ML19259C590
| ML19259C590 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/09/1979 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19259C588 | List: |
| References | |
| NUDOCS 7907170073 | |
| Download: ML19259C590 (69) | |
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0,. M.f* 4 Georgia Power n n? r(
[ N0 [b 790717007 3
EDWIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 SPENT FUEL POOL MODIFICATION 2 )k b
TAB'.E 3F CONTENTS Sectinn Title Page
- 1. 0 Introduction . . . . 1-1 2.0 Overall Description. . . . . . 2-1 3.0 Design Bases . . . . . . . . ... 3-1 4.0 Hechanical and ctructural Considerations . ....... 4-1 4.1 Seismic Analysis . . . . . . . . . . . . . . 4-1 4.2 Stress Analysis. . . . . . . . . . . .. . .. . 4-3 5.0 Mrterial Considerations. . . . ... ........ 5-1 6.0 Installation . . . . . . ..... ..... 6-1 7.0 Nuclear Considerations . . . . . . . . . ..... 7-1 7.1 Neutron Multiplication Factor. . . . .. ....... 7-1 7.2 Input Parameters . . . . . . .. .... 7-1 7.3 Geometry, Bias, and Uncertainty. . . ... .. 1-2 7.4 Postulated ficcidents . . ... .. 7-5 8.0 Thermal Hydraulic Considerations . . ... 8-1 8.1 Description of the Spent f uel Pool Cooling System. 8-1 8.2 Heat Loads and Pool Temperatures for Present Storage Capacity . . 8-2 8.3 Heat Loads and Pool Temperatures for Increased Storage Capacity . . .. ... 8-3 8.4 Loss of Spent Fuel Pool Cooling. .. .. . . 8-6 8.5 Local Fuel Bundle Thermal Hydraulics . . . 8-7 8.6 Radiological Impact of Spent Fuel Pool Boiling . . . ... 8-10 9.0 Cost Benefit Assessment. . . . . . . ... 9-1 9.1 Need for Increased Storage Capacity. .. . 9-1
- 9. 2 Alternative to Increasing Storage Capacity .... .. 9-2 9.3 Capital Costs. . . .. . .... 9-5 9.4 Resource Commitment. . . . . . ..... .. 9-5 9.5 Environmental Impact of Expanded Spent Fuel Storage. . . . 9-6 10.0 Radiological Evaluation. . . . .... ....... 10-1 10.1 Spent Resin Waste. . . . . .... .... . 10-1 10.2 Noble Gases. . . . . . .... ... 10-1 10.3 Gamma Isotopic Analysis for Pool Water . .. . . 10-2 10.4 Dose Levels Over and Along the Sides of the Pool .... 10-2 10.5 Airborne Radioactive Nuclides. . .. ... ... 10-2 10.6 Radiation Protection Program . ......... .. 10-2 10.7 Disposal of Present Spent Fuel Racks . .. . . 10-3 11.0 Accident Evaluation. . . . .. .... .. 11-1 12.0 Conclusions. . . . . ...... 12-1 13.0 Notes and References . . ... . 13-1 i
2104 097
- 1. 0 INTRODUCTION This design report and safety evaluation considers the installation cf high density, poisoned fuel storage racks in the existing spent fuel pools of Edwin I. Hatch Nuclear Plant Units 1 & 2.
The Hatch 1 and 2 spent fuel pools currently conta . racks that can hold 840 and 1120 fuel assemblies, respectively. It was originally assumed that about one quarter of the core would be discharged annu-ally and that spent fuel would be removed from the plant for reproc-essing within approximately a year after discharge from the reactor.
Because the reprocessing option is not available at this time, the storage capacity of the spent fuel pools must be expanded by replacing the existing spent fuel storage racks with high density, poisoned racks.
It is desirable to have enough capacity in reserve to allow for a full core offload. Such capacity will not exist in the Unit 1 spent fuel pool subsequent to its 1979 refueling cycle. Storage space in the Unit 2 spent fuel pool must then be used to allow for a full core discharge from Unit 1. The high density spent fuel storage racks will provide 3171 storage sc aces in Hatch 1 and 2755 in Hatch 2. The modification will provide storage capacity up to the year 1997 with a full core reserve, assuming annual quarter reloads. Instal u ,,r.; of the high density storage modules is scheduled to commence in March 1980, first in the Hatch 2 spent fuel pool and then in Hatch 1.
This report describes the design of the high density fuel storage racks to be insta' led and contains a discussion of the environmental and radiological considerations of the installation. tie information contained herein has been prepared based on the recommendations pro-vided in " Operating Technical Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" which was issued by the Nuclear Regulatory Commission (NRC) on April 14, 1978 and later amended on January 18, 1979.
General Electric Company will design and supply the high density, poisoned spent fuel storage racks that will be installed at Plant Ha tc t. . Similar storage racks have previously been reviewed and appro<ed by the NRC on the Monticello and Browns Ferry Nuclear Plants.
2i04 098 1-1
2.0 OVERALL DESCRIPTION The location of the spent fuel storage pools within the plant is shown in Figures 2-1 through 2-10 (Unit 1 FSAR Figures 12.1-4 through 12.1-8 and Unit 2 FSAR Figures 3.8-28 through 3.8-32,respectively). The arrangement of the high density fuel storage system for the pools i.
shown in Figures 2-11 and 2-12.
The high density racks are a base-supported modular design that will replace the existing fuel storage and control rod storage racks.
Control rod storage will be provided by supplying a minimum of twenty storage hangers in each of 'Inits 1 and 2. There will be ten extra positions in each pool for storage of defective fuel.
The high density module provides storage spaces for fuel bundles, which do not include the flow channels, or fuel assemblies, which do include the flow channels (see Note 1), on approximately 6.5 inch center to center spacing. Six basic configurations of the basic module are contemplated, containing 13 x 11, 13 x 13, 13 x 15, 13 x 17, 13 x 19, and 15 x 19 storage cells. The combined pool capacity of 5926 fuel assemblies stored in high density fuel racF.s is made up as shown below:
Module Fuel Configuration Capacity Quantity Assemblies Unit 1 13 x 11 143 4 572 13 x 13 169 4 676 13 x 15 195 5 975 13 x 17 221 3 663 15 x 19 285 1 285 3171 Unit 2 13 x 15 195 8 1560 13 x 17 221 3 663 13 x 19 247 1 247 15 x 19 285 1 285 2755 TOTAL 5926 An additional 80 spaces are included in the Unit 2 pool for spent fuel storage by using four of the existing storage racks. Together with the ten defective fuel locations in each pool, the maximum combined pool storage capacity is 6026 (3181 in Unit 1 and 2845 in Unit 2).
Each free standing fuel storage module is fabricated from fuel storage tubes, made by forming an outer tube and an inner tube of 304 stain-less steel with an inner core of Boral (see Note 2) into a single tube. The outer and inner tubes are welded together after being sized to the required dimensional tolerances. The completed storage tubes are fastened together by angler welded along the corners and attached to a base plate to form storage modules. Figure 2-13 shows the 13 x 13 module schematically. This module is approximately 7 feet square 2-1 2104 099
and 14 feet high. Figures 2-14 and 2-15 provide additional informa-tion pertaining to the arrangement plan of the pools.
The base plate of each module is supported on all four corners by 2-inch thick foot pads. The foot pads rest on 6-inch thick corner-support pads which in turn rest on the fuel pool floor liner. This raises the base plate of the module a minimum of 8 inches above the floor of the fuel pool, allowing sufficient clear area to permit natural circulation of cooling water to the modules without taking credit for sources of forced cooling.
2104 100 2-2
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3.0 DESIGN BASES The new spent fuel storage system was designed to conform to the applicable provisions of the following codes, standards, and regula-tions:
- 1. General Design Criterion 2 (per 10CFR50, Appendix A) as related to components important to safety being capable of withstanding the effects of natural phenomena.
- 2. General Design Criterion 3 as related to protection against fire hazards.
- 3. General Design Criterion 4 as related to components being able to accommodate the effects of and to be compatible with the environ-mental conditions associated with normal operation and postulated accidents.
- 4. General Design Criterion 62 as related to the prevention of criti-cality by physical systems.
- 5. Regulatory Guide 1.13 as it relates to the fuel storage facility design to prevent damage resulting from the SSE and to protect the fuel from mechanical damage.
- 6. Regulatory Guide 1.29 as related to the seismic design classifica-tion of facility components.
- 7. Regulatory Guide 1.92 as related to combination of loads for seismic analysis.
- 8. 10CFR20.
- 9. ASME Section III.
- 10. Branch Technical Positior. ASB 9-2 contained in the Standard Review Plan.
- 11. Light-Gage Cold-Formed Steel Design Manual, 1961 Edition, American Iron and Steel Institute.
- 12. 10CFR100
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3-1
4.0 MECHANICAL AND STRUCTURAL CONSIDERATIONS The high density fuel storage system (HDFSS) module has been analyzed for both operating basis earthquake (0BE) and safe shutdown earthquake (SSE) conditions. A detailed stress analysis was then performed to check the design adequacy of the module against calculated loads. Results indicated that the HDFSS module design is ader;aate for the postulated combined loading conditions. 4.1 Seismic Analysis The HDFSS module has been analyzed for both OBE and SSE conditions. Critical damping ratios of 2 percent were used in the analysis for the SSE condition and 1 percent for the OBE condition. The design floor acceleration response spectra are given in Figures 4-1 through 4-6. These spectra are based on Hatch 2 which bounds the spectra for Hatch
- 1. Combination of the modal response and the effect of the three components of an earthquake were performed in accordance with the applicable provisions of US NRC Regulatory Guide 1.92.
The seismic analysis was performed in several steps. First, the hydrodynamic effect, which represents the inertial properties of the fluid surrounding the submerged modules, was calculated to obtain the hydrodynamic virtual mass terms based on the module and pool configur-ation. Three-dimensional end effects and leakage between modules are accounted for by modifying the calculated hydrodynamic mass. Figures 4-7 and 4-8 show the plan view of the two-dimensional model of the modules and pool used in the hydrodynamic virtual mass analysis. The model consisted of two rigid bodies. the modules and the pool walls. Water finite elements fill the spaces in between the walls and the modules. The total mass matrix of each module for the analysis is equal to its structural mass matrix plus the hydrodynamic mass matrix. Conservative structural damping values of 1 percent for the OBE and 2 percent for the SSE are applied withoct any added damping from fluid effects. The WATER-01 computer program, GE proprietary, was used to determine the hydrodynamic mass of one rectangular body inside another rectangu-lar body. This program has been design reviewed and meets NRC-QA requirements. The metnodology in calculating hydrodynamic mass has been presented in Reference 1. Second, the derived total mass of the module was used to perform dynamic analysis for the OBE and SSE. As seen in Figure 4-9, for a typical 13 x 13 module, when the added-mass terms from the hydro-dynamic mass effect were included, the fixed base frequency decreased. Third, both finite-element and lumped-mass models of a podule were then developed to provide a basis for selecting simplified module models to be used in the module and support system analysis and module sliding analysis. The finite-element model also was used to obtain the distribution of shear forces in the module plate elements. 4-1 2104 129
Fourth, an eleven-node lumped-mass model was then developed by lumping the tributary module mass to the corresponding node point and ini-tially selecting the stif fness properties based on beam theory. The stiffness properties of this model were based on matching the natural frequencies of the finite element model. In the nonlinear analysis to calculate the amount (.f sliding and tilting, a two-node lumped mass model was found to adequately repre-sent the module and support system analyses since the response to the module and support system was shown to be primarily a rigid body motion. Both the first mode and rigid body dynamic properties were simulated by this model. The effects of the corner supports were added to the model by including base springs and the final model was used in the sliding analysis. The horizontal spring represents the stiffness of the support pad and the vertical spring represents the stiffness of the fuel support plate, the foot pad, and the support pad. The mechanism for controlling the shear force in each module is the limiting of the coefficient of friction between the module and the support pad by the selection of a non galling, corrosion-resistant material with a low coefficient of friction to be used as the module foot pads which are in contact with the stainless-steel support pads. The range of friction coefficient for the selected materials was found to be between 0.145 and 0.203. The friction coefficient between the stainless-steel support pads and the stainless-steel liner is at least 0.349. This difference ensures that sliding will occur between the foot pad and the support pad, and not between the support pad and the floor liner. The sliding analysis was done using the two-dimensional, non-linear DRAIN-2D and SEISM computer codes. DRAIN-2D was originally developed at the University of California at Berkeley; SEISM was developed by GE. Both computer codes have been design reviewed and meet NRC-QA requirements. Sliding and overturning of the module were studied for the SSE and OBE conditions. All of the modules were found to be stable under the worst postulated seismic loading conditions, and the minimum 2-inch clearance between modules precludes contact during a seismic event. 4.2 Stress Analysis The HDFSS module has been designed to meet Seismic Category I require-ments. Structural integrity of the rack has been demonstrated for the load combinations below using linear elastic design methods. Analysis was based upon the criteria and assumptions contained in the following documents:
- a. ASME Boiler and Pressure Vessel Code Section III, Subsection NF.
- b. USNRC, Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis
*L ) O!k \3 4-2
- c. Hatch 2 Final Safety Analysis Report, Seismic Design Criteria.
OBE - Operating Basis Earthquake SSE - Safe Shutdown Earthquake
- d. Light-Gage Cold-Formed Steel Design Manual, 1961 Edition, American Iron and Steel Institute.
Acceptance criteria were based on:
- a. Normal and upset (0BE) Appendix XVII, ASME, Section III.
- b. Faulted (SSE) Paragraph F-1370, ASME Section III, Appendix F.
- c. Local buckling stresses in the spent fuel storage tubes were calculated according to " Light-Gage Cold-Formed Steel Design Manual" of American Iron and Steel Institute in lieu of Appendix XVII, ASME, Section III, because of its applicability to these light gage tubes. Only the strength of the outer wall thickness of 0.090 inch nominal is considered in the stress calculations.
The applied loads to the rack are:
- a. Dead loads which are weight of rack and fuel assemblies, and hydrostatic loads.
- b. Live loads - effect of lifting an empty rack during installation.
- c. Thermal loads - the uniform thermal expansion caused by pool temperature changes from the pool water and stored fuel.
- d. Seismic forces of OBE and SSE.
- e. Accidental drop of a fuel assembly from the maximum possible height.
- f. Postulated stuck fuel assembly causing an upward force of 1000 pounds.
The load combinations considered in the rack design are:
- a. Live loads.
- b. Dead loads plus OBE.
- c. Dead loads plus SSE.
- d. Dead loads plus fuel drop.
In accordance with ASME Section III, Subsection NF, Paragraph NF-3230, thermal stresses are not considered. Furthermore, thermal loads were not included in combinations because the design of the rack makes them negligible; i.e., the rack is not attached to the structure and is free to expand or contract under pool temperature changes. 4-3 nt '\ 3u 4 \ b k
Stress analyses were done for both OBE and SSE conditions, based upon the shears and moments developed in the finite element dynamic anal-ysis of the seismic response. These values were compared with allow-able stresses referenced in ASME Section III, Subsection NF (Table 4-1). Values given in Table 4-1 are based on the limiting module size. Stresses for the other module configurations are lower, and therefore, are not given here. Additional analyses were then per-formed to determine the dynamic frequencies, earthquake loading reac-tions, and internal forces in critical module and support system locations. Those values are summarized in Table 4-2. The force path in the module caused by a horizontal earthquake is shown schematically in Figure 4-10. This figure shows the path of the horizontally induced earthquake fuel element inertial forces from the fuel element to the module support pads. Part of the fuel bundle inertial forces induced by the motion of the module are transferred either through the water or directly to the tube walls perpendicular to the direction of motion (Point 1 in Figure 4-10). These walls then transfer the forces to the side tube walls, which carry the forces down the walls and into the fuel support plates (Point 2). The por-tion of the fuel bundle load which is not transferred to the fuel tube walls is transferred directly to the fuel support plate at the point where the lower end fitting of the fuel bundle is supported vertically (Point 3). The fuel support plates, acting as a relatively rigid diaphragm, transfer the in plane shear forces to the long casting which then transfers the shear forces to the module base assembly plate (Point 4). The forces are carried in the module base assembly (Point 5) until they are ultimately transferred to the foot pad and to the support pad and the pool slab (Point 6). The vertical forces caused by earthquake and gravity loads become axial forces in the foot pads. The critical location for the com-pression forces from the foot pads is in the long castings and tubes directly above the foot pads. For stress analysis purpose, these compression forces are considered to be resisted by four fuel tubes sitting directly above the support pad. Fuel assembly drop accidents were analyzed. The results are sum-marized in Table 4-3. The HDFSS design does not require any different fuel handling procedures from those discussed in the Unit 1 and Unit 2 FSAR. The loads experienced under a stuck fuel assembly condition are less than those calculated for the seismic condition and have therefore not been included as a load combination. 2104 132 4-4
TABLE 4-1 Comparison of Calculated Stress vs. Allowables (psi) OBE Condition SSE Condition l Location / fype Calc Stress AllowableslCalc Stress Allowables Tube wall bending Will be pro- 20,630 Will be pro- 41,250 Tube wall shear vided by 11,000 provided by 22,000 Tube wall tension July 31,197914,880 July 31, 1979 29,760 Tube weld throat shear 11,000 22,000 Angle, weld throat shear 11,000 22,000 Casting bending 20,630 41,250 Casting wall shear 11,000 22,000 Casting wall compression 16,500 33,000 Fuel support plate bending 20,630 41,250 Support plate weld throat 20,630 41,250 bending Closure plate bending 20,630 41,250 Closure plate shear 11,000 22,000 Closure plate weld bending 20,630 41,250 Closure plate weld shear 11,000 22,000 Corner tube local compressive - - 17,224 stress check for local buckling 1 Allowable stresses referenced in ASME Section III, Subsection NF 2"{04 )
TABLE 4-2 Will be provided by July 31, 1979 e 2104 134
Table 4-3 High Density Spent Fuel Storage System Assembly Drop Accident Case Summary No. Case Description Effect on Reactivity
- 1. A fuel assembly drops 27 inches Analysis indicates that localized tube vertically and impacts the top damage or fuel support member damage will of a fully loaded HDFSS occur, but neutron absorber material will module. The dropped assembly not be removed from its position between comes to rest horizontally on adjacent fuel assemblies. A fuel assembly top of the HDFSS. resting horizontally atop the HDFSS does not increase the system reactivity because the reactivity assumes an infinite vertical length of fuel (no neutron leakage in the vertical dimension). keff < 0.90
- 2. A fuel assembly drops from 27 Structural analysis indicates that local-inches above the HDFSS, enters ized tube damage will occur and one neutron an empty storage position, and absorber plate may be damaged. A reactivity falls to the bottom of the analysis of this case, with the neutron storage position. absorber plate between two fuel assemblies totally missing, shows that k remains eff less than 0.90.
- 3. A fuel assmebly drops from 27 Same as Case 2 inches above the HDFSS and strikes a tube wall at an oblique angle.
- 4. A fuel assembly drops from 27 It is not possible for a fuel assembly inches above the top of a drop of 27 inches to drive four stored fully loaded module and assemblies through the bottom of the strikes the upper tie plates of module. Even so, the reactivity effect of 2, 3, or 4 fuel assemblies in this postulated event was calculated as a storage. limiting value. An 18-inch section of fuel in four bundles in an unpoisoned square array was found to have a k e approximately equal to that of the system.ffThere would be no increase in the overall reactivity kef f < 0. 90.
- 5. A fuel assembly drops from 27 This case was analyzed for normal handling inches above the HDFSS, falls conditions; kef f < 0. 90.
outside of the loaded HDFSS, and lodges adjacent and parallel to an unpoisoned, occupied fuel storage position. r 3 ) ,D 3 n/. 'I u $
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-- Without Mo'el Without Model With Model With Module with CJ Added Mass Added Mass Added Mass Added Mass Support System x. A *f = Fundamental frequency n Figure 4-9 Typical 13 x 13 Sequence of .tdule Modeling
i EARTHQUAKE FORCE-4l l l
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! SHORT CASTING
@ FUEL SUPPORT PLATE MODULE LONG BASE CASTING 5 ASSEMBLY FOOT PAD SUPPORT PAD Figure 4-10 Path of Earthquake Horizontal Forces in Module 2i04 145
5.0 MATERIAL CONSIDERATIONS Most of the structural material used in fabrication of the new HDFSS is type 304 stainless steel. This material was chosen because of its corrosion resistance and its ability to be formed and welded with consistent quality. The only structural material employed in the structure that is not 304 stainless steel is a special low-friction material used as a foot pad between the module and the support pad. Boral plates, used as a neutron absorber, are an integral non-structural part of the basic fuel storage tube. These plates are sandwiched m between the inner and outer wall of the storage tube and are not subject
- to dislocation, deterioration.or removal. The inner and outer walls of the storage tube are welded together at each end for mechanical rigidity. Small openings are formed in the top and bottom of each tube assembly by leaving gaps in the weld to allow for the venting of the envelope between the inner and outer tube walls. At norrt.u pool water operating temperature there is no significant deterioration or corrosion of stainless steel or Boral.
Specifications were developed specifically for the HDFSS which impose quality control requirements during the design, procurement, fabri-cation, installation, and testing of the storage system. Periodic audits of the various facilities and practices are performed by certi-fied quality assurance personnel to ensure that these QA/QC require-ments are being met. All welding and nondestructive examination (NDE) is done in accordance with the applicable provisions of the ASME Boiler
& Pressure Vessel Code, Section IX, and the American Society for Nondestructive Testing (ASNT).
Storage module components are assembled and welded in special fixtures to maintain close control of dimensional tolerances. Each storage position is checked with full length gauges to assure proper clearance between stored fuel bundles and storage tube walls. To provide assurance that specification Boral sheet is used in tube f abrication, a special quality control program is in effect at the manufacturer's facility. The concentration and distribution of the neutron absorbing material (BgC) are verified by chemical analyses and/or neutron transmission tests, and each sheet is dimensionally ' inspected. Before each piece of Boral is inserted into a tube assembly successful performance of the required inspections is verified. The presence of the neutron absorber material in the fabricated fuel storage module will be verified at the reactor storage pool site by scanning each storage tube in the modules with a neutron source and neutron detectors. The recorded results provide a comparison between neutron absorption through each Boral sheet and neutron absorption measured through the stainless steel without Boral. A significant increase in neu-tron absorption verifies the presence of Boral. 5-1 n\3,4 L u
}kb
Boral's corrosion resistance is similar to that of standard aluminum sheet. Corrosion data and industrial experience confirm that aluminum and Boral are acceptable (Reference 2) for the proposed application. Although experience indicates that it is unnecessary, an inservice test program will be conducted, consisting of periodically removing and examining samples of Boral plate which will be suspended in the storage pool. Pool water quality will be maintained as specified in the Hatch 2 FSAR,
, Section 9.1.3.2.4. No changes to water quality are expected as a result i of the planned modification to the spent fuel storage capacity (see Section 10-1 of the Radiological Evaluation).
2;04 147 i 4 5-2
6.0 INSTALLATION The HDFSS modules are a f ree-standing, bottom-supported design, resting on support pads placed onto the floor of the fuel storage pool. The installation program will consist of removing the low-density aluminum racks in the pools, placing the new support pads into prescribed positions, and lowering the new modules into position on their respective support pads. The initial installation will be in the Hatch 2 pool with the pool wet or dry. Special load-tested lifting fixtures, designed with a minimum safety factor of 3,are used to handle the support pads and the storage modules to minimize dropping any materials. The single-failure proof reactor building crane will be used to remove the old racks and to lower the new equipment into place. The Hatch 1 pool, which is filled with water and contains spent fuel, will be reracked similarly af ter the initial installation of modules in Hatch 2 has been completed. Stored fuel may be transferred to the Hatch 2 pool through the transfer canal to empty the Hatch I pool, or may be concentrated away from the rcrack work locations. The sequence of the rerack work will be such that no heavy equipment will be transferred over stored spent fuel at any time. The installation equipment is designed to allow installation of modules and pads into a water-filled pool. Following the installation, verification of neutron absorbers will be completed. O
/_ 1l u? )I }Ob r
6-1
7.0 NUCLEAR CONSIDERATIONS 7.1 Neutron Multiplication Factor The criticality analysis calculations were performed with the MERIT (Reference 3) computer program, a Monte Carlo program which solves the neutron transport equation as an eigenvalue or a fixed source problem including the effects of neutron shielding. This program is especially written for the analysis of fuel lattices in thermal nuclear reactors. A geometry of up to three space dimensions and neutron energies between 0 and 10 MeV can be handled. MERIT uses cross sections processed from the ENDF/B-IV library tapes. The qualification of the MERIT program rests upon extensive qualifi-cation studies including Cross Section Evaluation Work Group (CSEWG) thermal reactor benchmarks (TRX-1, -2, -3, -4) and B&W U0 and Pu0 criticals, Jersey Central experiments, CSEWG fast reactor benchmard (GODIVA, JEZEBEL), the KRITZ experiments, and in addition, comparison with alternate calculational methods. Boron was used as solute in the moderator in the B&W U0, criticals, and as a solid control curtain in the Jersey Central exp&iments. The MERIT qualification program has established a bias of 0.005 + 0.002 (la) ok with respect tc the above critical experiments. Therefore, MERIT underpredicts "eff by approximately 0.5 percent ak. The storage space (cell) infinite multiplication factor (k. ) was calculated for the high density fuel storage system as defined by the assumptions below and the exact geometry specifications.
- 7. 2 Input Parameters
- a. Standard BWR fuel configurations
- b. Maximum BWR fuel bundle multiplication factor (k. ) of 1.35 in standa.d core geometry at 20 C. The use of a maximum fuel km as, a criti ality base eliminates the need to analyze the multiplicity of U235 enrict'nent and burnable poison combinations.
3
- c. Storage space pitch of 6.563 in.
- d. Minimum allowable boron (B 10) concentrationg/cmqui alent to a homo-geneous areal concentration of 0.013 grams B
- e. Analysis conservatively performed using 2-dimensional infinite lattice (X,Y) model (no credit taken for axial or radial neutron leakage).
- f. Credit taken for double wall stainless steel tubes that separate fuel bundles.
The results of the calculations for several cases are in Table 7-1. The model geometry, bias, and uncertainity for each of the cases is described below. nd 3 !v\
\Y
7.3 Geometry, Bias, and Uncertainty The repeating cell geometry in Figure 7-1 is the exact geometry model, with the exception of squared corners, used in cases 1, 2 and 3 of Table 7-1. This model has the minimum allowable corner gap (storage cells touching), using the nomimal dimensions shown in Figure 7-2. No geometry bias is associated with this model. The MERIT program bias is 0.005 i .002 (lo) ak. The same basic geomet y model, but with the maximum axial average gap as shown in Figures 7-2 and 7-3, was used for case 4 of Table 7-1. The pitch was increased '.o 6.8324 in. , resulting in a gap spacing of 0.381 in. Note that this gap can occur only along one diagonal of the module with all storage tubes bowed at a maximum. This model has the same bias as the above; i.e. , no geometry bias and MERIT program bias of 0.0051 0.002 (lo) ak. An approximate geometry model, shown in Figure 7-4, was used for case 5 in Table 7-1. The model geometry bias relative to the exact model for the same conditions was 0.0087
- 0.0050 (le) ak. The MERIT program bias remains the same at 0.005
- 0.002 (lo) ak. In all cases the reported value of km includes the sum of all biases and the root-mean-square of the uncertainties.
The maximum k of a storage cell occurs at 20 C with the fuel bundles centered and no flow channels present. Any variation, such as increas-ing the cell pitch, eccentric bundle positioning, reducing moderator density, and increasing the temperature to 650C decreases the K. . Table 7-2 shows the maximum k. of the storage cell broken down into contributing bias and uncertainty values. The sensitivity of the cell km to decreasing moderator density is shown graphically in Figure 7-5. Since the cell is under-moderated, the optimum k. occurs at 1.0 g/cc. The design of the HDFSS has alternating spaces on the periphery of each module fabricated with unpoisoned closure plates. The unpoisoned locations are also di e-tly opposite each other on adjacent modules. The effect of the partially unpoisoned storage locations is small and insensitive te the inter-module water gap, as shown in Table 7-3. The maximum module k. occurs at the minimum possible water gap (1.244 in.) and is less than that of an infinite array of storage cells with no water gap. The module calculations in Table 7-3, were done with the model shown in Figure 7-6. Some of the details in the exact model were homogenized and simplified to reduce the input preparation in the module calculations. The model geometry bias was determined from an infinite array of simplified storage cells (Figure 7-7) relative to the exact geometry model. The module geometry model bias was determined to be 0.0017 1 0.0051 (la) ak. The same MERIT program bias applies. For all calculations the fuel bundle was discretely represented by fuel pellets, cladding, water rods, channels (when present), and fuel 7-2 2}Ok k
rod enrichment and burnable poison distributions within the bundle. Fuel pin spacers were not included (a conservative exclusion). The nominal bundle dimensions were used for all cases. The HDFSS includes defective fuel storage spaces attached externally to some of the storage modules. The geometric layout is shown in Figures 2-11 and 2-12. Analyses have demonstrated the HDFSS keff<0.95 with all defective fuel storage locations occupied with fuel. The sensitivity of km analyses to various changing parameters is j implied above. More specific relationships are as follows:
- a. Bundle Reactivity (percent U235) - Calculations are based on maxi-mum k. thereby obviating enrichment sensitivity considerations.
- b. Stainless steel thickness - Neutron absorption by the two layers of stainless steel comprising the storage tube was included in the criticality calculations using the nominal thicknesses of 0.0355 and 0.090 inch for the inner and outer tubes, respectively. The nominal inner tube thickness has been reduced to 0.0300 inch, and Monte Carlo calculations shown that the change in k is within the statistical uncertainty of the calculation (Case 2, Table 7-1).
- c. Water density - Figure 7-5 shows the variation of k with moderator (water) density. Since the cell is under-moderated the optimum
- k. occurs at 1.0 g/cc.
- d. Storage lattice pitch - An analysis was done using a minimum fuel pitch, represented by the storage tubes touching. Material toler-ances in the tubes were taken to maximize the km of the storage lattice. The result of this analysis is given as Case 5 in Table 7-
- 1. The results in Table 7-1 show that the nominal pitch (Case 2) has a higher k. result than the minimum pitch case (Case 6).
- e. The HDFSS and the BWR fuel to be stored in it are designed and fabricated to prevent significant quantities of air or other gas from being entrapped. Thus, no areas of reduced ef fective moderator density are created. But even if air were trapped, the effect of reduced density on the under-moderated fuel bundles is to reduce the k eff f the system.
7.4 Postulated Accidents Several fuel assembly drop accidents have been analyzed. The results are summarized in Table 4-3. The handling of heavy objects in the spent fuel pool area is addressed in Section 11.0 of the accident evalution. A tornado generated missile model has been used for the Hatch spent fuel pools (refer to the response to Question 130.19 in the Unit 2 FSAR) that could result in impacting the storage module. The angles in the structural grill system associated with the reactor building tornado relief vent openings have been postulated as a secondary missile source resulting from impact of a plank missile. A maximum of three angles 7-3 e nLiv
's 3 ') h3\
could be generated as secondary missiles with a maximum energy of 2000 f t-lb each. Analysis shows that the HDFSS module can withstand such impact energy without resulting in a nuclear hazard. Loss of all cooling in the spent fuel pool, resulting in boiling of the pool water, is an accident that has been analyzed in Section 8.4. The effect of such boiling on the undermoderated fuel bundle is to reduce the system k eff. No criticality accident will result. The fuel storage module design has been evaluated for the accepta-
, bility of stresses from several combined loads, including earth-
! quake-induced loads, as discussed in Section 4.2. Resultant stresses are within allowable limits, assuring the integrity of the modules under the combined loading. This precludes a criticality accident resulting from an earthquake. 2 Od k 7-4
TABLE 7-1 Single Cell High-Density Fuel Storage Criticality Results Case Pascription K. (+ 2 o )1 2 1 Nominal Rack Dimensio s 0.8668 + 0.0075 With Flow Channel @20 C 2 Nominal Rack Dimensions 0.8674 + 0.0086
- Without Flow Channel @20*C t
3 Sume as Case 2 except 0.8561 + 0.0084
@65 C 4 Increased Pitch without 0.8364 + 0.0106
~
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- k. 0.8624 Calculational Convergence Ak + 0.0038 MERIT Bias and Uncertainty t>k 0.005 + 0.002 j Model Bias and IJncertainty Ak None Total 0.8674 + 0.0086 (20)
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TABLE 7-3 HDFSS Criticality Analysis Module Interaction Description k. ( 20) Minimum gap between modules 0.8593 + 0.0131 (2A = 1.244 in.) Intermediate gap between 0.8579 1 0.0130 modules (2A = 2.100 in.) Nominal gap between modules (2A = 2.967 in.) 0.8506 1 0.0134 o/. ',i o 1 4' 155
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8.0 THERMAL HYDRAULIC CONSIDERATIONS 8.1 Description of the Spent Fuel Pool Cooling System The Spent Fuel Pool Cooling (SFPC) systems are described in detail in Hatch Unit 1 FSAR Section 10.4 and Unit 2 FgR Section 9.1.3. The Hatch Unit 1SFPCsystemincludestwo4.2gx10 Btu /hr capacity cooling trains. Unit 2 has a single 4.25 x 10 Btu /hr cooling train. One of the two Unit 1 trains is devoted to Unit 1 cooling and the other functions as a standby swing cooling train which is designed to operate as part of either the Unit 1 or the Unit 2 SFPC system. Normal make-up water to the spent fuel pool is provided from each unit's condensate storage tank. Plant service water provides a manually initiated backup Seismic Category I make-up source to each spent fuel pool. The SFPC system is not designed to meet Seismic Category I requirements; however, this design was justified, reviewed in detail, and approved by the NRC prior to issuance of the Unit 2 Operating License. InterconnectionoftheResidualHeatRemoval(RHR)systemtogheSFPC system is possible and provides a Seismic Category I, 31.3x10 Btu /hr capacity backup cooling system; i.e. , the RHR system and the SFPC system piping exposed to RHR flow comprise a Seismic Category I method of cooling. Only a fraction of the capacity of the RHR system is required in this mode of operation such that a restricting orifice is provided to limit the amount of water delivered to the fuel pool for cooling. This flow of approxime.tely 1700 gpm uses one RHR pump and one RHR heat exchanger to maintain the pool water under 150 F for maximum heat load conditions when the entire core is discharged to the pool. Under normal conditions, a closed valve and a blind flange provide dual isolation to the inlet and outlet lines connecting the SFPC system to the RHR system. 8.2 Heat Loads and Pool Temperatures for Present Storage Capacity Three design conditions were postulated for the design of the SFPC system:
- 1. Normal Condition The water in the pool is held to 125 F or less with a heat load of 4.25 x 106Btu / hour generated by stored fuel consisting of a 25 percent core that was unloaded from the reactor 30 days before and a 25 percent core that has been in storage for one year from a previous refueling operation. Thirty days after unloading the fuel to the pool, the rate of heat generation from the spent fuel is assumed to approach a constant level. It is also assumed that the 25 percent core unloaded at each refueling outage has had a maximum residence time in the reactor of four years. Under normag conditions, a single cooling train with a capacity of 4.25 x 10 Btu / hour will be suf ficient to maintain the pool water at or below 125 F.
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- 2. Refueling Condition 6
The pool water is held at 125 F or less with a heat load of 8.5 x 10 Btu /hr generated by stored fuel consisting of a 25 percent core that has decayed for 150 hours since reactor shutdow9 plus a 25 percent core in storage for one year from a previous refueling operation. The minimum projected time af ter reactor shutdown to accomplish cooling and opening of the reactor vessel and completion of transferring the spent fuel to the pool is 150 hours. With the assistance of the stydby swing cooling train, a combined cooling capacity of 8.5 x 10 Btu /hr is available to cope with the heat generated by newly unloaded fuel and to hold the pool water at or below 1250F.
- 3. Maximum Condition Thg pool water is held to 150 F or less with a heat load of 31.3 x 10 Btu /hr generated by stored fuel consisting of a 100 percent core unloaded from the reactor plus a 25 percent core held over for one year from a previous refueling. The 125 percent core load is assumed to have undergone the following exposures:
25 percent of core: 4 year exposure + 1 year decay 25 percent of core: 4 year exposure + 150 day decay 25 percent of core: 3 year exposure + 150 hour decay 25 percent of core: 2 year exposure + 150 hour decay 25 percent of core: 1 year exposure + 150 hour decay Under the maximum condition postulated, it is assumed that approxi-mately 150 hours after reactor shutdown the entire core in the reactor will have been transferred to the pool. Thus, tee RHR system will be free for cooling the large fuel load in the pool. With the full core offload plus one quarter core remaining from a previous refueling, a single train of the RHR system, without the assistance of SFPC, will maintain the spent fuel pool temperature at or below 150 F 150 hours af ter the shutdown. Operating experience with Hatch Unit I has indicated that calculated spent fuel pool heat loads and temperatures for the design basis are conservative and the actual heat loads have been approximately 15 percent less than the heat loads calculated. - 8.3 Heat Loads and Pool Temperatures for Increased Storage Capacity 8.3.1 To re-evaluate the Plant Hatch spent fuel pool cooling capabilities with the enlarged storage capacity, the decay heat loads were calculated using methods described by Branch Technical Position ASB 9-2 of the Standard Review Plan. 8.3.2 The pool capacity for the increased storage capacity heat load evalu-ation is assumed to be 5.83 cores. The 5.83 core capacity is arrived 8-2 2 ^\ b
at by assuming 1/4 core yearly offloads to the spent fuel pool up to 5-1/2 cores (22 batches) plus an additional batch (batch 23) of 1/3 core. All batches are assumed to have operated at full power for 90 percent of their four year exposure time. The three design conditions postulated in Section 8.2 are similarly evaluated below. 8.3.2.1 Normal Condition The heat load analysis for the normal operating condition assumed that there were 22 batches in the pool that had decayed from 1 to 22 years, and the latest batch (23) decayed for 30 days. A single spent fuel pool cooling system train was used for decay heat removal. The analysis showed that the heat load was 7.24 x 106 Btu /hr and bulk pool water temperature was at or below 139 F. Heat loads and pool temperatures as a function of refueling batches are shown in Figure 8-1. 8.3.2.2 Refueling Condition The assumptions for the refueling mode analysis were the same as those for the normal mode except that the late: t batch was assumed to have decayed for only 150 hours and two spent fuel pool cooling trains were in service. 6 The analysis showed the heat load was 11.57 x 10 Btu /hr and the bulk water temperature at or below 133 F. Heat loads and pool temperatures as a function of refueling batches are shown in Figure 8-2. 8.3.2.3 Maximum Condition The analysis for the heat load following full core discharge assumed that the pool already had 19 quarter core batches in storage that had decayed from 1 to 196years. The calculated heat load from the 19 batches was 2.39 x 10 Btu /hr. The additional decay heat load at 150 hogrs after shutdown for a full core offload was calculated to be 26.3 x 10 Btu /hr. Therefore, the cumulgtive heat load in the pool at 150 hours after shutdown is 28.69 X 10 Btu /hr. With a single train of the RHR system aligned for fuel pool cooling duty without the assistance of the SFPC system, the system will maintain pool water temperature at or below 145 F. Figure 8-3 shows the heat load as a function of time after shutdown for the full core discharge. As cn alternative to aligning the RHR system to the spent fuel pool for a full core offload, the fuel may be allowed to decay in the reactor vessel until the heat load of the core has decreased to a point where the SFPC system can maintain a temperature less than the design maximum temperature. A waiting time of 500 hours (approximately 21 days) is required in this case prior to full core offload. Af ter this time, two fuel pool cooling trains can maintain the poolwatertemgeratureator below 150 F (i.e. , a heat removal capability of 18.77 x 10 Btu /hr). 8-3 E
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8.3.3 For each design condition analyzed in 8.3.2, completely utilizing the expanded spent fuel pool storage capacity, the present SFPC systems or a single train of the RHR (for the full core of fload condition) are capable of maintaining pool water temperatures less than the design maximum temperature of 150 F. Considering the conservative assumptions used in the calculations and past operating experience, the actual temperatures for each condition are expected to be lower than those calculated and described above. 8.4 Loss of Spent Fuel Pool Cooling The consequences of a loss of the SFPC systams has been evaluated for the following two conditions:
- 1. Concurrent loss of the SFPC systems.
- 2. Maximum heat load.
8.4.1 Concurrent Loss of SFPC Systems Both spent fuel pools are assumed to be loaded as delineated in Section 8.3.2. Unit 1 and Unit 2 are assumed to be shut down for refueling 21 days apart, with Unit 2 being shut down first. Also, 21 days is assumed to be the minimum time required to complete a refueling operation. Therefore, Unit 2 will be operating while Unit 1 is shut down. Subsequently, both units' SFPC systems are postulated to be lost 150 hours af ter Unit 1 is shut down. Calculations using pool water volumes of 38,640 3ft each indicate that the time to boil for the Unit 1 pool is 14.7 hours and that the time to boil for the Unit 2 pool is 22.8 hours. The makeup water requirement following coiling was calculated to be 24 gpm for the Unit 1 pool and 15 gpm for the Unit 2 pool. During transition to boiling, no credit is taken for evaporative heat losses. Water level is maintained by the Seismic Category I Plant Service Water system. Conservatisms are included in the analysis by assuming that all decay heat is rejected to the pool water and none is rejected to the structures. Also, the heat capacity of the makeup water is neglected. Af ter approximately 150 hours following Unit 1 shutdown, the decay heat contributed by 2/3 core in the Unit 1 reactor pressure vessel has decreased enough to allow aligning one train of the RHR to provide spent fuel pool cooling and reactor pressure vessel cooling. With tl a reactor vessel head and the spent fuel pool gates removed, the RHR system can ae aligned for spent fuel pool and reactor pressure vessel cooling by installation of two spectacle flanges and operation of four isolation valves. The time required for realignment is estimated to be 8 hours. Subsequent to loss of the SFPC systems, Unit 2 will be brought to cold shutdown. A radiological analysis has been performed assuming that both pools boil simultanecusly. The consequences are presented in Section 8.6. c '\ u.1 d } b b O
8.4.2 Maximum Heat Load A full core offload creates the highest heat load in the spent fuel pool. However, with no fuel in the reactor pressure vessel, the RHR system is available for unrestricted spent fuel pool cooling. The redundant Seismic Category I design of the RHR system provides a high degree of assurance that it operates satisfactorily in the spent fuel poel cooling mode. 8.5 Local Fuel Bundle Thermal Hydraulics The bounding thermal-hydraulic conditions were calculated for fuel stored in a HDFSS module in the Hatch pools. Bases for the calculations for typical current generation fuel were the following: Maximum bundle burn-up 35,300 MWD /MTU Specific Power 36.6 kW/kgU 20% of time 48.3 kW/kgU 60% of time 60.0 kW/kgU 20% of time The ORIGEN Code (Reference 4) was used to calculate the decay heat for the bundle defined by these bases. The result was: Actinide Contribution 9,500 W/MTU Fission Product Contribution 152,000 W/MTU TOTAL 161,500 W/MTU With the bulk water temperature of the spent fuel storage pool constant at 140 F, the maximum fuel cladding temperature will be 186.1 F. The maximum water temperature associated with the hottest fuel bundle will be 163. 2 F. These temperatures and the maximum storage tube wall temperature of 157.5 F are low relative to structural integrity or corrosion limiting temperatures of the structural components of the storage system and fuel. A second set of calculations bracketed the thermal hydraulic conditions expected in potential future fuels. The bases used for these calculations were: Maximum bundle burn-up 44,000 MWD /MTU Specific Power 20.3 kW/kgU 20% of time 40.2 kW/kgU 60% of time 60.0 kW/kgu 20% o' time 2104 lo7 8-5
The ORIGEN Code was used as before, but the initial U 235 content was adjusted to 3.6 weight percent to correspond to the higher burn-up value. The decay heat calculated was: Actinide Contribution 10,700 W/MTU Fission Product Contribution 155,000 W/MTU TOTAL 165,700 W/MTU These values result in a maximum fuel cladding temperature of 186.6 F. The maximum water temperature will be 163.6 F and the maximum storage tube wall temperature till be 157.7 F. There is no thermal-hydraulic problem presented by potential future high burn-up fuels. Continuing efficiency of the exchange of heat from the spent fuel to the pool water depends on the convection flow of wa ar through the storage tube and flow channel, if present, encompassing a fuel bundle. The floc-like crud that adheres to the surfaces of the spent fuel bundles was studied to determine whether it is a potential mechanism for blocking flow through the channel. The floc was found to be extremely fine; pieces that spall off of the aggregate are not disposed to settle, but will flow upward with the convection current. Additionally, the floc is so fine that some of it will pass through conventional laboratory fil'er papers. Growth of crud in fuel storage conditions has not been observed ia commercial facilities. The potential for channel plugging by se,"- " or by blockage of ficw passages is therefore negligible. 8.6 Radiological Impact. 7 eat Fuel Pool Boiling The radiological impact of spent fuel pool boiling is maximized by assuming simultaneous failures of the SFPC systems for both Units 1 and 2 as described in Section 8.4.1. A radiological analysis nas been performed to determine the thyroid dose at the site boundary /LPZ, assuming that the pools boil and that there has been an iodine spike in the pools. The assumptions used are as follows:
- 1. The time to reach boiling is 14.7 hours for Unit 1 and 22.8 hours for Unit 2.
- 2. Boilina rate of the pool water is 11,955 lb/hr for Unit 1 and 7700 lb/hr for Unit 2.
'2 Volume of water in each pool is 38,640 f t3,
- 4. All failed fuel rods of the full core (average 1 percent of the core) are present in the 1/3 core discharged to each pool.
- 5. ThenormalI-131releasf0 rate.poefficientforleaking_gdsinthe y Unit 1 pool is 4.6 x 10 sec at 150 hours and 1 x 10 sec 8-6
for leaking rods in the Unit 2 pool at 27.25 days (21 days + 150 hours) using the methods described in Reference 5. These release rate coefficients are conservatively assumed to be constant during the heatup and boiling periods.
- 6. The above releas rate coefficien+ i spiked by a factor of 100 to simulate the hei tup conservatively.
- 7. The decontamination factor for I-131 during boiling is conserva-tively assumed to be unity.
- 8. No credit is taken for iodine plate-out or filtration by the standby gas treatment system.
- 9. Conservative ground level accident X/Q values are assumed for the dose calculation.
The results are summarized below: Site boundary /LPZ thyroid dose (0-2 hrs. ) 1.5 rem Site boundary /LPZ thyroid dose (0-4 days) 9.3 rem The above results,which are based on boiling of both Unit 1 and 2 pools, compared to the results presented in the Hatch 2 FSAR (response to Question 20.20 - 1.3 rem for 0-2 hours and 8.3 rem for 0-4 days), which are based on boiling of the Unit 2 pool only, support the Applicant's position that the SFPC system need not be upgraded to meet Seismic Category I design requirements. O
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9.0 COST BENEFIT ASSESSMENT 9.1 Need for Increased Storage Capacity The present spent fuel storage facilities at Hatch Units 1 and 2 were designed for temporary storage of spent fuel until the fuel had de-cayed enough for safe transport to a reprocessing facility. The absence of activity in the construction of new fuel reprocessing facilities and the cessation of operation of exisiting reprocessing
. facilities have created the need for increased on-site storage of spent fuel to permit long-term power plant operation.
The terms of Georgia Power Company's 1968 fuel supply contract with General Electric provide for the buy back and removal of the first two cores of Hatch-1 fuel by General Electric. Georgia Power and General Electric have agreed to defer removal of this fuel and to temporarily store this fuel at Plant Hatch. Georgia Power also entered into a reprocessing contract with General Electric covering spent fuel dis-charged for reprocessing through 1983. In 1974, General Electric informed Georgia Power that its Morris, Illinois, reprocessing facil-ity was inoperable and that the contract was being terminated. The anticipated spent fuel discharge schedule for the Hatch Nuclear Plant is described in Table 9-1. A review of the schedule indicates that, with the present storage rack configuration, full core storage reserve caLaoility will be lost in 1983 and all storage capacity will be expended in 1985. This prediction is based on maintaining reserve storage for a single core using the combined storage capacities of both spent fuel pools. This is possible because Unit 1 and Unit 2 share 6 common refueling floor and a transfer canal which connects the two spent fuel pools. Expansion of the storage capacity in both pools by using the General Electric designed high-density, poisoned storage racks will provide enough reserve storage capacity for off-loading a full core until 1997 and will provide spent fuel storage without a full core reserve until 1999. Presently, the Hatch spent fuel pools contain the following items in addition to the fuel and fuel racks: 2 control rod assemblies (Unit 1 pool) 8 control rod blade guides (Unit 1 pool) 140 control rod storage locations, Unit 1 40 control rod storage locations, Unit 2 Test weights for the fuel handling bridge, Unit 1 and Unit 2 (Unit 2 weights to be removed) 2104 173 9-1
Underwater vacuum cleaner (Unit 1) Miscellanc,us other equipment such as fuel sipping ca., sters which are temporarily located in the pool for outage work. 9.2 Alternative to Increasing Storage Capacity 9.2.1 Several alternatives to the expansion of the storage capacities of the Hatch Unit 1 and Unit 2 spent fuel pools to alleviate the spent fuel a storage space storage were considered. In summary, the alternatives were:
- a. shipment to a fuel reprocessing facility.
- b. shipment to an independent spent fuel storage facility.
- c. shipment to another reactor site.
- d. shutting down the reactor.
9.2.1.1 Shipment to a Fuel Reprocessing Facility There are currently .w cammercial spent fuel reprocessing facilities in operation in the United States. In April 1977, the President of the United States announced a spent nuclear fuel policy which included the indefinite deferral of commercial reprocessing in the U.S. nuclear power program. Reprocessing of spent fuel is not a viable alternative to the expansion of the Hatch spent fuel pools. Storage of the Hatch spent fuel at the existing (although not operating) reprocessing facilities is also not a viable alternative to the expansion of the Hatch spent fuel pools since the facility owners are not offering to provide comparable storage capacity. 9.2.1.2 Shipment to a Storage Facility Spent fuel storage at a private or government operated independent spent fuel storage facility is not currently available. The alternative of constructing a facility to serve Plant Hatch would not be economically viable. The Department of Energy has estimated that construction of a 5000 MTU independent spent fuel storage facility would cost
$200,000,000 (DPE/ET-0055 " Preliminary Estimates of Charge for Spent-Fuel Storage and Disposal Services", July 1978) or about $40/kg. A smaller facility designed to serve Plant Hatch would be expected to have a higher cost per kg. These costs are significantly larger than the estimated cost of the increased storage capacity which will be obtained by expanding the present reactor pools (approximately $12.5/kg).
9.2.1.3 Shipment to Another Reactor Site The only available reactor site which could be used as an alternative for Plant Hatch spent fuel storage facility within Georgia Power 9-2 3 a 17 4 e i
Company is Vogtle Nuclear Plant (a PWR) Unit 1 which has an expected inservice date of November 1984. This schedule cannot prevent Plant Hatch from losing its full core reserve capacity in 1983; neither can it alleviate the Plant Hatch spent fuel storage problem until the back-end-of-fuel-cycle problems are resolved. However, even if Plant Vogtle were used as an alternative site for Plant Hatch spent fuel storage, the estimated cost would be greater than that of expanding the Hatch pools, as shown below. The costs do not reflect the loss of storage space at Plant Vogtle.
- 1. Cost of BWR spent fuel storage racks $1,300/ assembly Installation (9%) 120 Contingencies (10%) 130 Engineering, supervision, and overhead (including licensing) (20%) 250
$1,800/ assembly
- 2. Cost of transportation (with cask rental) $1,200/ assembly
- 3. Total Cost $3,000/ assembly (approximately $16/kg) 9.2.1.4 Plant Shutdown Shutdown of the Hatch Nuclear Plant would require the purchase of power from substitute sources and/or production from less economical sources within the system. The figures shown in Table 9-2 are the increased production costs (actual year dollars) to the Southern electric system for replacement power if Unit 1 and Unit 2 are closed af ter the 1983 refueling. These figures do not include any capital (fixed) cost dollars that still would have to be amortized whether the plant is operating or not. Also not included is the cost of maintaining the plant in a shutdown condition and maintaining site security.
9.3 Capital Costs Costs incurred by expanding the spent fuel storage capabilities at the Hatch Plant are summarized on Table 9-3. These costs represent the current prediction of the total project costs, including the installa-tion of the high density spent fuel storage racks and disposal costs of the presently installed racks. Indirect capital costs other than those specified have not been considered. The overall scope of the project will include the following: o, 'i n
.a \ 17 5 9-3
- a. Design feasibility study.
- b. License avendment preparation and submittal.
- c. Engineering studies to support license amendment including nuclear analysis, seismic analysis, and thermal-hydraulic analysis.
- d. Installation preparation, including removal and disposal of origi-nal racks, hold-down clips, seismic restraints, etc..
- e. Installation of new racks.
- f. Development and implementation of poison verification procedures.
9.4 Resource Commitment The relatively small quantities of material resources that would be committed to the proposed modification would not significantly fore-close the alternatives available with respect to any other licensing actions designed to ameliorate a possible shortage of spent fuel storage capacity. The material rescerces that would be consumed by the proposed modification are listed below. Hatch Modification Material Quantity (lb) 304 Stainless Steel 5.8 x 10 5 Boron Carbide 1.4 x 10 4 Aluminum 5.1 x 10 4 9.5 Environmental Impact of Expanded Spent Fuel Storage An analysis of the Hatch Unit 1 spent fuel pool heat load when filled to the present 1.5 core capacity, 30 days after the last refueling shutdown, indicates that the bulk spent fuel pool temperature will be approximately 127.5 F. The bulk temperature of the Unit 2 spent fuel pool when filled to its 2.0 core capacity will be approximately 128 F. For the proposed expanded capacity, assuming that the spent fuel pools are filled to their expanded capacity 30 days af ter the last refueling shutdown, as previously discussed, each reactor building closed cooling water (RBCCW) heat exchanger inlet temperature can be expected to rise less than 1.5 degrees. The total evaporation rate of the two spent fuel pools can be expected to increase by 340 lb/hr. Each unit has a once-throcqn refueling floor ventilation system with a 30,000 cfm capacity for a cabined ventilation capacity of 60,000 cfm. The increased evaporation rate will have negligible effect on the refueling floor ventilation systems and, therefore, no effect on the environment. j)Qk \ 9-4
Assuming that all additional heat transferred to the RBCCW system is ultimately transferred to the plant service water system and assuming that no heat is lost through piping or components, the plant service water discharge temperature will be increased by approximately 0.6 F. Therefore, under normal conditions, the spent fuel pool storage expan-sion will have negligible effect on the operation of installed plant components and negligible impact on the environment as a result of increased heat loads. 9
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Table 9-1 Estimated Spent Fuel Discharge Schedule Annual Discharge Schedule (No. of Assemblies) Cumulative Discharges Year Unit 1 Unit 2 Combined (No. of Assemblies) 1977 92 92 92 1978 168 168 260 1979 164 164 424* 1980 140 168 308 732 1981 140 140 280 1012 1982 140 140 280 1292 (1) 1983 140 140 280 1572 1984 140 140 280 1852 (2) 1985 140 140 280 2132 1986 140 140 280 2412 1987 140 140 280 2692 1988 140 140 280 2972 1989 140 140 280 3252 1990 140 140 280 3532 1991 140 140 280 3812 1992 140 140 280 4092 1993 140 140 280 4372 1994 140 140 280 4652 1995 140 140 280 4932 1996 140 140 280 5212 (3) 1997 140 140 280 5492 1998 140 140 280 5772 (4) 1999 140 140 280 6052 2000 140 140 280 6332
- Presently in storage (1) Existing storage capacity - loss of full core reserve (3) Existing storage capacity - filled (3) Expanded storage capacity - loss of full core reserve (4) Expanded storage capacity - filled
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Table 9-2 Replacement Power Costs in Actual Year Dollars Differences In Cost . - If Generated In The Difference In ,, Total Cost Southern Electric System + Emergency Energy X Combustion Turbine X 112% = Difference Year x $1,000 Purchased * (GWH) Generation ($/MWH) x $I,000 s - 3/83-12/83 $150,805.00 157.0 80.86 $165,023.00 1984 207,193.00 72.1 86.46 214,175.00 1985 197,139.00 57.0 92.57 203,049.00 1986 108,872.00 171.3 98.59 227,787.00 1987 213,569.00 236.4 105.39 241,473.00 1988 226,721.00 231.1 112.93 242,665.00 1989 226,721.00 473.6 120.25 290,505.00 1990 239,770.00 563.8 128.97 321,209.00
*This energy would be purchased outside the Southern electric system, if available.
- Price is assumed to be 112% of Georgia Power's most expensive combustion turbine generation.
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Table 9-3 Capital Costs New Fuel Storage Racks for Spent Fuel Pool $6,100,000 Installation (including Disposal Costs) $ 552,000 g Contingencies (10%) $ 665,000 Engineering, Supervision, and Overhead (including Licensing and Legal Fees) (20%) $ 1,330,000 Subtotal: $ 8,647,000 Assuming expenditures of 25 percent in 1979, 50 percent in 1980, and 25 per-cent in 1981, escalation should result in the following additional changes: 1979 Escalation (10%) $ 648,000 1980 Escalation (10%) $ 432,000 1981 Escalation (10%) $ 216,000 SUBTOTAL: $1,296,000 Allowances for funds used during construction are calculated on a cumulative percentage basis and result in the following additional charges: 1979 $ 67,500 1980 $275,500 1981 $484,000 SUBTOTAL $ 827,000 Adding all of the above costs results in a total budget projection for the project of:
$10,770,000 3
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10.0 RADIOLOGICAL EVALUATION 10.1 Spent Resin Waste The fuel pool filter-demineralizer units are designed to maintain a water conductivity of less than 0.5 micro mho/cm. The units are backwashed when either the differential pressure across the demineral-izers is greater than 10 psi or the effluent conductivity is greater than 5 micro mhos. Hatch Unit 1 experience indicates that the filter-demineralizer was backwashed 41 times during 1978. Each backwash cycle amounts to 2.5 cubic feet of spent resin. The dose attributed to handling of the spent fuel pool resin in the radwaste system is approximately 0.3 man-rem /yr. The increase in the spent fuel pool storage capacity is not expected to appreciably affect the annual amount of solid radwaste or the annual man-rem dose. 10.2 Noble Gases Krypton-85 is released to the pool water and subsequently to the refueling floor atmosphere from the leaking fuel assemblies. For normal operating conditions, most of the krypton comes from the most recently discharged batch of fuel. After the most recent batch has cooled in the pool for 12 months, the pressure buildup in a fuel pin which causes the release of krypton has become very small. Thus, the increase in krypton-85 activity attributed to the increase in spent fuel pool storage capacity will be small compared to the total quan-tity of all noble gases released form the pools and negligible when compared to the annual plant noble gas releases. Despite the presence of some defective fuel bundles in the Unit 1 pool, krypton-85 activity levels in the refueling floor ventilation exhaust are below the mini-mum detectable level of approximately 108uCi/cc. 10.3 Gamma Isotopic Analysis for Pool Water Hatch Unit I has undergone three refuelings. Typical radioactive isotope concentrations in the Unit 1 spent fuel pool water are pre-sented in Table 10-1 at various dates. 10.4 Dose Levels Over and Along Sides of Pool Dose surveys at Hatch Unit 1 indicate that after every refueling outage the radiation field over the pool surface has returned to an apparent equilibrium of approximately 1 mr/hr. Local areas show 4 mr/hr (e.g., around the fuel grapple). Measurements taken during the May 1979 refueling outage show that the radiation levels along the sides and center of the pool are essen-tially the same (approximately 2 mr/hr). This indicates there has been no significant crud build up around the sides of the pool and the radiation levels are as low as reasonably achievable. 10-1 }}Od kb
10.5 Airborne Radioactive Nuclides Air samples taken from the Unit 1 refueling floor atmosphere during and after each refueling showed activity levels below the lower level of detection. Storage of additional fuel is not expected to increase the airborne activity on the refueling floor since the major contribution of airborne activity is attributed to the most recent batch of spent fuel that is placed in the pool. 10.6 Radiation Protection Program The Radiation Protection Program is described in Section 12.5 of the Hatch 2 FSAR. This program will be adhered to during the removal of the old racks and installation of the new racks. 10.7 Disposal of Present Spent Fuel Racks There are at present 42 aluminum racks in the Unit 1 pool and 56 in Unit
- 2. Each rack weighs about one ton. Presently, there is no fuel stored in the Unit 2 spent fuel pool. The racks removed from Unit 2 will be prepared and stored in the warehouse for future sale or use. The racks from the Unit 1 pool will be decontaminated, crated and shipped offsite to a licensed burial location. A reasonable effort will be made to limit personnel exposures to as low as reasonably achievable during this work.
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Table 10-1 Radioactive Isotopic Concentrations in the Spent Fuel Pool Water Isotope Fuel Pool Activity (uCi/cc) 7/11/78 1/15/79 5/8/79 5/29/79 I-131 LLD* LLD 6.76E-5 LLD Xe-133 LLD LLD 6.06E-5 LLD Mo-Tc-99m LLD LLD 1.25E-5 LLD Cr-51 LLD LLD 6.15E-4 LLD F-18 LLD LLD 3.49E-5 LLD Cs-134 1.03E-5 5.73E-6 1.74E-4 3.58E-5 Cs-137 1.89E-6 7. 64 E-6 1.46E-4 3.72E-5 Zr-95 LLD LLD 1.40E-4 7.4E-6 Nb-95 LLD LLD 1.55E-4 9.7E-6 Co-58 4.19E-6 8.29E-7 5.0E-5 1.09E-5 Mn-54 LLD LLD 6.77E-5 7.33E-6 Fe-59 LLD LLD 5.68E-5 LLD Zn-65 3.00E-5 4.1E-5 5.65E-4 6.94E-5 Co-60 5.1E-6 5.31E-6 1.58E-4 8.88E-6
- Lower level of detection
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11.0 ACCIDENT EVALUATION The spent fuel shipping cask drop analysis is described in the Hatch Unit 1 FSAR Question 10.3.4 response. The referenced drop analysis is applicable to Unit 2. Since the cask will not be handled over or in the immediate vicinity of either the Unit 1 or the Unit 2 spent fuel pool, the consequences of the cask drop are not affected by the installation of additional spent fuel storage capacity. Protection against the cask drop is afforded by the licensed single failure proof crane described in Hatch Unit 1 FSAR Section 10.20, by the single failure proof cask yoke described in Hatch Unit 2 FSAR Subsection 9.1.4.2.2, and by the interlocks and administrative con-trols described in the same subsection which limit the cask height over the refueling floor during cask handling operations. The Hatch Nuclear Plant design also incorporates several levels of protection against the drop of other crane loads into the spent fuel pool and onto stored spent fuel. The overhead crane is interlocked to prohibit operation over the spent fuel pool. The interlocks can bi overridden, but only under strict administrative controls. The only postulated loads which would require bypassing the interlocks which prohibit movement over the spent fuel pool are the handling of the spent fuel pool plugs (9 tons) and gates, and removal and installation of the old and new spent fuel racks, respectively, as discussed in Section 6.0. The spent fuel pool gates and plugs will be handled only under strictly controlled adminis-trative procedures. Additional information pertaining to the control of heavy loads near spent fuel has previously been discussed in Ref-erence 6. If unanticipated load handling should occur, the size of the load that can be handled over stored spent fuel, by any means, is limited to 1600 pounds by Hatch 2 Technical Specification 3/4.9.7. A proposed change to the Unit 1 Technical Specifications will be submitted to incorporate this same requirement. 21u,5 }Ok 9 11-1
12.0 CONCLUSION
S The information contained in this document to support the proposed modification satisfies the necessary applicable regulatory requirements to allow NRC approval for Georgia Power Company to rerack the Plant Hatch Units 1 and 2 spent fuel pools and demonstrates that the proposed modification can be safely accomplished. This proposed modification is the ,,iost cost effective and desirable alternative, and is in the best interest of the public. The proposed modification does not significantly change cr impact any previous determinations which are dccumented in the Hatch 1 and 2 Safety Evaluation Reports and Final Environmental Statements, and therefore precludes the need for preparation of an environmental impact statement.
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13.0 NOTES M4D REFERENCES Notes:
- 1. For the purposes of this report the term " fuel bundle" will imply configuration either with or without flow channels unless the term
" fuel assembly" is specifically and distinctly intended.
- 2. Boral is a product of Brooks and Perkins, Inc. , consisting of a layer of boron carbide-aluminum (8 C-A1) 4 matrix bonded between two layers of aluminum.
References:
- 1. L. K. Liu, " Seismic Analysis of the Boiling Water Reactor," Sym-posium on Seismic Analysis of Pressure Vessel and Piping Component, First National Congress on Pressure Vessel and Piping, San Francisco, California, May 1975.
- 2. U.S. NRC Safety Evaluation for Yankee Rowe, dated December 29, 1976, Page 4, Structural and Material Considerations.
- 3. C. M. Kang and E. C. Hanson, ENDF/B-IV Benchmark Analysis with Full Spectrum Three-Dimensional Monte Carlo Models, ANS Meeting, November 1977.
- 4. M. J. Bell, "0RIGEN Code - The ORNL Isotope Generation and Deple-tion," 0RNL-4628.
- 5. N. Eickelpasch and R. Hock, " Fission Product Release After Reactor Shutdown," I AEA-SN-178/19.
- 6. Letter from W. E. Ehrensperger, Georgia Power Company, to U. S Nuclear Regulatory Commission, dated July 24, 1978.
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