ML20003F532

Spent Fuel Storage Expansion Rept.

ML20003F532
Person / Time
Site:	Brunswick
Issue date:	04/16/1981
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20003F529	List: ML20003F528 ML20003F531 ML20003F532
References
NUDOCS 8104220306
Download: ML20003F532 (88)

Text

.

BRUNSWICK STEAM ELECTRIC PLANT UNIT NOS. 1 AND 2 SPENT FUEL STORAGE EXPANSION REPORT S

g10422 0306

TABLE OF CONTENTS

~

Page No.

1.0 INTRODUCTION

............................................... 1-1 2.0 OVERALL DESCRIPTION........................................ 2-1 3.0 DESIGN BASES............................................... 3-1 4.0 MECHANICAL AND STRUCTURAL CONSIDERATIONS................... 4-1 4.1 SEISMIC ANALYSIS........................................... 4-1 4.2 STRESS ANALYSIS............................................ 4-3 4.3 _ FUEL BUNDLE / MODULE IMPACT EVALUATION....................... 4-6 5.0 MATERIAL CONSIDERATIONS.................................... 5-1 6.0 INSTALLATION............................................... 6-1 7.0 NUCLEAR CONSIDERATIONS..................................... 7-1 7.1 NEUTRON MULTIPLICATION FACT 0R.............................. 7-1 7.2 INPUT PARAMETERS........................................... 7-1 7.3 GEOMETRY, BIAS, AND UNCERTAINTY............................ 7-2 7.4 INTERACTION WITH EXISTING STORAGE SYSTEM................... 7-3 7.5 POSTULATED ACCIDENTS....................................... 7-4 1

8.0 THERMAL ANALYSIS........................................... 8-1

8.1 DESCRIPTION

OF THE SPEF_g'y. POOL COOLING SYSTEM. . . . . . . . . . 8-1 8.2 HEAT LOADS AND PORn. .1.*,:4,Pr.'.iTURES FOR PRESENT STORAGE l

i

, CAPACITY...........s........................................ 8-2 i

8.3 HEAT LOAD AND POOL TEMPERATURE FOR EXPANDED STORAGE CAPACITY................................................... 8-2 1

l l

TABLE OF CONTENTS (Continued) l l

Page No.

8.4 LOSS OF SPENT FUEL POOL C00 LING............................. 8-4 8.5 LOCAL FUEL BUNDLE THERMAL HYDRAULICS....................... 8-6 8.6 RADIOLOGICAL IMPACT OF SPENT FUEL POOL BOILING............. 8-6 9.0 COST BENEFIT ASSESSMENT.................................... 9-1 9.1 NEED FOR INCREASED CAPACITY................................ 9-1 9.2 ALTERNATIVES TO INCREASED CAPACITY........ ................ 9-1 9.3 CAPITAL C0STS.............................................. 9-2 9.4 RESOURCE COMMITMENT........................................ 9-3 9.5 ENVIRONMENTAL IMPACT OF EXPANDED SPENT FUEL STORAGE........ 9-3 10 0 RADIOLOGICAL EVALUATION.................................... 10-1 10.1 SPENT RESIN WASTE.......................................... 10-1 10.2 NOBLE GASES................................................ 10-1 10 3 GAMMA ISOTOPIC ANALYSIS FOR POOL WATER..................... 10-1 10.4 DOSE LEVELS................................................ 10-2 10.5 AIRBORNE RADIOACTIVE NUCLIDES.............................. 10-2 10.6 RADIATION PROTECTION PR0 GRAM............................... 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 SPERI FUEL P00L. . . . . . . 11-1 11 L . - - -_. - . ,. - __ _ _ _ _ -. - . - _ - _ _ _ . _ . _ _ _ ___

TABLE OF CONTENTS (Continued)

Page No.

11.3 OTHER CRANE L0 ADS.......................................... 11-1 11.4 RADIOLOGICAL IMPACT........................................ 11-1

12.0 CONCLUSION

S................................................ 12-1 13 0 NOTES AND REFERENCES....................................... 13-1 I

i I

iii

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 can id 2088 BWR and 304 PWR fuel assemblies. It was originally assumed that aaout 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 1977 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 modules.

It is desirable to have enough capacity in reserve to allow for a full-core discharge. Such capacity was lost in the Unit 1 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 modules to be installed and contains a discussion of the environmental and radiological considerations of the installation. The information contained herein hac been prepared based on the recommendations provided in " Operating Technical Position for Reviev 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 18, 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.

1-1

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.

j The HDFS module (free standing on the floor of the fuel pool) provides storage l

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 j

storage module sizes, 13x15, 13x17, 13x19, 15x17, are planned. 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 l additional s' paces to store 442 more BWR assemblies in the area presently used for control rod storage. (See Figure 2-5).

Module Fuel Configuration Cap city Quantity Assemblies Unit 1 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

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 TOTAL 2346 l Following installation of the above new modules, the maximum combined pool l storage capacity will be 3642 BWR assemblies (1803 in Unit 1 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 after 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 l storage modules. Figure 2-6 shows a typical HDFS storage module

! schematically. The smallest module (13x15) is a rectangular array approximately 7 feet by 8.25 feet and 14 feet high and contains 195 storage spaces.

The module support system consists of a module base plate, four fo.ot pad l assemblies and foer support pads. Figure 2-7 illustrates the module support system. The support pads rest on the pool floor and are elev,.ced 22 inches l above the floor to bridge the existing support grid. No additional loading is 2-1

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. At the east end of the pool, the er'.iting pipe braces and crossbeam may be modified to lower them and provide c' :rance for the support pads.

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.

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.

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• • 1 2 8  : a  ! 3 33 r ,i . .. d BRUNSWICK STEAM FIGURE ELECTRIC PLANT Proposed Hfgh Density Fuel Carolina od Power & Lignt Company e

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SPENT FUEL POOL d< l' l . =

STORAGE EXPANSION

COP' ER CLOSURE CLOSURE PLATE TUBE F

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BRUNSWICK STEAM FIGURE ELECTRIC PLANT Schematic of Typical Power & L g t Company HDFS Medule 2-6 SPENT FUEL POOL STORAGE EXPANSION l

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BRUNSWICK STEAM FIGURE ELECTRIC PLANT Carolina Module Support System Power & Light Company 2-7 SPENT FUEL POOL STORAGE EXPANSION

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_ / _/ M SUPPORT PAD ASSEMBLY BRUNSWICK STEAM FIGURE ELECTRIC PLANT Carolina Module Support System Detail Power & Light Company 2-8 SPENT FUEL POOL STORAGE EXPANSION

3.0 DESIGN BASES' .

l 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 v).th the environmental conditions associated with normal ope *;atics. 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 AS3 9-2 contained in the Standard Review Plan.

i

11. Light-Gage Cold-Formed Steel Design Manual, 1961 Edition, American Iron and Steel Institute.

12. 10CFR100.

3-1

r 4.0 MECHANICAL AND STRUCTURAL CONSIDERATIONS The high density fuel storage (HDFS) module is being analyzed for both Operating Basis Earthquake (OBE) and Design Basis earthquake (DBE) conditions.

A preliminary stress analysis has been performed to check the design adequacy of the module against calculated loads. Preliminary results indicate that the HDFS module design is adequate for the postulated combined loading conditions.

This section is written on the assumption the final analyses will verify the preliminary analyses results. The final analyses (including figures and tables) will be provided approximately May, 1981.

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 the 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 the 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 (to be provided approximately May, 1981) shows the plan view of the two-dimensional model of the modules and 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 significantly 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 l and displacement of the module. Water finite elements fill the spaces in between the walls and the modules. The modelling 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 NRC-QA requirements. The methodology in calculating hydrodynamic mass has been presented in Reference 4-1.

4-1

i i

),

I 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 j in the module base plate.

The sliding analysis for the high density fuel storage module model is represented by a triangle with three masses. This model preserves the 1

overturning and tilting moment of the rectangularly shaped module. A l 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 preserved. 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 stiffness of the fuel support plate, the foot pad, and the support pad.

j 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 0.349. This difference ensures that i I sliding will occur between the foot pad and the support pad, and not between l the support pad and the floor liner (References 4-2 and 4-3).

i The sliding analysis was done using the two-dimensional, non-linear DRAIN-2D and SEISM computer codes. DRAIN-2D was originally developed at the University of California at Berkeley; SEISM was developed by GE. Both computer codes have been design reviewed and meet NRC-QA requirements. Sliding and overturning of the module were studied for the DBE and OBE conditions. All of the modules were found to be stable under the worst postulated seismic loading i

4-2

dl precludes contact conditions, and the minimum 2-inch clearance between mo u es during a seismic event.

4.2 STRESS ANALYSIS i ments.

The HDFS module has been designed to meet Seismic Category I requ re h load Structural integrity of the module has been demonstrated d for t e combinations below using linear elastic design metho s.

i d in the Analysis was based upon the criteria and assumptions conta ne

'following documents: NF.

ASME Boiler and Pressure Vessel Code Section III, Subsection a) il USNRC, Regulctory Guide 1.92, Combining Modal Responses and Spat a b)

Components in Seismic Response Analysis, Seismic Design Criteria.

BSEP-142 Final Safety Analysis Report, c)

OBE - Operating Basis Earthquake d) DBE - Design Basis Earthquake i Iron e)

Light-Gage Cold-Formed Steel Design Manual, 1961 Edition, Amer can and Steel Institute.

Acceptance criteria were based on:

Normal and upset (OBE) Appendix XVII, ASME,Section III.

a)

Faulted DBE Paragraph F-1370, ASME Section III, Appendix F.

=. b) l l ted Local Buckling stresses in the spent fuel storage tubes were ca cu al" of Am c) of its according to " Light-Gage Cold-Formed Steel Design Manua h stress calculations.

applicability to these light-gage tubes. thickness of 0.090 inch nominal The applied loads to the module are:

d Dead loads which are weight of module and fuel assemblies, an a) hydrostatic loads. i Live loads - effect of lifting an empty module during installat on.

b)

Thermal loads - the uniform thermal expansion fuel. caused by pool c) temperature changes from the pool water and stored d) Seismic forces of OBE and DBE. height.

Accidental drop of a fuel assembly from the maximum possibl.e l e) l l

4-3 l

f) Postulated stuck fuel assembly causing an upward force of 2000 pounds and a horizontal force of 1000 pounds.

The load ccabinations 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 Handling 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 Type 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 Ft Bending Stress 0.66 Sy Bearing 0.9 Sy Note: Sy and Fe 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 strass between loaded and unloaded cells is less than 11,800 psi. This is well within the allowable compressive stress in the tube wall. Furthermore, 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 thermal 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-indu.ced stresses are not additive from module to module because each module is independent of the others.

4-4

^

l Stress analyses were done for both OBE and DBE 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 has 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 forcea to the side tube walls, which carry the forces down the walls and into the fuel aupport 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 the in-plane shear forces to the long casting (see Figure 2-7) which then transfers the shear forces to the module base plate (Point 4).

The forces are carried in the module base plate (Point 5) until they are l

ultimately transferred to the foot pad, support pad, and the pool slab (Point 6).

l The vertical forces caused by earthquake and gravity loads become forces on l 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 purpose, this compression force 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 Handling 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.

Also evaluated was the damaging effect of a fuel bundle drop through an empty i

storage position along the outer rows of the module, impacting the base plate.

l It was determined that the fuel bundle will not possess enough ene'rgy to perforate the 1-inch thick base plate. The resulting configuration of the l

l 4-5

module will be adequate to maintain the fuel in a safe condition. This case is less critical than the cases discussed in Table 4-3.

The loads that may be carried over the spent fuel pool are listed in Table 4-4. A free fall of the BWR associated loads onto the fuel pool liner plate and storage modules was evaluated. It was determined that of the loads thus far evaluated, a BWR fuel assembly drop causes the most damaging effect due to its weight and geometrical configuration. Also, none of the other loads can be lifted to a position higher than that of a fuel assembly above the liner plate and storage modules. A free falling BWR fuel assembly dropping from a height extending 27.5 inches above the height of a module with 0 ft./sec. initial velocity is calculated to have a final velocity of 26.5 ft./sec. when it comes in contact with the slab liner plate after traveling through the water. Based on this velocity, the liner plate that is provided for the pool slab will not be perforated. The presence of concrete below the liner plate was conservatively neglected in the computation.

Regarding the integrity of the fuel and storage modules, the consequences of

, dropping any of the BWR associated items listed in Table 4-4 are no more severe than that of the BWR fuel assembly drop accidents summarized in Table 4-3. An evaluation of dropping a PWR fuel assembly is in progress; the results of this evaluation will be avaiable approximately May, 1981. 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 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 i model is that the fuel bundle is separated from the module and is free to 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 lumped element model is identical to the gapped element model shown in Figure 4-5 except that the gaps between the fuel bundle and the module are ignored.

4-6 1

The objectives of this evaluation are:

a) To asses.s the difference in maximum internal forces in the module as determined from a gapped element model and a lumped element model, and b) To assess the effects 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 history to use in conjunction with rigid body displacements. Gap 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 p+= (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 differences 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 A11owables (psi)

OBE Condition SSE Condition Calc

• Calc

• Location / Type Stress Allowablesl Stress Allowablesl

-Tube wall shear 11,000 22,000 Tube wall compression 14,880 29,760 Tube weld throat shear 11,000 22,000 Angle, weld throat shear 11,000 22,000 Casting wall shear 11,000 22,000 Casting wall compression 16,500 33,000 Casting base weld shear 11,000 22,000 Support plate weld throat 11,000 22,000 shear Closure plate compression 14,880 29,760 Closure plate shear 11,000 22,000 Closure plate weld shear 11,000 22,000 Corner tube local compressive - - 17,224 stress check for local i

buckling l

l A Allowable stresses referenced in ASME Section III, Subsection NF

• Results to be provided approximately May, 1981 l -

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TABLE 4-2

DYNAMIC FREQUENCIES, EARTHQUAKE LOADING REACTIONS, AND MAXIMUM AMOUNT OF SLIDING Module Size Direction Fundamental Frequency (Hz) Max. Reaction (ibs) Max. Sliding (in) f e

i I

• Results to be provisled approximately May, 1981 ,

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TABLE 4-3 High Density Spent Fuel Storage Systea Assembly Drop Accident Case Summary No. Case Description ,,

Effect on Reactivity

1. A BWR fuel assembly drops 27.5 Analysis indicates that localized inches vertically and impacts the damage or fuel support member damage top of a fully loaded HDFSS will occur, but neutron absorber 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 HDFSS does not increase the system reactivity because the reactivity assumes an infinite vertical length of fuel (no neutron leakage in the vertical dimension). keff < 0.90

2. A BWR fuel assembly drops from Structural analysis indicates that 27.5 inches above the HDFSS, localized tube damage will occur and 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 BWR fuel assembly drops from Same as Case 2.

27.5 inches above the HDFSS and strickes a tube wall at an oblique angle.

4. A BWR fuel assembly drops from It is not possible for a fuel 27.5 inches above the top of a assembly drop of 27.5 inches to 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 overr.11 reactivity keff < 0.90.

5. A BWR fuel assembly drops from This case was analyzed for normal ,

27.5 inches above the HDFSS, handling conditions; keff < 0.90. J falls outside of the loaded .

HDFSS, and lodges adjacent and parallel to an unpoisoned, occupied fuel storage position. 1 I

4-10 l

TABLE 4-4 Items That May be Moved Over The Spent Fuel Pool Racks Item Approximate Weight (ib)

BWR Fuel Assembly (Including Channel) 687 BWR Channel 62 BWR Control Rod 235 Fuel Sipping Equipment

• Defective Fuel Cannister PWR Fuel Assembly (Including Control Rod) 1605 l

4 l

l l

1 l

• Results to be provided approximately May,1981 4-11

TABLE 4-5 Normal'ized Rigid Body Displacement of Lumped And Gapped Models Friction Coefficient Gapped Element Model/

(p) Lumped Element Model W

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• Results to be provided approximately May , 1981 4-12

- _ . . . . . .-. - ._ - _ _ _ ~ _ _ - _ . . _.

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Friction Caefficient Gapped Model Lumped Model (u) Force (lbs.) Force (lbs.)

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Carolina 4-2 Power & Light Company Design Basis Earthquake SPENT FUEL POOL STORAGE EXPANSION

BRUNSWICK STEAM, - - - -

ELECTRIC PLANT FIGURE Ca olina (To Be Provided Later) 4-3 Power & Light Company SPENT FUEL DOOL STORAGE E7PANSION l

EARTHQUAKE FORCE- l l- ~

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@ Fixed point Element Definitions: 1 Channel beam element (nodes 1 & 2) 1 2 Fuel bundle element (nodes 1 & 2) '

1 3 Gapped, hydrodynamic element (nodes 2 & 3) 4 Elastic-plastic spring element (nodes 1 & 4)

BRUNSWICK STEAM FIGURE ELECTRIC PLANT Carolina Capped Element Model 4-5 i Power & Light Company i SPENT FUEL POOL '

STORAGE EXPANSION l

5.0 MATERIAL CONSIDERATIONS Most of the structural material 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 (B 4C) are verified by chemical analyses and/or neutron transmission tests, and each sheet is dimensionally inspected. Before each piece of Boral is inserted into a tube assembly successful performance of the required inspections is verified.

The presence of the neutron absorber material in the fabricated fuel storage module will be verified prior to first use at the reactor storage pool site by scanning each storage tube in the modules with a neutron source and neutron detectors. The recorded results provide a comparison between neutron absorption through each Boral sheet and neutron absorption measured through the stainless steel without Boral. A significant increase in neutron absorption verifies the presence of Boral.

Boral's corrosion resistance is similar to that of standard aluminum sheet.

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

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, and 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 life of the facility.

Pool 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).

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5-2

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 High 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 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 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 ar:e located outside l 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. Modification of the pipe bracing and cross beam which spans

this area may be required prior to installation of the support pads to ensure that fuel in the HDFS module can be sufficiently submerged.

The following is the sequence of events for performing the work associated with the removal of existing modules and installation of HDFS modules:

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 hardware.

d) Relocate stored control rods.

e) Remove existing control-rod storage hardware. This work will require

! divers.

6-1

f) Make necessary modifications to the grid bracing if required. This work will require divers.

g) Remove waste materials 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 l assistance from divers as required.

k) Survey the installed support base pads to determine if shimming of module 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; 12 vel 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 shimming cf the module foot pads is necessary to ensure the modules are installed plumb.

l 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 man-hours include not only construction and operating personnel u

1 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: 1 1

Above the Divers in Decon. Non-rad. I Pool the pool Work Work Total l

Manhours 5,350 650 400 2,700 9,100 1

' Millirem exposures were escimated for each category of work described above. ,

Measurements based on experience indicate that the average exposure rate above '

the pool normally does not exceed 8 mrem / hour, with an average rate expected at 5 mrem / hour. For purposes of calculating the total man-rem, 5 nrem/ hour was assumed. Radiation exposure at the bottom of the pool in the area in question, with allowances for vacuuming and moving of storea 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).

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l 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/B-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 Puo2 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 10.002 (10) Ak with respect to the above critical experiments. Therefore, MERIT underpredicts 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 enrichment and burnable poison combinations.

c) Storage space pitch of 6.563 in.

d) Minimum allowable boron (B10) areal concentration of 0.013 grams B10/em2 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.

7-1

i 7.3 GEOMETRY, BIAS, AND UNCERTAINTY The repeating cell geomet'ry 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 j 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.

i 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 10.002 (la) ak.

An approximate geometry model, shown in Figure 7-4, was used for case 5 in Table 7-1. The model geometry bias relative to the exact model for the same conditions was 0.0067 + 0.0050 (lo) ak. The MERIT program bias remains the same at 0.005 + 0.002 T1o) 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/ce.

i The design of the High Density Fuel Storage (HDFS) System has alternating l spaces on the periphery of each module fabricated with unpoisoned closure plates. (The two opposite non-tube corners of each module contain Boral 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 Table 7-3. The maximum module k. occurs at the minimum possible water gap (1.244 in.) and is less than that of an infinite array of storage cells with no water gap.

The module calculations in Table 7-3 were done with the model shown in Figure 7-6. Some of the details in the exact model were homogenized and simplified to reduce the input preparation in the module calculations. The model geometry bias was determined from an infinite array of simplified storage cells (Figure 7-7) relative to the exact geometry model. The module geometry model bias was determined to be 0.0017 2 .0051 (lo) 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 m

enrichment and buenable poison distributions within the bundle. Fuel pin spacers were not included (a conservative exclusion). The nominal bundle dimensions were used for all cases.

The sensitivity of k. analyses to various changing parameters is implied above. More specific relationships are as follows:

a) Bundle Reactivity (percent U235) - Calculations are based on maximum

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 l 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-moderated the optimum k. occurs at 1.0 g/cc.

d) Storage lattice pitch - An analysis was done using a minimum fuel pitch, represented by the storags tubes touching. Material tolerances in the tubes were taken to maximize the k. of the storage lattice. The result of the analysis is given as Case 6 in Table 7-1. A comparison 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 designad and fabricated to prevent significant quantities of air or other gas from being l 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 k gfe of the system.

7.4 INTERACTION WITH EXISTING STORAGE 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 systems 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 communication between systems.

, Calculations were made to determine the interaction between the faces of two l

high density modules, with partially unpoisoned storage locations directly l opposite each other. These calculations (described in Section 7.3 and 7-3

tabulated in Table 7-3) support the conclusion that neutron multiplication factor is insensitive to intermodule water gap.

~

The maximum neutron multiplication factor for combined systemr, is therefore, the value calculated for PWR fuel storage in the existing system, k egg = 0.905.

7.5 POSTULATED ACCIDENTS 7.5.1 Fuel Drop 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 of the modules or alongside the modules at the support grid.

7.5.2 Loss Of Pool Cooling The normal cooling of the spent fuel pool is accomplished by the spent fuel pool cooling system which consists of two parallel cooling loops uharing common supply and discharge piping. Loss of one cooling loop subsequent to the last expected refueling would reso't in a maximum pool temperature of 182.6*F. A single failure in the sht piping could cause loss of both cooling loops allowing the pool to re; .i 150*F in 57 minutes, and ultimately permitting pool Solk boiling in 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, if no actions were taken.

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9 7-4

TABLE 7-1 Single Cell High-Density Fuel Storage Criticality Restits Case Description L (+ 2a)1 1 Nominal Rack Dimensions 2 0.8668 1 0.0075 With now Channel @20'C 2 Nominal Rack Limensions Without now Qiannel @20*C 0.8674 1 0.0086 3 Same as Case 2 except 065*C 0.8561 1 0.0084 4 Increased Pitch without now Channel @20*C 0.8364 1 0.0106 5 Same as case 2 but with Eccentric Bundle Position 0.8276 1 0.0123 6 Minimum Pitch 3 without now Channel @20'C 0.8650 1 0.0088 l

l l b includes MERIT Program Bias aad Uncertainty 26.563 inch Pitch with Nominal Material Thickness 36.503 inch Pitch with Minimum Storage Ibbe Material Tolerances to Maximize L l

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7-5

TABLE 7-2 Bias and Uncertainty Components for Maximum b of a Storage Cell L 0.8624 Calculational Convergence ak 0.0038 MERIT Bias and Uncertainty ak 0.005 0.002 Model Bias and Uncertainty ak None Total 0.8674 0.0086 (2a)

)

7-6

TABLE 7-3 HDFSS Criticality Analysis Module Interaction Description k. (+ 2a)

' Minimum gap between modules 0.8593 1 0.0131 (2A = 1.244 in.) .

Intermediate gap between 0.8579 1 0.0130 modules (2A = 2.100 in.)

Nominal gap between modules 0.8506 2 0.0134 (2A = 2.967 in.)

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BRUNSWICK STEAM FIGURE ELECTRIC PLANT Carolina Simplified Cell Model 7,7 Power & Light Company SPENT FUEL POOL STORAGE EXPANSION

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.

The pumps take suction from the skimmer surge tanks, circulate the water i

through the heat exchangers and the filcer demineralizer, and discharge l through the diffusers at the bottom of the spent fuel pool. The cooled water j 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 l reactor building closed cooling water and service water systems are discussed l in Sections 10.6 and 10.8 of the BSEP FSAR.

l The residual heat removal (RRR) 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 gpm can be maintained in the RHR loop in this cooling mode.

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 under these conditions are provided in Tables 8-1 and 8-2, respectively.

l 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.

l 8-1 l

2. The RHR system, operated alone or in conjunction with the SFP cooling system, shall maintain the SFP bulk temperature at or below 150'F folloVing a full core unload.

8.2 HEAT LOADS AND POOL TEMPERATURES FOR PRESENT S,TORAGE CAPACITY The design conditions in this seerdon are those selected from the best data available at the time of the 1977 license submittal. Because of this and additional system modifications for the expansion of the spent fuel pool storage capacity, the results based on this data (shown in Table 8-3) are not comparable with the results in Section 8.3.

The present spent fuel pool for each BSEP unit can accommodate 35 full-size storage racks for BWR or PWR assemblies and 7 half-size storage racks for BWR assemblies. Each full-size BWR rack can store 36 BWR assemblies and each PWR rack can store 16 PWR assemblies.

Three design conditions were postulated for the spent fuel pool cooling system.

a) Worst Case Condition In this analysis the pool receives a full core discharge in addition to the inventory that existed prior to the start of the core discharging.

This pre-unload discharge consists of the three most recent refueling discharges from BSEP and the three most recent refueling discharges of PWR assemblies from H. B. Robinson.

b) BSEP-1 Expected Case, Normal Refueling In this analysis the pool has been receiving a normal refueling discharge of BWR assemblies each year since September 1978 and PWR assemblies from H. B. Robinson between October 1976 and March 1980 and is full by September 1983.

c) BSEP-2 Expected Case, Normal Refueling In this analysis the pool has been receiving a normal refueling discharge of BWR assemblies each year since April 1977 and PWR assemblies from H. B. Robinson between October 1976 and March 1979 and is full by April 1983.

The results of the pool bulk temperature analysis are shown in Table 8-3.

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

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 l system. In order to verify the adequacy of the existing cooling ,

system in maintaining the SFP water temperature within allowable limits, two l cases, as described below, have been considered. '

l Case 1 - Refueling Case in this case, the spent fuel pool heat load corresponds to the pool inventory l immediately following the refueling, which leaves a reserve storage enough for i 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 eind 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 November 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 Spent Fuel Pool 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 211 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 poal would be 59 hours6.828704e-4 days <br />0.0164 hours <br />9.755291e-5 weeks <br />2.24495e-5 months <br /> after shutdown for the refueling condition and 164 L urs after shutdown for the core unload condition.

Figure 8-2 and Figure 8-3 show the total decay heat loads for BSEP 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following reactor shutdown.

B-3

i j 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 perfdrmed 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 l cases in question. The results are presented below:

, 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 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after fuel transfer starts and eventually drops below 125'F at about 35 days after reactor shutdown. It should be noted that in these plots, time zero denotes the time when the fuel transfer is started, i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after unloading started and the maximum pool temperature would be 197.2'F. However, under the core upload concition, the RHR system is available for cooling the SFP as discussec in Section 8.1. Because of the high heat removal capacity of the RHR system, the SFP temperature will drop rapidly as soon as one train of the redundant RRR 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the unloading is started, the pool temperature can be kept below 150'F.

If the SFP system and one train of the RHR 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 effect 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

8.4.1 Refueling Case For either ugit, if a failure occurred within the SFP cooling system during the refueling operation, the RRR system would be available for cooling the spent fuel pool as in the Core Unload Case and the pool temperature could be maintained below 150*F. Therefore, significar.t increases in t'he pool ,

temperature could result only if the SFP cooling system failed after the I 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 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> and the maximum temperature is 182.6*F.  !

l The maximum pool temperature that would result depends on the time of '

failure of the SFP components. If the failure occurs 35 days after l shutdown, the pool temperature can be maintained below 150'F with only 1 one SFP heat exchanger / pump operating. I 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 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The make-up water requirement following boiling is 28 gpm.

The radiological 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 t ime . BSEP-2 is refueled much later than BSEP-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 would have decreased significantly by that time. However, for conservatism, a i

radiological analysis has been performed assuming that the conditions in the BSEP-1 pool are identical to those in the BSEP-2 pool. That is, both spent 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 Unload Case For the core unload situation, a single train of the redundant, seismic 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 B-5

is, therefore, inconsequential so far as the 150'F pool temperature limit and pool boiling are concerned.

The pool temperature 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 fue], 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.

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 MWe/ Assembly Total Core Power Level 2436 MW e Assemblies per Core 560 Transfer Time Core to Pool 15 minutes l

Cooldown Time Prior to Fuel Movement 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Fuel Storage Pool Bulk Water Temperature 150*F The ORICEN 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 toth BSEP l

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.

i

! 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 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> for both units.

2. The boiling rate of the pool water is 1.38 x 104 lbm/hr for each unit.

! 8-6

3. The volume of water in each pool is 45,360 cu. ft.

4. All failed feel roda of the full core (average 1 percent of the core) are present in the 1/4 core discharge to each pool.

5. The normal :-131 release rate coefficient for leaking rods to the pool is 4.6x10-10 sec -1 at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 hours6.828704e-4 days <br />0.0164 hours <br />9.755291e-5 weeks <br />2.24495e-5 months <br /> 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 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.

l I

l l

1 1

8-7 i

l

TABLE 8-1 SPENT FUEL POOL HEAT EXCHANGERS PERFORMANCE DATA No. of Units 2 Heat Transfer Area Per Unit, ft2 1285 Design Temperature, 'F 150 Heat Transfer Coefficient, Btu /hr-ft 2 *F 261 Heat Transfer Rate Per Unit, Btu /hr 3.75x106 Shell Side Tube Side (Closed Cooling Water) (SFP Water)

Design Pressure, psig 150 200 Flow Rate Per Unit, gpm 800 500 Inlet Temperature, 'F 100 125 8-8

TABLE 8-2 RESIDUAL HEAT REMOVAL HEAT EXCHANGERS PERFORMANCE DATA No. of Units 2 Heat Transfer Area Per Unit, ft2 3,280

' Design Temperature, 'F 400 Design Pressure, psig 450 Heat Transfer Coefficient, Btu /hr-ft 2 _oF 281 Heat Transfer Rate Per Unit, Btu /hr 24.5x106 l

Shell Side Tube Side (SFP Water) (Service Water)

Flow Rate Per Unit, gpm 4,950 8,000 Inlet Temperature, 'F 125 90 l

i l

l

{

l l

8-9

TABLE 8-3 RESULTS OF BULK TEMPERATURE ANALYSIS FOR PRESENT STORAGE CAPACITY Maximum Cooling Heat Load Maximum Days Cooling Case Mode 106 BTU /HR Pool Temp. To 125'F a SFP & RHR 27.5 128 -

b SFP 10.9 136 8 c SFP 11.4 138 11 l

This analysis was performed originally for the 1977 license submittal with higher safety margins in the calculation of heat loads and other system parameters in order to bound uncertainties. Because of this and additional system modifications for the expansion of the spent fuel pool storage capacity, the results presented in Table 8-8 are not comparable with those shown above.

8-10

TABLE 8-4 l -

SFP INVENTORY (LAST REFUELING) FOR BSEP UNIT NO. 1 Assembly Irradiation Cooling Discharge No. of Power Time Time Date Batch Assemblies (MWt) (day) (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 %3848 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 9ES 100128

5/74 13 102 PWR 14.65 985 112632 i

3/73 14 7 PWR 14.65 985 122592 l

1226 BWR 160 PWR CAPACITY 1803 BWR 160 PWR

, CAPACITY LESS FULL CORE RESERVE = 1243 BWR I

1 l

I 8-11

TABLE 8-5_

SFP INVENTORY (LAST REFUELING) FOR BSEP UNIT NO. 2 Assembly Irradiation Cooling Discharge No. of 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 PWR 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 I l

l 8-12

TABLE 8-6 SEP INVENTORY (FULL CORE UNLOAD) FOR BSEP UNIT NO. 1 Assembly Irradiation Cooling Discharge No. of 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/ * ,4 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 5/80 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 1

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 B-13

TABLE 8-7 SFP INVENTORY (FULL CORE UNLOAD) FOR BSEP UNIT NO. 2 Assembly Irradiation Cooling Discharge No. of Power Time Time Date Batch Assemblies (MWt) (day) (hr) 11/87 1 140 BWR 4.35 299 24 l

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 11/83 8 140 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 PWR 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 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 -

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 Ex/ Pump 182.6 1.9 Never None -

1.0 13.5 Core Unload SFP System 197.2 46.2 Never RRR 6 SFP Systems 124.6 -

Never RHR System 131.7 -

Never l

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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' Barnwell 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 need, 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 Brunswick 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 No. I was lost after the 1980 refueling and Unit 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 ALTERNATIVES TO INCREASED CAPACITY Several alternatives to the expansion of the storage capacities of the Brunswick spent feel pool to alleviate the spent fuel storage space problem were considered.

In summary, the alternatives were:

a) Shipment to a fuel reprocessing facility.

b) Shipment to an independent spent fuel storage facility.

c) Shipment to another reactor site.

d) Shutting down the reactor.

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 States announced a spent nuclear fuel policy which included the indefinite deferral of commercial reprocessing in the U. S. nuclear power program.

9-1

~ -

l Reprocessing of spent fuel is not a viable alternativa to tha exp:nsion of tha 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 MTU 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 $40/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 J storage problems as Brunswick with only a slight variation on crucial dates.

The Harris Plant has an expected commercial inservice date of September 1985, 1 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 withf 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 refoelings. 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 ~

Brunswick 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.

9-2

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 Modification Material Quantity (ib.)

304 Stainless Steel 2.36 x 105 Supports .44 x 105 Boron Carbide 5.9 x 103 Aluminum 2.2 x 104 9.5 ENVIRONMENTAL IMPACT OF EXPANDED SPENT FUEL STORAGE For both BSEP units, the additional heat load in the Refueling Case as imposed by the increased capacity of the spent fuel pool will increase the pool bulk temperature from 125'F, the original design value, to 145'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 storage capacity increases the maximum spent fuel pool heat load from 27.5x106 BTU /HR (based on the worst case analysis for the 1977 license 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 load becomes equal to the previous maximum spent fuel pool heat load of 24x106 BTU /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

effect on the operation of the original plant components and negligible impact on the environment.

4 9-4

TABLE 9-1 BRUNSWICK UNIT 1 AND 2 ANTICIPATED FUEL DISCHARGE SCHEDULE POOL LIMIT: UNIT 1 CURRENT 1026 BWR 160 PWR PROPOSED 1803 BWR 160 PWR UNIT 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 J AN 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: I 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.

9-6

TABLE 9-3 SPENT FUEL POOL EXPANSION COST ESTIMATE Racks and Equipment $2,870k Installation S 963k Engineer Supervision & OH $ 599k

Contingency $1,074k Allowance for Funds During Construction s 342k TOTAL $5,848k i

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10.0 RADIOLOGICAL EVALUATION 10.1 SPENT RESIN WASTE i

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 mho/cm and the filter is not effective in reducing the conductivity level.

Brunswick experience indicates that the filter-demineralizer was backwashed approximately 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 after refueling. Therefore, only a negligible increase, if any at all, in the spent fuel pool work area is expected as a result of the increased number of assemblita stored in the pool. The only significant foreseeable increase in routine 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 GASES Krypton-85 is released to the pool water and subsequently to the refueling floor atmosphere from the leaking fuel assemblies. For normal operating conditions, most of the krypton comes from the most recently discharged batch of fuel. 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 fueel pool storage capacity will be small compared to

• the total quantity of all noble gases released from the pools and negligible when compared to the annue.1 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 GAMMA ISOTOPIC ANALYSIS FOR POOL WATER Brunswick Unit I has undergone two refuelings and Unit 2 has undergone three.

Typical radioactive isotope concentrations in the Unit 1 spent fuel pool water are presented in Table 10-1 at various dates as are gross beta measurements.

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No long term tends of substantially increasing activity are noticable in these measurements although these 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 rates may vary f rom approximately 2 to 25 mi/hr but no increasing trends have been identified.

10.5 AIRBORNE RADIOACTIVE NUCLIDES Air samples taken from t'.ne refueling floor atmosphere during and after each refueling showed activity levels below the lower level of detection. Storage of additional fuel is not expected to increase the airborne activity on the refueling floor since the major contribution of airborne activity is attributed to the most recent batch of spent fuel that is placed in the pool.

10.6 RADIATION PROTECTION PROGRAM The Radiation Protection Program is described in Section 12.5 of the Brunswick FSAR. Ihis 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.

10.8 IMPACT ON RADIOACTIVE EFFLUENTS The Spent Fuel Pool has its own filter-demineralizer units and, under normal 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 concentrations in the Spent Fuel Pool area is expected. Therefore, negligible impacts on gaseous effluents from the plant is anticipated as a result of the increase in spent fuel storage capacity.

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 Systen, as discussed in Section 9.3.4 of the FSAR.

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TABLE 10-1 BRUNSWICK STEAM ELECTRIC PLANT ISOTOPIC ANALYSIS OF SPENT FUEL POOL WATER UNIT #1 UNIT #1 ISOTOPE 2/23/79 5/2/80 ISOTOPE 2/23/79 5/2/80 Mn - 54 6.118E-4 2.309E 4 Mn - 54 1.09E-4 8.35E-5 Fe - 59 LLD* LLD* Fe - 59 .LLD* LLD*

Co - 58 2.14E-3 LLD* Co - 58 4.21E-5 ttp*

Co - 60 2.457E-4 4.053E-3 Co - 60 6.163E-4 1.033E-3 Zn - 65 5.01E-5 LLD* Zn - 65 LLD* LLD*

Cs - 134 3.847E-5 3.069E-4 Cs - 134 1.164E-3 3.368E-4 Cs - 137 4.574E-5 1.178E-3 Cs e 137 1.307E-3 4.868E-4 Sb - 122 LLD* 4.113E-5 Sb - 122 1.36E-4 3.315E-5 measurements in microcuries/ml GROSS BETA MEASUREMENTS DATE 1/11/78 7/?2/78 1/17/79 7/18/79 1/15/80 7/17/80

)

UNIT #1 1.95E-5 7.62E-5 1.15E-4 1.45E-2 5.493E-2 8.16E-4 )

UNIT #2 4.68E-5 1.33E-4 1.60E-5 3.51E-4 1.735E-3 5.93E-4

• LLD - Lower Limit of Detection l

10-3

11.0 ACCIDENT EVALUATION 11.1 SPENT 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 S.F.P.,

as recommended in NRC Regulatory Guide 1.13, " Fuel Storage Facility Design Basis."

11.3 OTHER CRANE LOADS 11.3.1 Interlocks 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. Lift heights of fuel hoists are limited to a maximum of 16 ft. by cut-out switches on the winch and also by mechancial design of the boom.

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 tue to a postulated loss of SFP cooling is examined in Section 8.4. ~

The antlysis indicates that even with the increase in spent fuel inventory the c"fsite doses are still well within the 10 CFR 100 limits.

9 11-1

12.0 CONCLUSION

S The information contained in this document to support the proposed modification satisfies the necessary applicable regulatory requirements to allow NRC approval for 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 Brunswick Steam Electric Plant Safety Evaluation Reports and Final Environmental Statements, and therefore precludes the need for preparation of an environmental impact statement.

S 12-1

13.0 NOTES AND REFERENCES Notes:

1. For the purposes of this report the term " fuel bundle" will imply configuration either with or without flow channels unless the term " fuel assembly" is specifically and distinctly intended.

1

2. Boral is a product of Brooks and Perkins, Inc., consisting of a layer of boron carbide-aluminum (B4C-A1) matrix bonded between two layers of aluminum.

1

References:

1 4-1 L. K. Liu, " Seismic Analysis of the Boiling Water Reactor," Symposium on Seismic Analysis of Pressure Vessel and Piping Component, First National Congress on Presrure vessel and Piping, San Francisco, California, May 1975. i 4-2 W. C. Wheadon, " Friction Test of Graphite Base Materials Sliding Against Type 304 Stainless Steel Plates", GE report No. C5445-TR-02, dated April 19, 1976. (Proprietary) 4-3 E. Rabinowicz, " Friction Coefficient Value for A High Density Fuel Storage System", GE Report VPF No. V5455, dated Janucy 3, 1978.

5-1 U.S. NRC Safety Evaluation for Yankee Rowe, dated December 29, 1976, 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.

5-3 Brooks & Perkins Report, "The Suitability of Brooks & Perkins Spent Fuel Storage Module For Use in BWR Storage Pool", Report No. 577.

5-4 A. J. Jacobs, "Boral Corrosion Test: 2022-Hour Results", GE Report No. 77-688-120, dated December 15, 1977. (Proprietary) 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.

8-1 M. J. Bell, "0RIGEN Code - The ORNL Isotope Generation and Depletion," -

ORNL-4628.

8-2 N. Eickelpasch and R. Hock, " Fission Product Release Af ter Reactor Shutdown," IAEA-SN-178/19.

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