ML20041G432
| ML20041G432 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/16/1982 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20041G431 | List: |
| References | |
| NUDOCS 8203220291 | |
| Download: ML20041G432 (184) | |
Text
{{#Wiki_filter:4 ATTAc.sNssNr i 4 BRUNSUICK STEMi ELECTRIC PLANT UNIT NOS. 1 AND 2 SPENT FUEL STORAGE EXPANSION REPORT 8203220291 820316 PDR ADOCK 05000324 p PDR
TABLE OF CONTENTS Page No. 1.0 I NTR0 DU C I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 OVERALL DESCRIPTION........................................ 2-1 3.0 DESIGN BASES............................................... 3-1 4.0 MECHANICAL AND STRU CURAL CONSIDERATIONS................... 4-1 4.1 SEISMIC ANALYSIS........................................... 4-1 4.2 STRESS ANALYSIS............................................ 4-3 4.3 FUEL BUNDLE / MODULE IMPAC EVALUATION. . . . . . . . . . . . . . . . . . . . . . . 4-6 5.0 MATERIAL 0)NSIDERATIONS.................................... 5-1 6.0 INSTALLATION............................................... 6-1 7.0 NUCLEAP CONSIDERATIONS..................................... 7-1 7.1 NEUTRON MULTIPLICATION FA COR... .......................... 7-1 7.2 INPUT PARAMETERS........................................... 7-1 7.3 GEOMETRY, BIAS, AND UNCERTAINTY............................ 7-2 7.4 INTERACION WITH EXISTING STORAGE SYSTEM. . . . . . . . . . . . . . . . . . . 7-3 7.5 POSTULATED ACCIDENTS....................................... 7-4 8.0 THERMAL ANALYSIS........................................... 8-1
8.1 DESCRIPTION
OF THE SPENT FUEL POOL COOLING SYSTEM.......... 8-1 8.2 S E CT IO N DELET E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2 8.3 HEAT LOAD AND POOL TEMPERATURE FOR EXPANDED STORAGE CAPACITY................................................... 8-2 i Revision 1
TABLE OF CONTENTS (Continued) Page No. 8.4 LOSS OF SPENT FUEL POOL C00LINC............................ 8-4 8.5 LOCAL FUEL BUNDLE THERMAL HYDRAULICS....................... 8-6 8.6 RADIOLOGICAL IMPACT OF SPENT FUEL POOL B0ILINC............. 8-6 9.0 COST BENEFIT ASSESSMENT.................................... 9-1 9.1 NEED FOR INCP. EASED CAPACITY................................ 9-1 9.2 ALTERNATIVES TO INCREASED CAPACITY......................... 9-1 9.3 CAPITAL'CdSTS.............................................. 9-2 9.4 RESOURCE COMMITMENT........................................ 9-3 9.5 ENVIRONMENTAL IMPACT OF EXPANDED SPENT FUZu STORACE........ 9-3 10.0 RADIOLOGICAL EVALUATION.................................... 10-1 10.1 SPENT RESIN WASTE.......................................... 10-1 10.2 NOBLE CASES................................................ 10-1 10.3 CAMMA ISOTOPIC ANALYSIS FOR POOL WATER..................... 10-1 10.4 DOSE LEVELS................................................ 10-2 10.5 AIR 30RNE RADIOACTIVE NUCLIDES.............................. 10-2 10.6 RADIATION PROTECTION PR0 CRAM............................... 10-2 10.7 DISPOSAL OF PRESENT SPENT FUEL RACKS....................... 10-2 10.8 IMPACT ON RADIOACTIVE EFFLUENTS............................ 10-2 11.0 ACCIDENT EVALUATION........................................ 11-1 11.1 SPENT FUEL SHIPPING CASK DROP - OUTSIDE OF FUEL P00L. . . . . . . 11-1 11.2 SPENT FUEL SHIPPING CASK DROP - OVER SPENT FUEL P00L....... 11-1 11
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4 i l TABLE OF CONTENTS (Continued) 1 l Page No. 11 3 OTHER CRANE L0 ADS.......................................... 11-1 11.4 RADIOLOGICAL IMPACT........................................ 11-1 i 1 4
12.0 CONCLUSION
S................................................ 12-1 i 1 i j 13.0 NOTES AND REFERENCES....................................... 13-1 4 i t e i i e i i ? i i i 1 ,I 1 i ) ,Y a 1 111 h
1.0 INTRODUCTION
This design report and safety evaluation considers the installation of high density, poisoned fuel storage modules in the existing spent fuel pools of Brunswick Steam Electric Plant (BSEP) Units 1 and 2. The BSEP 1 and 2 spent fuel pools combined currently contain modules that c an I hold 2088 BWR and 304 PWR fuel assemblies. It was originally assumed that about one quarter of the core would be discharged annually and that spent fuel would be removed from the plant for reprocessing within approximately a year after discharge from the reactor. When reprocessing was not available in 1377 and 1978, the original modules were replaced with the present modules. Because the reprocessing option is still not available at this time, the storage capacity of the spent fuel pools is proposed to be expanded by replacing some of the existing spent fuel storage modules with high density, poisoned codules. It is desirable to have enough capacity in reserve to allow for a full-core discharge. Such capacity was lost in the Unit I spent fuel pool subsequent to its 1980 refueling. The high density spent fuel storage modules will provide a total of 1803 storage spaces in BSEP 1 and 1839 in BSEP 2 for BWR assemblies. The modification will provide storage capacity until 1988 for BSEP 1 and 1987 for BSEP 2 with a full-core reserve, assuming annual quarter core reloads. This report describes the design of the high density fuel storage modo'es to be installed and contains a discussion of the environmental and radiological considerations of the installation. The information contained herein has been prepared based on the recommendations provided in " Operating Technical Position for Review and Acceptance of Spent Fuel Storage and Randling Applications" which was issued by the Nuclear Regulatory Commission (NRC) on April 14, 1978 and later amended on January 13, 1979. General Electric Company will design and supply the high density, poisoned spent fuel storage modules that will be installed at BSEP 1 and 2. Similar storage modules have previously been reviewed and approved by the NRC. I 1-1
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2.0 OVERALL DESCRIPTION The location of the spent fuel storage pool within the plant is shown in Figures 2-1 through 2-2. The existing spent fuel storage module arrangement and module support grid are shown in Figures 2-3 and 2-4. The arrangements of the proposed new high-density fuel storage (HDFS) modules for the pools are shown in Figure 2-5. The HDFS module (free standing on the floor of the fuel pool) provides storage spaces for fuel bundles or fuel assemblies (See Note 1, Sec. 13.0) on approximately 6.6 in. center to center spacing. For each pool, four basic l storage module sizes, 13x15, 13x17, 13x19, 15x17, are planned. The expansion allows 10 full spacas and 2 half spaces of the existing 38 1/2 spaces in the I support grid to accommodate 731 instead of 396 BWR assemblies and also adds additional spaces to store 442 more BWR assemblies in the area presently used for control rod storage. (See Figure 2-5). Module Fuel Configuration Capacity Ouantity Assemblies Unit 1 13 x 15 195 1 195 13 x 17 221 1 221 13 x 19 247 1 247 15 x 17 2$$ 2 510 SUBTOTAL 1173 l Unit 2 13 x 15 195 1 195 13 x 17 221 1 221 13 x 19 247 1 247 15 x 17 255 2 510 SUBTOTAL 1173 l TOTAL 2346 Following installaticn of the above new modules, the maximum combined pool storage capacity will be 3642 BWR assemblies (1803 in Unit I and 1839 in Unit 2), and 304 PWR assemblies (160 in Unit 1 and 144 in Unit 2). Each HDFS module is fabricated from fuel storage tubes, made by forming an outer tube and an inner tube of 304 stainless steel with an inner core of Boral (see Note 2, Section 13.0) into a single tube. The outer and inner tubes are welded together af ter being sized to the required dimensional tolerances. The completed storage tubes are fastened together by angles welded full length along the corners and are attached to a base plate to form storage modules. Figure 2-6 shows a typical HDFS module schematically. The l smallest module (13x15) is a rectangular array approximately 7 feet by 8.25 feet and 14-feet high and contains 195 storage spaces. l The module support system consists of a module base plate, four foot pad assemblies and four support pads. Figure 2-7 illustrates tne module support system. The support pads rest on the pool floor and are elevated 22-inchas l above the floor to bridge the existing support grid. No additional loading is 2-1 Revision 1
applied to the grid as a result of the installation. The support pads are fabricated from 2-inch-thick stainless steel plates and are designed to be rigid under the module load. No modification is necessary to the existing grid restraint system bracing and crossbeam. The foot pad assemblies are bolted to the module base plate at each of the four corners. The foot pad consists of a 3/4-inch-thick by 15-inch-diameter plate made from special low-friction material pressed into a 1/2-inch deep circular recess in a 2-inch-thick stainless steel foot pad base. The 2-inch stainless steel base is 19-3/4-inches square with 10-1/2-inches at 45' cut off one corner to match the module base plate. The foot pad bears on the support pads. These are the only sliding surfaces for the modules. See Figure 2-8. The module base plate is fabricated from 1-inch-thick stainless steel plate. l The long castings and short closure plates on the perimeter of the module are welded to the base plate. See Figure 2-7. The gap between modules is a minimum of 2 inches. There are no other gaps in the module construction. i 2-2 Revision 1
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4 BRUNSWlCK STEAM ELECTRIC PLANT FIGURE Proposed High Density e.uel Carchna Power & Lignt Company Storage Modules Arrangement , . SPENT FUEL POOL STORAGE EXPANSION
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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 regulations:
- 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 environmental conditions associated with normal operation and postulated accidents.
- 4. General Design Criterion 62 as related to the prevention of criticality 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 classification 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 Position ASB 9-2 contained in the Standard Review Plan.
- 11. Light-Cage Cold-Formed Stael Design Manual,1961 Edition, American Iron and Steel Institute.
- 12. 10CFR100.
3-1
4.0 MECHANICAL AND STRUCTURAL CONSIDERATIONS The high density fuel storage (HDFS) module has been analyzed for both Operating Basis Earthquake (OBE) and Design Basis earthquake (DBE) conditions. A stress analysis has been performed to check the design adequacy of the module against calculated loads. Results indicate that the HDFS module design is adequate for the postulated combined loading conditions. 4.1 SEISMIC ANALYSIS The HDFS module has been analyzed for both OBE and DBE conditions. Critical damping ratios of 2 percent were used in the analysis for the DBE condition and 1 percent for the OBE condition. The design floor acceleration response spectra are given in Figures 4-1 and 4-2. Combination of the modal response and the effect of th, three components of an earthquake was performed in accordance with the applicable provisions of US NRC Regulatory Guide 1.92. The seismic analysis of the module 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 cae hydrodynamic virtual mass terms based on the module and pool configuration. Three-dimensional end effects and leakage between modules were accounted for by modifying the calculated hydrodynamic mass. Figure 4-3 shows the plan view of the two-dimensional model of the modules and l pool used in the hydrodynamic virtual mass analysis. The model consists of two rigid bodies: the module and the pool walls. The walls are considered rigid because their substantial thickness makes them considerably stiffer than the module and the water in the pool. The distance between the modules and the walls of the pool is very large compared to the magnitude of the deflection of the module walls and the pool walls during a seismic event. Consequently, the assumption that both bodies are rigid does not sf gnificantly affect the hydrodynamic mass contribution. In addition, ignoring the flexibility of the wall will result in higher hydrodynamic mass. This will result in a lower natural frequency of the module. Because of the shape of the floor spectra, underestimating the natural frequency of the module provides a conservative estimate of stresses and displacement of the module. Water finite elements fill the spaces in between the walls and the modules. The modeling considered the effect upon the hydrodynamic mass of the adjacent existing storage baskets which are fixed to the upper grid support on the floor of the pools. 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 DBE are applied without 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 rectangular body. This program has been design reviewed and meets NRO-QA requirements. The methodology in calculating hydrodynamic mass has been presented in Reference 4-1. 4-1 Revision 1
y Second, the derived total mass of the module was used to perform dynamic analysis for the OBE and DBE. Third , both finite-element and lumped-mass models of a module were 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 base plate. The sliding analysis for the RDFS module model is repr3sented by a triangle with three masses. This model preserves the overturning and tilting moment of l the rectangularly shaped module. A rectangular model with more mass will not produce higher effects. Thus, there would be no difference in results if a rectangular model was used. In the nonlinear analysis used to calculate the amount of sliding and tilting of a HDFS module, a two-node lumped-mass model was found to adequately represent the module and support system analyses, since the response of the module support system was shown to be primarily first mode and rigid body motion and both the first mode and rigid body dynamic properties could be simulated. The lumped mass at the top of the two-mass model was selected so that the base shear force of the first mode was pr . .rved. The height of the model was selected to preserve the overturning moment at the base of the module for both the first mode response and rigid body motion. The summation of the two lower masses and the upper mass used in the model equals that total mass of the module. The distance of the two lower masses was selected to preserve the mass moment of inertia of the module. This ensured that the shear force at the base was preserved for rigid body motion. Finally, the stiffness of the structural element was selected to preserve the fundamental frequency of the module. 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 stif fness of tha 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 0.108 and 0.222. The friction coefficient between the stainless-steel support pads and the stainless-steel liner is at least 0.3 49. 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 (References 4-2 and 4-3). The sliding analysis was done using the two-dimensional, non-linear DRAIN-2D anc SEISM computer codes. DRAIN-2D was originally developed at the University of California at Berkeley; SEISM was developed by CE. Both computer codes have been design reviewed and meet NRC-0A requirements. Sliding and overturning of the module were studied for the DBE and OBE conditions. All of the modules were found to be stable under the worst postulated seismic loading 4-2 Revision 1
conditions, and the minimum 2-inch clearance between modules precludes contact during a seismic event. 4.2 STRESS ANALYSIS The HDFS module has been designed to meet Seismic Category I requirements. Structural integrity of the module 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. c) BSEP-1&2 Final Safety Analysis Report, Seismic Design Criteria. d) OBE - Operating Basis Earthquake DBE - Design Basis Earthquake e) Light-Gage Cold-Formed Steel Design Manual, 1961 Edition, American Iron and Steel Institute. Acceptance criteria were based on: a) Normal and upset (OBE) Appendix XVII, ASME, Section III. b) Faulted DBE 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 KVII, 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 module are: a) Dead loads which are weight of module and fuel assemblies, and hydrostatic loads. b) Live loads - effect of lif ting an empty module 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 DBE. e) Accidental drop of a fuel assembly from the maximum possible height. 4-3
f) Postulated stuck fuel assembly causing an upward force of 2000 pounds and a hori:ontal force of 1000 pounds. The load combinations considered in the module design are: a) Live loads. b) Dead loads plus OBE. c) Dead loads plus DBE. d) Dead loads plus fuel drop. The allowable stresses for each loading combination follow ASME Boiler and Pressure Vessel Code Section III, Subsection NF, per " Operating Technical Position for Review and Acceptance of Spent Fuel Storage and Randling Applications." Only an elastic analysis was considered. The two controlling loading combination equations were found to be D + L + OBE and D+ L + DBE. DBE was also considered to check for elastic buckling per ASME Section III, Subsection NF. The allowable stresses are given in Table 4-1 based on the following equations, and they are consistent with the requirements specified in Regulatory Guide 1.124 Stress Tvee D+L + OBE D+L + DBE Tension (w/o pin hole) 0.6 Sy (w/ pin hole) 0.45 Sy Shear 0.4 Sy Increased by 1.2 SY k Bending Stress 0.66 Sy Bearing 0.9 Sy Note: Sy and F e are specified minimum yield strength and allowable tensile stress, respectively. . Thermal loads were not included in combinations because the design of the module makes them negligible. Assuming the boundaries of the module are completely fixed and the module is not allowed to expand, the maximum thermal stress between loaded and unloaded cells is less than 11,800 psi. Ihis is well within the allowable compressive stress in the tube wall. Furthe rmo re , according to the ASME Section III, Subsection NF, Paragraph NF-3230, Appendix XVII Article F-1370, thermal stresses need not be considered in the stress calculation but only in the buckling analysis for the DBE condition. This is consistent with industrial practice for piping stress analysis where ! ther=al stress is treated as secondary stress. Therefore, under the cooling water flow conditions in the modules, the maximum temperature gradient between a loaded and an empty cell is no more than 44*F. Temperature-induced stresses are not additive fron module to module because each module is independent of the others. l 4-4
Stress analyses were done for both OBE and DRE conditions, based upon the shears and moments developed in the finite-element dynamic analysis of the seismic response. These values were compared with allowable stresses referenced in ASME Section III, Subsection NF (Table 4-1). Values given in Table 4-1 are based on the maximum stresses calculated for the high density modules. A dynamic load amplification factor hes been applied to stresses due to the horizontal seismic load to account for the effects of impact between the fuel and the module. A derivation of this factor is given in Section 4.3. Additional analyses were then performed to determine the dynamic frequencies, earthquake loading reactions, and the maximum amount of sliding. The stability of the modules against overturning was also checked and they were found to be stable. Those values are summarized in Table 4-2. The force path in the module caused by a horizontal earthquake is shown schematically in Figure 4-4. This figure shows the path of the horizontally induced earthquake fuel bundle inertial forces from the fuel bundle to the base plate. 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 4-4). 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 portion 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 t ie in plane shear forces to the long casting (see Figure 2-7) which then transfers the chear forces to the module base plate (Point 4). The 'f7 eces are carried in the module base plate (Point 5) until they are ultimately transferred to the foot pad, support pad, and the pool slah ( Point 6). The vertical forces caused by earthquake and gravity loads become forces on the support pads and to the base plates. The critical location for the compression forces from the foot pads is in the long castings and tubes directly above the grid. For stress analysis purposes, this compression force l is assumed to be resisted by four fuel tubes above the support pad. Fuel assembly drop accidents were analyzed using analytical methods in accordance with the " Operating Technical Position for Review and Acceptance of Spent Fuel Storage and Hardling Applications." In estimating local damages in the module, the maximum strain energy resulting from plastic deformation is equated to the maximum potential energy of the fuel. Energy dissipation attributable to the viscosity of the water and plastic deformation of the fuel , bundle was ignored for conservative results. The stainless steel for the module is assumed to exhibit a bi-linear hysteresis relationship, with yield stress and ultimate stress as the two control points. The results are summarized in Table 4-2. i Also evaluated was the damaging effect of a fuel bundle drop through an empty storage position along the outer rows of the module, impacting the base plate. It was determined that the fuel bundle will not possess enough energy to perforate the 1-inch thick base plate. The resulting configuration of the 4-5 Revision 1
module will be adequste to maintain the fuel in a safe condition. This case is less critical than the cases discussed in Table 4-3. The significant loads that may be carried over the spent fuel storage modules are listed in Table 4-4. Both the existing, nonpoisoned fuel storage modules and the new HDFS modules were designed to protect the stored fuel assemblies in the unlikely event of a fuel assembly being dropped onto the modules. In addition, the HDFS modules were analyzed for structural damage from a BWR fuel assembly drop from a height where the bottom of the fuel assembly was 72 inches above the top of the module. The results, presented in Table 4-3, show that there would be no increase in the overall reactivity. The impact energies for the actual maximum drop heights (15.5 inches for a BWR fuel assembly and 24.5 inches for a PWR fuel assembly) are less than the example used in the analysis; therefore, they would cause less damage to the modules. The provisions employed to prevent movement of heavy objects over the spent fuel pool are discussed in Section 11.0. The HDFS system design does not require any different fuel handling procedures from those discussed in the Brunswick FSAR. The loads experienced under a stuck fuel assembly condition are less than those calculated for the seismic condition and have, therefore, not been l included as a load combination. 4.3 FUEL BUNDLE / MODULE IMPACT EVALUATION An analysis was performed to evaluate the effect of an impact load that is possible because of gaps between the fuel bundles and the fuel storage module. In the seismic analysis for the BSEP high density spent fuel storage module (results in Table 4-2), gaps were not considered and the fuel bundle was treated as an integral part of the module in addition to the hydrodynamic mass due to surrounding water. A gapped element model was prepared to study the effect of impact loads on the module. This model is shown in Figure 4-5. The distinct feature of this model is that the fuel bundle is separated from the module and is free to l vibrate within the confines of the storage position in the module. The fuel bundle is modeled as being pinned supported at the base and the entire module is submerged under water and free to slide. For comparison purposes regarding the impact load effect, a lumped element model was also constructed. The j lumped element model is identical to the gapped element model shown in i Figure 4-5 except that the gaps between the fuel bundle and the module are ignored. 4-6 Revision 1
1 I i The objectives of this evaluation are: a) To assess the difference in maxi =um internal forces in the module as determined from a gapped element model and a lumped element nodel, and b) To assess the ef fects on rigid body displacements, the two models were subjected to a constant 1.0g base acceleration for a period of 0.8 seconds. This acceleration was applied for two cases, corresponding to friction coefficients of 0.108 and 0.222. The use of a constant 1.0g base acceleration was mandated by the lack of a definitive time hi' story to use in conjunction with rigid body displacements. Cap effects on internal forces were evaluated by subjecting both models to the Brunswick time history. This was done for three cases: u = 0.108, u = 0.222, and u+= (fixed base). The results of these analyses are presented in Tables 4-5 and 4-6 for rigid body displacements and internal forces, respectively. Table 4-5 shows the displacement ratio between the gapped and the lumped element model. It indicates that there are no significant dif ferences between the rigid body displacements as determined from the gapped and lumped element models for both u = 0.108, and u = 0.222. Thus, it can be concluded that gap effects on the rigid body motions can be neglected and that the results provided in Table 4-2 are adequate for design purposes. Table 4-6 indicates that the internal forces (or spring loads) in the module determined from the gapped models are significantly less than the corresponding forces in the lumped models for the two cases u = 0.108, and u = 0.222. For the case where u + = this situation is reversed, however, and the internal force in the gapped model exceeds the internal force in the lumped model. Thus, it can be concluded that where rigid body motion is permitted and friction forces are within the range of interest, the internal forces are conservatively determined from the lumped model. The ratio between the spring forces in the gapped model and the lumped model (fixed base case shown in Table 4-6) is treated as the dynamic load amplification factor and used in the stress analysis comparison in Table 4-1. This approach is conservative since results for the sliding model indicate that there is a reduction in internal stresses for the gapped element model. 4-7
TABLE 4-1 Comparison of Calculated Stress vs Allowables (psi) OBE Condition SSE Condition Calc. Calc. Location / Type Stress Allowablesl Stress Allowables I Tube wall shear 5,520 9,260 8,230 15,400 Tube wall compression 6,470 13,900 9,380 23,100 Tube weld throat shear 5,520 9,260 11,640 15,400 Angle, weld throat shear 7,810 9,260 11,640 15,400 Casting wall shear 3,340 9,260 9,170 15,400 Casting wall compression 8,900 15,300 14,220 23,100 Casting base weld shear 3,830 9,260 7,660 15,400 Support plate weld throat 3,870 9,260 15,330 15,400 shear Upper Closure Plate Compression 5,820 13,900 7,880 23,100 Shear 4,470 9,260 5,260 15,400 Weld Shear 6,320 9,260 7,440 15,400 Lower Closure Plate Compression 4,000 13,900 5,660 23,100 Shear 7,340 9,260 11,490 15,400 Weld Shear 7,340 9,260 13,58 0 15,400 Corner tube local - - 9,120 23,100 compressive stress check for local buckling i Allowable stresses referenced in ASME Section III, Appendix F, Section F-1370 l 4-8 Revision 1
.- - - = . - - . .
i. TABI.E 4-2 l E DYNAMIC FREQUENCIES, EARTHOUAKE LOADING REACTIONS, AND MAXIMUM AMOUfff 0F SLIDING Module Size Direction Fundamental Frequency (Hz) Max. Reaction (lbs) Max. Sliding (in) 13x15 13-(NS) 7.81 113,800 0.43 4 15-(EW) 9.26 106,300 0.35 13x17 13-( NS) 5.89 125,200 0.31 17-(EW) 9.13 115,200 0.41 13x19 13-(NS) 6.77 143,200 0.39 19-(EW) 11.80 121,000 0.53 . I5x17 15-( NS) 7. 52 141,000 0.46 7 e 17-(EW) 8.70 128,700 0.45 l 1
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- 8
l I l l l TABLE 4-3 High Denet't y Spent Fuel Storage System Assembly Drop Accident Case Summary No. Case Description Effect on Reactivity 1, A fuel assembly drops 72 Analysis indicates that localized } inches vertically and impacts the damage or fuel support member damage top of a fully loaded HDFS will occur, but neutron absorber l module. The dropped assembly material will not be removed from comes to rest horizontally on top its position between adjacent fuel of the HDFSS. assemblies. A fuel assembly resting horizontally atop the HDFS module does l not increase the system reactivity because the reactivity assumes an infinite vertical length of fuel (no neutron leakage in the vertical dimension). k,ff < 0.90. l
- 2. A fuel assembly drops from Structural analysis indicates that 72 inches above the HDFS module, localized tube damage will occur and l enters an empty storage position, one neutron absorber plate may be and falls to the bottom of the damaged. A reactivity analysis of storage position. this case, with the neutron absorber plate between two fuel assemblies totally missing, shows that keff remains less than 0.90.
- 3. A fuel assembly drops from Same as Case 2.
72 inches above the HDFS module and l strikes a tube wall at an oblique angle.
- 4. A fuel assembly drops from It is not possible for a fuel 72 inches above the top of a assembly drop of 72 inches to l fully loaded module and strikes drive four stored assemblies through the upper tie plates of 2, 3, or the bottom of the module. Even so, 4 fuel assemblies in storage. the reactivity effect of this postulated event was calculated as a limiting value. An 18-inch section of fuel in four bundles in an unpoisoned square array was found to have a keff approximately equal to that of the system. There would be no increase in the overall reactivity keff < 0.90.
- 5. A fuel assembly drops from This case was analyzed for normal 72 inches above the HDFS module, handling conditions; keff < 0.90.
falls outside of the loaded module, and lodges adjacent and parallel to an unpoisoned, occupied fuel storage position. 4-10 Revision 1
- i i
f I i i ! r i TABLE 4-4 i Items That May be Moved Over The Spent Fuel Fool Racks L 1 , Item Approximate Weight (lbs.) l BWR Fuel Assembly (Including Channel) 735 l i 1 l BWR Otannel 62 l l BWR Control Rod 235 I l PWR Fuel Assembly (Including Control Rod) 1605
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TABLE 4-5 Normalized Rigid Body Displacement of Lumped And Gapped Models Friction Coefficient Gapped Element Model/ (u) Lumped Element Model 0.132 1.01 0.200 1.02 l 4-12 Revision 1
l l TABLE 4-6 Spring Forces In Lumped And Capped Models Friction Coefficient Gapped Element Model Force / Lumped (a) Element Model Force (lbs.) 0.132 0.647 0.200 0.493 a
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b HRUNWsICK STI AM Stodules and Pool ,odel {or .. FiffuRE (Ltcrmc ew.r cui.e. liydredynamic Virtual :taus ,3 Po , 8. L-pst Car %.ny , gg SPtNT FUEL PCOL sfol:M;r E APANSION -,
i EARTHOUAKE FORCE- l l~ l si N-i i FUEL BUNDLE sl I sl l
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#N FUEL SUPPORT g PLATE MODULE A TING 5 BASE PLATE ASSEMBLY FOOT PAD
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SUPPORT PAD BRUNSWICK STEAM ELECTRIC PLANT FIGURE Path of Earthquake Hori:ontal , Power & L t Company SPENT FUEL POOL STORAGE EXPANSION
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@ Generalized channel-beam mass
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@ Fixed point Element Cefinitions: @ Channel beam element (nodes 1 & 2) 2 Fuel bundle element (necas 1 & 2) 3 Gapped, hydrodynamic element (nedes 2 & 3) 4 Elastic-plastic spring element (nodes 1 & 4)
BRUNSWICK f. TEAM ELECTRIC PLANT FIGURE Carolina Gapped Element Model Power & Light Company 4-5 SPENT FUEL POOL STORAGE EXPANSION r--- -
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l 5.0 MATERIAL 00NSIDERATIONS Most of the structural mate used in fabrication of the new High Density Fuel Storage (HDFS) System 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 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 normal pool water operating temperature there is no significant deterioration or corrosion of stainless steel or Boral.
! Specifications were developed specifically for the HDFS System which impose quality control requirements during the design, procurement, fabrication, installation, and testing of the HDFS System. Periodic audits of the various facilities and practices are performed by certified quality assurance personnel to ensure that these QA/QC requirements 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 Boral sheet used in tube fabrication meets specification, a special quality control program is in effect at the manufacturer's facility. The concentration and distribution of the neutron absorbing material (B4C) 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. Neutron transmission testing will be utilized to detect the presence, ! distribution, and neutron absorption capability on each side of the cells in the HDFS modules prior to first use for spent fuel storage. The recorded results will be used to verify that the minimum uniform concentration of B10 is at least that used in the criticality analysis.
- Boral's corrosion resistance is similar to that of standard aluminum sheet.
l Corrosion data and industrial experience confirm that aluminum and Boral are acceptable (References 5-1, 5-2, 5-3 and 5-4) for the proposed application. Although experience indicates that it is unnecessary, an inservice test program will be conducted, consisting of periodic examination of surveillance l 5-1 Revision 1
samples which will be suspended underwater in the fuel storage pool. These samples consist of two types; the first being 8-inch x 8-inch coupons of Boral plate with stainless steel sheet formed to both sides, tnd the second consisting of 6-inch square samples of Boral without stainless " cladding.' The stainless " clad" coupons have two sides open to permit water access. Sufficient samples are included to permit destructive examination of a sample on inspection intervals of 1 to 5 years over the 1.ife of the facility. Fool water quality will be maintained as specified in the Brunswick FSAR. No changes to water quality are expected as a result of the planned modification to the spent fuel storage capacity (see Section 10-1 of the Radiological Evaluation). 1 i i I i I I 5-2 I
1 6.0 INSTALLATION The Brunswick Steam Electric Plant, Unit Nos. I and 2, spent fuel pools are filled with water and contain spent fuel which must be relocated within the pool prior to installation of the new Righ Density Fuel Storage (HDFS) modules. The fuel cask crane has adequate capacity on its main hook to handle all loads that will be encountered. The maximum capacity of the auxiliary hook on this crane is five tons. Loads in excess of five tons will be handled by the main hook. The work will be planned so that no heavy equipment will be transported over stored spent fuel. The HDFS modules are designed to be free-standing with a bottom-supported i design. They rest on support pads placed on the floor of the fuel storage pool which have legs to bridge interferences and obstructions, and a plate to form a level surface. Foot pads are bolted to the bottom of the module and rest on the support pads. A 13-cell by 19-cell module and a 13-cell by 15-cell module will be installed in the east end of both units' spent fuel pools. This area is presently being used only for storage of control rod assemblies. Two 15-cell by 17-cell modules and one 13-cell by 17-cell module will replace two grid rows of existing modules in the west end of each unit's spent fuel pool. Where free-standing modules are located over a grid at the west end of the pool, existing modules will be removed, support pads will be placed at design locations, and the new modules will be lowered into position onto the new 4 support pads. The existing grid truss and bracing on the west end will not be removed. The new modules will be installed above the grid and bracing. In the east end of the pool where free-standing modules are located outside the grid area, existing hardware supports being utilized for control-rod hangers will be removed and replaced with wall hangers to provide space for the new modules. No modification is necessary to the existing grid restraint system bracing and crossbeam. The following is the sequence of events for performing the work associated with the removal of existing modules and installation of HDFS modules: 4 l a) Shuffle spent fuel in the east end of the pool as necessary to ensure that the maximum possible separation between divers and stored spent fuel will be maintained during the removal of control rod storage hardware and
- modification of the grid bracing.
b) Vacuum pool to the extent possible in the area of proposed work. c) Install new control-rod storage tardware. d) Relocate stored control rods. e) Remove existing control-rod storage hardware. This work will require divers. 6-1 Revision 1
f) Make necessary modifications to the grid bracing if required. This work will require divers. g) Remove waste =aterials from the pool, h) Re-vacuum pool in areas where cutting was performed.
- 1) Survey the spent fuel pool floor to locate obstructions. This can be accomplished from above the pool.
j) Install support pads; level and shim feet as necessary. This can be accomplished from above the pool using a TV camera and extension tool with assistance from divers as required. , , k) Survey the installed support base pads to determine if shimming of sodule foot pads is necessary to ensure the modules are installed plumb.
- 1) Bolt foot pads to bottom of modules with necessary shimming.
m) Place protective covers over the support pads. HDFS modules will be placed later as needed. n) Shuffle fuel elements at west end of pool. o) Unlatch and remove existing modules from the two grid rows in the west end of the pool. p) Move modules to decontamination area, decontaminate and prepare for storage. q) Survey the floor to locate obstructions. This can be accomplished from above the pool. r) Install support pads; level and shim as necessary. This can be accomplished from above the pool using a TV camera and extension tool with assistance from divers as required. s) Survey the installed support pads to determine if shi= ming of the module foot pads is necessary to ensure the modules are installed plumb. t) Bolt foot pads to bottom of modules with necessary shimming. u) Place HDFS modules on support pads. v) Decontaminate tools and temporary equipment. It is estimated that this work will require approximately 75 working days per spent fuel pool. Manpower requirements will fluctuate with each task, but it is estimated 9,100 man-hours will be utilized. For estimating radiation exposure, this work has been further delineated into four categories: work above the pool water surface, work requiring divers in the pool, decontamintion work, and work in areas where no radiation exposure is expected. All nan-hours include not only construction and operating personnel 6-2
but contingency for health physics, engineering support, and Quality Assurance personnel but does not include time required for shuffling of spent fuel. The following table represents the estimated man-hours in each category: Above the Divers in Decon. Non-rad. Pool the cool Work Work Total Manhours 5,350 650 400 2,700 9,100 Millirem exposures were estimated for each category of work described above. Measurements based on experience indicate that the average exposure rate above the pool normally does act exceed 8 mrem / hour, with an average rate expected - at 5 mren/ hour. For purposes of calculating the total man-rem, 5 mren/ hour was assumed. Radiation exposure at the bottom of the pool in the area in question, with allowances for vacuuming and =oving of stored fuel to the maximum distance from this area, should result in a dose rate between 20 and 75 mrem / hour. The maximum expected dose rate was used in determining total man-rem. Decontamination work will be accomplished using hydrolasers or similar equipment in the existing cask decontamination area. It has been estimated that the maximum dose rate to be experienced will be 15 mrem / hour. Using the assumptions discussed above, it is estimated that the maximum total man-rem for this will be 81 man-rem. Because the nature of this work will require specialized personnel, this dose will be spread over several distinct crews. Every feasible means will be utilized to keep radiation exposure as low as reasonably achievable (ALARA). 6-3
7.0 NUCLEAR CONSIDERATIONS 7.1 NEUTRON MULTIPLICATION FACTOR The criticality analysis calculations were performed with the MERIT (Reference 7-1) 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/3-IV library tapes. The qualification of the MERIT program rests upon extensive qualification studies including Cross Section Evaluation Work Group (CSEWG) thermal reactor benchmarks (TRX-1, -2, -3, -4) and B&W UO2 and Pu02 criticals, Jersey Central experiments, CSEWG fast reactor benchmarks (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 UO2 criticals, and as a solid control curtain in the Jersey Central experiments. The MERIT qualification program has established a bias of 0.005 20.002 (le) Ak with respect to the above critical experiments. Therefore, MERIT under,":dicts keff 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 standard-core geometry at 20*C. The use of a maximum fuel k. as a criticality base eliminates the need to analyze the multiplicity of U235 enrichnent and burnable poison combinations. c) Storage space pitch of 6.563 in. d) Minimum allowable boron (310) areal concentration of 0.013 grams 310/cm2 distributed homogeneously. 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 tabulated in Table 7-1. The model geometry, bias, and uncertainity for each of the cases is described below. l { l l 7-1 l l
7.3 CEOMETRY. BIAS. AND UNCERTAI5"rY 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 nominal dimensions shown in Figure 7-2. No geometry bias is associated with this model. The MERIT program bias is 0.005 2.002 (la) ak. The same basic geometry model was used for case 4 of Table 7-1, but with the maximum axial average gap as shown in Figures 7-2 and 7-3. The pitch was increased to 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 0.005 20.002 (la) ak. , 4 An approximate geometry model, shown in Figure 7-4, was used for case 5 in i Table 7-1. The model geometry bias relative to the exact model for the same conditions was 0.0087 + 0.0050 (Ic) ak. The MERIT program bias remains the ' same at 0.005 + 0.002 Tle) ak. In all cases the reported value of k. 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 increasing the cell pitch, eccentric bundle positioning, reducing moderator density, or increasing the temperature to 65'C, 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 k. 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 High Density Fuel Storage (HDFS) System has alternating spaces on the periphery of each module fabricated with unpoisoned closure plates. (The two opposite non-tube corners of each module contain Boral i plates in them so the geometry is the same as the adjacent tubed corners). The unpoisoned locations are also directly opposite each other on adjacent modules. The effect of the partially unpoisoned storage locations is small and insensitive to the inter-module water gap, as shown in Table 7-3. The maximum module k. occurs at the minimum possible water gap, as shown in l 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 modal 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 : .0051 (Ic) 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 rod 7-2
enrichment and burnable poison distributions within the bundle. Fuel sin spacers were not included (a conservative exclusion). The nominal bundle dimensions were used for all cases. The sensitivity of k. analyses to various changing oarameters is imolied above. More specific relationships are as follows: a) Bundle Reactivity (percent U235) - Calculations are based on maxinun
- 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 show 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-coderated 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 tolerances in the tubes were taken to maximite the k. of the storage lattice. The result of the analysis is given as Case 6 in Table 7-1. A conparison of Cases 2 and 6 in Table 7-1 shows that within the statistical error bounds there is no significant difference between the results. e) The HDFS modules and the BWR fuel to be stored in it are designed and fabricated to prevent significant quantities of air or other gas fron being entrapped. Thus, no areas of reduced effective moderator density are created. Even if air were trapped, the effect of reduced density on the under-moderated fuel bundles is to reduce the keff of the system. 7.4 INTERACTION WITH EXISTING STORACE SYSTEM The proposed high density (poisoned) fuel storage system will be located in the existing spent fuel pools adjacent to the existing fuel storage system. The minimum separation between the proposed and existing systems will be six inches. The proposed module and the existing storage system have individually been shown to have a neutron multiplication factor lower than the nuclear criticality safety criterion of 0.95. Each system has incorporated neutron absorber materials (stainless steel for the existing; boral and stainless steel for the proposed) in its design. The adjacent faces that the two systeas present to one another are either stainless steel or stainless steel and boral. In this configuration and with a separation of six inches of water, there is no significant neutron connunication between systens. Calculations were made to determine the interaction between the faces of two high density modules, with partially unpoisoned storage locations directly opposite each other. These calculations (described in Section 7.3 and 7-3
1 I tabulated in Table 7-3) support the conclusion that neutron multiplication factor is insensitive to intermodule water gap. , I The maximum neutron multiplication factor for combined systems, is therefore, the value calculated for PWR fuel storage in the existing system, k,fg = 0.905. 7.5 POSTULATED ACCIDENTS 7.5.1 Fuel Droo Fuel handling within the fuel pool is limited to the movement of a single fuel assembly at any one time. The accidental dropping of an assembly will not result in a critical mass situation whether the assembly is laying across the top cl the modules or alongside the modules at the support grid. 7.5.2 Loss Of Pool Cooling i The normal cooling of the spent fuel pool is accomplished by the spent fuel I pool cooling system which consists of two parallel cooling loops sharing common supply and discharge piping. Loss of one cooling loop subsequent to the last expected refueling would result in a maximum pool temperature of 182.6*F. a single failure in the shared piping could cause loss of both cooling loops allowing the pool to reach 150*F in 57 minutes, and ultimately permitting pool bulk boiling in 13.5 hours, if no actions were taken. I t i i t l
, 7-4 1
TABLE 7-1 Single Cell High-Density Fuel Storage Criticality Results Boron Congneration Case Description g /sq.cm. K, (+ 20)I 1 Nominal Rack Dimensiono 2 a. 0.013 0.8668 + 0.007 5 With Flow Channel @20*C b. 0.010 0.8749 _T 0. 0080 2 Nominal Rack Dimensions 0.013 0. 867 4 + 0. 0086 Without Flow Channel 620*C 3 Same as Case 2 except R65'C 0.013 0. 8 561 + 0. 0084 4 Increased Pitch without Flow 0.013 0.8364 + 0.0106 Channel @20'C 5 Same as Case 2 but with 0.013 0. 8276 + 0. 0123 Eccentric Bundle Position 6 Minimum Pitch without Flow 0.013 0.86 50 + 0.0088 Channel @20'C I k, includes MERIT Program Bias and Uncertainty 26 . 563-inch Pitch with Nominal Material Thickness
! 36 . 503-inch Pitch with Minimum Storage Tube Material Tolerances to i
Maximize k, 7-5 Revision 1
TABLE 7-2 Bias and Uncertainty Components for Maximum k. of a Storage Cell
- k. 0.8624 Calculational Convergence ak 0.0038 MERIT Bias and Uncertainty ak 0.005 2 0.002 Model Bias and Uncertainty ak None Total 0.8674 2 0.0086 (2c) 7-6
. - . . - - . _ - . - - . . . - . . ~ . - - . _ . . -.-- .-. _ -. - . - = . . . - - . .. .
4 ) TABLE 7-3 i 4 HDFSS Criticality Analysis i Module Interaction-
- Description k. (+ 2c)
Minimum gap between modules 0.8593 2 0.0131 (2A = 1.244 in.) i i Intermediate gap between 0.8579 2 0.0130 l modules (2A = 2.100 in.)
- Nominal gap between modules 0.8506 2 0.0134 (2 A = 2.967 in.)
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8.0 THERMAL ANALYSIS
8.1 DESCRIPTION
OF THE SPENT FUEL POOL COOLING SYSTEM The spent fuel pool (SFP) cooling system for BSEP Units 1 and 2 is described in detail in FSAR Section 10.5. It consists of two fuel pool cooling pumps, two heat exchangers, two filter demineralizers, two skimmer surge tanks, and associated piping, valves, and instrumentation. The two fuel pool pumps are connected in parallel, as are the two heat exchangers. 4 The pumps take suction from the skimmer surge tanks, circulate the water through the heat exchangers and the filter demineralizer, and discharge through the diffusers at the bottom of the spent fuel pool. The cooled water traverses the pool, picking up decay heat and debris before flowing over the skimmer weirs and scuppers into the skimmer surge tanks. Makeup water for the system is provided from the demineralized water system to the skimmer surge tanks. The SFP heat exchangers use the reactor building closed cooling water to cool the SFP water. In turn, the reactor building closed cooling water heat exchangers are cooled by the service water system. An adequate supply of service water is available during all modes of operation to maintain the reactor building closed cooling water supply temperature below 100*F. The reactor building closed cooling water and service water systens are discussed in Sections 10.6 and 10.8 of the BSEP FSAR. The residual heat removal (RHR) system, which is discussed in Section 4.8 of the BSEP FSAR, can be connected to the SFP cooling system by means of a seismic category I cross tie to provide an alternative mode of cooling for the SFP and the reactor well. This is depicted in the flow diagram shown in Figure 8-1. In this mode of cooling, a RHR pump takes suction from the recirculation line, circulates the water through a RHR heat exchanger and discharges through the SFP diffusers, whereas the SFP pumps circulate the water from the skimmer surge tanks through the SFP heat exchangers and the filter demineralizers, and discharge through diffusers in the reactor well. The two cooling systems can be operated independently. A flow of approximately 4950 gom can be maintained in the RRR loop in this cooling node. For the SFP system, a minimum total flow of 1000 gpm can be obtained in either cooling mode. , The performance data for the SFP and RHR heat exchangers uader these conditions are provided in Tables 8-1 and 8-2, respectively. The RHR heat exchangers use the service water as the cooling medium. The maximum service water temperature is assumed to be 90*F, consistent with the original design bases given in the FSAR. The following criteria, as stated within the BSEP FSAR, are applicable to the evaluation of the adequacy'of the present SFP cooling system in handling the heat load corresponding to the expanded fuel pool storage capacity:
- 1. The SFP cooling system alone shall maintain the SFP bulk temperature at or below 150*F following a refueling.
8-1
. _ . - _ -- ~-. - . . - _ _ _
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- 2. The RRR sys?.em, operated alone or in conjunction with the SFP cooling system, shall maintain the SFP bulk temperature at or below 150*F following a full core unload.
8.2 SECTION DELETED. 8.3 HEAT LOAD AND POOL TEMPERATURE FOR EXPANDED STORAGE CAPACITY
- 8.3.1 Spent Fuel Pool Inventory The expansion allows 10 full spaces and 2 half spaces of the existing 38 1/2 spaces in the support grid to accommodate 731 instead of 396 BWR assemblies and also adds additional spaces to store 442 more BWR assemblies in the area presently used for control rod storage. The actual volume of water in the 8-2 Revision 1 i
y ,, -
i spent fuel pool after the expansion is 45360 cu. ft. The increased storage capacity can be expected to further increase the heat load on the spent fuel pool cooling system. In order to verify the adequacy of the existing cooling system in maintaining the SFP water temperature within allowable limits, two cases, as described below, have been. considered. Case 1 - Refueling Case In this case, the spent fuel pool heat load corresponds to the pool inventory immediately following the refueling, which leaves a reserve storage enough for a full core unload but not enough for an additional refueling plus a full core unload. Based on the expanded capacity and assuming no fuel shipments off the BSEP site, this condition would likely occur in March 1987 for BSEP Unit 1. The-pool would then contain 1226 BWR assemblies and 160 PWR assemblies. For BSEP Unit 2, this condition would likely occur in November 1986 and the pool would then contain 1190 BWR assemblies and 144 PWR assemblies. Case 2 - Core Unload Case In this case, the SFP heat load corresponds to the pool inventory considered in Case 1, plus a full core of BSEP fuel assemblies. This condition is assumed to occur at the end of the fuel cycle following the last refueling described in the Refueling Case. For BSEP-1, this would occur in March 1988. For BSEP-2, this would occur in "ovember 1987. The spent fuel pool inventories for the Refueling Case are given in Table 8-4 and Table 8-5 for BSEP-1 and BSEP-2, respectively. For the Core Unload Case, they are presented in Tables 8-6 and 8-7. Also included in the tables are parameters pertinent to calculations of the decay heat load, including the continuous assembly power level, the irradiation time at that power level, and the cooling time. 8.3.2 Seent Fuel Fool Heat Load The decay heat due to each irradiated fuel assembly has been calculated utilizing the formulation in the Standard Review Plan, Branch Technical Position ASB 9-2. The spent fuel pool heat load at any time is obtained by summing up the decay heat of all fuel assemblies actually present in the pool at that time. The cooling times for all fuel assemblies include 24-hour cool-down period per current technical specifications plus the time required , to transfer the fuel assemblies that are to be unloaded to the pool during ] refueling or core unload. The time for fuel movement has been conservatively estimated to be 15 minutes per fuel assembly. Based on this model, the time the last fuel assembly enters the spent fuel pool would be 59 hours after shutdown for the refueling condition and 164 hours after shutdown for the core unload condition. Figure 8-2 and Figure 8-3 show the total decay heat loads for 3SEP Units 1 and 2 as functions of time for the Refueling Case and the Core Unload Case, respectively. It should be noted that in these plots, time zero denotes the time when the fuel transfer is started, i.e., 24 hours following reactor shutdown. 3-3
Figures 8-2 and 8-3 show that, in both cases , the transient heat load for BSEP-1 is only slightly higher than that for BSEP-2. The following analysis has been performed based on the heat load for BSEP-1. 8.3.3 Spent Fuel Pool Bulk Temperature The spent fuel pool bulk temperature has been calculated for each of the two cases in question. The results, which indicate coapliance with the FSAR temperature limit criteria, are provided below and summarized in Table 8-3. 8.3.3.1 Refueling Case For this case, the transient response in the spent fuel pool bulk temperature with the SFP cooling system operating is shown in Figure 8-4. The pool temperature would reach a maximum of 145.2*F at approximately 50 hours af ter fuel transfer starts and eventually drops below 125'F at about 35 days af ter reactor shutdown. It should be noted that in these plots, time zero denotes the time when the fuel transfer is started, i.e., 24 hours following reactor shutdown. 8.3.3.2 Core Unload Case For this case, if only the SFP cooling system were available during the core unload operation, the pool bulk temperature transient would be that shown in Figure 8-5. The pool temperature would reach 150*F at 46.2 hours af ter unloading started and the maximum pool temperature would be 197.2*F. However, under the core unload condition, the RER system is available for cooling the SFP as discussed in Section 8.1. Because of the high heat removal capacity of the RRR system, the SFP temperature will drop rapidly as soon as one train of the redundant RHR system is used to cool the SFP. As the SFP heat load gradually increases, the pool temperature rises again and goes through a maximum. As long as the RHR system is brought into service within 46.2 hours af ter the unloading is started, the pool temperature can be kept below 150*F. If the SFP system and one train of the EHR system are operated simultaneously in this cooling mode, the pool temperature goes through a maximum of 124.6*F. If one train of the RHR system is operated alone, the subsequent maximum pool temperature is 131.7*F. 8.4 LOSS OF SPENT FUEL POOL COOLING The ef fect of a loss of the SFP cooling system on the pool temperature of either unit has been evaluated for the following conditions:
- 1. Refueling Case a) Loss of one SFP heat exchanger / pump, b) Loss of the SFP cooling system.
- 2. Core Unload Case a) Loss of the SFP cooling system.
For failure condition (la), the flow rate in the SFP system is conservatively assumed to be half of the flow rate without failure. 8-4 Revision 1
1 8.4.1 Refueling Case For either unit, if a failure occurred within the SFP cooling system during the refueling operation, the RHR system would be available for :ooling the spent fuel pool as in the Core Unload Case and the pool temperature could be maintained below 150*F. Therefore, significant increases in the pool temperature could result only if the SFP cooling system failed after the reactor operation has been resumed. a) Partial Loss of SFP Cooling System Figure 8-6 shows the transient response in the SFP water temperature resulting from the failure of one SFP heat exchanger / pump immediately after the reactor operation has been resumed. It has been conservatively assumed that the reactor operation is started as soon as the last fuel assembly has been placed in the pool. The figure shows that the pool temperature reaches 150*F in 1.9 hours and the maximum temperature is 182.6*F. The maximum pool temperature that would result depends on the time of failure of the SFP components. If the failure occurs 35 days af ter shutdown, the pool temperature can be maintained below 150*F with only one SFP heat exchanger / pump operating. b) Loss of the SFP Cooling System - Figure 8-7 shows the transient response in the spent fuel pool temperature for a postulated failure of the entire SFP cooling system immediately after the reactor has resumed operation. The pool temperature reaches 150*F in 57 minutes and bulk boiling starts in 13.5 hours. The make-up water requirement following boiling is 28 gpm. The radiolog! cal impact due to boiling in the spent fuel pool is most severe if both BSEP-1 and BSEP-2 pools are boiling simultaneously. This would occur if the SFP cooling systems in both units were assumed to fail at the same time. BSEP-2 is refueled much later than 3SEP-1. The above pool temperature transients apply directly to BSEP-2 if a concurrent loss of the SFP cooling system for both units is postulated to occur at ' completion of the refueling of BSEP-2. The situation in BSEP-1 is less severe because the SFP heat load l would have decreased significantly by that time. Howeve r , for conservatism, a j radiological analysis has been performed assuming that the conditions in the l 3SEP-1 pool are identical to those in the BSEP-2 pool. That ta, both spent i fuel pools start boiling simultaneously, each with a boiling rate of 1.38 x 104 lbm/hr. The consequences are presented in Section 8.6. 8.4.2 Core Unioad Case l For the core unload situation, a single train of the redundant, seismic l category I RHR system alone, operating in the alternative cooling mode to cool the spent fuel pool, can maintain the spent fuel pool temperature at or below 131.7'F as discussed in Subsection 8.3.3. A loss of the SFP cooling system l l 8-5 i
~
is, therefore, inconsequential so far as the 150*F pool temperature limit and pool boiling are concerned. The pool :emperature results for the various cases analyzed in Sections 8.3 and 8.4 are summarized in Table 8-8. All transient analyses performed are conservative, as no credit has been taken for cooling by evaporation from the pool surface or the thermal inertia of the steel module components or pool walls. Additional conservatism includes the high continuous power level assumed for the H. B. Robinson fuel, minimum available cooling time, and maximum service water and closed cooling water temperatures. It is therefore expected that actual pool water temperatures would be lower than those calculated. c 8.5 LOCAL FUEL BUNDLE THERMAL HYDRAULICS . The bounding thermal hydraulic conditions were calculated for fuel stored in a HDFS module or basket in the BSEP pools. Bases for the calculations for typical current generation fuel were the following: Maximum Burnup 35000 mwd /MTU Continuous Power Level 4.35 MW:/ Assembly Total Core Power Level 2436 MW e Assemblies per Core 560 Transfer Time Core to Pool 15 minutes Cooldown Time Prior to Fuel Movement 24 hours Fuel Storage Pool Bulk Water Temperature 150*F The ORIGEN Code (Reference 8-1) was used to calculate the decay heat for the bundle defined by these bases. The result was a heat generation rate of 385,000 BTU / hour per assembly. The maximum fuel cladding temperature will be 227'F. The maximum water temperature associated with the hottest fuel bundle will be 194*F. These temperatures and the maximum storage tube wall temperature of 189'F are low relative to structural integrity or corrosion limiting temperatures of the structural components of the storage system and fuel. 8.6 RADIOLOGICAL IMPACT OF SPENT FUEL POOL BOILING The radiological impact of bulk boiling in the spent fuel pools of both BSEP units is most severe when the SFP cooling systems for both units are postulated to fail simultaneously following a refueling, as described in Subsection 8.4.1. A radiological analysis has been performed to determine the thyroid dose at the site boundary /LPZ. The following assumptions were made:
- 1. The time to reach boiling is 13 hours for both units.
- 2. The boiling rate of the pool water is 1.38 x 104 lba/hr for each unit.
8-6
r
- 3. The volume of water in each pool is 45,360 cu. ft.
4 All failed fuel rods of the full core (average 1 percent of the core) are present in the 1/4 core discharge to each pool.
- 5. The normal I-131 release rate coefficient for leaking rods to the pool is 4.6x10-10 3ee -1 at 24 hours after reactor shutdown using the methods described in Reference 8-2. It is conservatively assumed that the release rate coefficient is constant at this value until completion of unloading when the SFP cooling system fails, i.e., 59 hours after reactor shutdown.
- 6. The above release rate coefficient is spiked by a factor of 100 following loss of cooling to simulate the heatup conservatively.
7 The decontamination factor for I-131 during boiling is conservatively t assumed to be unity.
- 8. The iodines are then vented through the Standby Gas Treatment System (SGTS) and released from the stack to the atmosphere. The filters in the SGTS are assumed to be 95 percent efficient for iodines.
The results obtained for the simultaneous failure of the SFP cooling systems of both units are summarized below: Site boundary /LPZ thyroid dose (0-2 hrs.) 0.3 rem . Site boundary /LPZ thyroid dose (0-4 days) 1.1 rem These results support the Applicant's position that the SFP cooling system need not be upgraded for the proposed expansion of the SFP storage capacity. 1 i a 8-7
l l TABLE 8-1 SPENT FUEL POOL HEAT EXCHANGERS PERFORMANCE DATA No. of Units 2 Heat Transfer Area Per Unit, ft 2 1285 Design Temperature, 'F 150 Heat Transfer Coefficient, Btu /hr-ft 2 _.F 261 Heat Transfer Rate Per Unit, Btu /hr Sr10 6 g Shell Side Tube Side (Closed Cooling Water) (SFP Water) Design Pressure, psig 150 200 Flow Rate Per Unit , gpm 800 650 Inlet Temperature, 'F 95 125 c 1 8-8 Revision 1
TA3LE 8-2 RESIDUAL HEAT REMOVAL HEAT EXCHANGERS PEPJ010!ANCE DATA No. of Units 2 Heat Transfer Area Per Unit, ft2 3,280 Design Temperature, *? 400 Design Pressure, psig 450 Heat Transfer Coefficient, Btu /hr-f t 2 *F 281 Heat Transfer Race Per Unit, Btu /hr 24.5x106 Shell Side Tube Side (SFP Water) (Service Water) Flow Race Per Unit, gpm 4,950 8,000 Inlet Temperature, *F 125 90 k 8-9
i 1 l l l TABLE 8-3 COMPARISON OF BULK TEMPERATURE ANALYSIS RESULTS FOR EXPANDED STORAGE CAPACITY TO FSAR CRITERIA Core Unload Last Refueling Case Case i Cooling Mode SFP & RHR SFP Maximum Heat Load 29.2 x 10 6 14.1 x 10 6 (BTU /hr) Maximum Calculated 124.6 145.2 Pool Temperature ('F) FSAR Maximur. 150 150 Allowed Temperature (*F) Days Cooling - 35 to 12 5*F 1 1 1 l l l 8-10 Revision 1
i
. l TABLE 8-4 SFP INVENTORY (LAST REFUELING) FOR BSEP UNIT NO. 1 Assembly Irradiation Cooling Discharge No. of Power Time Time Date Batch As semblies (dav)
(MWt) (hr) 3/87 1 140 BWR 4.35 1197 24 3/86 2 140 BWR 4.35 1197 8784 3/85 3 140 BWR 4.35 1197 17544 3/84 4 140 BWR 4.35 1197 26304 3/83 5 140 BWR 4.35 1197 35088 3/82 6 140 BWR 4.35 1197 43848 5/80 7 245 BWR 4.35 1197 59400 1/79 8 87 BWR 4.35 1197 71424 1/78 9 6 PWR 14.65 985 79704 9/77 10 50 BWR 4.35 1197 82344 3/76 11 4 BWR 4.35 1197 96120 11/75 12 45 PWR 14.65 985 100128 5/74 13 102 PWR 14.65 985 112632 3/73 14 7 PWR 14.65 985 122592 1226 BWR 160 PWR CAPACITY 1803 BWR 160 PWR CAPACITY LESS FULL CORE RESERVE = 1243 3RR 8-11
TABLE 8-5 SFP INVENTORY (LAST REFUELING) FOR BSEP UNIT NO. 2 Assembly Irradiation Cooling Discharge No. o f Power Time Time Date Batch Assemblies (MWt) (day) (hr) 11/86 1 140 BWR 4.35 1197 24 11/85 2 140 BWR 4.35 1197 8784 11/84 3 140 BWR 4.35 1197 17544 11/83 4 140 BWR 4.35 1197 26328 11/82 5 140 BWR 4.35 1197 35088 11/81 6 136 BWR 4.35 1197 43848 3/80 7 132 BWR 4.35 1197 58800 4/79 8 4 PWR 14.65 985 66600 3/79 9 132 BWR 4.35 1197 67560 1/78 10 40 PWR 14.65 985 77136 9/77 11 90*BWR 4.35 1195 79680 10/76 12 51 PWR 14.65 985 88008 11/75 13 6 PWR 14.65 985 97464 3/73 14 43 FWR 14.65 985 119880 1190 BWR 144 PWR CAPACITY 1839 BWR 144 PWR CAPACITY LESS FULL CORE RESERVE = 1279 BWR
*90 assemblies shipped back from Unit 1 8-12
TABLE 8-6 SFP INVENTORY (FULL CORE UNLOAD) FOR BSEP UNIT NO. 1 Assembly Irradiation Cooling Discharge . No . o f Power Time Time Date Batch Assemblies (MWt) (day) (hr) 3/88 1 140 BWR 4.35 299 24 3/88 2 140 BWR 4.35 599 24 3/88 3 140 BWR 4.35 898 24 3/88 4 140 BWR 4.35 1197 24 3/87 5 140 BWR 4.35 1197 8808 3/86 6 140 BWR 4.35 1197 17568 3/85 7 140 BWR 4.35 1197 26328 3/84 8 140 BWR 4.35 1197 35088 3/83 9 140 BWR 4.35 1197 43872 3/82 10 140 BWR 4.35 1197 52632 5/S0 11 245 BWR 4.35 1197 68184 1/79 12 87 BWR 4.35 1197 80208 1/78 13 6 PWR 14.65 985 88488 9/77 14 50 BWR 4.35 1197 91128 3/76 15 4 BWR 4.35 1197 104928 11/75 16 45 PWR 14.65 985 108912 5/74 17 102 PWR 14.65 985 121416 3/73 18 7 PWR 14.65 985 131376 1785 BWR 160 PWR CAPACITY 1803 BWR 160 PWR 8-13
TABLE 8-7 SFP INVENTORY (FULL CORE UNLOAD) FOR BSEP UNIT NO. 2 Assembly Irradiation Cooling Discharge No. o f Power Time Time Date Ba tch Assamblies (MWt) (day) (hr) 11/37 1 140 BWR 4.i5 299 24 11/87 2 140 BWR 4.35 599 24 11/87 3 140 BWR 4.35 898 24 11/87 4 140 BWR 4.35 1197 24 11/86 5 140 BWR 4.35 1197 8784 11/85 6 140 BWR 4.35 1197 17544 11/84 7 140 BWR 4.35 1197 26304 11/33 3 140 BWR 4.35 1197 35088 11/82 9 140 BWR 4.35 1197 43848 11/81 10 136 BWR 4.35 1197 52608 3/80 11 132 BWR 4.35 1197 67560 4/79 12 4 PVR 14.65 985 75360 3/79 13 132 BWR 4.35 1197 76320 1/78 14 40 PWR 14.65 985 85896 9/77 15 90 BWR 4.35 1197 88440 10/76 16 51 PWR 14.65 985 96768 l 11/75 17 6 PWR 14.65 985 106224 3/73 18 43 PWR 14.65 985 128640 1750 BWR 144 PWR CAPACITY 1839 BWR 144 PWR l 8-14
TABLE 8-8 ;
SUMMARY
OF RESULTS FOR SPENT FUEL POOL BULK TEMPERATURE ANALYSIS Maximum Time to Time to Temperature Reach 150*F Reach 212*F Case Cooling Mode (*F) (hr) (hr) Refueling SFP System 145.2 Never Never 1 SFP Rx/ Pump 182.6 1.9 Never None -- 1.0 13.5 Core Unioad SFP System 197.2 46.2 Never RER & SFP Systems 124.6 - Never RHR System 131.7 - Never S-15
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9.0 COST BENEFIT ASSESSMENT 9.1 NEED FOR INCREASED CAPACITY The spent fuel storage facility at the Brunswick Plant was originally designed for temporary storage of spent fuel until the fuel had cooled enough for transportation to a reprocessing facility. Delays in the licensing of the AGNS' 3arnwell fuel reprocessing plant, and the cessation of operation of existing facilities created the need for increased storage capability to permit continued plant operation. In recognition of this n-eed, additional capacity was added in 1977 and 1978 to provide a total storage capacity of 2,088 BWR and 304 PWR fuel assemblies. The indefinite deferral of reprocessing in the U.S., the uncertain availability of away-from-reactor storage, and the continued slippage in the scheduled, availability of a geological repository have made it necessary once again to seek interim relief by a further expansion of the Brunswick spent fuel pool capacity. The anticipated fuel discharge schedule for 3runswick Unit Nos. I and 2 is described in Table 9-1. A review of this schedule indicates that with our present system storage capacity, full core discharge capability for Unit do. I was lost after the 1980 refueling and Unic 2 will lose this capability in 1982. By combining storage capacity in both pools, it is possible to have a full core discharge until 1982 but for Unit No. I this would be a very time-consuming process. With our present storage capacity, both Unit Nos. I and 2 would be forced to shutdown in 1986. Expansion of the storage capacity by the use of General Electric high density poisoned storage modules to a total of 3642 BWR spaces will produce enough capacity to provide for a full core reserve for Unit No. 1 until 1988 and Unit No. 2 until 1987. Unit operations would be permitted until 1992 and 1991 respectively. 9.2 ALTERVATIVES TO INCREASED CAPACITY Several alternatives to the expansion of the storage capacities of the Brunswick spent fuel pool to alleviate the spent fuel storage space problem were considered. In sunmary, 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. a) Shipment to a Fuel Reprocessing Facility There are currently no commercial spent fuel reprocessing facilities in operation in the Unites States. In April 1977, the President of the United i States announced a spent nuclear fuel policy which included the indefinite deferral of commercial reprocessing in the U. S. nuclear power program. 9-1
Reprocessing of spent fuel is not a viable alternative to the expansion of the Brunswick spent fuel pools. Storage of the Brunswick spent fuel at the existing (although not operating) reprocessing facilities is also not a viable alternative to the expansion of the unit spent fuel pool since the facility owners are not offering to provide comparable storage capacity. b) 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 the CP&L system would not be economically viable. The Department of Energy has estimated that construction of a 5000 EnXI independent spent fuel storage facility would cost $200 million (DPE/ET-0055 " Preliminary Estimates of Charge for Spent Fuel Storage and Disposal Services," July 1978) or about S40/kg. A smaller facility designed to serve our needs 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 $14.75/kg). c) Shipment to Another Reactor Site The only available reactor sites which could be used as alternative spent fuel storage facilities within the CP&L system are the H. B. Robinson Plant and the Shearon Harris Nuclear Power Plant. The Robinson unit has the same fuel storage problems as 3runswick with only a slight variation on crucial dates. The Harris Plant has an expected commercial inservice date of September 1985, and will thus be unavailable in time to prevent the loss of full core reserve or possibly the forced shutdown of the units. d) Plant Shutdown Shutdown of the Brunswick 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 CP&L electric system for replacement power if the Brunswick units are shut down after the 1986 refuelings. 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 3runswick Plant are summarized on Table 9-3. These costs represent the current prediction of the total project costs, including the installation of the high density spent fuel storage modules and disposal costs of the presently installed modules. Indirect capital costs other than those specified have not been considered. The overall scope of the project will include the following: a) Design feasibility study. o, -
b) Design amendment 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 original modules, latching device, seismic restraints, etc. e) Installation of new modules. 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 foreclose the alternatives available with respect to any other licensing actions designed to ameliorate a possible shortage of spent fuel storage capacity. The material resources that would be consumed by the proposed modification are listed below. Brunswick Modificatioa Material Quantity (lb.) 304 Stainless Steel 2.36 x 10 5 Supports . 44 x 10 5 Boron Carbide 5.9 x 10 3 Aluminum 4 2.2 x 10 9.5 ENVIRONMENTAL IMPACT OF EXPANDED SPENT FUEL STORAGE For both RSEP units , the additional heat laod in the Refueling Case as imposed by the increased capacity of the spent fuel pool will increase the pool bulk temperature form 125*F, the original design value, to 14 5* F. The ventilation system for the spent fuel pool area in each unit has a capacity of 27,000 scfm. The evaporation rate of the spent fuel pool water from the surface of each pool increases by 348 lbs/hr, due to the increased water temperature. This small increase in the evaporation rate will have an insignificant effect on the ventilation system. In the core unload case, the expanded gtorage capacity increases the maximum spent fuel pool heat load from 27. 5x10 BTU /HR (based on the worst case analysis for the September 23, 1976 license amendment submittal) to 29.2x106 [ BTU /HR. The increase drops rapidly and vanishes about 12 days after reactor shut down at which time the actual heat laod becomes equal to the previous maximum spent fuel pool heat load of 24x10 BTU 6 /HR. Assuming that all the additional decay heat is ultimately transferred to the water in the service ' water system, the service water discharge temperature will be increased about 0.43*F. Therefore, under all expected and postulated conditions, the increased heat load as a result of the spent fuel pool storage expansion will have negligible 9-3 Revision 1
effect on the operation of the original plant components and negligible impact on the environment. 1 i 9-4
TABLE 9-1 BRUNSWICK UNIT 1 AND 2 ANTICIPATED FUEL DISCHARGE SCHEDULE I POOL LIMIT: UNIT 1 CURRENT 1026 BWR 160 PWR PROPOSED 1803 BWR 160 PWR tmIT 2 CURRENT 1062 BWR 144 PWR PROPOSED 1839 BWR 144 PWR UNIT 1 UNIT 2 TOTAL ASSEMBLIES IN POOL TOTAL ASSEMBLIES IN POOL DATE BWR PWR DATE BWR PWR JAN 1, 1980 320 154 JAN 1, 1980 132 105 JUL 1, 1980 476 154 APR 1, 1980 264 105-MAR 7, 1982 616 160 NOV 15, 1981 400 144 MAR 1, 1983 526 160 NOV 15, 1982 540 144 MAR 7, 1983 666 160 MAR 1, 1983 630 144 MAR 7, 1984 806 160 NOV 15, 1983 770 144 MAR 7, 1985 946 160 NOV 15, 1984 910 144 MAR 7, 1986 1086 160 NOV 15, 1985 1050 144 MAR 7, 1987 1226 160 NOV 15, 1986 1190 144 MAR 7, 1988 1366 160 NOV 15, 1987 1330 144 MAR 7, 1989 1506 160 NOV 15, 1988 1470 144 MAR 7, 1990 1646 160 NOV 15, 1989 1610 144 MAR 7, 1991 1786 160 NOV 15, 1990 1750 144 9-5
TABLE 9-2 INCREASES IN ANNUAL PRODUCTION COSTS IN ACTUAL YEAR 1000'S DOLLARS Scenario 1 Scenario 2 Scenario 3
- Robinson 2
- Brunswick 1 & 2
- Robinson 2, Nuclear Nuclear Brunswick 1 & 2 Off Off Off 1984 -16,296.6** -
-16,296.6**
1985 141,761.6 125,257.1 110,840.3 1986 201,446,0 168,248.0 454,166.0 1987 205,870.0 594,965.0 1,052,197.0 1988 161,730.0 452,018.0 815,048.0 1989 226,920.0 679,936.0 1,118,864.0 1990 184,271.0 540,174.0 954,911.0
- NOTES:
Robinson 2 forced to shutdown on November 1,1984 Brunswick 1 forced to shutdown on March 1, 1986 Brunswick 2 forced to shutdown on November 1, 1986
**The increase in production cost is negative because units are not scheduled for maintenance overhaul in the year they are being retired.
a 9-6
TABLE 9-3 SPENT FUEL POOL EXPA'iSION COST ESTIMATE Racks and Equipcent $2,870k Installation S 963h Engineet Saperv'.rior. & Od S 599k Contingenc7 $1,074k Allowance for Fands carir.g Construction S 342k TOTAL $5,848k ! 1 i i 1 9-7
10.0 RADIOLOGICAL EVALUATION 10.1 SPEVT 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 the spent fuel pool conductivity is above 1 micro cho/cm and the filter is not effective in reducing the conductivity level. Brunswick experience indicates that the filter-demineralizer was backwashed approximacely 10 times per year per unit. Each backwash cycle amounts to about 2.5 cubic feet of spent resin. The dose attributed to handling of the spent fuel pool resin in the radwaste system is less than 0.3 man-rem /yr. The increase in occupational exposure to personnel from the additional fuel assemblies themselves which could be stored as a result of the increased storage capacity resulting from this modification is negligible because of the depth of water shielding the fuel and the decay of the more active isotopes. Routine exposure increases resulting from radionuclide concentrations in the spent fuel pool water should not be significant, since the fuel pool filter-demineralizer units are capable of maintaining the design pool water cleanliness. The concentrations of airborne radionuclides in the spent fuel pool area result mainly from the most recently discharged batch of fuel and will decrease rapidly af ter refueling. Therefore, only a negligible increase, if any at all, in the spent fuel pool work area is expected aa a result of the increased number of assemblies stored in the pool. The only significant foreseeable increase in routino operational exposures is the possible increase in frequency of backwashing the fuel pool filter-demineralizers and the associated man-rem exposures of these operations. A very conservative estimate would be that the spent resin volume would double. Based on past experience, this would result in an addition of less than 0.3 man-rem / year to the total routine operational exposure for the Brunswick Plant. 10.2 NOBLE CASES 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. Af ter 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 quantity of all noble gases released from the pools and negligible l 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 minimum detectable level. 10.3 CAMMA ISOTOPIC ANALYSIS FOR POOL WATER Brunsick Unit I has undergone two refuelings and Unit 2 has undergone three. Typical radioactive isotope concentrations in the Unit I spent fuel pool water are presented in Table 10-1 at various dates as are gross beta measurements. 10-1 Revision 1
i l t i No long term trends of substantially increasing activity are noticeable in these measurements although there are some short term variations due to work in the pool. 10.4 DOSE LEVELS Routine dose level measurements at both units on the operating deck level have returned to equilibrium levels. Throughout this level, dose *zates may vary l from approximately 2 to 25 mr/hr but no increasing trends have been l identified. 10.5 AIRBORNE RADIOACTIVE NUCLIOES Air samples taken from the refueling floor atmosphere during and af ter 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. j 10.6 RADIATION PROTECTION PROGRAM The Radiation Protection Program is described in Section 12. 5 of the Brunswick 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 The present spent fuel racks being replaced by this modification will be decontaminated and stored on-site or disposed es low-specific-activity material. 10.8 IMPACT ON RADIOACTIVE EFFLUENTS The spent fuel pool has its own filter-demineralizer units and, under normal I circumstances, the pool water is not transferred to the liquid waste system. Therefore, negligible impact on liquid effluents from the plant is anticipated as a result of the increase in spent fuel storage capacity. As discussed in Section 10.5, negligible increase in the airborne radionuclide incentrations in the spent fuel pool area is expected. Therefore, negligible pacts on gaseous effluents from the plant is anticipated as a result of the _ncrease in spent fuel storage capacity. i The annual amount of spent resins to be handled will be increased, as discussed in Section 10.1. The resins will be processed through the Solid Radwaste System, as discussed in Section 9.3.4 of the FSAR. 10-2 Revision !
TABLE 10-1 BRUNSWICK STEAM ELECTRIC PLANT ISOTOPIC ANALYSIS OF SPENT FUEL POOL WATER UNIT 41 UNIT di ISOTOPE 2/23/79 5/2/90 ISOTOPE 2/23/79 5/2/80 Mn - 54 6.118 E-4 2.309E 4 Mn - 54 1.09E-4 8.?5E-3 Fe - 59 LL'
- LLD* Fe - 59 LLD* LLD*
Co - 58 2.14E-3 LLD* Co - 58 4.21E-5 tto. Co - 60 2. 4 57 E-4 4.053E-3 Co - 60 6.163E-4 1.033E-3 Zn - 65 5.01E-5 LLD* Zn - 65 LLD* LLD* Cs - 134 3.847g-5 3. 069 E-4 Cs - 134 1.164E-3 3.368E-4 Cs - 137 4.574E-3 1.178E-3 Cs - 137 1.307g-3 4.868E-4 Sb - 122 LLD* 4.113E-5 Sb - 122 1.36E-4 3.315E-5 measurements in microcuries/ml GROSS 3 ETA MEASUREMENTS DATE 1/11/78 7/12/78 1/17/79 7/18/79 1/15/80 7/17/80 1 1 UNIT 11 1.95E-3 7.62E-5 1.15E-4 1.45E-2 5.493E-2 8.16E-4 UNIT d2 4.68E-3 1.33E-4 1.60E-5 3.51E-4 1.735E-3 5.93E-4
*LLD - Lower Limit of Detection 10-3 i
11.0 ACCIDENT EVALUATION 11.1 SPE.VT FUEL SHIPPING CASK DROP - OUTSIDE OF FUEL POOL It is extremely improbable that the spent fuel cask could be dropped, inadvertently or otherwise, due to the redundancy of vital crane components, conservative design margins, periodic testing and inspecting, and operator qualifications and administrative operating procedures. 11.2 SPENT FUEL SHIPPING CASK DROP - OVER SPENT FUEL POOL Despite the high degree of improbability of a cask drop, there are limiting stops in the crane control system to prevent unwanted travel over the SFP, l as recommended in NRC Regulatory Guide 1.13, " Fuel Storage Facility Design Basis." 11.3 OTHER CRANE LOADS 11.3.1 In terlocks A series of interlocks protects against inadvertent motions which could possibly cause an accident, e.g. the BWR hoist blocks out the PWR hoist and vice versa. Control rods not fully inserted prevent installation or removal of fuel assemblies. The refueling platform cannot travel over the core when the reactor is in the start-up mode. Lif t heights of fuel hoists are limited by cut-out switches on the winch and also by mechancial stops. 11.3.2 Load Limiting and Load Slack Cut-Outs Each fuel assembly hoist is provided with a load limiting cut-out which is set at 50 lbs. above the weight of the fuel assembly for which it is used. Each hoist is also provided with a load slack cut-out to indicate binding during insertion. 11.4 RADIOLOGICAL IMPACT The radiological impact of bulk boiling in the spent fuel pool due to a l postulated loss of SFP cooling is examined in Section 8.4. The analysis indicates that even with the increase in spent fuel inventory the offsite doses are still well within the 10 CFR 100 limits. 11-1 Revision 1
l
12.0 CONCLUSION
S '. The information contained in this document to support the proposed modification satisfies the necessary applicable regulatory requirements to a allow NRC approval for Carolina Power & Light Company to rerack the Brunswick Steam Electric Plant, Units 1 and 2 spent fuel pools and demonstrates that the
, proposed modification can be safely accomplished. This proposed modification
- is the most cost effective and desirable alternative, and is in the best
! interest of the public. The proposed modification does not significantly change or impact any previous determinations which are documented in the 3runswick Steam Electric Plant Safety Evaluation Reports and Final Environmental Statements, and therefore precludes the need for preparation of an environmental impact statement. i i 1 4 1 l l t I f 12-1
{
~ \
j 13.0 NOTES AND REFERENCES J l 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. I
- 2. Boral is a product of Brooks and Perkins, Inc., consisting of a layer of boron carbide-aluminum (B4 C-A1) matrix bonded between two layers of
- aluminum.
+
References:
4-1 L. K. Liu, " Seismic Analysis of the Boiling Water Reactor," Symposium ay, Seismic Analysis of Pressure Vessel and Piping Component, Ft st National Congress on Pressure vessel and Piping, San Francisco, Califorot.2, May 3 1975. i i 4-2 W. C. Wheadon, " Friction Test of Graphite Base Materials Sliding Against Type 304 Stainless Steel Places", GE report No. C5445-TR-02, dated i I April 19, 1976. (Proprietary) ! 4-3 E. Rabinowic:, " Friction Coefficient Value for A High Density Fuel Storage System", GE Report VPF No. V5455, dated Janury 3, 1978. I
- 5-1 U.S. NRC Safety Evaluation for Yankee Rowe, dated December 29, 1976, i j Page 4, Structural and Material Considerations.
- 5-2 U. E. Wolff, "Boral From Long-Term Exposures at BNL and Brooks &
Perkins", GE Report No. 78-212-0079, dated December 14, 1978. i 5-3 3 rooks & Perkins Report, "The Suitability of Brooks & Perkins Spent Fuel Storage Module For Use in BWR Storage Pool", Report No. 577. l 5-4 A. J. Jacobs, "Boral Corrosion Test: 2022-Hour Results", GE Report l No. 77-688-120, dated December 15, 1977. (Proprietary) I j 7-1 C. M. Kang and E. C. Hanson, ENDF/B-IV Benchmark Analysis with Full Spectrum, Three-Dimensional Monte Carlo Modes, ANS meeting, November ! 1977. i ! 8-1 M. J. Bell, "0RIGEN Code - The ORNL Isotope Generation and Depletion," ORNL-4628. l 8-2 N. Eickelpasch and R. Rock, " Fission Product Release Af ter Reactor Shutdown," IAEA-SN-178/19. I 13-1
J ATTACHMENT 2 SPENT FUEL POOL STORAGE MODIFICATION RESPONSES TO NRC LETTER DATED JULY 14, 1981 Request for Information
- 1. Information needed for the subject review is:
A. Production drawings of the rack and rack components that show: material used, overall configuration, details, fabrication, welding, and inspection notes (if any). B. Drawings that show the construction features and details of the spent fuel pool and its liner including material, fabrication, and welding. C. Drawings of the current positioning of all racks (BWR and PWR) in the pool. D. Drawings showing the proposed positioning of all racks (Hi-Density, BWR and PWR) in the pool and the location of attachments on the walls (include dimensions). E. Key calculations of the seismic analysis of the rack and the spent fuel pool. F. Key calculations of the natural convection analysis. G. A list of plants where these racks have been previously licensed (NEDO-24076) . l
Response 1. A The drawings of the rack and rack components listed below show the material used, overall configuration, details, fabrication, welding, and inspection notes (if any). The drawings listed below are provided in Enclosure 1 to Attachment 2. Vendor Print No. GE Drawing No. Title VP-GEPA-030 C5472-E-104 Arrangement Units 1 & 2 VP-GEPA-031 C5472-E-105 Support Base Interface VP-GEPA-032 C5472-E-106 Support Base, West VP-GEPA-033 & 034 C5472-E-107 (162) Support Base, East VP-GEPA-035 C5472-E-108 Support Base Arrangement VP-GEPA-007 C5472-D-109 Slider Pad Assembly VP-GEPA-039 C5472-E-ll8 Temp. Control Rod Hanger VP-GEPA-040 C5472-E-119 Control Rod Hanger Inst'n. VP-GEPA-022 C5472-C-120 Control Rod Hanger Bars VP-GEPA-023 C5472-C-121 Slider Pad Shims VP-GEPA-041 C5472-E-142 Control Rod Transfer Station VP-GEPA-010 & 011 C5472-E-123 (162) EDFSS Module Assy. (13 wide) VP-GEPA-014 & 015 C5472-E-124 (l&2) HDFSS Module Assy. (15 wide) VP-GEPA-016 C5472-E-125 Base Assy. 13x15 VP-GEPA-017 C5472-E-126 Base Assy. 13x17 VP-GEPA-018 C5472-E-127 Base Assy. 13x19 VP-GEPA-019 C5472-E-128 Base Assy. 15x17 VP-GEPA-001 C5472-C-129 Fuel Support Plate VP-GEPA-002 C5472-C-130 Fuel Support Plate VP-GEPA-003 C5472-C-131 Closure Plate i VP-GEPA-004 C5472-C-132 Identification Plate
- VP-GEPA-005 C5472-C-133 Lift Lug VP-GEPA-006 C5472-D-134 Tube & Fitting Assembly VP-GEPA-042 C5472-E-135 Support Base, West VP-GEPA-043 C5472-E-136 Support Bases, West VP-GEPA-044 C5472-E-137 Support Bases, West VP-GEPA-045 C5472-E-138 Fuel Space Identification VP-GEPA-020 C5472-E-141 Spent Fuel Storage Tube l
Response 1.B: The drawings listed below show the construction features and details of the spent fuel pool and its liner including material, fabrication, and welding. The drawings listed below are provided in Enclosure 2 to Attachment 2. Drawing No. Rev. Date Title 9527-F-ll56 5 5/9/74 Reactor Building - Unit 2 Concrete - Sheet 1 Fuel Fool Girders and Walls 9527-F-1157 2 11/7/72 Reactor Building - Unit 2 Concrete - Sheet 2 Fuel Pool Girders and Walls 9527-F-1158 3 8/8/72 Reactor Building - Unit 2 Concrete - Sheet 3 Fuel Pool Girders and Walls 9527-F-1159 3 8/8/72 Reactor Building - Unit 2 Concrete - Sheet 4 Fuel Pool Girders and Walls 9527-F-ll60 0 3/27/72 Reactor Building - Unit 2 Concrete - Sheet 5 Fuel Pool Girders and Walls 9527-F-1161 1 5/31/72 Reactor Building - Unit 2 Concrete - Sheet 6 Fuel Pool Girders and W a lls 9527-F-1162 0 3/24/72 Reactor Building - Unit 2 Concrete - Sheet 7 Fuel Pool Girders and Walls
9527-F-ll63 5 5/22/81 Reactor Building - Unit 2 Concrete - Sheet 8 Fuel Pool Girders and Walls 9527-F-1164 3 8/8/72 Reactor Building - Unit 2 Concrete - Sheet 1 Fuel Pool Girders and Walls 9527-F-1165 3 8/8/72 Reactor Building - Unit 2 Concrete - Sheet 2 Fuel Pool Girders and Walls 9527-F-1166 3 8/8/72 Reactor Building - Unit 2 Reinforcing - Sheet 3 Fuel Pool Girders and Walls 9527-F-1167 0 2/28/72 Reactor Building - Unit 2 Miscellaneous Aluminum Fuel Pool Girders and Walls 9527-F-1168 4 2/16/77 Reactor Building - Unit 2 Miscellaneous Aluminum Fuel Pool - Sheet 1 9527-F-1169 1 3/6/72 Reactor Building - Unit 2 Miscellaneous Aluminum Fuel Pool - Sheet 2 9527-F-1170 4 2/16/77 Reactor Building - Unit 2 Miscellaneous Steel Pool Liners - Sheet 7 9527-F-il89 4 4/18/73 Reactor Building - Unit 2 Miscellaneous Steel Pool Liners - Sheet 1
i i
- o~
I ! 9527-F-1190 6 12/31/73 Reactor Building - Unit 2 I Miscellaneous Steel I Pool Liners - Sheet 2 1 9527-F-1191 5 4/18/72 Reactor Building - Unit 2 { Miscellaneous Steel Pool Liners - Sheet 3
- 9527-F-1192 2 6/7/72 Reactor Building - Unit 2 j Miscellaneous Steel d
Pool Liners - Sheet 4 9527-F-1198 2 6/7/72 Reactor Building - Unit 2 Miscellaneous Steel Pool Liners - Sheet 5-1 i j 9527-F-1199 6 2/9/81 Reactor Building - Unit 2 Miscellaneous Steel l l Pool Liners - Sheet 6 9527-F-1255 3 8/8/72 Reactor Building - Unit 2 Reinforcing - Sheet 4 Fuel Pool Girders and Walls 9527-F-1267 0 6/14/72 Reactor Building - Unit 2 Reinforcing - Sheet 16 Fuel Pool Girders and Walls 9527-F-250 7 8 2/12/75 Reactor Building General Arrangement i Section "B-B" Unit No. 2 ] t 4 4 y
- ,, ,-- - ,,,- ,- ,,,- ,- ,,, ---- -- , - - - , - - - , - , - -. - - - - , , - - - - - - - - - - - - - - - . - - . . - ~ - . - - - - -
Response 1.C Drawings of the current positioning of all PWR and BWR racks in the spent fuel storage pool are provided in Enclosure 3 to Attachment 2. It should be noted that the PWR and BWR racks are interchangeable and their positioning may be changed between now and the time of proposed reracking. I l L
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Response 1.D The proposed positioning of all high density BWR and PWR racks in the spent fuel storage pool and the location of all attachments on the walls are shown in GE drawing no. C5472-E-104. This drawing is included in the drawing package provided in Enclosure 1 to Attachment 2 (See Response 1. A).
1 l Response 1.E i A copy of key calculations of the seismic analysis of the rack and the spent fuel pool is provided as Enclosure 4 to Attachment 2.
Response 1.F Key calculations of the natural convection analysis will be provided in a subsequent submittal.
Response 1.C General Electric high-density fuel storage modules have been previously licensed for use at the following plants: Plant Utility
- 1. Monticello Northern States Power Monticello, Minnesota
- 2. Browns Ferry 1,2, & 3 Tennessee Valley Authority Decatur, Alabama
- 3. Hatch Units 1 & 2 Georgia Power Company Baxley, Georgia
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ATTACHMENT 2. ENCt.OSURE 4 Table of Contents Part 1: Configuration of Rack / Cask Arrangement Part 2: Loads Definition for Floor Slab A. Free Standing Rack Module G. E. Letters and Comments by UE&C B. UE&C Leading Generation on Existing Racks Part 3: Finite Element Model for Fuel Pool Slab and Walls Part 4: Analysis,and Results 9 -e 'N++ **+ + o w me -o . , = . . . . ..=,ww-ow. ..-we.- . , .=% , s_ _
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_ _ __ _y ,f 3 0 Id 1 4 d.!*4 a 33 h h ! !@ i' .! l BRUNSWICK STEAM ELECTRIC PLANT URE Carolina l Power & Light Company I SPENT FUEL POOL ^ I STORAGE EXPANSION
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Part 2: Loads Definition for Floor Slab - A. Free Standing Rack Module G. E. Letters and Coments by UE&C
c. Carolina Power & Light Company 75z7-683
; File: 0011-500-XXX-XIXB CU-10708 MAY 151981 f gg{
Mr. L. R. Scott, Project Manager United Engineers & Constructors, Inc. 30 South 17th Street Philadelphia, Pennsylvania 19101 CAROLINA POWER & LIGHT COMPANY BRUNSWICK PROJECT 1975-1977 - 1,600 MW - UNITS 1 A'ID 2 BRUNSWICK STEAM ELECTRIC PLANT MAXIMUM P0dL FLOOR LOADS ,
Dear Mr. Scott:
GE has completed their seismic axial:Irsis for BSEP high-density fuel storage modules. They found horizontal floor loadings did not change from preliminary esti=ates given to UE&C in July 1980; however, the
- vertical floor loadings were lower as shown below:
DBE Max. Vertical Reactor (lbs)
' Module Size Calculated Estimated Earlier 13x15 64,200 120,000 13x17 56,500 130,000 13x19 66,100 150,0,00 15x17 . 65,600 150,000 If you have any questions on the above, please do not hesitt.te to call me.
ABC/CER/bkp (0908) Yours very uly. cc: Mr. W. M. Biggs
- Mr. N. J. Chiang1 Mr. F. R. Coburn
.f .b A. B. Cutter - Vice President Mr. J. Craven Nuclear Plant Engineering Mr. C. R. Dietz .
Mr. E. A. Grimm Mr.. R. J. Groover '. Mr. J. M. Rucki UOB FILE 11 !.'9 Mr. R. L. Sanders LR SCOTT 11UO Mr. D. R. Sponseller 11US Mr. L. V. Wagoner BJ HUSF.' . TON g r ,. ,.y. p .. ;qg gjgg Mr. J. M. Waldorf - Mr. T. H. Wyllie r i?"IiMAN 11U6
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[ NU CLE/ R FUE! { '"U! F C COTDI 45 { ]I ts { lh*g h { k w a n a a { AND SERVICES
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CENGRtd. ELECTmc COMPANY.175 CURTN&R AVE., SAN JOSE, CALIFCRtJiA e6125 0 i ViSIO r4 M/C 859 , j . August 12, 1980 . I Mr. W. M.'Biggs l Nuclear Power Plant Services . l' Carolina Power & Light - P.O. Box 1551 . j Raleigh, fl.c. 27602 , d
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Subject:
Estimated Maximum 03E Seismic Leads on Brunswick Pool Floor -
Dear Mr. Biggs:
; Per Mr. A. Yeh of UE&C reauest, I have reviewed the seismic incut at j the pool floor level for the CBE condition in order to estimate ficar
- loadings due to die free standing HDFS modules. Because tt:e 05E floor i spectrum was fcund.to be almost equal to the SSE floor spectrum with
. certain narrow frequency range tne OBE spectrum even higher. it is my i[.,. judgecent that E5t of the SSE floor loadings can be used for the OSE
- condition floor evaluation. The estimated SSE floor loadings h:ve been l provided to you earlier on July 23, 1980.
l If you have any questions on the above, please do not hesitate to call. i Very truly yours, , GEriERAL ELECTRIC COMPANY
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P. C. Nun Y.1nacer l , Project Engineeiing ; PCS/cr cc: B. Huselton (UELC) F. H. Shadel . . 1 i. D. R. Sponseller . R. G. Tunell E. A. Grtrn (. . r .
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G E N E R A L h E L E CT P. I C AND SERVICES GENERAL ELECTRIC COMPANY 175 CURTNER AVE., SAN JOSE, CAUFORNIA 95125 DIVISION M/C 549 July 23,1980 Mr. W. M. Biggs - Nuclear Power Plant Services Carolina Power & Light P.O. Box 1551 Raleigh, N.C. 27602
Subject:
. Estimated Maximum !aismic Loads on Brunswick Pool
. Floor From HDFS Module
Dear Mr. Biggs:
The following tabulation sumarizes the subject estimated loads per UE&C request. , DBE DBE Module Size Max. Reaction (lbs) Max. Hor. Force (1bs) 13x15 120,000 27,000 13x17 130,000 29,000 13x19 150,000 33,000 15x19 150,000 33,000 The above loads include the weight of the module and are to be applied at each of the two supporting points on the front side of the module along the direction of earthquake motion. The horizontal and vertical forces are to be applied concurrently. A preliminary sketch showing pool arrangement is attached for your reference. Please note that these loads were estimated from Hatch application and no analysis has been performed because of tight time constraint. Please do not hes". ate to call us if you have any questions. Very truly yours,
, GENERAL ELECTRIC COMPANY Attachment
- E. A. Grimm, Manager Fuel Storage Projects EAG/PCS/mr ,
cc: B. Huselton (UE&C)
- F. H. Shadel .
D. R. Sponseller P. C. Sun , R. G. Tunell
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G E N E R A l.' h E I.E C T R IAND C SERVICES GENERAL ELECTRIO COVPANY,17s CURTNER AVE., SAN JOSE. CALIFORNIA 95125 DIVISION - M/C 859 July 23, 1980 Mr. W. M. Biggs - Nuclear Power Plant Services : Carolina Pcwer & Light .". P) 'i] j,g,3, P.O. Box 1551 Raleigh, N.C. 27502 j
Subject:
Estimated Maximum !aismic Leads on Brunswick Pool - l Floor From HDFS Module . - - . j
, I Dear Mr. Biggs-The following tabulation su:nari:es the subject estimated loa'ds per UE&C request. . . ,_ , , , _
,on. + Pacq r DBE kus DBE . . Max. Reacticn (lbs) Max. Hor. Force (lbsl ' Y Suoy, Module Size 53 l46 3 13x15 120,000 - 27,000
#66 2 13x17 130,000 29,000 f g y,o 13x19 150,000 33,000 b,3 33,000 150,000 S.o 214.o 15x19 The above loads include the weight of the module and are to be applied .
at each of the two supporting points on the front side of the module . along the direction of earthquake motion. The horizental and vertical forces are to be applied concurrently. A preliminary sketch showing pool arrangement is attached for your reference. [
.I Please note that these loads wers estimated frem Hatch application and l no analysis has been performed because of tight time constraint. Please i do not hesitate to call us if you have any questions. 1 St5C t .0 erwn Surers / 3 - 7'-2" 5-to Very truly yours, , / 5 - 8 '- I , io - Il GENERAL ELECTRIC COMPANY / 7 - f -Mf 7-0 .
I f - #'N' g ,, 3t 3c3,,nt 3-I Md.14.b E. A. Grimm, Manager g(,3 , * - Fuel Storage Projects 3 W EAG/PCS/ar 51 'Cc /ho /* rjgywgy. g Q '764-Con cc: B. Huselten'(N&C) 4 .To i .Wr., U SCN F. H. Shadel 8AcA::,S 4pp - D. R.'Sponseller -
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. t cbm soe7 Cev. 3-n GENERAL COMPUTATION SHEET
* " """ C ALC. SET NO.
mited acon en vucunn:.gineers mu. gez7-b b F lN AL f l-)2.,3 -)=$ Og- j"' waue on coua*Nv- unires I k volo suocer MgWWNE SIMPMr40Q . SHEET 229 OF 327 J.O. 9527. Ces-
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a p p w c. -
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Fo P_ b S E ' 1 , .w cmi Pacs. JL.c b AT CCM G. boe 7o N : zG/V"X = y._= z y. zo Xf z/o/</ W W P oF zvsy'
^
of SE'" *- ML'/M ~ r tue m t : z co t-/ "e , s/ = aq.zo cf e- z,24, I bJe7oV: //. Z_ /c = .;2. - 2 . Q. K fh (con awTh / 3D3* y '+ f 2+ a * = (zy.Zf+(2$2.f t-@.9Y = 3 caaimig By or - acs : ua,- ss.7 m (sta=x p.C) - ecoyaucye s.zxcegeup.Q-E r-crreura.=EFc 34 54 x - 3 3.'7 e. - 3.2 x.} .= zG ,7a K ^ T m- coe>L=e. L I Sl2Ss. = .34,34 e y l UJe:,- Fh = 7 65 g h - O + E Q = 34 54 + 7 63 = 92 x y /
L Form 5007 n 2.n GENERAL COMPUTATION SHEET
***" C ALC. SET NO.
.mited
-- engineers eaEuu. esz7-F lN A L f I-2B-FP-ccf- P 1 saus o, cou,4sv __.O b N unirn b votD
,,,,,,,_enae. sr~or susz. Pau <4per e. SHEET 23OOF 327 a.o. crs27. ces "ty COMP. BY CHK'D BY MO AAA bu e ro V : n.# = 2.ex tm *h/a "4h/ro 4
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due n u)3-B - 53.@32?,7.(,3g} T s =.c sedare. oy = z 2-7.(,3 = f.as f - sEsmc_ t:cuu =. z.e+7.cos = /o.% s-cHecc smR. SHE4R =zL/] = 23.to2 r = //. 2/c g4ce O- '
//
l // , . M$ Q.' c$ g AE. bSE g la Ceu Sc)E R4cK
-ME4t' =
C' . 4r cozeust. a.' : DOE D #t /307 "f = / 2. / X f / z<se l D o E 70 l '. /Y75 * = 27 3#f s Zs2.'7" DOE D \ 8. l.0A =. /,'/ ^ by - C
- ses.s = (qz,fy#cz7 3yyo,y)z- =
zS P f (ceu w ) /
*'* 5**7 "" 3 77 GENERAL COMPUTATION SHEET IOISCIPUNE) mited engineers a constncors oc PaELiu.
N27 FIN AL fgd.pf.Od- g
.- Naue on counanv I/.2f $$RP ,,, y m y,,,, (h L veio l I
susacevh..tE.irT Qi gm r CA f1 v' -'Eg5#p,yg SHEET Z(44 OF 327 J.o.qC27.c e r "t, COMP. BY CHK'O BY o SY MM Sh_ EdYlo N ML Boc.W . DATE DATC - WU ~O- & Q Q y N L R ec,rr e n 9 5 - _ C 2hM Ad.A(sgg e,.. CM
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' c"" C A LC. S E T N_O.
intted a en ar_gineers constructors PacLiu. esz7-FINAL /-gB.FP-of F C' NAME OP COMPANY. 4 W UNIT 4 g fyr. VOID su eJE cT. EbeT_%.67.c.Mqs Op.ig Tid {en.d..i.tdMT 7 "r y COMP. SY CHK'DBY sAY AM
?cy. o> c+ww,,.a
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% AL w m HT :
l W r - W g.+ W pt% C Ea + F*L+ wm _ ,# GnulcPO b%0Ms msTtu MTE2 cuycnMLy 24 THG Ge s-ic. u n e L - Beoone coc>i r . J FLLi ecycepiac, nq WM,e -
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porm scop new. M7 GENERAL COMPUTATION SHEET C ALC. SET NO. mited a en oc.gineers consvuctors Pacciu. I esz7-FlNAL fll-RB-F'P ?
;b hI6f u nit /s.d 5. h V010 l SHE E T 2% OF 37,7 sua;echf,dT h.Ed. b'TS.8.6.kE C/2y.ACd.{ M ?Ed.MfMf.!J T
[ J.O. fC C 27, o K "c y COMP. BY CHK'O BY shy Art #1 U31-Ro Ehsko
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Perm soor nm. 2.n GENERAL COMPUTATION SHEET (DISCIPLIN E) ~ ONQlO M PRELIM. l fg7 FINAL [l 1 -f3.f-g.0 Y F Naut of COMPANY L> S UNITS a VOID l sus;Ect .W .tbMcB[*y ( -- C.tT._bpgg,dg,q SHEET 2'l7OF 37,7 a.o4co.oJT "t y COMP. BY CHK'D BY sM AAA U'?I-% Nl/5~/80 DATE DATE NAM ?. ]?6A F JMy\.AcEHeJJ , VGbotTy , AccclerKFod (_4R.% lAe) Al $yer2.p=9TeoO) Ccp p, ptpg g S fAck pcw C F26 - Herat E c GA L e E - NO- 79pGr2 Ve L(,54/qcc) keel (cyg 28 2 0.co (19 g,l 8 g c.e l 76* 3 oco342> o.d 8 I o.3I g4+ 4 oco653 o.867_ o.39 i l t s o. o ' o -
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Porm soor oev.3.n GENERAL COMPUTATION SHEET (DISCIPLIN E) hh0 M PRELIM. fgg7 8 FIN AL l l-)g8.j jogy- f NAus os COMPANY .q.L UNIT /S-l VOID susJacT k2.6 d I f.) ? b M e f46,. p. , ,d dC1TV T/ApkVgn{ SHEET 2YR OF 37JJ
' I I J.O.@E),o g "t y COMP. BY CHK'D BY
%Y AAM b'k']-Se %s/se DATE DATE TAsts s : H4 < <m .pg.cs s A w mcq.
CW% LtepAL S0pcsr=%Ti O CoAf,.. 8 O N 'i 5 ~ - he To,e 08.e - He e # rA L (~ 5te v- 2. %Ae (.ege seu eag.a.p76 o" / l C. d- 1 + 17 % ,o < w z ld-G 3 i377 o
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'* " 5007 "25 3 77 GENERAL COMPUTATION SHEET (osselPUN E)
C ALC. SET NO. mited engineers a constructors m nEuu. e&/
#D C' FlN AL l (-g8.pf.QG ..f k ~. . o, coo,*~v C P4L CBsEFb _
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J.o. 9sz7-o8C
% Cac. SuJP_ R4ck FC*.OBE "'" ' " " ' ' ' **
MWD AAM_ Szz/e: Tk/go XIN S77C c UG2.s ] ke r = k xdc O#) c.= z. WNs2es /Gc = _L z. %\)t J YI p7*< #E 7 /# bcom gel, jyo" - - - C. s*4 geg itz " l h' * #Ey 2V" l F<el. mes b3 a= #Es 56" l 1 l s z. <>< $22 zg"
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Foez. c4d 3 _ N u.n w pos7. ,= z70364 l wik = 270% /&2' 0 /(so,96 l Uh.~ vh s Q =0& ~ Q = /60 9Ax zg"- 9s' os,z.# uJ, - u)7 = /Go.9 %, (ty"= z 2sz. 6,4 1
ro.m soor em. 3 77 GENERAL COMPUTATIOM SHEET l
* ' Sc""
CALC. SET NO. I mited a en w.gineers constructors P R E LIM. fSE7 - FINAL M /-fB-FP-c4 F NAus or COMPANY - U NIT /= VOID SUSJECT _ ! J.O.9527 020 "E y COMP. BY CHK'O BY
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! DATE DATE A
KE z== L f 9s' os.z - - 2 \. 3 & . y 4 /s\ec ' /, (, ,g g g /
- 2-s.a3 (./gsf .2ce"*
#E3 = 5,23 (. c/&/)' = /.Sc/9 "O' gey - s.es ( .ecz)' - y.33 z "d meg = s 23 ( I. 299)' = 9 232 "N
(' ge g = s.es (/.%Gy - /s./ez "* Ke = z.9z ( z.z%f = N.Go "4 XET
- NE'SD" ~
Esctio3 Prece (Etee s*M boe m r.e : GC) r c., !/z. [Hi
%g. = GSQ e //e = }uc(geQ} h.
m>u wcc : Cacahu; m) Kt= 5 E.1 (us/,'Q v M
Fem un am 3.n GENERAL COMPUTATION SHEET
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}I g1erigilleets PRELIM. fgg~/
, p FIN AL I /-ES-FF-c4 -F saue or counasy C P 8 (- ( O b E D \ ._ --
unirn / O 2-
. voto l suoscr 6!A.B 5fsyJrpcs_ Mg /Mg t.c4p:q__ 8 SHEET JS/ OF 327 J.O. 9527,o26 "t y COMP. BY CHK'D BY Q = 3(29eo@(329(d = 2.27do' ,} g W g Ehz/s G r/rc
= a,27xto"( t'.elo-5) = l ~61 xto* '/A.
= l G3*l6' %
%~L.22sid'(
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#9 = 2,&7 tic L w.\)
Kg - 2.ehto* / \ \ . z .aq< to' % . l ( oz') ! c N = z 27s tdz-Q/ y/ / o') N =
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o b M7 = 2,&7xId' __.L_ - 6.osx ior A/;a .
/G2" S H e A R. p e g:a s s :
Rz.= Iz Kz.(seg Yt. L 2
' {a(f.sico")(.zm]N = 7.zp rio"#
S4 6 - [z(/.63sio')(/.sw)}i- G.G3v.lo k = 6.# xto'#
@ ' [z(4.s6 lo')(v.332.T 2
Ac - {z(2. okie')(9.2w)]L= G.systo 4 L' .4,. = {z (l.oy v.iv)(/a.lez)]'/z.= 6. /s x!o*4 47=- a (G.osx b")( /Y co) } E = f 24 *Io" # . l
re,m soor c.v. >n GENERAL COMPUTATION SHEET (DISCIPLIN E) ~
,UMI0flglD88tB P R E U M.
952.7 -
., FIN AL f f hf.pf.QG f NAME OF CO M P A N Y.--- U NIT /= VOID l
$UBJECT ,,,,,, [_ ,,,, k . ,__ SHEET ZS2.0F %7 J.O.9527,02C "Ey, COMP. BY CHK'O BY
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gg j)g
'EIzzl20 h ' '/5/8a CATE DATE k4p
- 7. zq) *+ (G.G3)*+(6. W)*+. (G. 34')% CG ./s)*-(4 2)*] (l o'*) *
= /5304 7D7N BA5:5. SN E>hE. b]E 7e /Aly.I BASE MOMEJJr bOE m r_6 : ( M)
A giz P = ue Mz' Y' m ML = Wiki k M z. = b . [ z.
= (7.Z( x to* @)(zE") = z.o3 xto#"O M3 = CG.ssuoSd')(s/) = 3.71 x.to "4 ^
s Mg = (GW2 uo54)(9fD = 5,W uo* d4 M c = (6.3Y A10*#)( ll ?.") " 7. /c x10# #O M3= (G./5 xlON(/(lo")= 2 6/ y.lOf 'I M 7 = (%zo xio**)(/68")= 7,o6 x 1o3 "4 { 'T ' i.o3f+6 71)4(s.W)%(7./)L +('E,6l)*4(7.(S)']( lo P) [
/Y90 ## TC*rAl &*WE NOM &JJl' 'h0E '?o iM(T .
poem soo7 c.v.3.n GENERAL COMPUTATION SHEET C ALC. SET NO. talted en gineers s- PRELW. 952,7-
.,s FIN AL /-es-/:F-c4 F N AME OF COMPANY CP ,L (SS@b UNIT /S [. 2- VOto susJEcT__ d E J.o. 9sz7. 02s-
"E, COMP. BY CHK'O SY Ef M UE NUL{ W CP / D $CT kl$
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, , , DATE DATE rFssua c P bisTetBO77oo i
% - 6e.z7 h : mecuax/tc4c desi@ %
s H'ysnsa/s A-(207N NAratecoc S?ST fp l-lo9 p /-//o AI = fo227(39') = z4.(s (th fhcT. A-T THE SAME T5te) eer. posz. 4r. .'y ( InW-CT SAeAR h htoMSUT ODh No. cf infacr-Pce- Ass' 4 kg = /5.30 ic -
- f. 8 .
Muip = /t'90 p. 2 l 1 ADIUSTEb INDACT-SMR dMOMMT [G2 INM) All P *'3 #(##7)^ ' ' Y # M'ay = /490 K.(A2'D = lOl'1.2
- sas cous,bmbo : (selsuic.+14pce)
Av A Os.yly+ 00,ydf]'/z. = le.ez x C %. M = (n 9sy+ 0e n.2y i = ass z ""
Form soop an. 3 77 GENERAL COMPUTATION SHEET C ALC. SET NO. hinited engineers eaEtiu. I 9sz7-FINAL f( l-[8-ff.cg-.f NAME OF COMPANY - UNITg 2- Vol0 l
,u,,,n juJR. "SCGOT FdC- W kDjfr Mats SHEETES'/ OF 337 i
J.o. 9 5 2 7 o e r "E y COMP. SY CHK'D BY 5(o CELL BuJi? R&ClC Fce CGS 16+h, AA A4
- _ . - . . . %ho ' Dss 1 C., DATE DATE Y -
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c b AT COYfJ62. L 1 boe m // : 20 G,3, E " - % - 19.lO Ki z' 54" Ca DOER To 1 - 2.olo 3. Z "5 = = /1 /O *: [
/02 "
DCx5 7o Q ; M,& E z. 6 h $ ( W SH. /4 OF 28-F P- o t) 1 SlESS*Qf+yz+e a ' = ( 09.i)2+G9 lf+ t* ' 27 09
- I FFCM L0 3 - B = 33. 7- 3 2 = 7, G 3 'N J
1 l l E.Fa = 27 09 - 7.G3 = w
- 19. W i
.tr ca.2M;D?. % *-
Sl2SS = 2. 7,cq e h
% - S = 74 % \
(, y EFg = 27 0947,(,3 = 34 72xy l l
parm soor n s.n GENERAL COMPUTATION SHEET (DISCIPLINE)
.mited
~n engineers PaEtiu.
oAz7 i g~ FIN AL k / 28-FP-of- F' NAME OF COMPANY UNITn; VOID l su aJ a cT-.- ,2, b ST P4 M,,,,,,$ /_ cad >$, , SHEET 256 OF 37,7 J.O.95Z7 $
, "E y COMP. BY CHK'D BY k%b .AA b1 boe u V: 7 A z.ox } f h /a 'fbf/f.
1 doe = 4-a : (3s.7- 3.z.) . 7 53 e 4 7' EYc = SeiSMIb u b = 765 -2.o = Salo % n. Y v . DSMic BounJ = 7,GS+2 0 = 9,635 h 1 seem .L g . 12.roze,- g ate /m z r_. l C
,orm stor am 3 n GENERAL COMPUTATION SHEET I C ALC. SET NO. l
* ' Sc' * "
inited en wrgineers aconstructors PREUK l9%- FINAL ki-25-i=f-QG -{
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1 BUJh SHE E T ZS 2. OF 32,7 sus;Ec7 SFt=>Jr
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/ za,5.2"5
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If.hc h DusTo 9 5 E = M = 2. K y (SHTli or R B - F F-ol) V U)s-S . 7,6 br 1 i
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% - B , 7, G 2,c y 4
l Form 5007 Rev. 3 n GENERAL COMPUTATION SHEET
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mited engineers a consvuew. m eaEuu. esz7-
#lNAL l=k*ff= 0 '! ~ f
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& cor.>mt c : .
xsh /,AAf bus 7o f+: 1. = 17.la f slidVJ V7A/fe bue va 1.: y = 19,In f ..r. ..r. tutvav:awzetf
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/
EFa = 34.7%d/ l
Fe m so n a*v.3 n GENERAL COMPUTATION SHEET
""**" C ALC. SET NO.
mited engineers acm.vucun m eaEtiu. psz7-FlN AL f j~28-PP-Q y ~f NAus or couPANv hC UNiin Voto SUBJECT _ bh3N[ PAcC b., ..h([ . & H T % F 3g,7 l J.O. 9Sz7g "E y COMP. BY CHK'O BY
%CeL M EM-X
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fM K'l-R) }}A Yllr/sv Y)I'6/Co
- a. J DATE DATE 1
H C b (SEE .SljR.&3?f8SO THis Sfr) H: V 2+2n f {g A: y= 24 zc 4 4=E-=22eh( (S HT 1*b o .c-55 -i=P-OI)
%- % 7, Case k
. 4.83 4 EFg mmc 9 c-E-std / 0. 4.2 schie. towa= 9:ss.b er coeccg. k.
H-: )(.= 24 2m h 1 y = zY,tk 1 V os z,m 4f Ub -b s 'SS k
~~4 Q.m.
,. +. o EF b' 5"NC 'f ' #**N p. o SGSMIC tou))J = &c h_
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O so'o7 pev. [-F7 GENERAL COMPUTATION SHEET ,
*C"'" C ALC. SET NO. l mited engineers
& Consvuctors anc P R E U M. l 95g7-FINAL f/-28~Ff-Q) -F NAME OF COMPANY. -- UNITS ! VOID l
$UBJECT - SHEET 2860F 327 J.O. P r 2 7 5 5 \
"ty COMP. SY CHK'O BY T ** ' '
- vnh mat i+:p 2q.zat
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y,s z.e 4 , Sl2SSs 39,3[g p (consvD DJd -IS. 7,G3M H E Vc - zc Jr 1 y' c ymz d . (_e,
& x.= zV,2 c.4
.1 p zV Z5 [ .
V E=2&w4t sess - 39M (mos@ DJo -$ 7,G3rd q EFg = pgf_ _ _
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C
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Part 3: Finite Ele =ent Model for Fuel Pool Slab and Walls
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l . ATTACHMENT 3 SPENT FUEL POOL STORAGE RESPONSES TO NRC LETTER DATED NOVEMBER 5, 1981 i i i REQUEST NO.1: i What is the low friction material referred to on page 2-2 of the l ! April 6, 1981 submittal? 4
RESPONSE
Enclosed as Enclosure 1 to Attachment 3 is a General Electric report I entitled " Friction Tests Slider Pad-Support Base Materials." Also, enclosed as Enclosure 2 to Attachment 3 is a report to General Electric i
! entitled " Friction Coefficient Value for a High Density Fuel Storage i System." These studies relate to the identification and testing of materials to be used as slider pads and support bases and the determination of friction coefficients for these materials.
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! REQUEST NO. 2:
1 Provide copies of references 4-2 and 4-3. i I RESPONSE: A non-proprietary version of reference, 4-2 entitled " Friction Tests Slider Pad-Support Base Materials" by W. C. Wheadon, dated April'1981, is provided as Enclosure 1 of Attachment 3 in lieu of the proprietary report' requested. The primary dif ference between the non-proprietary and proprietary reports
! is the deletion of the actual test data from the non-proprietary report. A copy of reference 4-3 entitled " Friction Coefficient Value for a High Density Fuel Storage System" by E. Rabinowicz, GE Report VPF No. V5455, dated -
January 3,1978, is. provided as Enclosure .2 of Attachment 3. I a i e 4 i i 4
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REQUEST No. 3: l i Provide a discussion of the additional heat to be discharged to the i environment as a result of the expansion. i l
RESPONSE
1 Based on the following calculations, the impact of the additional heat to l be discharged to the environment as a result of the spent fuel storage
- expansion is negligible.
Assumptions: System Existing Heat Load Per Unit 6 Service Water 64 x 10 BTU /hr. 9 i Circulating Water 5.5 x 10 BTU /hr. Calculations: s AL = Fractional additional heat load to environment H , = 1981 Maximum Heat Load - 1976 Maximum Heat Load Existing Heat Load i t 4 6 AL H
= 29.2 x 106 BTU /hr - 27.5 x 10 BTU /hr (64 x 10D BfU/hr. + 55 x 100 BTU /hr) i 6
= 1.7 x 10 BTU /hr 5.564 x 109BTU /hr
-4 23.06 x 10 ALH = 0.0306%
l Maximum Heat Loads referenced in the Spent Fuel Storage Expansion Report. l Section 9.5, enclosed as Attachment 1. _ . , . . . - . - . _ _ _ . _ _ _ . - _ , _ _ . . _ _ . . .m _ _
REQUEST NO. 4: Provide the maximum heat load (BTU /HR) and' i ays cooling to 125'F per Table 8-3 for the analyzed refueling ant. core unload cases.
RESPONSE
The requested information is provided in Table 8-3, Revision 1 of the report entitled " Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Spent Fuel Storage Expansion Report." A copy of the revised report is enclosed (refer to Attachment 1).
r I w REQUEST NO. 5: What volume of solid radioactive waste is expected to be generated as a result of re-racking of the Brunswick, Unit Nos.1 and 2, spent fuel pool? Estimate the volume, as prepared for shipment (in drums, baxes or wrapped) to the disposal site. Estimate the curie content of this waste.
RESPONSE
This response will be provided in a subsequent submittal. l I
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1 REQUEST NO. 6: Our records indicate that the spent fuel storage pool cleanup system contains two 500'gpm, 300 ft. head, 60 HP pumps upstream of two heat exchangers and two filter /demineralizers. At what inlet water temperature do the pumps ! cavitate? Please indicate if your design for the proposed modifications includes any change to the automatic backwash of the filter /demineralizers or automatic bypass to maintain a minimum total flow of 1000 gpm through the SFP system (Figure 8-1). Also, indicate if-the proposed expansion is expected to have any impact on the anticipated annual volume of demineralizer resin waste from this source; if so, provide your estimate of the annual volume and curie content.
! RESPONSE:
i The spent fuel pool clean-up system pumps will not cavitate unless the inlet water temperature rises above 195'F. The following data has been used for these calculations:
- 1. The reactor building is actually operated at 0.25 inches of water negative pressure. Because this is too i
small a pressure to be significant, the reactor building , is as umed to be 14.7 psia (atmospheric pressure). ! .. The pressure drop calculation is based on the piping configuration as shown in drawing numbers 9527-D-2607 and 9527-D-2610 (not enclosed); also shown is the centerline of pump suction at elevation 82' 0". I
- 3. The valves, fittings, and pipe friction losses are calculated in accordance with Crane Technical Paper 410, x 1976 Edition.
- 4. The strainer friction loss is assumed to be 1.82 ft.
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- 5. The friction loss calculation is based on a pump flow rate of 500 gpm at 300 feet head, as shown in System Description SD-13.
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- 6. An allowance of 15% for piping old age / clogging and J
other contingency is added to the friction loss calculation.
- 7. Pumps operating at skimmer surge tank low-low level pump trip point of elevation 94'-0" as shown on Drawing Number 9527-LL-7044. (not enclosed).
The calcuation is summarized as follows: 4 ? Original pressure: 14.7 psia f Vapor pressure at 195*F: 10.385 psia Specific gravity of water at 195*F: 0.966 ) j Static head = 94-82 = 12 Feet Piping friction loss = 3.72 psi = 8.89 feet NPSH available = (14.7 - 10.385) x 2.31/0.966 + 12 - 8.89
= 13.42 feet i
Since the NPSH available at 195'F water temperature has a very small I margin over the NPSH requirement of 13 feet in accordance with System i Description SD-13, pump cavitation will occur when the inlet water i temperature rises above this temperature. Response to the remaining questions will be provided in a subsequent
- submittal.
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