 22. U metal cylinders 93.2 bare plexiglas 15.74 1.010 1 .003
- 23. U metal cylinders 93.2 paraffin plexiglas 24.43 1.006 1 .004
- 24. U metal cylinders 93.2 bare plexiglas 21.74 0.999 1 .003
- 25. U metal cylinders 93.2 paraffin plexiglas 27.94 0.994 1 .005.
- 26. U metal cylinders 93.2 bare steel 14.74 1.000 1 .003
!?
- 27. U metal cylinders 93.2 bare plexiglas steel 16.67 1.006 1 .003
TABLE 3-3
SUMMARY
OF COOLING SYSTEM ANALYSIS RESULTS
- 1) Heat-Exchanger Capability One exchanger in service = 3.76 x 106 BTU /HR.
Two exchangers in service = 7.52 x 106 BTU /HR. Three exchangers in service = 11.28 x 106 BTU /HR.
- 2) Maximum Pool Heat Load to insure exit temperature is below 1500F.
One exchanger in service = 8.66 x 106 BTU /HR. Two exchangers in service = 17.33 x 106 BTU /HR. Three exchangers in service = 26.0 x 106 BTU /HR.
- 3) Normal Refueling; 1/3 core every 18 months.
a) Maximum heat load - 13.14 x 106 BTU /HR at 312 hours after shutdown b) Maximum pool outlet temperatures One exchanger in service = 180.00F Two exchangers in service = 135.00F Three exchangers in service = 120.00F , c) With one exchanger in service - time to reach 1500F = 7.3 days *
- 4) Full core offload just before normal refueling.
a) Maximum heat load = 23.12 BTU /HR. x 106 b) Maximum pool outlet temperatures One exchanger in service = 2120F Two exchangers in service = 169.00F Three exchangers in service = 143.00F c) Time to reach 150,0F One exchanger in service = 7.3 days
- Two exchangers in service = 26.9 days
- d) Time to reach 2120F One exchanger in service = 27.5 days *
- Assuming initial pool temperature of 1100F j
3-20 Rev. 1
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$ PENT FUEL POOL NATURAL CIRCULATION MODEL (ELEVATION VIEW)
FIGURE 3-1 3-21 Rev. 1 0674C/0207C 1
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_., ..g. . .', SPENT FUEL POOL NATURAL CIRCULATION MODEL (PLAN VIEW) FIGURE 3-2 l 3 22 Rev. 1 0674C/0207C ( A _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - __
_m ...m._ ___m___ _..___..m _ ._.... _ _ . u _ _ _ _ . . . _ _ _ - . _ _ _
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i
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i 1 i i SPENT FUEL RACK INLET FLOW AREA (PLAN VIEW) FIGURE 3-3 3-23 Rev. 1 0614C/0201C
-- l - - . _ . _ _ _ _ - . _ _ _ - _ _ - . _ _ _ , _ . - - , _ _ , . _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ . _ _ _ . . _ . - _ . _ . _ . _ _ _ _ _ _. _ _ _ _ - _ _ , _ , - _ _ _ _ _ - - _ -
4.0 MECHANICAL, MATERIAL AND STRUCTURAL CONSIDERATIONS
4.1 DESCRIPTION
OF STRUCTURE 4.1.1 Description of Spent Fuel Pool Structure A description of the spent fuel storage pool is provided in Section 10.3.4.2 of the PBAPS Updated FSAR.(1) The spent fuel has been designed to meet Seismic Category I requirements. The walls and floor of the spent fuel pool are lined with an eight gage thick stainless steel liner. This liner serves only as a water tight boundary, it is not a structural member. Steel embedments are provided in the pool walls and slab to attach the liner to the pool structura. Monitoring trenches are provided behind the liner for detecting and collecting any leakage. Any leakage is directed to the liquid radwaste system. Individual liner plates are connected with full penetration welds. The spent fuel pools are located inside the Reactor Buildings in an elevated position adjacent to the North (Unit 2) and South (Unit 3) sides of the drywell shield walls, shown on Figures 4-1 through 4-3. The spent fuel pool has been evaluated structurally for the additional loading due to the increased number of fuel elements and modified rack design in accordance with Standard Review Plan (SRP) Section 3.8.4(2) and the NRC position paper entitled, "0T 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(3), and other applicable codes and standards identified in Section 4.2. The new spent fuel storage racks are designed so that the floor loading from racks filled with spent fuel assemblies will not exceed the structural capacity of the spent fuel pool structure. The analysis of the spent fuel pool slab for the effects of a cask drop as described in the PBAPS Updated 41 Rev. 1 a
FSAR is unchanged by the storage of additional spent fuel assemblies. Critical areas of the structure, such as the junction between the wall and the floor, have been checked for the additional loading. 4.1.2 Description of Spent Fuel Racks The function of the spent fuel storage racks is to provide for storage of new and spent fuel assemblies in a flooded pool, while maintaining a coolable , geometry, preventing criticality, and protecting the fuel assemblies from excessive mechanical or thermal loadings. A list of design criteria is given below:
- 1. The racks are designed in accordance with the NRC #0T Position for Review and Acceptance of Spent Fuel Stordge and Handling Applications." dated April 14, 1978 and revised January 18, 1979,
- 2. The racks are designed to meet the nuclear requirements of ANSI /ANS-57.2-1983, Sections 6.4.2.1 and 6.4.2.2. The effective multiplication factor, Kerr, in the spent Fuel pool is less than or equal to 0.95, including all uncer-tainties and under all credible conditions as described in Section 3.0.
1
- 3. .The racks are designed to allow coolant flow such that boiling does not occur in the water channels between the fuel assemblies in the rack.
1
- 4. The racks are designed to Seismic Category I requirements, '
and are classified as ANS Safety Class 3 and ASME Code Class 3 Component Support structures. The structural evaluation and seismic analyses are performed using the specified loads and load combinations in Table 4.2. I 4-2 Rev. 1 i
- 5. The racks are designed to withstand loads which may result from fuel handling accidents and from the maximum uplift force of the fuel handling crane without violating the criticality acceptance criterion.
- 6. Each storage position in the racks is designed to support and guide the fuel assembly in a manner that will minimize the possibility of application of excessive lateral, axial and bending loads to fuel assemblies during fuel assembly handling and storage.
- 7. The racks are designed to preclude the insertion of a fuel assembly in other than design storage locations.
- 8. The materials used in construction of the racks are compatible with the storage pool environment and do not contaminate the fuel assemblies.
The spent fuel pool storage rack arrangements for Peach Bottom Units 2 and 3 are shown in Figures 4-4 and 4-5. Each storage location is capable of storing 7x7,88,and8x8(R)BWRfuelassemblies(withorwithouttheirfuel channels) at a design enrichment equal to or less than 3.50 maximum weight percent U235-4.1.2.1 Design of New Spent Fuel Racks The high-density rack details are shown in Figure 4-6. The rack modules are free-standing and self supporting. The modules are neither anchored to the floor nor braced to the pool walls. The storage cells within a rack module are assembled in a checkerboard pattern and welded to a common base plate. The vertical corners of adjacent cells are welded together to form an integral structure (Figures 4-6 and 4-7). Each rack module is provided with remotely adjustable leveling pads which are located at the center of the four corner 4-3 Rev. 1
~
cells and at interior module locations. T:tese pads are used to level the racks during installation and distribute loads to the pool floor. The rack module consists of two major sections; the base support assembly and the cell assembly. Figure 4-6 illustrates these sections. The racks are constructed from type 304LN stainless steel except for the leveling screws which are type 17-4 PH stainless steel. The major components of the base support assembly are the leveling block assembly, the support pad, and the leveling screw. The top of the leveling block assembly is welded to the base plate. The support pads transmit the loads to the pool floor and provide a sliding contact. The leveling screw permits remote leveling adjustment of the rack. The major components of the cell assembly are the fuel assembly cell, the Boraflex (neutron absorbing) material, and the wrapper. The wrapper is attached to the outside of the cell by intermediate spot welding along the ! entire length of the wrapper. The wrapper, which holds the Boraflex material I in place, provides for venting of the Boraflex to the pool environment. Depending on the cell location in the rack module, and the associated criticality requirements, some cells have a Boraflex wrapper on all four sides, some on three sides, and some on two sides. Cells with four wrappers are located in the interior of the rack, cells with three wrappers are located ontheperipheryoftherack,andcellswithtwo(adjacent)wrappersare located at the corners of the rack. Rack module data for both units are described in Table 4.1. 4.1.2.2 Fuel Handling The storage of additional spent fuel assemblies in the spent fuel pool will not affect tFe analysis and consequences of the design basis fuel handling accidents as presented in the PBAPS Updated FSAR. The spent fuel storage racks are designed to withstand the design basis fuel handling accident. The resulting criticality and radiological consequences of a postulated fuel l 4-4 Rev. I f JL_____.____.____
assembly drop are addressed in Sections 4.6 and 5.3, respectively, of this safety analysis report. The parameters of the postulated fuel assembly drop accident are contained in Section 4.6. 4.2 APPLICABLE CODES, STANDARDS, AND SPECIFICATIONS l The spent fuel storage racks are designed and fabricated to applicable provisions of the following codes, standards, and NRC Regulatory Guides:
- a. NRC "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and revised January 18, 1979.
- b. Code of Federal Regulations 10CFR50, Appendix A. " General Besign Criteria for Nuclear Power Plants," Criteria 61 and 62.
Code of Federal Regulations 10CFR21 " Reporting of Defects and Noncompliance." ,
- c. NRC Regulatory Guides R.G. 1.13 Spent Fuel Storage Facility Design Basis, Rev. 2, Dec. 1981 (Draft).
R.G. 1.25 Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling 1 Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water
- Reactors, March 1972. ,
R.G. 1.29 Seismic Design Classifications, Rev. 3, Sept. 1978. R.G. 1.31 Control of Ferrite Content in Stainless Steel l Welding Metal, Rev. 3 April 1978. 4-5 Rev. 1 A__._____________________________ _ _ _ . _ _ _ _ _ _
R.G. 1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants, Rev. 2, April 1973. R.G. 1.70 Standard Format and Content of Safety Analysis Report for Nuclear Power Plants, Rev. 3 Nov. 1978. R.G. 1.92 Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev. 1. Feb. 1976. R. G. 1.142 Service Limits and Loading Combinations for ' Class I Linear-Type Component Supports, Rev. 1. Jan. 1978.
- d. NRC Standard Review Plans - NUREG-0800 SRP 3.7 Seismic Design, Rev. 1. July 1981 SRP 3.8.4 Other Category I Structures, Rev. 1. July 1981 SRP 3.8.5 Foundations, Rev. 1. July 1981 SRP 9.1.2 Spent Fuel Storage, Rev. 3, July 1981 SRP 9.1.3 Spent Fuel Pool Cooling and Cleanup System, Rev. 1. July 1981 SRP 9.2.5 Ultimate Heat Sink, Rev. 2, July 1981
- e. Industry Codes and Standards American Society of Mechanical Engineers, Boiler and Pres-sure Vessel Code 1983 Edition, Winter 1983 Addendum except as noted below:
- a. Section !!, " Material Specification."
- b. Section !!!, Division 1. " Nuclear Power Plant Components," Subsection NF, " Component Supports," 1980 Edition, Sumner 1982 Addendum l
4-6 Rev. 1 ik
l c. Section V, " Nondestructive Examination."
- d. Section IX, " Welding Requirements."
American National Standards Institute, ANSI-57.2-1983, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations." American National Standards Institute ANSI-52.1-1983, " Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants." American National Standards Institute, N45.2-1971, " Quality l Assurance Program Requirements for Nuclear Facilities." 1 American liational Standards Institute, N45.2.1-1973, " Cleaning of Fluid Systems and Associated Components for Nuclear Power Plants." American National Standards Institute, N45.2.2-1972,
" Packaging, Shipping, Receiving, Storage,and Handling of Items for Nuclear Power Plants."
American National Standards Institute, ANSI-8.1-1983,
" Nuclear Criticality Safety in Operations with Fissionable i
Materials Outside Reactors." l ACI Committee 318 " Building Code Requirements for Rein-forced Concrete (ACI 318-3), "American Concrete Institute, Detroit, Mich., 1983. Also, ACI Committee 318. " Commentary on Building Code Requirements for Reinforced Concrete (ACI 318-83)," American Concrete Institute, Detroit, Mich., 1983. 4-7 Rev. 1 A
e ACI Committee 349, " Code Requirements for Nuclear Safety Related Concrete Structures (ACI 349-80)," American Concrete Institute, Detroit, Mich., 1980. AISC, " Specification for the Design. Fabrication, and Erection of Structural Steel for Buildings," American Institute of Steel Construction, New York, N.Y., 1978. AISC, " Specification for the Design, Fabrication and Erection of Steel Safety-Related Structures for Nuclear
,- facilities." American Institute of Steel Construction, New York,N.Y.,FinalDraftofjnitialissue,9/14/83.
v' s
- f. U.S. Nuclear R'equlatory Commission (USNRC) Branch Technical Position:
,. a. CPB 9.1-1, " Criticality in Fuel Storage Facilities." '
I . b. APCSB9.2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling." 4.3 SEISMIC AND IMPACT LOADS The new spent fuel racks are designed, and the spent fuel pool structure p 'eva,1uated, using the seismic loading described in this section. Earthquake loading is predicated upon an operating basis earthquake (OBE) at h the sit'e having a horizontal ground acceleration of 0.05 g. In addition, a
%% safe. shutdown earthquake (SSE), having a horizontal ground acceleration of 0.12 g, is used to check the design to ensure no loss of function. The' site ground accelerations are providediin PBAPS Updated FSAR Appendix C.3. The OBE andSSEdesignationsusedherei$correspondtoFSARdesignationsofdesign 2
h[, earthquake (DE) and caximum credible earthquake (MCE) respectively. 1 ,, h
\
4-8 Rev. 1 O' _% _
Horizontal response spectra applicable for the spent fuel pool structure are applicable for both orthogonal horizontal directions. The vertical component of acceleration is taken as two-thirds of the horizontal ground acceleration. The seismic loads for the spent fuel pool structure are generated using the equivalent static load method. Seismic analysis of the fuel storage racks is being performed by the time I history method. Where time histories are used, the three orthogonal time histories are statistically independent. The time histories and response spectra utilized in these analyses represent the responses of the pool structure to the specified ground motion. .The seismic analysis of the racks ( is being performed with a damping value of 2 percent and 5 percent for OBE and SSE, respectively. Increased damping due to submergence in the water has not been considered in this analysis. Maximum dynamic forces and stresses are being calculated for the worst condition as determined by combination with forces and stresses computed in accordance with Section 4.4. The analysis includes the effects of the water in the pool, such as fluctuation of pressure due to acceleration, and sloshing. Deflection or movements of racks under earthquak'e loading is limited by design such that the racks do not touch each other or the spent fuel pool walls, and the fuel assemblies are not damaged. The interaction between the fuel elements and the rack has been considered. particularly gap effects. The resulting impact loads are of such small magnitudes that there is no structural damage to the fuel assemblies. 4-9 Rev. 1 O
4.4 LOADS AND LOAD COMBINATIONS 4.4.1 Spent Fuel Racks The loads and load combinations to be considered in the analysis of the spent fuel racks are shown in Table 4-2 and include those given in the NRC, "0T 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. The major structural loads are produced by the operational basis earthquake (OBE) and safe shutdown earthquake (SSE) events. The loads or stresses from the nonlinear seismic analysis are adjusted by peaking factors from the structural model to account for the stress gradients through the rack module. Consequently, the maximum loaded rack components of each type are analyzed. Such an analysis envelops the other areas of the rack assembly. For each component, the maximum seismic stress is combined with the dead weight stress to produce the total stress. 4.4.2 Spent Fuel Pool Structure All loading combinations required by USNRC Regulatory Guide 1.142 (4), USNRC StandardReviewPlan3.8.4(2),ACI(5,6),andAISC(7,8)wereevaluated. The number of combinations to be analyzed were reduced by eliminating combinations governed by others. Final governing equations for the spent fuel pool l structure are shown in Table 4-3 for concrete structures using strength design methods and for structural steel using plastic design methods. The dead load includes the weight of the spent fuel racks, stored fuel, spent fuel pool (SFP), and contributing weight of the adjacent floor slabs, roof, and walls. 4-10 Rev. 1 e t
The live load includes the roof snow load, distributed live loads on the adjacent floor slabs, crane loads, and a buoyant weight of a loaded spent fuel storage cask. Hydrostatic loads consist of the lateral water pressure exerted on the SFP walls and slab. Thermal loads are based on the pool water temperatures of 150 degrees F resulting from a full core discharge under normal operating conditions, and saturation temperatures for accident conditions varying from 252 degrees F at the bottom of the pool to 212 degrees F at the free water surface. In all cases, a conservative Reactor Building indoor ambient air temperature of 68 degrees F is used. A stress free temperature of 70 degrees F is used. 4.5 DESIGN AND ANALYSIS PROCEDURES 4.5.1 Design and Analysis Procedures for Spent Fuel Storage Racks The seismic and stress analysis of the spent fuel rack modules considers the various conditions of full, partially filled, and empty rack modules. The racks have been evaluated for both operating basis earthquake (OBE) and safe shutdown earthquake (SSE) conditions and meet Seismic Category I requirements. A detailed stress analysis has been performed to verify the acceptability of the critical load components and paths under normal and faulted conditions. The racks are freestanding and have been evaluated to show that under all loading conditions they do not impact each other nor do they impact the pool walls. The rack dynamic response of the fuel rack assembly during a seismic event is the condition which produces the governing loads and stresses on the structure. The seismic analysis of a free-standing fuel rack is a time-history analysis performed on a non-linear model using the dynamic capabilitiesoftheWestinghouseElectricComputerAnalysis(WECAN) Code (9,10). This is a general purpose finite element code with a great variety of elements which have static and dynamic capabilities. 4-11 Rev. 1 l
The time history analysis is performed on a 3-0 single cell non-linear model with the effective properties of an average cell within the rack module. The effective single cell structural properties are obtained from a 3-D structural model of the rack module, as shown in Figure 4-8. The details of the structural model and the seismic model are discussed in the following paragraphs. The structural model, shown in Figure 4-8, is a quarter section representation ; of the rack assembly consisting of beam elements interconnected at a finite number of nodal points. and general mass matrix elements. The beam elements model the beam action of the cells, the stiff. M ng effect of the cell to cell welds, and the supporting effect of the suppot pads. The general mass matrix elements represent the hydrodynamic mass of the rack module. The' beams which represent the cells are loaded with equivalent seismic loads and the model produces the structural displacements and internal load distributions necessary to calculate the effective structural properties of an average cell within the rack module. In addition to the stiffness properties, the internal load and stress distributions of this model are used to calculate stress peaking factors to account for the load gradients within the rack module. The nonlinear seismic model, shown in Figures 4-9 and 4-10, is composed of the effective properties from the structural model with additional elements to account for hydrodynamic mass of the fuel, the gap between the fuel and cell, and the support pad boundary conditions of a free standing rack. The elements of the nonlinear model are as follows: The fuel assembly is modeled by beam elements which represent the structural and dynamic properties of the fuel rod bundle and grid support assemblies. The cell assembly is represented by three-dimensional beam elements which have structural properties of an average cell within the rack structure. The water within the cell and the hydrodynamic mass of the fuel assembly are modeled by a general mass matrix element connected between the fuel and cell. 4-12 Rev. 1 l
The gaps between the fuel and cell are modeled by three-dimensional dynamic elements which are composed of a spring and damper in parallel, coupled in series to a concentric gap. The properties of the spring are the impact stiffness of the fuel assembly grid or nozzle and cell wall. The properties of the damper are the impact damping of the grid or nozzle. The properties of the concentric gap are the clearance per side between the fuel and cell. The hydrodynamic mass of a submerged fuel rack assembly is modeled by general mass matrix elements connected between the cell and pool wall. The sloshing movement of the pool water occurs above the top of the racks. Therefore, no sloshing loads are imposed on the rack structure. The support pads are modeled by a combination of three-dimensional dynamic friction elements connected by a " rigid" base beam arrangement which produces the spacing of support pads. The cell and fuel assemblies are located in the center of the base beam assembly and form a model which represents the rocking and sliding characteristics of a rack module in both directions on a plane. Vertical grounded springs at the support pad locations are used to model and account for the interaction between the racks and the spent fuel pool structure. The friction elements are capable of reversing the direction of the restraining force when sliding changes direction. This model is run for a range of friction coefficients (0.2 and 0.8) to obtain the maximum values. The results from these runs are fuel to cell impact loads, support pad loads, support pad liftoff, rack sliding. and fuel rack structure internal loads and moments. These values are searched through the full time in order to obtain the maximum values. The internal loads and stresses from the seismic model are adjusted by peaking factors from the structural model to account for the stress gradients through the rack module. In addition the results are used to determine the rack response for full, partially filled, and empty rack module loading conditions. 1 4-13 Rev. 1
( 4.5.2 Design and Analysis Procedures for Spent Fuel Pool Structure The spent fuel pool (SFP) structure was analyzed to determine the maximum allowable fuel rack loads which could be imposed on the pool slab. A comprehensive structural analysis was performed and results were evaluated in accordance with current codes, standards and regulatory requirements. 1 4.5.2.1 Computer Program and Finite Element Model The MSC/NASTRAN Version 62A (11) general purpose finite element program is used for this work. The three dimensional finite element model (FEM) includes the entire SFP structure plus adjacent key structural members to the extent where suitable boundary conditions can be assumed. See Figures 4-11 and 4-12 for the structural dimensions and the finite element model. i Figure 4-12 also shows the superelement configuration. Superelement 1 comprises the pool slab and the lower portion of the pool walls. Superelement l 2 comprises the drywell shield structure and the balance of the model is in Superelement 3. The residual structure, or Superelement 0, consists of only the grid points on boundaries between the other superelements. A computer plot of the FEM is shown in Figure 4-13. 4.5.2.2 Geometry, Boundary Conditions, and Element Types
? -
The ,modeled structurc~1nc'ludes the walls and slab of the SFP, the North and i West exterior walls of the Reagtor Building between E1. 165' and 234', drywell shield wall between El. 165' and 234', portions of the floor slabs at E1. 195'-3" and 234'-0" and interior wall between E1. 195'-3" and 234'-0". The steel superstructure of the Reactor Building and concrete floors and walls South of the drywell shield wall are not included in the FEM.
- Floor slabs and walls immediately adjacent to the SFP are modeled to simulate l the proper lateral restraint on the pool structure. Complete fixity against i
3 f 4-14 Rev. 1
translation and rotation is assumed at the base of the drywell shield wall. Cut-off boundaries of adjoining walls and slabs were restrained with translational springs. These springs permit the model to simulate the cantilever mode deflected shape of the Reactor Building under horizontal seismic loading. Translational springs simulate lateral stiffness of the remainder of the Reactor Building walls which were not included in the model. In-plane rotations of all interior grid points on slabs and walls are restrained. A summary breakdown of the model connectivity is given below. The overall model contains an estimated 11,000 independent degrees of freedom. MSC/NASTRAN FINITE ELEMENT MODEL DATA Element Type Number
- ELAS2 134 QUAD 4 1,836 BAR 192 TRIA3 479 RBAR, RTRPLT, RBEl, RBE2 234 Total = 2,875 i
Superelement No. of Grids No. of Elements 0 163 11 1 596 823 2 374 579 3 1.133 1.462 Total = 2,266 Total = 2,875 4-15 Rev. 1 _u-_-_-_-------
4.5.2.3 Analysis Method raoa1 ne al fo ed r o a en a a ng t ern i Concrete cracks under low tensile stresses at approximately 10 percent of its l ultimate compressive strength. In reinforced concrete structures, localized l concrete cracking reduces the overall structural stiffness. In structures such as this SFP, concrete stresses are usually tension with low compression. j Under these conditions concrete cracking is the major source of the nonlinearity as compared with nonlinear stress-strain relations and formation of plastic hinges. For this reason it is important to develop a logical procedure for establishing cracking patterns. It is also important to note that as long as equilibrium conditions are satisfied and the yield criterion is not exceeded anywhere, a safe lower bound solution will result regardless of the loading history used in the analysis. Therefore, the sequence of loading and the resulting crack pattern is not critical for strength evaluations. An iterative incremental approach is used to arrive at the final solutions. l Checking stresses against the cracking criterion and adjustment of element material properties are done manually at the end of each iteration. Each iteration is a linear elastic analysis, however, the cumulative effect is a nonlinear analysis that has taken into account load redistribution after concrete cracking due to mechanical loads and thermal loads. At the end of each iteration, the cumulative load which includes the latest load increment, is applied to the structure. 4.5.2.4 Section Properties After Cracking i In order to arrive at a reasonable final condition with regard to cracking patterns under mechanical loads, cracking criteria must be implemented. Initially, i.e., prior to any load application all elements are considered to be uncracked with linear elastic properties. After load increments are applied, the cracking criteria are checked manually and local element i 4-16 Rev 1 Ll________________-___-_
elastic and shear moduli are modified accordingly to represent the cracked condition. The cracking criteria are applied primarily to only those elements comprising the pool slab and the lower portions of the pool walls up to approximately 12 feet above the top of the pool slab. Cracking due to mechanical loads in the balance of the structure does not have significant effects on the results for the pool slab. l Extreme fiber tensile stresses in the element orthogonal local 1 and 2 directions are compared against the concrete modulus of rupture. Properties of elements having extreme fiber tensile stresses exceeding this value are
, modified. This is done by inputting the effective elastic modulus Et or E2 based on the cracked section moment of inertia.
The inplane shear modulus G12 is calculated using the new Eg and E2 values from the principles of orthotropic plate behavior. In addition, the reduced transverse shear moduli G g, and G22 are used in the cracked directions. 4.5.2.5 Slab Shear Transfer to Pool Walls A typical section through the intersection of the pool slab and wall is shown
; in Figure 4-14.
Transverse shear forces in the suspended slab are eventually developed by vertical axial tension forces in the pool walls. The critical section for slab shear and bending is taken at the face of the walls in accordance with ACI Code provisions. Similarly, the critical section in the wall is taken on l the horizontal plane at the top of slab elevation. Two additional critical sections at the wall / slab joint were evaluated in addition to the ones required by the ACI Code. These two sections are shown in Figure 4-15. Postulated concrete cracks along Critical Sections 1 and 2 shown in this figure represent the most severe conditions for load transfer through the wall / slab joint. I 4 17 Rev. 1 A , - _ - _ _
Calculations of concrete shear strength, Vc, and stirrup shear strength, Vs, are in compliance with Chapter 11 of ACI 349 and ACI 318. Concrete shear strengths are reduced to account for effects of axial tension forces. Also, increased concrete shear strength is gained from the presence of axial compressiveforcesusingEQ.(11-4)ofACI318andACI349. Shear capacities of the steel beams and connections are determined in accordance with Part 2 of theAISCspecifications(7)forplasticdesign. 4.5.2.6 Thermal Moment Relaxation Thermal stresses are introduced by internal and external restraints. In reinforced concrete structures, cracking at a section reduces its rigidity in addition to the total stiffness of the member which relieves thermal stresses. Ideally an analysis would consider effects of concrete cracking on internal force distributions throughout the loading history. One alternate approach however, in compliance with ACI 349 Appendix A (6), is to assume the structure is uncracked for mechanical loads and cracked only for thermal loads. This second approach was used in this safety investigation for local areas of the SFP structure away from the pool slab. The method used to establish cracking patterns in lower areas of the SFP structure and the pool slab is described in Subsection 4.5.2.3. 4.5.2.7 Alternate Supporting Calculations This section provides an estimate of the pool slab's ultimate load carrying capacity. The predicted slab strength is used to determine an overall margin of safety under certain loading conditions. 'In addition, a nonlinear finite element analysis of a simplified pool slab structure is made. Design loads used in these alternate calculations are listed in Table 4-4. These loads are equal to the loads used for the MSC/NASTRAN finite element analysis and they are used to estimate the maximum allowable loading on the SFP. 4-18 Rev. 1
The total capacity of the slab system to support vertical loads is controlled by shear which is comprised of the concrete shear capacity, stirrup capacity, and shear strength of the six W36 x 230 beams directly beneath the pool slao. Since thermal loads do not contribute to the gross shear forces around the perimeter of the slab, there are only two controlling loading combinations to consider for checking total shear capacity, as sumarized below.
SUMMARY
OF TOTAL LOADS VERSUS SLAB SHEAR CAPACITY Maximum Allowable Calculated b
& Load Combination Load (kip) $Vn(kip * )V 3 U = 1.40 + 1.4F + 1.7L + 1.9E 17,211 21,436 7 Y=1.7(0+F+L+E) 19,324 23,508 where, ( = ACI strength reduction factor Vn = Shear strength of concrete plus stirrups Vb = Shear capacity of six W36 X 230 beams Simplified finite element models with nonlinear solution techniques can be very effective in taking advantage of the inherent strengths of structures such as the SFP. Also, it is possible to account for internal arching action if elements are modeled through the thickness of the slab. It had not been possible to account for the beneficial effects of arching action in hand calculations or in the MSC/NASTRAN finite element analysis. The purpose of this nonlinear finite element analysis is to predict the collapse load as accurately as possible.
ThepoolslabwasmodeledusingtheADINA(12)finiteelementprogramforthe purpose of demonstrating an overall load carrying capacity in excess of code requirements. The slab is idealized as a system of twistless beam strips spanning across the pool in each direction at the slab centerlines. Each concrete strip is 4-19 Rev. 1 x
modeled using plane stress elements arranged in several layers through the thickness of the slab. Plane stress elements were chosen because of the nonlinear concrete material model available for this element. The material model describes the nonlinear stress-strain relations, tensile failure, I compression crushing and post failure behavior of concrete. Reinforcement was modeled using nonlinear truss elements. The W36 X 230 beam is idealized using l plane stress elements for the web and nonlinear truss elements for the flanges. l Boundary conditions are assumed fixed at the supports and a uniformly distributed load is applied incrementally to the top of the slab. No , reinforcement yielding and very little concrete cracking were predicted at the allowable design load. 1 The analysis was halted when the applied load approached three times the factored design load. Final results indicated some concrete cracking at supports and at midspan and top bar yielding at the supports would occur, but collapse was not yet imminent. The ADINA model is used here as an alternate analysis to lend support and to increase the level of confidence in results of I the more comprehensive analyses using the MSC/NASTRAN finite element model. 4.6 STRUCTURAL ACCEPTANCE CRITERIA 4.6.1 Structural Acceptance Criteria for Spent Fuel Storage Racks The fuel racks are analyzed for the normal and faulted load combinations of Section 4.4 in accordance with the NRC "0T Position for Review and Acceptance ofSpentFuelStorageandHandlingApplications."(3) The major normal and upset condition loads are produced by the operating basis i earthquakes (OBE). The thermal stresses due to rack relative expansion are calculated and combined with the appropriate seismic loads in accordance with the NRC, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," (with clarifications as noted in Table 4.2). l l 4-20 Rev. I l__ ____ __-_ _
1 I The faulted condition loads are produced by the safe shutdown earthquakes (SSE) and a postulated fuel assembly drop accident. The computed rack structure stresses are within the acceptance limits f identified in the NRC, "0T Position for Review and Acceptance of Spent Fuel i Storage and Handling Applications " (with clarifications as noted in l Table 4.2). Table 4-5 provides the calculated rack stress margin of safety i (M.S.). Margin of safety is defined as the follows: [ M.S. = Allowable stress 'I Design stress In sumary, the results of the seismic and structural analysis show that the l PBAPS spent fuel storage racks meet all the structural acceptance criteria [ adequately. r 4.6.2 Fuel Handlina Crane Uplift Analysis The objective of this analysis is to ensure that the rack can withstand the ! maximum uplift load of 4,000 pounds and a horizontal force of 1,100 pounds of the fuel handling crane without violating the criticality acceptance criterion. The maximum uplift load is approximately two times the capacity of the fuel handling crane. In this analysis the loads are assumed to be applied to a fuel cell. Resulting stresses are within acceptable stress limits, and there is no change in rack geometry of a magnitude which causes the ' criticality acceptance criterion to be violated. 4.6.3 Fuel Assembly Drop Accident Analysis The objectives of this analysis are to ensure that, in the unlikely event of ; dropping a fuel assembly on a storage rack, accidental deformation of the rack : will not cause the criticality acceptance criterion to be violated, ard the i spent fuel pool liner will not be perforated. 1 ij t e 4 21 Rev. 1 t A
Three accident conditions are postulated. The first accident condition assumes that the weight of a fuel assembly and handling tool impacts the top end fitting of a stored fuel assembly or the top of a storage cell from a conservative drop height of 2 feet in a straight attitude. The second accident condition is similar to the first except the impacting mass is at an inclined attitude. The impact energy is absorbed by the dropped fuel assembly, the stored fuel assembly, the cells and the rack base plate assembly. Under these faulted conditions the criticality acceptance criterion is not violated and the pool liner is not perforated. L The third accident condition assumes that the dropped assembly falls straight through any empty cell and impacts the rack base plate from a conservative drop height of 2 feet above the top of the rack. The results of this analysis show that the impact energy is absorbed by the fuel assembly and the rack base plate. The spent fuel pool liner is not perforated. Criticality calculations show that Keff <0.95 and the criticality acceptance criterion is not violated. In each of these accident conditions, the criticality acceptance criterion is not violated and the spent fuel pool liner is not perforated. 4.6.4 Fuel Rack Sliding and Overturnino Analysis Consistent with the criteria of "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," the racks are evaluated for overturning and sliding displacement due to earthquake conditions under the various conditions of full, partly filled, and empty fuel rack loadings. The nonlinear model described in paragraph 4.5 is used in this evaluation to account for fuel-to-rack impact loading, hydrodynamic forces and the nonlinearity of sliding friction interfaces. The horizontal resistive force at the interface between the rack module and pool floor is produced by friction. A low coefficient of friction (u a 0.2) produces maximum rack sliding while a high value (u a 0.8) produces maximum rack horizontal overturning force, l t 4 22 Rev. 1 d
The fuel rack nonlinear time history analysis shows that the maximum fuel rack sliding during a SSE event is .0098 inches. This distance combined with structural and maximum thermal displacements is less than the rack-to-rack, rack-to-floor obstruction, or rack-to-aall clearances; thus, impact between adjacent rack modules, rack module and floor obstructions, and rack module and pool wall is prevented. l The energy which produces lift-off was applied to the fuel rack module for various fuel loading conditions in order to obtain the loading which resulted in the minimum factor of safety against overturning. The 9 x 14 rack in the 9 cell direction was analyzed since it has the minimua resistance to overturn. The minimum factor of safety against overturning is calculated to be /3 which is much greater than that permitted by Section 3.8.5.!!.5 of the Standard Review Plan. 4.6.5 Fuel Bundle / Module Impact Evaluation The nonlinear seismic model includes the gap between the fuel assembly and storage ceII. and provides the fuel to cell impact load and the associated fuel storage rack response during the seismic event. The fuel assembly and fuel storage cell impact forces obtained from the nonlinear analysis were used to evaluate the effects on the fuel rack structure and fuel assembly structure. These loads are within the allowable limits of the fuel rack module materials and fuel assembly materials. Therefore, there is no damage to the fuel assembly or fuel rack module due to impact loads. 4.6.6 Structural Acceptance Criteria for Spent Fuel Pool Structure The available section strengths for reinforced concrete elements are calculated by the strength design method in accordance with ACI 318 and ACI 349(5,6). Axial force / moment and axial force / shear interaction diagrams are generated for the entire SFP structure. These interaction diagrams were then used to manually check each critical section. The axial force / shear interaction diagram for the SFP floor includes the transverse shear strength of the steel beams. The available section strengths for structural steel 4 23 Rev. 1 A
members for axial loads plus bending are determined by plastic design methods in accordance with AISC (7). The section strengths required to carry the increased loading are based on results from the MSC/NASTRAN finite element analyses. Required strengths in terms of shear forces and bending moments are determined for each element in l the SFP structure and for each of the governing load combinations. l 4.6.6.1 Pool Slab and Walls The SFP load carrying capacity is limited by transverse shear capacity of the pool slab for Load Combination Nos. 3 and 7. Large reserve flexural capacities exist in the pool walls and slab for all load combinations. A reduced transverse shear capacity was used in the pool slab to reflect the small amount of membrane tension generated by the lateral fluid pressures on the pool walls. This shear capacity was compared againse peak transverse shear forces from the MSC/NASTRAN finite element analysis results and is adequate. The load transfer capacity of the wall / slab joints on the East and West sides of the pool were evaluated and found to be adequate. 4.6.6.'2 Structural Steel Members Simple beam moments due to factored slab dead load are manually superimposed on the FEM results. The resulting total bending moments were used to evaluate the beams for combined axial load and bending and all beams were found to be adequate. 4.6.6.3 Adjacent Reinforced Concrete Structural Members Adjacent structural members evaluated in this safety investigation include a portion of the Reactor Building exterior wall and a section of the drywell shield wall. The increase in the spent fuel pool floor design loads result in a small percentage increase of the design loads for structures adjacent to the 4 24 Rev. 1 O
pool as the loads are distributed throughout the Reactor Building. The changes in stress conditions of adjacent structures due to increased spent fuel storage capacity have been calculated to be minimal. The shear stresses for the original design of the drywell shield wall are contained in FSAR Table C.4.5 and are sumarized in Table 4-6. Original design concrete shear stresses at EL. 180'-0" are 89 percent of the allowable stress for dead load plus operating basis earthquake (OBE) and 69 percent of the allowable stress for dead load plus thermal load plus safe shutdown earthquake (SSE). All other reported stresses are less than 50 percent of the corresponding allowable stresses. Additional shear stresses due to increased spent fuel storage capacity are calculated to be 0.0020 kip /in2 and 0.0032 kip /in2 at EL. 180'-0" for OBE and SSE respectively. These shear stress increments are based on the MSC/NASTRAN finite element analysis results. These increments represent increases in total shear stresses from 89 percent to 92 percent of the allowable for OBE and from 69 percent to 70 percent for SSE. The resulting total concrete shear stresses are less than the allowable shear stresses. Local areas of the North exterior wall of the Reactor Building were also evaluated due to the increased loads. The areas checked are the support points of the East and West walls of SFP. These areas are adequate for combined axial load and bending. Shear forces are also less than the shear capacity. 4.6.7 Spent Fuel Pool Serviceability Checks 4.6.7.1 Deflections Slab centerline deflections resulting from the MSC/NASTRAN analysis are used as the basis for checking code requirements on allowable deflections. Table 9.5(a) of ACI 349 indicates deflections that could be expected in twelear Safety Related concrete structures but which are not necessarily maximum e l 4 25 Dev. 1 i
allowable deflections. Maximum allowable deflections are normally specified in the plant design criteria. However, since specific deflection criteria are not available for this SFP structure, the ACI criteria are used as the basis for determining maximum allowable deflections. The ACI criteria apply to unfactored loads. Effects of concrete creep on deflections were considered and the calculated long term deflections are less than the allowable values. 4.6.7.2 Concrete Crack Widths Crack widths were calculated at the midspan and at the support edges of the SFP floor for unfactored loading conditions. An allowable crack width of 0.016inchforoperatingconditionsisrecommendedbyACICommittee224(13). Calculated crack widths are less than the allowable value. 4.6.8 Pool Slab / Fuel Rack Interface loads The actual total pool floor loads from the racks are calculated from the results of the rack nonlinear seismic model. The distribution of the loads on the individual support pads is calculated from the overall load from the seismic model times the peaking factor for the individual pad as determined from the structural model. The actual fuel rack / pool floor interface loads are within the maximum allowable limits specified by the spent fuel pool structural analysis. Maximum allowable loads for the SFP floor as calculated by the SFP structural analysis are given in Table 4-7. Allowable local bearing pressures are calculated based on concrete code allowables defined in Section B.3.1 of Appendix B " Alternate Design Method" of ACI 318 for all loading combinations, except No. 7. A one-third increase in allowable bearing stress ispermittedforseismicandaccidentloadingconditions,(i.e., load combinationsNo.5andNo.6). Allowable bearing stress for load combination No. 7, which includes an accidental fuel drop condition, is determined in accordance with the Strength Design provisions of ACI 318 and ACI 349. , 4-26 Rev. 1 e i _ . _ _ _ . _ _ _ _ . . _ - . ~,__m_,.___ _____ -- .-
i ! ) l Table 4-8 summarizes pool floor loads calculated as a result of the rack l j nonlinear seismic analysis, and the margin of safety for various load !
< combinations. l I 4.6.9 Evaluation of Scent Fuel Pool Slab for Fuel Handling Accidant I l
t Impact loads resulting from a dropped fuel assembly do not control the design l of the pool slab. Also, concrete scabbing off the bottom of the slab will not ll occur and concrete penetration is limited to the topping concrete above the j structural slab.
! l The maximum leakage rate due to a postulated perforation of the liner is I one-fifth of the available makeup water capacity. There is no danger of l
) draining the spent fuel pool or uncovering the spent fuel. l 4.6.10 Scent Fuel Pool Liner Plate Evaluation f ) Typical wall and floor liner plates and anchorage details were evaluated. , j Effects of increased loading on the pool slab due to reracking were included i in the liner evaluation in addition to normal and accident thermal effects. _ l [ The liner function is to retain the SFP fluid. The liner is not required to ! j provide support for any member or loading, and therefore is not a structural - i element. Acceptability criteria for the liner is the prevention of tearing or ! leakage. Consequently, the liner was evaluated for liner strain only. l ! Adequate margins of safety exist on the liner plates and anchorages to ensure j i the functionability of the liner system. !
- i
! l i, l t
i 4 i i i j 4-27 Rev. 1 l I
- e {
o __ 9
4.7 MATERIALS, QUALITY CONTROL, AND SPECfAL CONSTRUCTf0N TECHN! QUES i 4.7.1 Construction Materials Construction materials conform to the requirements of ASME Boiler and Pressure Vessel Code, Section !!!, Subsection NF. All the materials used in the rack construction are compatible with the storage pool environment and do not contaminate the fuel assemblies or the pool water. The racks are constructed i from type 304LN stainless steel except the leveling screws which are type 17-4 PH stainless steel. Plates placed on the spent fuel pool floor to avoid rack ; support pad interferences are Type 304 stainless steel. 4.7.2 Neutron Absorbing Material The neutron material, Boraflex, used in the Peach Bottom spent fuel rack construction is manufactured by Grand Industrial Services, Inc., and I fabricated to safety related nuclear criteria of 10CFR50, Appendix 8. ' Boraflex is a silicone based polymer containing fine particles of boron carbide in a homogeneous, stable matrix. Boraflex contains a minimum 810 areal density of 0.021 gm/cm2, Boraflex has undergone extensive testing to study the effects of gamma and neutron irradiation in various environments, and to verify its structural integrityandsuitabilityasaneutronabsorbingmaterial.(14)Testswere 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 attentuation capabilities before and after being subjected to an environment of borated water and 1.02 x 1011radsgammaradiation.(15) l i 4-28 Rev. 1 !
l Long term borated water soak tests at high temperatures were also l conducted.(16) It was shown that Boraflex withstands a borated water l imersion of 2400F for 260 days without visible distortion or sof tening. Boraflex maintains its functional performance characteristics and sho.s no l evidence of swelling or loss of ability to maintain a uniform distribution of l boron carbide. During irradiation, a certain amount of gas may be generated. A conservative evaluation of the effect of gas generation on the spent fuel pool building atmosphere indicates that the maximum gas generation would be less than 0.01 percent of the total room volume. Additionally, the majority of gas generation is nitrogen. oxygen and CO2 - The testing performed to date verifies that Boraflex maintains a long-term material stability and mechanical integrity, and can be safely utilized as a poison material for neutron absorption in spent fuel storage racks. 4.7.3 Quality Assurance Proaram The PBAPS spent fuel racks are "Q-Listed" items. PECo's Quality Assurance Program is applicable to the specification, procurement, design, fabrication, inspection, handling, and installation of the new storage racks. l The rack design, analysis, material procurement, and fabrication were performed under the Westinghouse Nuclear Components Division's Ouality Assurance Program. This program conforms to the requirements of the American SocietyofMechanicalEngineers(ASME) Boiler &PressureVesselCode, Section!!!&VI!!(SubsectionNCA-4000),10CFR50,AppendixB,ANS!N45.2, MIL-!-45208, MIL-Q-9858A, and RDT F2-2. 4-29 Rev. 1
T The Westinghouse Nuclear Components Division holds the following ASME Boiler & Pressure Vessel Code certificates with the scope authorizations as indicated below: Certificates Scope Authorizations N Class 1, 2, 3 vessels, pumps, piping systemst Class 2, 3 storage tanks; and Class CS core support structures l NPT Class 1, 2, 3 components parts and appurtenances, tubular parts welded with filler metalt and Class CS core support parts. U Pressure vessels - The preparation of the specification for rack design, and the analysis of the spent fuel pool structure was performed under the Gilbert / Commonwealth, Inc., (G/C)NuclearQualityAssuranceProgram. This program, implemented through l l the G/C " Nuclear Quality Assurance Manual" is described in the G/C Corporate Topical Quality Assurance Report GAI-TR-106, Rev. 3. " Quality Assurance Program for Nuclear Power Plants." This program has been approved by the NRC and is in compliance with the requirements of 10CFR50 Appendix B. USNRC Reg. Guide 1.28, ANS! N45.2, and USNRC Reg. Guide 1.64. 4.7.4 Construction Techniques 4.7.4.1 Administrative Controls Buring Manufacturing and Installation The Peach Bottom Units 2 and 3 new spent fuel storage racks will be manufactured at the Westinghouse Nuclear Components Division, Pensacola, Florida. This facility is a modern high-quality shop with extensive experience in forming, machining, welding, and assembling nuclear grade equipment. Forming and welding equipment are specifically designed for fuel rack fabrication and all welders are qualified in accordance with A$ME Code Section IX. l 4 30 Rev. 1
To avoid damage to the stored spent fuel during rack replacement, all work on the racks in the spent fuel pool area will be performed by written procedures. These procedures preclude the movement of the fuel racks over the stored spent fuel assemblies. Radiation exposures during the removal of the old racks from the pool will be controlled by written procedures. Water levels will be maintained to afford adequate shielding from the direct radiation of the spent fuel. Prior to rack replacement, the cleanup system will be operated to reduce the activity of the pool water to as low a level as can be practically achieved. 4.7.4.2 Procedure The following sequence of events is anticipated for the spent fuel storage rack replacement for PBAPS Units 2 and 3:
- a. Design and fabricate new spent fuel storage racks.
- b. Prepare modification procedure.
- c. Fabricate and test all special tooling.
- d. Receive and inspect new spent fuel storage racks.
The final configuration of the new rack modules in the spent fuel pool is shown in Figure 4 4 and 4-5. The installation of these racks will be accomplished in accordance with the following considerations and guid61tness o The removal of old racks and installation of new racks will be performed using plant approved procedures. o At no time will a rack module be carried directly over another module installed in the spent fuel pool, containing spent fuel. 4 11 Rev. 1 l _ _ _ . _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . .
o All load handling operations for the new high density fuel storage racks in the spent fuel pool area will be conducted in accordance with the criteria af Section 5.1.1 of NUREG-0612
" Control of Heavy Loads at Nuclear Power Plants" o Spent fuel relocations within the pool will be performed as required to maintain separation between the stored fuel and the rerack operations.
4.8 TESTING AND IN-SERVICE SURVE!LLANCE 4.8.1 initial Verification The neutron absorber rack design includes a neutron absorber verification view hole in the cell wall 50 that the presence of neutron absorber material may be visually confirmed at any time over the life of the racks. Upon completion of rack fabrication, such an inspection is performed. This visual inspection, coupled with the Westinghouse quality assurance program controls and the use of 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 additional on-site neutron absorber verification beyond normal receipt inspection activities. 4.8.2 Periodic Verification Surveillance The neutron absorber coupons used in the surveillance program will be representative of the material used. They will be of the same composition, produced by the same method, and certified to the same criteria as the production lot. The sample coupons will be of a similar thickness as the neutron absorption material used within the storage system. Each specimen will be encased in a stainless steel jacket of an identical alloy to that used in the storage system, formed so as to encase the neutron absorption material and fix it in a position similar to that c Tigned into the storage system. The jacket will be mechanically closed without welding in such a maance as to 4 12 Dev. 1 0
l retain its form througheut the use period yet allow rapid and easy opening without contributing mechanical damage to the specimen contained within. A series of not less than 24 of the jacketed specimens shall be suspended from rigid straps so designed as to be hung on the outside periphery of a rack ' module. There are 2 sets of these straps. The specimens will be located in the spent fuel pool such that they will receive a representative exposure of gama radiation. The specimen location will be adjacent to a designated storage cell with design ability to a110w for removal of the strap, providing access to a particular specimen. As discussed in Section 4.7.2, irradiation tests have been previously performed to test the stability and structural integrity of Boraflex in boric acid solution ender irradiation. These tests have concluded that there is no evidence of deterioration of the suitability of the Boraflex poison material through a cumulative irradiation in excess of 1.03 x 1011 rads gamma i radiation. As more data on the service life performance of Boraflex becomes available in the nuclear industry in the coming years through both experimentation and operating experience, PECo will evaluate this information l and will modify the surveillance program accordingly. PEco plans to perform an initial surveillance of the specimens after approximately five years of exposure in the pool environment. During this survet11ance, several specimens will be removed from the pool and examined. This examination is expected to include visual inspection as well as other tests determined necessary to verify that the performance of the Boraflex is consistent with the reported test results. Based on the results of this initial surveillance. PECo will determine the scheduling and extent of additional surveillances 50 as to assure acceptable material performance throughout the life of the plant, i t 4 33 Rev. !
4.9 REFERENCES
- 1. Philadelphia Electric Company; " Peach Bottom Atomic Poacr Station Units 2 and 3. Final Safety Analysis Report (FSAR)," 1972 and as amended by UFSAR, Docket No. 50-277 and 50-278.
- 2. U. S. Nuclear Regulatory Commission, " Standard Review Plan 3.8.4,
'Other Seismic Category ! Structures'," Rev. 1, NUREG-0800, July 1981.
- 3. U. S. Nuclear Regulatory Commission, letter from B. R. Grimes to All Power Reactor Licensees, 4/14/78, with enclosure entitled, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," including Supplement dated 1/18/79.
- 4. U. S. Nuclear Regulatory Commission, " Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments)," Regulatory Guide 1.142 Rev. 1, October 1981.
- 5. ACI Comittee 318. " Building Code Requirements for Reinforced Concrete (ACI318-83),"AmericanConcreteInstitute, Detroit,Mich.,
1983. Also, ACI Committee 318. " Commentary on Building Code Requirements forReinforcedConcrete(AC!318-83),"AmericanConcretoInstitute, , Detroit, Mich., 1983. )
- 6. ACI Committee 349, " Code Requirements for Pudtar Safe 5y Related Concrete Structures Detroit,Mich.,1980(.ACI 349-80)," American Comritt institute, ,
- 7. I A!SC, " Specification for the Destgr. Fr.brication, and Erection of Structural Steel for Buildings," American 'rstitute of Steel Construction, New York, N.Y., 1978. '
(
- 8. A!SC, " Specification for the N si$n, Fabrication and Erection of Steel Safety Related Structures for Nucletr Facilities," American Institute of Steel Construction, New York,ll.Y., Final Draft of '
initial issue, 9/14/83. (
- 9. WECAN, " Documentation of Selected Westinghouse Strwtural Analysis Computer Codes," WCAP-8252.
- 10. WECAN, " Benchmark Problem So.utier t'raloyed for Verifitation of the WECAN Computer Program," WCAP 0929.
- 11. C.W.McCormick(Ed.),MSC/WTHN r's Kinual, MacNeal-SchwendlerCorp.,LosAngobj!,p,C4 j ts, if.,I6y 1981, 2 vol.
1 4-14 pev. t
-- . - u
- 12. ADINA. A Finite Element Program for Automatic Oynamic Incremental Nonlinear Analysis. ADINA Engineering, Inc., Watertown, Mass.,
Report AE 81-1 September 1981.
- 13. ACI Committee 224. " Control of Cracking in Concrete Structures,"
American Concrete Institute, Detroit, Mich., Report No. ACI 224R-80, October 1980. i
- 14. J. S. Anderson, "Boraflex Neutron Shielding Material -- Product Performance Data," Brand Industries, Inc., Report 748-30-2 (August 1981).
- 15. J. S. Anderson, " Irradiation Study of Boraflex Neutron Shielding Materials," Brand Industries, Inc., Report 748-10-1 (August 1981).
- 16. J. S. Anderson, "A Final Report on the Effects of High Temperature
( Borated Water Exposure on BISCO Boraflex Neutron Absorbing
) Materia's," Brand Industries, Inc., Report 748-21-1 (August 1978).
f
+
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ls b ( l
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j 4-35 pey, 1 l
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+
iti < TABLE 4-1 l is 1 . RACK MODULE DATA (PER UNIT) l
.w Storage' Rack Assy. Dry Weight (1bs)
,) h Array Locations Dimensions (inches) Per Rack Assy.
1 9 x 14 126 54 x 89 x 180 10,000 s 2 10 5 14 , 280 64 x 89 x 180 11,200
, Ys 1 11 x 14 Mod. 119 70 x 89 x'180 9,500 s ,s 1 12 x 15 180 76 x 95 x 180 14,400
, 1 12 x 17 204 76xlb7x180 16,300
+
2 12 x 20 480 76 x 126 x 180
- 19,200
'~ 2 15 x 19 570 95 x 120 x 180 ' 22,800
. ' 1 17 x 20 340 107 x 126 x 180 27,200
_ 4_ 19 x 20 1,520 120 x 126 x 180 30,400 15 racks 3,819 s Il l: s 4' L
;, Storage locations center-to-center spacing (inches) 6.28 Storage cell inner dimension (inches), 6.07 Intermediate storage location inner dimension (inches) 6.12 Type of Fuel BWR 8 x 8 BWR 8 x 8 (R)
BWR 7 x 7 \ \ , a ' e I i-
,e '*'
\
l' . k a f s t 4-36 Rev. 1 M,. ' O. .
TABLE 4-2 s. STORAGE RACK LOADS AND LOAD COMBINATIONS Load Combination Acceptance Limit D+L Normal limits of NF 3231.la 0+L+Pf Normal limits of NF 3231.la D+L+E Normal limits of NF 3231.la 0+L+To Lesser of 2Sy or Su stress range D + L + To + E Lesser of 2Sy or Su stress range D+L+Ta+E Lesser of 2Sy or Su stress range D + L + To + Pr Lesser of 2Sy or Su stress range D + L + Ta + E' Faulted condition limits of NF 3231.lc (see Note 3) D+L+Fd The functional capability of the fuel racks shall be demonstrated Notes: h
- 1. The abbreviations in the table above are those used in Standard Review Plan (SRP) Section 3.8.4 where each term is defined except r
' for Ta, which is defined here as the highest temperature associated with the postulated abnormal design conditions. Fd is the force caused by the accidental drop of the heaviest load from the maximum possible height, and Pr is the upward force on the racks caused by a postulated stuck fuel assembly. l 2. The provisions of NF-3231.1 of ASME Section III, Division I, shall l be amended by the requirements of Paragraphs c.2, 3 and 4 of i Regulatory Guide 1.124, entitled " Design Limits and Load f Combinations For Class A Linear-Type Component Supports." I 3. l 5 For the faulted load combination, thermal loads were neglected when they are secondary and self-limiting in nature and the material is ductile. r-s O 4-37 Rev. 1
,f , I .
1 TABLE 4-3 SPENT FUEL P0OL GOVERNING DESIGN LOAD COMBINATIONS ) Reinforced Concrete
- 1. U = 1.40 + 1.4F + 1.7To
- 2. U = 1.40 + 1.4F
- 3. U = 1.40 + 1.4F + 1.7L + 1.9E 4.
U = D + F + L + E' + Ta
- 5. U=0+F+L+E' 6.
U = 1.05D + 1.05F + 1.3L + 1.43E + 1.3To Structural Steel
- 7. Y = 1.70 + 1.7F + 1.7L + 1.7E
> 8. Y = 1.30 + 1.3F + 1.3L + 1.3E + 1.3To 9. Y = 1.1 (D + F + L + E' + Ta) Notation: D = dead load , E = OBE (design earthquake) E'= SSE (maximum credible earthquake) L = live load Ta= thermal load produced by accident condition To= thermal load during normal operation U = section strength required to design loads based on the Strength Design method for reinforced concrete Y = section strength required to resist design loads based on Plastic Design method for structural steel s 4-38 Rev. 1
TABLE 4-4 SLAB DESIGN LOAD
SUMMARY
MECHANICAL LOADS TOTALS (KIP) Dead 1,475.5 Live 252.8 Hydrostatic 3,373.0 Fuel Rack Buoyant Weight 3,823.1 Vertical OBE 2,442.4 Vertical SSE 5,001.7 THERMAL LOADS TEMPERATURE (OF) Accident, mean 98.5 Accident, gradient 140.2 (*70.1) , Normal, mean 113.0 Normal, gradient 63.2 ( 31.6) Notes:
- 1. Dead load includes weight of slab and six W36 X 230 beams.
- 2. Live load includes weight of a buoyant loaded cask plus miscellaneous floor loads.
- 3. Seismic loads include structural, hydrodynamic, and fuel rack contributions.
- 4. Fuel rack buoyant weights and fuel rack seismic loads used in the SFP analyses represent the maximum alloweble loads.
4-39 Rev. 1
I TABLE 4-5
SUMMARY
OF DESIGN STRESSES AND MINIMUM MARGINS OF SAFETY Normal & Upset Conditions Design Allowable Margin ) Stress Stress of (psi) (psi) Safety 3
- 1. Support Pad Assembi,y 1.1 Support Pad Shear 1595 11000 5.90 Axial and Bending 10479 16500 Bearing .57 13645 27500* 1.02 1.2 Support Pad 3 crew Shear 7958 11000 .38 1.3 Support Structure Axial and Bending 17626 27500*
Shear .56 1233 11000 7.92 Weld Shear 19072 275000* .44 2.0 Cell Assembly 2.1 Cell Axial and Bending .816 1.0** .23 2.2 Cell to Bnse Plate Weld Weld Shear- 19082 24000 .26 2.3 Cell to Cell Weld Weld Shear 16286 21000 Pin Shear .29 7384 9260 .25 2.4 Cell to lirapper Weld ) Weld Shear 8300 11000 i 2.5 Cell Seam Weld .33 l Weld Shtar 3501 4516*** i
.29 l 2.6 Cell to Cover Plate Welds Weld Shaar 11854 24000 1.03
** Thermal Plus OBE Stress is Limiting
*** Allowable per Appendix XVII -2215 Eq (24)
Design LoaJ and Allowable Load in Lbs is shown 4-40 Rev. 1 (
TABLE 4-6 DRYWELL SHIELD WALL ORIGINAL DESIGN
SUMMARY
1 ALLOWABLE 2 MAXIMUM PERCENTAGE LOAD STRESS STRESS SECTION OF ALLOWABLE COMBINATION (XSI) (KSI) USED f E1. 145'-0" D+E 24.0 8.2 34 Circumferential tension reinforcement D + T + E' 54.0 23.5 < 44 E1. 145'-0" D+E 1.80 0.45 25 Concrete vertical compression D + T + E' 3.40 1.57 46 E1. 180'-0" D+E 0.07 0.0625 89 Concrete shear D + T + E' O.253 0.174 69 Notes:
- 1. Based on FSAR Table C.4.5.
- 2. Based on ACI 318-63 Working Stress Design Methods.
4-41 Rev. 1 (
.a
F TABLE 4-7 MAXIMUM ALLOWABLE FUEL RACK / POOL FLOOR INTERFACE LOADS TOTAL LOADS VERTICAL HORIZONTAL NO. LOCAL BEARING LOAD COMBINATION (KIP) (KIP) (KSI) 7 1. D+L 3,900.01 N/A 2.4
- 2. D+L+T o 3,900.01 N/A 2.4
- 3. D + L + To + E 5,700.0 1,900.0 2.4
- 4. D + L + Ta + E 5,700.0 1,900.0 2.4
- 5. D+L+To+Pf 5,700.0 N/A 3.2
- 6. D + L + Ta + E' 8,000.0 3,000.0 3.2
,, 7. D+L+Fd 8,000.0 N/A 4.76
. Alternatel
- 8. 1.4 (D + L + To)
+ 1.9E 8,900.0 3,600.0 See Note 2
- 9. 1.4 (D + L + Ta)
+ 1.9E 8,900.0 3,600.0 See Note 2 10, 1.7 (D + L + To
+ E) 9,700,0 3,200.0 See Note 2
- 11. 1.7 (D + L + Ta
+ E) 9,700.0 3,200.0 See Note 2 Notes:
1. Additional structural limits specified in Load Combination No. 8, 9, 10, and 11 shall be satisfied if total vertical loads calculated for Load Combination No. 1 and 2 are less than 3,700.0 kip. Otherwise, Load Combination No. 8, 9, 10, and 11 may be used in l'eu of Load Combination No. 1, 2, 3, 4, and 5.
- 2. When total loads are evaluated using Load Combination No. 8, 9,10, and 11, local bearing pressures shall satisfy Load Combination No. 1, 2, 3, 4, and 5.
3. Notations used in this table are the same as defined in SRP 3.8.4, Appendix D. 4-42 Rev. 1 5
TABLE 4-8 Pool Floor Loads Design Allowable Margin Stress Stress of Load Combination Condition
- or Load or Load Safety
- 1. D+L Local Bearing 1.76 2.4 .36
- 2. D + L + To Local Bearing 1.76 2.4 .36
- 3. D + L + To + E Local Bearing 1.94 2.4 .24
- 4. D + L + Ta + E Local Bearing 1.94 2.4 .24
- 5. D + L + To + Pf Local Bearing 1.76 3.2 .82
- 6. D + L + Ta + E' Vertical 6180 8000 .29
) Horizontal 1670 3000 .80 Local Bearing 2.63 3.2 .22
- 7. D + L + Fd Vertical 4130 8000 .94 Local Bearing 4.39 4.76 .08
[ 8. 1.4(D + L + To) + 1.9E Vertical 7730 8900 .15 Horizontal 1590 3600 1.26
- 9. 1.4(D + L + Ta) + 1.9E Vertical 7730 8900 .15
> Horizontal 1590 3600 1.26
- 10. 1.7(D + L + To + E) Vertical 8760 9700 .11 Horizontal 1420 3200 1.25
- 11. 1.7(D + L + Ta + E) Vertical 8760 9700 .11 l
) Horizontal 1420 3200 1.25
- Vertical refers to total pool floor vertical load in kips. Horizontal refers to total pool floor horizontal load in kips. Local bearing refers to pool floor bearing stress under the highest loaded support pad in ksi.
4-43 Rev. 1
I - OI - N I il (NEW FUEL STORAGE [ STEAM. SEPARATOR r yh-- 4 oevea sTo a^ se) .< O 3 3,w / j REA TOR , hlp l t i.A M l l 3.1 7I A L_j
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5.0 COST / BENEFIT ASSESSMENT 5.1 COST / BENEFIT ASSESSMENT The cost / benefit of the chosen reracking alteration is demonstrated in the following sections. 5.1.1 Need for Increased Storage Capacity
- a. PECo currently has no contractual arrangements with any fuel reprocessing facilities.
- b. At PBAPS, both the Unit 2 and Unit 3 spent fuel pools have been previously reracked to 2,608 cells each.
Table 5-1 includes proposed refueling schedules for both Unit 2 and Unit 3,.and the expected number of fuel assemblies that will be transferred into the spent fuel pools at each refueling until full core discharge capability is lost.
- c. As of May 1985, the Unit 2 spent fuel pool contained 1464 spent fuel assemblies. The Unit 3 spent fuel pool contained 1,497 spent fuel assemblies and one new assembly.
- d. Adoption of this proposed spent fuel storage expansion would not necessarily extend the time period that spent fuel assemblies would be stored on site. Spent fuel could be sent off site for final disposition under existing legislation. The government facility is expected to become available in 1998. As matters now stand and until alternate storage facilities are available, spent fuel assemblies on site will remain there.
- e. Table 5-2 references the spent fuel storage capacity for both the PBAPS Unit 2 and 3 spent fuel pools after reracking. Based on the present PECo fuel management policy, the Unit 2 spent fuel pool will lose Full Core 5-1 Rev. 1 i
Discharge Reserve (FCDR) in 1987. The Unit 3 pool will lose FCDR in 1988. In order to maintain full core discharge reserve capability during the new rack installation, the new racks require installation prior to the September 1986 refueling outage. 5.1.2 Estimated Costs The costs associated with the Units 2 and 3 proposed spent fuel pool modification are estimated to be in the neighborhood of 9.1 million dollars. This figure includes items such as; 1) extensive engineering studies of spent fuel disposal alternatives, 2) design, engineering, manufacture, and installation of new spent fuel storage racks, 3) removal of offsite disposal of the existing spent fuel storage racks, 4) project management and licensing, and 5) allowance for funds used during construction. Estimated value of uncertainty and cost escalation are not included in this sum. 5.1.3 Consideration of Alternatives
- a. There are no operational commercial reprocessing facilities available for PECO's needs, nor are there expected to be any in the foreseeable future,
- b. At the present time, there are no existing available independent spent fuel storage facilities. There are no firm commitments by either commercial firms or government agencies to construct or operate an independent spent fuel storage facility. In addition, cost and/or schedule considerations make an independent spent fuel storage facility on site unacceptable to meet the spent fuel storage needs at PBAPS Units 2 and 3.
- c. At present PECo has no license to transship fuel between the Limerick and the PBAPS sites, nor are presently installed storage racks at Limerick licensed to store fuel generated at PBAPS. There are no plans at the present time for transshipment within the PECo system, 5-2 Rev. 1 J
- d. Shutting down PBAPS would result in an economic hardship that would be imposed on PECo shareholders and customers. Moreover as indicated in NUREG-0575, " Final Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor fuel," the replacement of nuclear power by coal-generating capacity would cause excess mortality to rise from 0.59-1.7 to 15-120 per year for 0.8 GWY(e). Based on the above, shutting down PBAPS does not represent a viable alternative.
The subject of the comparative economics associated with various spent fuel options in the subject on Chapter 6 of NUREG-0575. Although the material presented is generic, it is of value in comparing the costs of the various i operations. Of the options presented in Chapter 6 of NUREG-0575, high-density spent fuel storage at the site is the most economic option. Should the lack of spent fuel storage cause a shutdown, energy generated by PBAPS would need to be purchased from the Pennsylvania, New Jersey, Maryland Interconnection at a cost significantly higher than could be generated at PBAPS. In addition, P8APS would need to be replaced with new generation facilities which would result in environmental effects and additional commitment of materials and resources. Plant shut down would place a heavy financial burden on PECo and its customers which cannot be justified. The average daily cost to PECo for a plant shutdown is approximately $500,000 per Unit. 5.1.4 Resources Committed Reracking of the spent fuel pools will not result in any irreversible and irretrievable commitments of water, land, and air resources. The land area now used for the spent fuel pools will be used more efficiently by safely increasing the density of fuel storage. The materials used for new rack fabrication are discussed in Section 4.7.1. These materials are not expected to significantly foreclose alternatives available with respect to any other licensing actions designed to improve tae possible shortage of spent fuel storage capacity, u m e
t I 5.1.5 Thermal Impact on the Environment Section 3.2 contains~a description of the folleving censioeratio s: IN: , additional heat load and the anticipated treximum tempefatare of ' water in the SFP that would result from the propose'd' expansion, the' additional heat load on component and/or plant cooling water systems. The proposed increase in storage capacity will result in an insignificant impact on the environment. 5.2 RADIOLOGICAL EVALUATION 4 The new spent fuel storage racks will allow storage of more spent fuel than is ' currently described in the FSAR. However, because the proposed racks will not affect such parr. meters as fuel burnup, the amount of fuel used by the reactors, or discharge. schedules, there will be no appreciable change in the amount of radioactivity released from the station. l 5.2.1 Solid Radaaste Filter-demineralizer units are provided'to minimize corrosion product buildup and control water clarity in'the spent fuel pool. The filter-demineralizers, arranged so that one is designated for eeen reactor ano a third is a common . spare for use by either unit, are normal,1y in continuous oper,stion. The basis ' for regeneration of the filter-demineralizers is chemical exhaustion of the " resin or high differential pressure across the filter unit.' Presently, each , filter-demineralizer is reqcnerated approximately every three' weeks resniting in approximately 100 cubic feet of solid wastelannually for each Unit. . 4 s Because most corrgsion products are partictdate%atter Which is introductd to the fuel pool during refueling and fuel , transfer operations, the amount of solid wa"ite generated by the cleanup systb.: is not a furction of spent fuel storage capacity. Therefore no significant increase in the quantity of solid radioactive wastes generated by the spent fuel pool cleanup system is anticipated. Replacement of the filter-demineralizer resins is accompli,;Md-remotely resulting in very low personnel doses. ',
- 1 1
, 5-4 '
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5.2.2 , Gasecus ,Radweste , Gaseous radicactive ' releases from the sta9on M ll not increase due to-
. 'c . Operating pitnt experience with dense fuel storage has shown no noticeable increases in airborne radioactivity above the spent fuel pool or at the site boundary. No significant increases are expected from more '^ , dense storage. Historical data of airborne radioactivity over the pool shows levels less tnan'10-10 microcuries/cc.
- d. As stated in Se,ction 5.2.1 and based on op\erating experience with dense '
fuel storage racks, there is no significant increase in the radwaste
- generated by the spent fuel pool cleanup system. This is because j op'eratinge'xperiencehasshownthatwithhigh'dhnsitystorageracks,
,i '
l there is no significant increase in the radioactivity levels in tha spent sf Je1 pool water. Operating experience with high density storage racks j has shown no significant increase in the annual man-rem due to the increased fuel storage, including the changing of spent fuel pool cooling systerc resins arid filters. Changing the racks to an even higher density will rct ' change these conclusions.
=? e. Most of th'e crud associated with spent fuel storage is released soon
'( \
after fusi is removed from'the reactor. Once fuel is placed into the pool storage positions, additional crud contribution is minimal. l Crud sha*:en loose during fuel handling and fission products released through defects in the fuel account for most of the radioactivity in the pool water. As the decay heat in the defective fuel decreases, the release of fission products is also reduced. Therefore, most releases from failed fuel occur soon after refueling. This radioactivity will either decay or be removed by the spent fuel pool cleanup tystem w1 thin a year such that there will be no increase in, lor.g term activity levels.
, 4 e
i s 5-6 Rev. 1 q s ,C ,
The highest possible water level is maintained in the spent fuel pool to keep exposure as low as reasonably achievable. Should crud buildup ever be detected on the spent fuel pool walls around the pool edge, it could easily be washed down. ' 5.2.3.1 Personnel Exposure - Fuel Pool Shielding Radiation shielding for the spent fuel pool is provided by six-foot thick concrete pool walls, a six-foot three-inch thick concrete floor slab and approximately 24 feet of water above the spent fuel storage racks. In order to verify that the proposed fuel storage modification will meet existing radiation zone designations for areas surrounding the fuel pool, a three-dimensional shielding analysis was performed. The QADM00-G point kernel gamma shielding code was used for this analysis. The fuel pool was conservatively assumed to be filled to the proposed capacity (ie 3819 spent fuel assemblies) with a radiation source term based on activity inventories in the spent fuel assuming full power operation at the end of core life immediately preceding shutdown, a radial peaking factor of 1.65 and a shutdown decay period of 120 hours. The spent fuel activity inventories were calculated by the ORIGEN computer code using input applicable to the PBAPS l operating cycles. The gamma source volume (ie, fuel assemblies, fuel racks, internal cooling water) was homogenized into one QADMOD-G composition. Other regions in the model accounted for the external cooling / shielding water, concrete walls and floor, and areas surrounding the fuel pool. Since most of the material between the source and the receiver points is concrete, the build-up factors for concrete was used. This assumption is conservative for a receptor point above the pool surface since the concrete gamma build-up factors are higher than those for water. The mass attenuation coefficients contained in the QADM00-G library were used for all materials in the model. As an additional conservatism, the fuel pool steel liner was not included in the QADM00-G model. Also, there are no pool wall penetrations below or near the top of the stored fuel. The closest spent fuel pool wall penetrations correspond to the pool cooling system piping. These penetrations are located approximately 22 feet above the top of the fuel assemblies where they are of 57 Pev. I t _________
no concern from a pool re-racking / shielding standpoint. The QADMOD-G model is shown in Figures 5-1, 5-2, and 5-3, The region materials for this model are identified in Table 5-3. Results of this analysis show that radiation dose levels on the outside surface of the pool walls, below the spent fuel pool floor at elevation 165'-0" of the reactor building and at the pool surface will all be less than 2.5 mrem /hr. This radiation level meets the design radiation zoning requirements for areas surrounding the spent fuel pool as identified in Section 12.0 of the PBAPS UFSAR. Therefore, it is concluded that the proposed fuel storage modification will not increase radiation levels or personnel doses in the reactor building or surrounding areas above current design values. 5.2.4 Radiation Protection During Re-Rack Activities 5.2.4.1 General Description of Protective Measures The radiation protection aspects of the spent fuel pool modification will be controlled in accordance with existing health physics control procedure. Gamma radiation levels in the pool area are monitored by the station Area Radiation Monitoring System, which has a high level alarm feature. Additionally, periodic radiation and contamination surveys are conducted in work areas as necessary. Where there is a potential for significant airborne radionuclide concentrations, continuous air samplers can be used in addition to periodic grab sampling. Personnel working in radiologically controlled areas wear protective clothing and respiratory protective equipment, depending on work conditions, as required by the applicable Radiation Work Permit (RWP). Personnel monitoring equipment is assigned to and worn by all personnel in the work area. At a minimum, this equipment consists of a thermoluminescent dosimeter (TLD) and self-reading pocket dosimeter. Additional personnel monitoring equipment, such as extremity badges, are utilized as required. Contamination control measures are used to protect persons from internal exposures to radioactive material and to prevent the spread of contamination. 5-8 Rev. 1 I
Work, personnel traffic, and the movement of material and equipment in and out of the area are controlled so as to. minimize contamination problems. Material and equipment will be monitored and appropriately decontaminated and/or wrapped prior to removal from the spent fuel pool area. The station radiation protection staff closely monitors and controls all aspects of the work so that personnel exposures, both internal and external, are maintained as low as reasonably achievable (ALARA). Water levels in the spent fuel pool will be maintained to provide adequate shielding from the direct radiation of the spent fuel. Prior to rack replacement, the spent fuel pool cleanup system will be operated to reduce the activity of the pool water to as low a level as can be practically achieved. 5.2.4.2 Re-Racking Procedures Prior to re-racking, a detailed re-rack plan will be developed and approved by the PBAPS Plant Operation Review Committee. The plan which will identify step-by-step operations, will assure that personnel exposure will be maintained as low as is reasonably achievable (ALARA). It will require that spent fuel stored in the racks is not within the area of influence of a potential rack-drop accident during removal of existing rack modules or installation of new ones, and that operations or potential accidents (e.g., rack-drop) will not adversely affect any plant equipment needed to mitigate consequences of a reactor accident or necessary to maintain safe shutdown. Similar operations have been successfully accomplished by PECo and a number of utilities in the past, and a safe and acceptable re-racking plan will be developed. The Generic Environmental Impact Statement (NURGE-0575, August 1979) also suggests that re-racking may be safely accomplished subject to evaluation of specific rack designs and factors enumerated above. , 5-9 Rev. 1
5.2.5 Symmary of Re-Rack Operation The change out of racks will be accomplished using the standard practice of removing o'd racks, installing new racks, moving fuel into the new racks, and repeating the process until all of the old racks are removed and new racks installed. The spent fuel pool operations will be performed underwater in order to naintain shielding of personnel from the stored spent fuel. In order to describe the proposed operations, the following two areas are defined at; follows: Area I: The area.on the east side of the Unit 2 spent fuel pool; and the west side of the Unit 3 spent fuel pool. In general, this area currently contains the stored spent fuel. Area II: The area on the west side of the Unit 2 spent fuel pool; and the east side of the Unit 3 spent fuel pool. The segtence of removal and installation operations in each SFP will be:
- 1. Transfer existing spent fuel assemblies in existing Area I storage racks, leaving existing Area'II racks empty.
- 2. Remove existing Area II storage racks.
- 3. Install new Area II high density storage racks.
- 4. Relocate existing spent fuel assemblies to new Area II high density storage racks.
- 5. Remove existing Area I storage racks.
- 6. Install new Area I high density storage racks.
5-10 Rev. 1
5.2.5.1 Material Receipt and Handling The racks will be delivered by truck. Material receipt, receipt inspection, handling, and storage will be performed in accordance with approved PECo procedures. Each rack module will be positioned in the equipment access lock at Building Elevation 135' and the racks lifted individually, using the 125-ton reactor building crane and a specially designed rack lifting tool. Racks will be lifted to the refueling floor (Elevation 234') via equipment hatches at Elevations 165', 195', and 234'. The handling of all materials entering or leaving the spent fuel pool will be scheduled and controlled by PECo approved procedures to preclude movement over racks containing spent fuel assemblies. 5.2.5.2 Fuel Handling Spent fuel assemblies in the spent fuel pool at the time of installation will be transferred into Area I of Unit 2 and Unit 3, as appropriate. After inspection, acceptance, and installation of the new high density racks in Area II, the spent fuel assemblies will be moved from the Area I racks to the Area II high density racks. Fuel handling and placement will be performed in accordance with procedures prepared by PECo Electric Production Department. Work will be performed by PECo personnel. 5.2.5.3 Removal of Existing Racks The existing fuel racks are a free standing design. Using a specially designed lifting tool, the rack modules will be grappled and individually lifted from the pool. 5-11 Rev. 1
Work will be performed by divers, or by utilizing remote tools from above the pools as appropriate. As existing racks are removed from the pools, they will be surveyed by PECo Health Physics personnel to determine radiation / contamination levels. The dryer-separator pits are tentatively planned to be utilized as decontamination areas to prepare the racks for off-site shipment where subsequent decontamination is planned for free release. 5.2.5.4 Rack Disposal The spent fuel storage rack modules that will be removed from the spent fuel pool weigh between 8,700 and 16,350 pounds each. When the racks are removed from the pool, they will be rinsed with demineralized water or spent fuel pool water to remove loose contamination. Depending on the levels of loose contamination, this rinse will be with either a low or high pressure spray. Personnel involved in the operation, and others in the immediate area, will wear appropriate protective clothing and respiratory protective equipment if needed. The rinsing operation is expected to remove significant quantities of loose contamination from the racks while causing relatively low exposure to decontamination personnel. This procedure minimizes subsequent personnel exposures due to handling and packaging of rack sections for shipment and disposal. Disposal options for the old racks include decontamination and burial. Depending on the effectiveness of decontamination, the rack sections will eventually either be sold as scrap, or if decontamination is not possible, buried at a low-level burial site. 5.2.5.5 Removal of Interference Hardware The original design of the PBAPS spent fuel pool storage racks utilized swing bolts anchored to the spent fuel pool floor. The new rack support pads cannot 5-12 Rev. 1 t
avoid all of these bolts. It is anticipated that four swing bolts per pool will need to be cut and removed. In operations similar to what was performed at the last PBAPS reracking, the swing bolts will be cut to within 1 incn of the pool liner, and support plates installed around the remaining sections still attached to the liner. The rack support pads will rest on these support plates in order to avoid interferences with the removed swing bolts. _Some sections of the Spent Fuel Pool Cooling System discharge piping are also scheduled to be removed as they will interfere with the new storage racks. This discharge piping removal is similar to that performed at the Limerick l Generating Station and several other nuclear plants. The spent fuel pool cooling analysis performed has confirmed that the piping removal will not have an adverse effect on spent fuel cooling. It is anticipated that divers will be used to remove the piping and swing bolts described above. Prior to all diving operations, the spent fuel assemblies stored in the pool will be arranged in such a manner as to yield t the lowest practicable dose rates to divers and still minimize the amount of movement or rearrangement of the assemblies. These two criteria should lessen the combined effects of radiation directly from the fuel and radiation from activated particles in the water that are stirred up by fuel movement. Travel l paths will be established for divers to ensure that exposures received going l
, to and from the work areas are maintained ALARA. Health Physics personnel l will be in the immediate area at all times when divers are in the water.
Their duties will be to provide health physics support to minimize personnel exposure and to enforce good radiological work practices and adherence to RWP requirements. They, along with the diver supervisor who will be in direct comunication with the divers, will continually observe the divers while they are in the pool and ensure that the divers are warned if they approach high radiation / exclusion zones. Work zone boundaries will be utilized to indicate to the diver how close he is to the edge of his allowed work area. Divers may wear protective clothing items inside their rubber diving suits to protect them frcm contamination when they remove their diving suits and exit 5-13 Rev. 1
the spent fuel pool area. TLD's will be worn inside the diving suits on the head, chest, back, and extremities as prescibed by Health Physics personnel. Self-reading pocket dosimeters will be sealed in plastic bags and also worn , t inside the diving suit. The self-reading pocket dosimeters will be read and ; recorded after each dive. Tabulations of each individual's cumulative whole ; body dose will be prepared for each diver and will be reviewed by the diving ; supervisor and the cognizant Health Physics Supervisor. This information will be used primarily (1) to maintain doses ALARA within the limits and (2) to efficiently allocate exposure among the divers working in the pools. 5.2.5.6 Installation of New High Density Racks Racks will be lifted individually from the refueling floor lay-down area and lowered into position in the pools using the 125-ton reactor building crane and a remote lifting device provided by the rack vendor. Each rack will be aligned and levelled as it is placed in its proper pool location. Each rack is a free-standing unit which rests on the pool floor utilizing bearing pads attached to levelling screws. Levelling screws will be adjusted by remotely-controlled tools. The requirements of 10 CFR 50, Appendix 8 as included in the PBAPS, Units 2 and 3 Quality Assurance Plan, Volume I, will be implemented for receipt, storage and installation of the spent fuel racks. Work will be performed in accordance with procedures which are reviewed and approved prior to use. Inspections will be performed and documented by per,sonnel other than those who performed the work. Site activities will be subject to audit by the Quality Assurance Section of PECO's Engineering and Research Department. 5.2.6 Re-Racking Exposure Estimate Table 5-4 is a summary of estimated exposures for the reracking operation and for each group of workers. These estimates are based upon a per task evaluation of the man hours and dose rates being observed in the area near and over the spent fuel pools, and radiological surveys performed during the last l 5-14 Rev. 1 , l
reracking at PBAPS. It is anticipated that this estimate is conservative. All operations performed will follow ALARA principles. 5.3 ACCIDENT EVALUATION 5.3.1 Spent Fuel Handling Accidents 5.3.1.1 Fuel Assembly Drop Analysis The proposed spent fuel pool modifications will not increase the radiological consequences of fuel handling accidents previously evaluated in the PBAPS Updated FSAR. 5.3.1.2 Cask Drop Analysis The PBAPS Updated FSAR evaluates the potential for a cask drop over the spent fuel pool. The rerack program will not alter the cask handling procedures or the results of the cask drop evaluation. The cask handling crane meets the design and operational requirements of NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants". Existing technical specifications regarding allowable loads carried over spent fuel, and established heavy load paths will be followed during the removal and installation of storage racks. 5.3.2 Conclusions The proposed spent fuel pool modifications will not increase the radiological consequences of a heavy load or cask drop accident previously evaluated. Since there will be a negligible change in radiological conditions due to the increased storage capacity of the spent fuel pool, no change is anticipated in the radiation protection program. In addition, the environmental consequences of a postulated fuel handling accident in the spent fuel pool, described in Updated FSAR Section 14.0, remain unchanged. Therefore, there will be no 5-15 Fev. 1
change or impact to any previous determinations of the Final Environmental Statement. Based on the foregoing, the proposed modification will not significantly affect the quality of the human environment; therefore, under 10 CFR 51.5c, issuance of a negative declaration is appropriate. l l l l l l l l I l l l l l l S-16 Rev. 1 L
- TABLE 5-1 EXISTING SPENT FUEL STORAGE CAPACITY PBAPS Unit 2
-Refuel Number of Cumulative Total Storage Space Date Assm. Discharged Discharged Available To Date 1,464 1,144 9/86 276 1,740 863
- 12/87 276 2,016 592 **
i l PBAPS Unit 3 Refuel Number of Cumulative Total Storage Space Date Assm. Discharged Discharged Available 1 To Date 1,497 1,111 ! 3/87 276 1,773 835
- 9/88 276 2,049 559 **
- Storage Limit with Full Core Discharge Reserve (FCDR)
** FCDR Lost L
5-17 Rev. 1
f
. l TABLE 5-2 SPENT FUEL STORAGE CAPACITY AFTER RERACK l
PBAPS Unit 2 Refuel Number of Cumulative Total Storage Space Date Assm. Discharged Discharged Available To Date 1,464 2,355 9/86 276 1,740 2,079 12/87 276 2,016 1,803 6/89- 276 2,292 1,527 12/90 276 2,568 1,251 6/92 276 2,844 975
- 12/93 276 3.120 699 **
PBAPS Unit 3 Refuel Number of Cumulative Total Storage Space Date Assm. Discharged Discharged Available To Date 1,497 2,322 3/87 276 1,773 2,046 9/88 276 2,049 1,770 3/90 276 2,325 1,494 9/91 276 2,601 1,218 3/93 276' 2,877 942
- 9/94 276 3,153 666 **
Storage Limit with Full Core Discharge Reserve (FCDR)
** FCDR Lost 5-18 Rev. 1
TABLE 5-3 IDENTIFICATION OF REGION MATERIALS FOR QADM00-G MODEL OF SPENT FUEL POOL I-REGION NUMBER MATERIAL DESCRIPTION
- I air Area adjacent to pool wall 2 air Area adjacent to pool wall 3 air Area adjacent to pool wall 4 air Area adjacent to pool wall 5 air Area below pool floor 6 Concrete Pool Wall 7 Concrete Pool Wall 8 Concrete Pool Wall 9 Concrete Pool Wall 10 Concrete Pool floor slab 11 Water Pool Water 12 Water Pool Water 13 Water Pool Water 14 Water Pool Water 15 Water Pool Wcter 16 Source Homogenized spent fuel 17 Water Pool Water 18 air . Area adjacent to pool wall 19 air Area adjacent to pool wall 20 air Area adjacent to pool wall 21 air Area adjacent to pool wall 22 Concrete Pool Wall 23 Concrete Pool Wall 24 Concrete Pool Wall 25 Concrete Pool Wall 26 air Area above pool surface 27 air Area above pool 4
l 5-19 Rev. 1 l
TABLE 5-4 ESTIMATED DOSES DURING RERACKING (All doses are in man-rem per spent fuel pool) Fuel Handling Operators 7.5 Overhead Crane Operator 0.6 Health Physics 2.1 Construction Crew 15.8 Divers and Diver Support 6.4 Engineering 0.2 Miscellaneous
- 3.3 35.9 Estimated Range = 30-36 Man-Rem per pool
- 10% assumed for miscellaneous operations 5-20 Rev. 1
l l l FIGURE 5-1 I PSAPS SPENT FUEL POOL QADM00-G MODEL (SECTIONVIEWEAST-WEST)
- SOURCE ,
= SOCCARY O - = "o" 8 ? l f s '
g soir.s . !' V R L l
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d i 5-21 pey, i i l l * . _ _ _ _ _ ,
I FIGURE 5-2 PBAPS SPENT FUEL POOL QADM00-G MODEL (SECTION VIEW NORTH - SOUTH) l
- 501lRCE b - BGWDARY O-"""
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@ 4 l
rios.it ! T l ED@ l
@ r 4 W
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I g I 4 R I I o i Y -+Y 9 2 4 S $ 9 4 % i o Ehl N i N. !.k. I 5 22 Rev. 1
v. FIGURE 5-3 PBAPS SPENT FUEL POOL QADM00-G MODEL (PLANVIEW)
] - SotRCE d - BOUNDARY x
a 0 - ""
- l l
-8700.
m 9 O ( - m.a
- gg e gv.M, -
- =. w :. .
- 5) l45
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@ l j' O E 4
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*W $ 3% $
b i Y i
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