ML19312C823
| ML19312C823 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/12/1975 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 8001100734 | |
| Download: ML19312C823 (21) | |
Text
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.n DUKE P0WER C0MPANY 7CONEE NUCLEAR STATION Docket No. 50-287 License No. DPR-55 74 W4's f
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SPENT FUEL STORAGE FACILITY MODIFICATION SAFETY ANAIYSIS REPORT
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THE ATTACHED FILES AR: OFFICIAL RECORDS OF THE OFFICE OF REGULATION. THEY HAVE I
BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD ANS MUST BE RETURNED TO THE CENTRAL RECORDS STATION 008. ANY PAGE(S)
REMOVED FOR REPRODUCTION MUST BE RETURNED 3
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TABLE OF CONTENTS l
Section P_ age,
1.0 INTRODUCTION
1 2.0 DESIGN OF HIGH CAPACITY FUEL ASSDiBLY STORAGE RACK 3
2.1 DESIGN BASES 3
2.2 SYSTEM DESCRIPTION 3
I 2.3 SYSTEM EVALUATION 3
]
2.3.1 Structural and Seismic Analysis 3
2.3.2 Nuclear Criticality Analysis 5
j 2.4 MATERIAL, CONSTRUCTION, AND QUALITY CONTROL 6
i 3.0 INTERFACE OF HIGH CAPACITY FUEL SIORAGE RACK AND SPENT 7
l FUEL STORAGE POOL 4.0 THERMAL ANALYSES 9
4.1 DESIGN BASES 9
t 4.2 SYSTEM DESCRIPTION 9
4.3 DESIGN EVALUATION 10 4.3.1 Normal Operation 10 I
4.3.2 Reliability Considerations 11
)
5.0 SAFETY ANALYSIS 12
6.0 REFERENCES
14 I
LIST OF TABLES Table Pm 1
4.1 SPENT FUEL COOLING SYSTEM DATA, UNIT 3 15 LIST OF FIGURES Figure Page 2.1 SPENT FUEL STORAGE AREA FUEL MODULES ARRANGEMENT 16 j
2.2 SPENT FUEL STORAGE MODULES 6 X 8 MODULE ASSEMBLY 17 3.1 SPENT FUEL STORAGE MODULE POOL INTERFACE 18 4
l 4.1 SPENT FUEL COOLING SYSTEM 19 l
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1.0 INTRODUCTION
This report is prepared and submitted in support of Duke Power Company's application to amend the Oconee Nuclear Station Unit 3 Facility Operating License, DPR-55, for modification of the Unit 3 spent fuel storage facility.
The modification of the existing spent fuel storage facility, which has a capacity for 216 spent fuel assemblies, is necessitated because of the expected increase in the inventory of spent fuel assemblies above the capacity of the existing facility.
The proposed modification provides for accommodating up to 474 fuel assemblies in the Unit 3 spent fuel storage pool and consists of replacing the existing fuel assembly storage rack with the Combustion Engineering, Incorporated, supplied High Capacity Fuel Assembly Rack without changing the basic structural geometry of the spent fuel storage pool.
The design objective of the new spent fuel assembly storage rack is to provide storage capacity for up to 474 fuel assemblies, having a feed enrichment of up to 3.5 weight percent U-235 in UO2 or the equivalent, and to maintain them in a safe and suberitical configuration during normal and abnormal conditions.
The High Capacity Fuel Assembly Storage Rack consists of an array of one-quarter inch thick stainless steel storage cavities having a nominal center-to-center spacing of 14.090 inches; each storage cavity can accommodate one fuel assembly. The fuel assembly storage cavities are structurally con-nected to form ten (10) fuel assembly storage modules by means of dimensionally controlled steel channels which limit the structural deformations and maintain the required center-to-center spacing between adjacent fuel assembly storage cavities. All ten modules are interconnected and rest on the pool floor.
The High Capacity Fuel Assembly Storage Rack and its associated structures are designed to the qualifications of a Seismic Category I structure.
The mechanical design of the individual fuel assembly storage cavities is such that it provides a flow path for convective cooling of the spent fuel assemblies through natural circulation. Thermal analysis of the Unit 3 spent fuel 9001 shows that the existing Spent Fuel Cooling System has the capability to maintain the spent fuel pool at approximately 150 F or less while dissipating the decay heat from up to 474 fuel assemblies.
Seismic analysis of the High Capacity Fuel Assembly Storage Rack has shown that the structure will withstand Safe Shutdown Earthquake loadings.
The design and arrangement of the High Capacity Fuel Assembly Storage Rack provide for maintaining a multiplication factor of less than 0.95 when all 474 fuel assembly storage cavities are occupied by unirradiated fuel as-semblics of feed enrichment up to 3.5 weight percent U-235 in UO2 r the equivalent, even under worst-case situations.
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Safety analysis of the thermal conditions within the spent fuel pool indi-cates that even with the most severe expected heat load, the stored spent fuel can be adequately cooled during normal and abnormal conditions.
Duke Power Company is responsible for the design, purchasing, fabrication, installation, and testing of the High Copacity Fuel Assembly Rack, and for gross thermal-hydraulic analysis of the fuel pool. Duke has contracted with Combustion Engineering, Incorporated, to design, manufacture, and deliver to the site the High Capacity Fuel Assembly Storage Racks.
Com-bustion Engineering was also retained by Duke to perform the seismic, criticality, and detailed thermal-hydraulic analysis of the rack.
i on the basis of the information presented in this report and referenced material, Duke Power Company concludes that the proposed modification of the Oconee Unit 3 spent fuel storage facility will provide safe storage for up to 474 fuel assemblies and that the modification will not endanger the i
health and safety of the public.
J 2.0 DESIGN OF HIGH CAPACITY FUEL ASSEMBLY STORAGE RACK
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2.1 DESIGN BASES The High Capacity Fuel Assembly Storage Rack is designes to provide storage locations for up to 474 fuel assemblies and is designed to maintain the stored fuel, having a feed enrichment of 3.5 weight percent U-235 in UO2 or equivalent, in a safe, coolable, and suberitical configuration during normal and abnormal conditions.
2.2 SYSTEM DESCRIPTION The storage rack provides storage locations for 474 fuel assemblies in a rectangular array. The rack is comprised of eight modules, each containing 48 fuel assembly storage locations in a 5 x 8 array, and two modules each, containing 42 fuel assembly storage locations in a 6 x 7 array. Provision is also made to accommodate six failed fuel assemblies in oversized storage cavities. All ten modules are interconnected and are arranged as shown in Figure 2.1.
Each fuel assembly storage module is composed of rectangular storage cavities fabricated from one-quarter inch thick stainless steel plate, with each cavity capable of accepting one fuel assembly. The fuel assembly storage cavities have lead-in surfaces at the top to provide guidance for insertion of fuel assemblies. The cavities are open at the top and bottom to provide a flow path for convective cooling of spent fuel assemblies through natural circulation. The fuel assembly storage cavities are structural]; connected to form modules through the use of channels, plates, and angles which limit structural deformations and maintain a nominal center-to-center spacing of 14.090 inches between adjacent storage cavities during design conditions including the Safe Shutdown Earthquake. The design and configuration of a typical fuel assembly storage module (in this case, the 6 x 8 module) is shown in Figure 2.2.
- 2. 3 SYSTEM EVALUATION 2.3.1 Structural and Seismic Analysis Fuel assembly storage rack and associated structures are designed to with-stand the maximum forces generated during normal operation combined with the Safe Shutdown Earthquake according to the requirements of a Seismic Category I structure. For these conditions, the storage rack design is such that all stresses fall within the allowable stress limits specified in the AISC Specifications for Design, Fabrication, and Erection of Structural Steel. The stress analysis of the rack design demonstrates that the loads and load combinations are conservative relative to Standard Review Plan 3.8.4.
Normal operating loads include dead weight (in air) and thermal expansion loads. Lateral and vertical seismic loads along with the fluid forces generated by seismically generated pool water sloshing are considered to be acting simultaneously.
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d The seismic input spectra conform to the requirements of Regulatory Guide 1.'60, " Design Response Spectra for Seismic Design of Nuclear Power Plants."
Reference is made to Project 81 PSAR, Docket Nos. STN50-488 through -493, Section 3.7.
The smoothed response spectra shown on Figure 2E-2A were normalized to 10% g for Safe Shutdown Earthquake (SSE). An earthquake acceleration-time history compatible with these spectra, as shown in Figures 2E-2B through 2E-2E, was used as a base motion on the model of the Auxiliary Building.
a The seismic response of the Auxiliary Building to the base excitation is determined by a dynamic analysis. The dynamic analysis is made by idealizing the structure as a series of lumped masses with weightless elastic columns i
acting as spring restraints. The base of the structure is considered fixed.
The choice of the location of the mass-joints depends on the distribution of l
masses in the real structure.
I The modes and frequencies of natural vibration are obtained. A time-history analysis is performed using the modal superposition method, in which the responses in the normal modes are determined separately, then superimposed to provide the tota 1 response. The structural damping values used are 4 percent for SSE and 7 percent for SSE. These values conform to the requirements of Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants."
i The acceleration-time history response of the mass-joint at elevation 802 ft.
l (fuel pool floor) was then used as input for the time history analysis of l
the fuel rack model (see paragraph (c) below).
Seismic response loads for determining stresses in the racks are determined by the following procedure:
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(a) A modal extraction analysis of a space frame analytical model of the rack modules minus fuel but including the effects of the water sur-rounding and within the cans is performed using the STARDYNE(1) Computer i
Code.
(b) The modal parameters of the rack modules are used to derive a dynamically-equivalent model of the rack modules for incorporation into a lateral nonlinear model which includes both the racks and the fuel assemblies.
t (c) A time history analysis of the nonlinear model is performed using the pool structure response time histories as input. This analysis is performed with the SHOCK (2) computer code and the results include the impact force time histories between the fuel assemblies and cavities and the rack support reaction loads, i
(d) The fuel assembly / cavity impact load time histories and the rack support motion time histories are then applied to the STARDYNE space frame model of the rack module to obtain the detailed distribution of structural member loads.
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f (e) The above analysis is consistent with the three directional component excitation requirements of Regulatory Guide 1.92 with the assumption that the base of the racks is restrained against lateral motion. This conservative assumption is applied to account for the high coefficient of friction that exists for stainless steel in contact with stainless steel. Damping values of 2 percent and 3 percent are assumed for the rack structural members and fuel respectively.
An analysis was performed to determine the effects of seismic excitation of the pool water using the procedure outlined in Reference 3.
This analysis showed that the rack is below the depth where sloshing forces are effective and therefore should not experience excitation from this source. However, for conservatism drag forces on the rack were calculated using the peak pool fluid velocity due to sloshing combined with the peak rack response velocity. The results show total drag forces on the entire rack structure in the pool to be only three cercent of the weight of the empty rack.
The maximum uplif t load available from the fuel handling crane on the storage rack is limited to 3000 lbs or less by the hoist interlock.
This loading represents a less severe loading on the fuel rack structure than that resulting from the design basis earthquake.
Structural design precludes placing a fuel assembly between cells, and the rack will withstand the loadings imposed by a postulated dropped fuel as-sembly.
2.3.2 Nuclear Criticality Analysis The fuel assembly storage rack is designed to maintain a minimum edge-to-edge spacing of 3.950 inches between fuel assemblies under design conditions including the design basis earthquake.
Calculated values of Keff, including calculational uncertainties, are less than 0.95.
The effective multipli-l cation factors are calculated under the following conditions.
(a) Unitradiated fuel assemblies containing 3.5 weight percent U-235 in UO2 or equivalent in all 474 fuel storage locations.
(b) A minimum edge-to-edge spacing of 3.950 inches between adjacent fuel assemblies.
(c) Fuel assemblies assumed to be displaced diagonally off-center and in contact with the fuel assembly storage cavity corner wall resulting in the minimum separation of fuel assemblies in each cluster of four fuel storage cavities.
(d) Non-borated pool water at 68 F.
(e) No control rods inserted in stored fuel assemblies.
For the criticality evaluation, neutron cross section data for representative fuel rod cells and materials between and around fuel assemblies are generated.
with the CEPAK code (4, 5, 6, 7).
Spatial calculations are performed using the two-dimensional transport code DOT-2W (8) for a doubly infinite lattice of fuel assemblies at the designated minimum edge-to-edge spacing.
2.4 MATERIAL, CONSTRUCTION, AND QUALITY C011 TROL The entire fuel assembly storage rack is constructed of type 304 stainless steel. All welded construction is used in the fabrication of the fuel as-sembly storage rack. The all-welded construction ensures the structural.
integrity of the storage modules and provides assurance of smooth, snag-free passage in the storage cavities so that it is highly improbable that a fuel assembly could become stuck in the rack.
The material, construction and quality control procedures are in accordance with the quality assurance requirements of Duke Power Company, as described in Duke Power Company Topical Report, DUKE-1.
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3.0 INTERFACE OF HIGH CAPACITY FUEL STORAGE RACK AND SPENT FUEL STORAGE POOL Unit 3 Spent Fuel Pool is a reinforced concrete structure with a k inch thick stainless steel liner for leak tightness. The pool is an integral part of the Unit 3 Auxiliary Building and designed as a Class I structure in ac-cordance with Appendix SA, paragraph 5A.l.1 and 5A.2 of the Oconee Nuclear Station Final Safety Analysis Report (FSAR).
The pool floor will support the high* capacity storage rack as a free-standing structure during all design conditions. For installation iato the pool, pairs of rack modules are field connected by welding coupling plates (two on each side) to form four 8 x 12 assemblies and one 7 x 12 assembly. These assembled pairs have four floor bcams attached to their bases prior to being lowered into the pool. Each floor beam attached to the assembled pairs has bearing pads which are shimmed to ensure contact with the pool floor cor-responding to the model used for the seismic analysis. The assembled pairs are connected to adjacent assembled pairs at the tops. Figure 3.1 shows the layout of the pool-rack interfacing.
The pool is constructed of concrete with a compressive strength of 3000 psi at 28 days and reinforcing steel with a minimum yield strength of 60,000 psi. Maximum stresses in the pool floor prior to installation of the high-capacity rack were 830 psi for concrete and 16,725 psi for reinforcing steel.
With the high-capacity rack installed maximum stresses are 1000 psi for concrete and 19,948 psi for reinforcing steel for the loading condition of U = 1.4D + 1.7L + 1.9E.
These stresses are well within ACI allowables of 2550 psi for concrete and 54,000 psi for reinforcing steel.
For the free-standing rack structure, conservative analysis shows that under si ;1taneous forces from vertical and lateral seismic excitation, including pool fluid sloshing, the residual displacement of the rack relative to the pool floer is less than 1.9 inches for unloaded conditions and less than 0.7 inches for full-loaded condition.(i.e., much less than the minimum clearance of 6.1 inches to pool walls and installed equipment.)
The relative motion between the rack and pool floor is determined by the following procedure for both full and empty rack.
(a) The amplitude of the most severe of the two lateral time histories as determined from an inspection of their response spectra is increas.d by a factor of /2 to conservatively account for simultaneous excitation in the two horizontal directions. This time history is then input to the lateral SHOCK code models for full and empty rack.
(b) The coefficients of friction used in the analyses, 0.4 static and 0.2 dynamic, are well below minimum values found in the literature for stainless on stainless and similar metals in a water environment.
These low friction coefficient values account for the decrease in normal force between the rack base and pool floor due to vertical seismic response. The peak vertical seismic response was calculated te be 0.3 g based on the vertical natural rack frequency and vertical response spectrum. 11/14/75
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(c) The rack / pool floor normal force on which the lateral friction forces used in the analysis are based includes bouyancy effects due to sub-mergence of the rack and fuel.
The maximum lateral seismic force exerted by each rack module on pool floor I
is 111,200 pounds and results in a stress of 1,650 psi in the floor liner and 2,330 psi in the weld connecting the floor liner to embedments in the concrete. The maximum combined seismic and thermal stress in the floor liner.is 16,300 psi and 22,000 psi in the weld between liner and embedments.
The maximum stresses are below the design allowable stress of 27,000 psi.
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4.0 THERMAL ANALYSES 4.1 DESIGN BASES The Spent F,uel Cooling System for Unit 3 is designed to remove the dec'ay heat from the stored fuel in the spent fuel pool. The existing Unit 3 system duplicates the equipment used for the Unit 1 and 2 system, and has the capa-bility to maintain the spent fuel pool at approximately 1500F or less while removing decay heat from the stored fuel assemblies, even when all the fuel storage locations (474) are occupied. Therefore, the system is capable of cooling the increased inventories of stored fuel which will be present as a result of the modifications to the Unit 3 spent fuel storage racks.
In addition to its primary function, the system provides for purification of the spent fuel pool water, the fuel transfer canal water, and the contents of the borated water storage tank in order to remove fission and corrosion products and to maintain the water clarity for fuel handling operations. The system also provides for filling the fuel transfer canal and the incore instrument handling tank.
4.2 SYSTEM DESCRIPTION The Spent Fuel Cooling System (Figure 4.1) provides cooling for the spent fuel pool to remove fission product decay heat energy. System performance data are shown in Table 4.1.
Major components of the system are briefly described below.
Spent Fuel Coolers The spent fuel coolers are designed to maintain the temperature of the spent fuel pool as noted in Section 4.1.
There are two coolers for Unit 3, ar-ranged in parallel.
Spent Fuel Coolant Pumps The spent fuel coolant pumps take suction from the spent fuel pool and re-circulate the fluid back to the pool af ter passing through the coolers, dermineralizer,and filters in various combinations depending on conditions.
There are two pumps for Unit 3.
The spent fuel pumps are also used for filling the fuel transfer canal or incore instrumentation handling tank with borated water from the borated water storage tank.
Spent Fuel Coolant Demineralizer The spent fuel coolant demineralizer will process approximately one-half of the spent fuel pool volume in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. There is one demineralizer for Unit 3.
Spent Fuel Coolant Filters Tha spent fuel coolant filters are designed to remove particulate matter from the spent fuel pool water. They are sized for the same flow rate as the de-mineralizer (180 gpm). There are two filters for Unit 3. l
Borated Water Recirculation Pump This pump removes water from the borated water storage tank for deminerali-zation and filtering. The pump may also be used while demineralizing and filtering the water in the fuel transfer canal during a transfer of fuel.
It may also be used for emptying the fuel transfer canal if both spent fuel coolant pumps are unavailable for use.
There is one pump for Unit 3.
4.3 DESIGN EVALUATION 4.3.1 Normal Operation The normal operation of the Spent Fuel Cooling System serves two main functions. The first is to maintain the pool water at a temperature of ap-proximately 150 F or less with stored fuel at expected operating histories and with both coolers operating. The second function is to provide purifi-cation of the spent fuel pool coolant for clarity during fuel handling operations.
The first function is accomplished by recirculating spent fuel coolant water from the spent fuel pool through the pumps.and coolers and back to the pool.
The spent fuel pumps take suction from the spent fuel pool and deliver the water through the shell side of the two coolers arranged in parallel and back to the pool.
A bypass purification loop is provided to maintain the purity of the water in the spent fuel pool. This loop is also utilized to purify the water in the borated water storage tank following refueling, and to maintain clarity in the fuel transfer canal during refueling. Water from the borated water storage tank or fuel transfer canal can be purified by using the borated water recirculation pump.
Normally,1/3 of a core will be discharged into the pool at each refueling of Unit 3.
Transfers of spent fuel assemblies from the Unit 1-2 pool will also be made. When the first batch is introduced, one pump and one cooler will be capable of maintaining the pool at 125 F or less. As fuel accumu-lates, two pumps and two coolers will be utilized. The heat load will vary and be primarily dependent on the age of the most recently irradiated fuel in the pool.
For conservatism, the cooling system has been evaluated under the assumption that it is used only for Unit 3 fuel and without inputs of decayed batches from Units 1 and 2.
The postulated inventory is one full core discharge with the remainder of the storage locations occupied by batches previously discharged at one-year intervals. The resulting combination of assemblies is:
(a) 1/3 core irradiated 720 days and decayed 7 days.
(b) 1/3 core irradiated 410 days and decayed 7 days.
(c) 1/3 core irradiated 100 days and decayed 7 days.
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(d) 1/3 core irradiated 930 days and decayed 118 days.
(e) Four 1/3 core batches irradiated 930 days and decayed one, two, three and four years, respectively, longer than batch (d) above.
With this heat load, the system can maintain the Unit 3 pool at 150 F or less for two pumps and coolers, and at 205 F or less for postulated loss of one pump and one cooler.
4.3.2 Reliability Considerations The Spent Fuel Cooling System provides adequate capacity and component redundancy to assure the cooling of stored spent fuel, even when large quantities of fuel are in storage. Multiple component failures or complete cooling failures permit ample time to assure that protective actions are taken. The system is arranged so that loss of fuel pool water by piping or component failure is impossible. The system performs no emergency functions.
Alarms are provided to indicate pool level and temperature.
The Spent Fuel Cooling System has no process lines connecting to the Reactor Coolant System.
Its major penetration to the Reactor Building is through the fuel transfer tube.
The fuel transfer tube is isolated inside the Reactor Building by a blind flange connection in the fuel transfer canal.
On the outside of the Reactor Building, the fuel transfer tube is isolated by a closed, manually-operated valve in the spent fuel pool.
5.0 SAFETY ANALYSIS The storage rack is designed and constructed to retain the integrity of the structure under all anticipated loads, including the Safe Shutdown Earth-quake, with the maximum number of fuel assemblies occupying the storage locations.
The rack design provides protection against damage to the fuel and precludes the possibility of a fuel assembly being placed between cells. Although not required for safe storage of spent fuel assemblies, the spent fuel pool water is normally borated to a concentration of at least 1800 ppmb. The rack design also assures a Keff of less than 0.95 even when the entire array of fuel assemblies, assumed to be in their most reactive condition, are im-mersed in unborated water at room temperature. Under these conditions a criticality accident during refueling or storage is not considered credible.
As discussed in Section 4.0, the Spent Fuel Pool Cooling System design is such that no single component failure can lead to water losses that will uncover the spent fuel in storage. In the event one pump and one cooler were unavailable at the time of maximum thermal loading, the remaining cooling train would limit the bulk pool temperature to 205 F or less. In the unlikely event of total loss of all forced cooling, considerable time would be required to heat the large pool water mass to the boiling temperature.
With the most severe expected heat load and loss of all pumps and coolers, the Unit 3 spent fuel pool (assumed to be adiabatic) requires 6.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to reach 205 F.
This is consideced ample time to effect repairs or to provide alternate cooling.
An analysis of the maximum fuel cladding temperature has been performed for the postulated case of complete loss of coolant circulation to the pool. The analysis assumes maximum anticipated heat load in the pool, with the hottest assembly located in the least cooled storage area. The maximum cladding temperature will occur at the location of maximum heat flux. For a fuel assembly having the maximum value for decay heat power of 80 kw, and for an axial peak to average power density ratio of 1.2, the maximum local fuel rod heat flux is 1200 BTU /hr-ft2 Natural circulation flow rates within the storage tubes have been calculated which give confidence that convection film coefficients 2
I in excess of 50 ETU/hr-f t _oF can be expected. Assuming this low value for conservatism, the clad surf ace temperature is 24 F above the coolant l
temperature. Because the heat flux is small, very large uncertainties in the film coefficient are acceptable without causing prohibitively high clad temperatures. For example, a reduction by a factor of five in the film coefficient would result in a clad surface temperature of 120 F above the i
A reduction by a factor of ten, from 50 BTU /hr-ft
_O 2 p coolant tempergture.
to 5 BTU /hr-f t - F would result in a clad surface temperature of 240 F above the coolant temperature. These temperatures are below 650 F, which is the normal operating temperature of the' fuel clad in the core.
Safety provisions are designed into the fuel handling system La prevent the i
development of hazardous conditions in the spent fuel pool, it *he event of component malfunctions, accidental damage, or operational or administrative failures during refueling or transfer operations. Section 9.7 of the Oconee Nuclear Station FSAR describes the system design, operation, and safety provisions for the Fuel Handling System. The potential radiological con-sequences of a fuel handling accident involving mechanical damage to a fuel assembly are analyzed in FSAR Section 14.2.2.1 and in FSAR Supplement 1, page 1-20.
Although the new storage rack provider for accomodating a larger inventory of spent fuel, the parameters, assumptions, and condi-tions used in this accident analysis are not affected by this modification.
Therefore, the radiological consequences of a postulated fuel handling accident are no more severe than those reported in FSAR Section 14.2.2.1 or l FSAR Supplement 1.
The radiation level within the pool area is such that the maximum continuous dose rates are not greater than 2.5 mrem /hr in the working area and 10 mrem /hr at the surface of the pool water.
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6.0 REFERENCES
1.
MRI/STARDYNE User Manual, Computer Methods Department, Mechanics Research, INC., Los Angeles, California, January 1, 1970.
2.
Gabrielson, V.
K., " SHOCK-A Computer Code for Solving Lumped-Mass Dynamic Systems SCL-DR-65-34," January, 1966.
3.
TID-7024, " Nuclear Reactors and Earthquakes," United States Atomic Energy Commission, Division of Technical Information, August, 1963.
4.
CEPAK-A Combustion Engineering, Inc., lattice analysis code consisting of the FORM (Reference 5), THERMOS (Reference 6), and CINDER (Reference
- 7) progre.ms interlinked in a consistent way with inputs from differential microscopic cross section libraries.
5.
McGraff, D.
J., " FORM-A Fourier Transform Fast Spectrum Code for the IBM-7090," NAA-SR-Memo, September 1960.
6.
Honeck, H., " THERMOS-A Thermalization Transport Theory Code for Reactor Lattice Calculations," BNL-5816, July 1961.
7.
England, T. R., " CINDER-A One Point Depletion and Fission Product Program," WAPD-TM-334, Revised June 1964.
8.
Soltesz, R.
G., et al., " User's Manual for the DOT-2W Discrete Ordinates Transport Computer Code," W5L-TME-1982, December 1964.
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4 TABLE 4.1 SPENT FUEL COOLING SYSTEM DATA, UNIT 3 System Cooling Capacity, BTU /h 6
Normal 15.5 x 10 I
0 Design 28.6 x 10 i
System Design Pressure, psig 125 System Design Temperature, F 250 Spent Fuel Cooler Type Tube and shell Material Tube /Shell SS j
6 Capacity, BTU /h/ cooler 7.75 x 10 Cooling Water Flow, lb/h/ cooler 5 x 105 Code ASME VIII/III-C Spent Fuel Pump i
Type Horizontal, centrifugal Material SS Flow, gpm 1,000 Head, ft H O 100 2
Motor Horsepower, hp 40
{
Spent Fuel Pool Volume, ft 50,000 i
)
Spent Fuel Filter j
Design Flow Rate, gpm 180 i
Material SS Design Temperature, F 250 Design Pressure, ps'.g 125 Code AfME III-C Borated Water Retirculation Pump Type Vertical, inline, centrifugal Material SS Flow, gpm 180 Head, ft H O 140 2
Motor Horsepower, hp 15 Design Temperature, F 250 Design Pressure, psig 125 Spent Fuel Demineralizer Type Mixed bed Material SS j
Resin Volume, ft 21 Flow, gpm 180 Design Temperature, F 250 Design Pressure, psig 125 i
Code ASME III-C 1.
- 13'-11.375" -
CASK STORAGE PIT
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FAIED FUEL STORAGE 000000.
LOCATIONS 000000 OOOOOo 6x7 a
000000 MODUE '-
000000 c
000000 000000 6x8 6x8 MODUE MODUE 47'7. 50" l
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6x8 MODUE MODUE i
FUEL UPENDER+
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SPENT FUEL STORAGE AREA FUEL MODULES ARRANGFRENT OCONEE NUCLEAR STATlON UNIT 3 l
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