ML17317B109

Forwards Addl Info Re Spent Fuel Storage Capacity Expansion Program.Encl Info Includes Details of Structural Analysis, Revisions to General Description of New Spent Fuel Racks & to Thermal Analysis

ML17317B109
Person / Time
Site:	Cook
Issue date:	04/16/1979
From:	Dolan J INDIANA MICHIGAN POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AEP:NRC:00169, AEP:NRC:169, NUDOCS 7904230169
Download: ML17317B109 (37)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS}

ACCESSION NBR!7900230169 DOC ~ DATE: 79/00/16 NOTARIZED: NO BYNAME FACIL:50~315 DONALD C ~ COOK NUCLEAR POWER PLANTr UNIT 1i INDIANA 50 316 DONALD C ~ COOK NUCLEAR POWER PLANTg UNIT 2i INDIANA 5 AUTH AUTHOR AFFILIATION DOLANnJ ~ ED INDIANA 8 MICHIGAN POWER CO, .

REC IP ~ NAME RECIPIENT AFFILIATION DENTON H ~ R ~ OFFICE OF NUCLEAR REACTOR REGULATION SUBJECT! FORWARDS ADDL INFO RE SPENT FUEL STORAGE'APACITY

~ ENCL INFO INCLUDES DETAILS OF STRUCTURAL ANALYSIS' EXPANSION'ROGRAM REVISIONS TO GENERAL DESCRIPTION OF NEW SPENT FUEL RACKS 5 TO THERMAL ANALYSIS~

DISTRIBUTION CODE: A001S COPIES RECEIVED:LTR l ENCL M SIZE: "~

TITLE: GENERAL DISTRIBUTION FOR AFYER ISSUANCE OF OPERRRTIHG LIC.

NOTEgi.~+ E - M ~w ~L.L lI78 TL.

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCl ID CODE/NAME: LTTR ENCL ACTION: 05 BC ogpu& / 7 7 II INTERNALt 0 1 1 02 NRC PDR 1 2 2 14 TA/EDO 15 CORE PERF BR 1 1 16 AD SYS/PROJ .

17 ENGR BR 1 1 18 REAC SFTY BR 19 PLANT SYS BR 1 1 20 EEB 21 EFLT TRT SYS 1 1 22 BRINKMAN EXTERNAL! 03 LPDR 1 1 oc NSIC . 1 1 23 ACRS 16 16 APR 24 i97g TOTAL NUMBER OF COPIES REQUIREDt LTTR 38 ENCl 38

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INDIANA II MICHIGAN POWER COMPANY P. O. 8OX 18 80WLING GREEN STATION HEW YORK, N. Y. 10004 April 16, 1979 AEP:NRC:00169 Donald C. Cook Nuclear Plant Units No. 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. OPR-58 and DPR-74 Spent Fuel Storage Capacity Expansion Program-Structural Analysis Mr. Harold R. Oenton, Director Office of I'luclear Reactor Regulation U.S. Nuclear Regulatory Commission

. Washington, O.C. 20555

Dear Mr. Denton:

Please find attached additional information on the Spent Fuel Storage capacity expansion program for the Donald C. Cook Nuclear Plant. A brief description of each attachment follows:

Attachment No. 1 presents the details of the required structural analysis performed by the Exxon Nuclear Company for the new spent fuel racks and replaces the originally submitted sections 3.6 and 3.7 of Reference (1).

This analysis completes our submittal describing the spent fuel storage capacity expansion program at the Donald C.Cook Plant.

Attachment No. 2 includes revisions to the general description of the new spent fuel racks given as part of reference (1).

Attachment No. 3 includes revisions to the thermal analysis previously submitted in reference (2).

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~ .--.

Nr. Harold R. Dent AEP:NRC:00169 The fee f'r processing this and the previous submittals was remitted to the Nuclear Regulatory Commission on December 21, 1978, reference (3).

Very truly yours, JED:em ohn . Dolan Vice President Sworn and subscribed to before me this l<~ day of April, 1979 in New York County, New York Notary Public

HLEEN NOTARY I'UttliC, 'tale of New York Ho. 41-~frtii's'292 Qualified in (queens CountY Certificate filed in New York CountY Coauression axttir< and vertically with spring KV.

The stiffness of springs KT and K> were selected to duplicate the frequency and mode shape of the primary horizontal mode of the complete rack assembly as determined from the SAP- IV detailed elastic model. The stiffness of the vertical support spring KV .was selected to duplicate the frequency of the primary vertical mode.

Added hydrodynamic mass was treated in the same manner as with the SAP-IV model, i.e. it was included in the hori-zontal direction but not in the vertical direction. The mass of the rack structure was included at Nodes 4 and 8.

Sliding gap elements are provided between Nodes 1 and 20 and Nodes 3 and 21 to permit both rocking and sliding of the complete rack assembly. Analyses were performed for both minimum and maximum friction coefficient values.

The values used were respectively 0.30 and 0.75 and are taken from the results of experimental work reported in References (ll) and (12). The values used are based on statistical analysis of the experimental results and provide at least a 95/ confidence level that the actual friction coefficient at any contact point will not fall outside the range of the values used. Impact between adjacent racks was considered by providing gap elements between nodes 8 and 22 and nodes 19 and 23. The initial gap in these elements was set equal to half the gap between adjacent racks, and conservatively assumes that racks could be moving out of phase with each other.

The model was subjected to simultaneous, statistically independent horizontal and vertical time histories at the pool floor; i.e. Nodes 20 and 21. These were artificially generated time histories whose response spectra enveloped the floor level response spectra for the floor of the fuel storage pool.

The results from the nonlinear time history analysis were compared with the results of simplified elastic time history analyses to develop impact factors for use in the detailed stress analysis. This simplified elastic time history analysis was performed by closing all the fuel cell to fuel gaps and by fixing the rack feet sliding gap elements in the model shown in Figure 3.6.2.

3.6.5 Dro ed Fuel Assembl Accident An evaluation of the effects of a postulated dropped fuel assembly accident has been performed to confirm that there would be no effect on the spacing of fuel assemblies stored in the racks. The method of design and analysis was identical to that submitted by Public Service Electric and Gas Company in it's letter dated November 18, 1977 and subsequent submittals and as approved by the NRC in its Safety Evaluation Report for the Salem Station Unit No. 1 Spent Fuel Rack Modification dated January 15, 1979.

3.6.6 Resu1ts Structural Anal sis Computer plots of the primary vibrational mode shape in the horizontal and vertical directions are presented in

Figures 3.6.3, 3.6.4 and 3.6.5. The calculated natural frequencies and participation factors are given in Table

3. 6.2.

The limiting. load combinations and stress values for the rack members are presented in Table 3.6.3.

Time Histor Anal ses The results obtained from the time history analyses are as follows:

(a) The maximum horizontal motion at the top of a fully loaded rack resulting primarily from sliding motion with a friction coefficient of 0.3 is 0.46 inches under SSE loading conditions.

(b) The maximum horizontal motion at the top of a fully loaded rack, resulting primarily from rocking, with a friction coefficient of 0.75 is 0.64 inches under SSE loading conditions.

(c) The maximum horizontal motion of a worst case par-tially loaded rack, primarily resulting from rocking, with a friction coefficient of 0.75 is 3.29 inches under SSE loading conditions. This is the worst case loading condition for overturning and shows that there is a large factor of safety against rack overturning. This case could only occur for a peripheral rack rocking away from the center of the array. A worst case partially loaded rack within the array will impact against adjacent racks and the resulting impact loads have been included in the structural analysis of the rack members.

(d) The elastic lumped parameter analysis loads are increased by impact factors of 1.54 and 1.94 for SEE and OBE loads, respectively. Consequently, the loads developed from the SAP-IV elastic analysis were increased by these factors for comparison against the allowable stresses.

(e) The maximum loads on the fuel assembly from the nonlinear seismic impact analysis with the assembly modeled with three equally spaced spacer grids, are 840 lbs on the upper end fitting, 800 lbs, 1120 lbs and 520 lbs, respectively on the top, middle and bottom spacer grids, and no load on the bottom end fitting. A transverse compression test was performed as part of testing on the stainless steel clad Boral fuel cell. This test showed that the cell could elastically resist a uniform load of up to 105 lbs/in against the side of the cell. Therefore,,for the maximum impact load against the side of the cell, only a 1.3 inch length of cell would be reauired to resist this load, which is much less than the area imoacted by the fuel bundle. At the location of the fuel assembly upper end fittings, additional conservatism is provided by the fact that the fuel cells are supported by the intercell spacer grid bars.

Dro ed Fuel Assembl Accident Compression tests were performed on 2-ft long sections of the Boral poison spent fuel cell together with the flared lead-in section to determine the load-deflection characteristics of the cell. Two (2) cases were consi-dered. The first case is that of an assembly falling vertically directly on one cell but rotated 45'uch that the corners of the assembly hit the sides of the cell in a diamond pattern. This case produces maximum force and deflection of an individual cell. The second case is that of an assembly falling vertically at the center of a group of four (4) cells, resulting in a maximum force applied to the rack structure.

Each of the specimens was made of a 2 foot long section of the Boral poison spent fuel cell together with the flared insertion guide attached at the top. A steel bear-ing plate was attached to the bottom for support. The load was applied by a thick square steel loading plant mounted on the load head of a 120 kip static test machine. The plate, representing the bottom of a fuel assembly, con-tacted the top of the insertion guide in a manner which represented the alignment and orientation of the fuel assembly with respect to the fuel cell at the assumed time of impact. An increasing load was slowly applied and the load-deflection recorded until the insertion guide was significantly deformed. In each of the two (2) tests, the structure deformed elastically up to the onset of localized inelastic buckling. In the first case where

the four (4) corners of the loading plate contact the mid-length of the four (4) sides of the cell, initial inelastic deformation occurred at a load of 4,500 lbs.

The maximum load of 25,000 lbs occurred at a deflection of 2.5 inches. In the second case where the loading plate is centered on a group of four (4) fuel cells, initial inelastic deformation occurred at a load of 5,000 lbs for the one cell tested. The maximum load of 10,000 lbs occurred at a deflection of 0.8 inches. In both cases, local crushing of the cell was limited to the upper 7 inches of the insertion guide.

The mechanical properties of Type 304 stainless steel have been shown to be relatively insensitive to strain r ate over a wide range. It is, therefore, concluded that the static load tests, described above, adequately repre-sent the response to fuel cell impact at the low veloci-ties involved (approximately 8 ft per second).

The experimentally developed load-deflection curves were used to calculate the peak force and deflection resulting from the energy of a dropped assembly. The maximum drop height is 15 inches from the top of the fuel cells, since the fuel assemblies will not be moved over the fuel storage racks at a higher elevation. The maximum kinetic energy at the point of impact is 2,020 ft-lbs. For the first case (diamond pattern) the maximum force, on the impacted cell, is 25,000 lbs and the maximum deflection is 2.5 inches. For the second case, the maximum force on

each of the four (4) impacted cells is 10,000 lbs and the maximum deflection is 0.8 inches. In both cases, local crushing of the cell is limited to the upper 7 inches of the lead-in section, above the rack module upper grid structure and above stored fuel assemblies.

In the case where the fuel assembly is dropped inside the storage cell, the fuel assembly would impact the 1/4 inch base plate at the bottom of the rack module. The impact energy will be absorbed primarily by bending deformation of the 1/4 inch base plate and a small amount of bending distortion of the base assembly beam members. The total distortion of the base plate was conservatively calcu-lated to be approximately 1 inch. The spacing of adjacent stored fuel assemblies will not be affected.

The effects of a dropped assembly accident in which the assembly rotates as it drops, was also evaluated. In this case, the assembly impacts a row of storage cells and comes to rest laying on top of the rack modules. The maximum kinetic energy of impact on one cell is conserva-tively calculated to be 1,500 ft-lbs resulting in lower loads than the simple vertical drop case discussed above.

The dynamic response of the rack structure to the impulsive type loading of the fuel bundle drop accidents described above has been evaluated and the resulting rack member stresses compared against the allowables. The results of these analyses show that:

1) Inelastic deformations are limited to the immediate area(s) of assembly impact.

2) All other rack member stresses are within the limits specified in Section 3.6.2.

3.6.7 Fuel Stora e Pool H drod namic Effects The array of fuel storage racks are located within 11-inch or less of the fuel storage pool walls on the North, South and West walls of the fuel storage pool. Any effects of this small mass of water on the vibrational response or stress in the racks were, therefore, considered to be insignificant. At the East end of pool in the cask area, there is a 12 ft clear-ance between, the fuel storage racks and the pool wall. The seismic effects on the racks of the water in this area were analyzed in accordance with the methods given in Reference 13.

The added hydrodynamic loads increase the horizontal seismic loads in the East-West di rection by 8~, This results in a substantially smaller percentage increase in stresses when three axes simultaneous seismic loads are combined with the dead weight loads on the rack structure.

3.6.8 Fabrication and Installation The spent fuel storage rack modules are in conformance with the materials, fabrication, installation and examin-ation criteria of the 1977 Edition of the ASME Code,Section III, Subsection NF Articles NF-2000, NF-4000, and NF-5000, respectively, except as specifically noted below.

The fuel storage modules are not certified or stamped asSection III components; therefore, the documentation, certification and programs which are specifically concerned with production of Code certified components have been replaced with Exxon Nuclear Company's gA/gC program, which is in compliance with ANSI N45.2 and 10 CFR 50, Appendix B.

Exxon Nuclear does not require compliance with the materia'1 traceability requirements of NF-4122. The fabricator is required to demonstrate a material control program which will insure that only certified material is used in the storage module. The only exception to this rule is in the use of neutron poison materials, which are required to be in compliance with NF-4122.

TABLE 3.6.2 NATURAL FRE UENCIES AND t<ODAL PARTICIPATION FACTORS Participation Factor triode 8 Fre uenc Hz X Y z Notes 1 7. 19 0. 52 0.12 0. 02 2 7.46 21. 34 0. 56 0. 02 Primary X-ttode 3 7.63 0. 57 20. 87 0.01 Primary Y-Vode 4 7.91 0. 31 0.17 0.02 5 10.45 0.00 0. 01 0. 01 6 10.70 2.32 0.04 0.01 7 10.72 0.04 2.53 0.01 8 10.73 0. 11 0.11 0. 02 9 11.23 0. 01 0.01 0.02 10 11. 34 0.20 0. 01 0.02 11 11.35 0. 03 0. 11 0. 02 12 11.37 0. 01 0. 00 0. 02 13 11.44 0. 00 0. 00 0.00 14 11.50 0.21 0.00 0.03 15 11.51 0.02 0. 01 0. 01 16 11. 53 0. 00 0. 20 0. 01 17 11. 53 0.01 0. 01 0. 07 18 11. 57 0.00 0.01 0. 00 19 11.59 0.01 0.11 0. 01 20 11.78 0.01 0.00 0. 02 21 11. 82 0.00 0. 00 0. 01 22 11.85 0.01 0. 00 0. 01 23 11.85 0.01 0.00 0. 05 24 11. 86 0.00 0.01 0. 01 25 16. 26 0.25 2.62 0.08 26 16.37 2.85 0.27 0. 01 27 16.43 0.35 0.22 0.05 28 18.06 0.00 0.00 0. 02 29 22.52 5.93 0.01 0.12 30 22.82 0.00 4.73 0.00

TABLE 3.6.2 Continued)

Participation Factor Mode e Fre cene Hz) X Y Z Notes 31 23.28 0. 03 0. 01 0. 13 32 26.20 1. 31 0. 07 0.34 33 26.36 6.14 0. 10 0. 24 34 27.05 0.19 5. 41 0.19 35 30.29 2. 83 0. 12 0. 96 36 30.77 0. 21 0. 18 2.16 37 31. 10 0.20 0.39 19.12 Primary Z-t1ode

~-

TABLE 3.6.3 COMPARISON OF MOST LIMITING STRESSES AND ALLOWABLE STRESSES ON STRUCTURAL MEMBERS Most Limiting Str uctural Most Limiting Type Of Combined Stress Al 1 owa bl e Member Load Combination Stress Ratio Limit Upper Grid Bars D + E 1.01 1.0 Top Peripheral D + E Bending 0.55 1.0 Beam + Axial Mid-Height D + E Bending 0.07 1.0 Peripheral Beam + Axial Vertical Corner Angles D+ Ta+ E'ending E' Bending

+ Axial 0.49 1.0 Base Angles + E Bending 0.80 1.0

+ Axial Outer Base D+T + E' Bending 0.96 1.0 Channel + Axial Center Base D + Ta + E' Bending 1.0 1.0 Channel + Axial Outer Shear + Ta + E' Shear + 0.99 1.0 Diaphragms Compression Buckling Internal Shear + Ta + Shear + 0. 92 1.0 Diaphragm Compression Buckling Fuel Cells E'+T,+E'+Ta+

Bending' 0.82 1.0 Axial Feet Brackets E' Bending 1.0 1.0 Screw Feet + Ta + Bending 0. 89 1.0

+ Axial

Figere 3.6.1 SPENT FllEI. STORAGE RACK SAP IV SEISMIC MOOEI.

Figure 3.6.2 S)NGLE STORAGE RAIlK NON-LINEAR MODEL LNTEII-BACK GAP ELEMENT P

8 G5 G5 KP RACK STRRCTIIRE ~ G5 CG5 G4 I8 G4 CG4 FUEL CELL NODES "G3 CG3 63 FUEL ASSEMBLY HODES KG2 CG2 G2 GAP ELEMEHTS CG K

I io

't 3 V

~RACK FEET, SI.IOIHR RAP ELEIAEHTS~

KF KF 20

Figure 3.8.3 MODE SHAPE 2

[Primary X Mode)

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E

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Figure 3.6.4 MODE SHAPE 3

[Primary Y Mode)

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3.7 References (1) L. M. Petrie and N. F. Cross, "KENO IV: An Improved Monte Carlo Criticality Program," ORNL-4938, Oak Ridge National Laboratory (November 1975).

(2) N. M. Greene, et al, "AMPX - A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B," ORNL-TM-3706, Oak Ridge'National Laboratory (March

'1976).

(3) H. H. Porath, "CCELL Users Guide," BNH/JN-86, Pacific Northwest Laboratories (February 1972).

(4) G. E. Whitesides and N. F. Cross, "KENO - A Multigroup Monte Carlo Criticality Program," CTC-5, Union Carbide Corporation Nuclear Division (September 1969).

(5) V. E.,Grob, et al, "Multi-Region Reactor Lattice Studies Results of Critical Experiments in Loose Lattices of UO Rods in H20," HCAP-1412, Hestinghouse Electric Corporation (1960).

(6) K. D. Lathrop, "DTF-IY - A FORTRAN-IV Program for Solving the Multigroup Transport Equation with Anisotropic Scatter-ing," LA-3373, Los Alamos Scientific Laboratory (July 1965).

(7) Information obtained via personal communication with E. B.

Johnson and G. E. Whitesides, Oak Ridge National Laboratory, Oak Ridge, Tennessee (September 1976).

(8) S. R. Bierman, E. D. Clayton and B. M. Durst, Critical Separation Between Subcritical Clusters of 2.35 Ht.X U-235 Enriched UO~ Rods in Hater with Fixed Neutron Poisons," PNL-2438, Pacif'tc Northwest Laboratories (October 1977).

9. SAP-IV, "A Structural Analysis Program for Static and Dynamic Response of Linear Systems", K. J. Bathe, E. L. Wilson, F. E. Peterson; Earthquake Engineering Research Center Report No . 73-11, Revised April 1976.

10. ANSYS, Engineering Analysis System Users Manual, Swanson Analysis Systems, Inc., Houston, Pa., March 1, 1975.

ll. General Electric Report No. 60GL20, "Investigation of the Slid-ing Behavior of a Number of Alloys Under Ory and Water Lubri-cated Conditions," R. E, Lee, Jr., January 22, 1960.

12. Friction Coefficients of Water-Lubricated Stainless Steels For a Spent Fuel Rack Facility," Prof. Ernest Rabinowicz, MIT, November 5, 1976, performed for the Boston Edison Company.

13. TID-7024, Nuclear Reactors and Earthquakes, August 1963.

ATTACHMENT 2 Insert the attached page 2 and Figure 1.2 and delete the similarly numbered ones in to our transmittal No. AEP:NRC:

00105, dated November 22, 1978.

The replacement spent fuel storage racks are to be fabricated primarily from type 304 stainless steel. The individual fuel assem-blies will be stored in square fuel storage cells fabricated from stainless steel-clad Boral* material. The high density (poison) spent fuel module construction is essentially a replica of the design used in the replacement racks for the Salem Nuclear Generating Station, which the Commission approved in the Safety Evaluation Report dated 1-15-79.. The module is shown in Figure 1-2.

The design utilizes a stiffened module base and an upper box structure consisting of plate diaphragms and a top grid. The vertical loads are carried by the module base. Horizontal seismic loads are carried to the module base through the plate diaphragms. The 10xl0 modules are nominally 15'-3/4" high and 9'-l/2" square; the 10xll modules are nominally 15'-3/4" high, 9'-l/2" wide and 10'" long.

The modular base is approximately 1'-1/4" from the pool floor. The modules will be installed in the spent fuel pool which is 39'" 58'" long and wide.

The detailed design of the spent fuel storage cells is slightly different from the design for the Salem Nuclear Generating Station.

Their basic function and construction, however, are similar. Figure 1-3 illustrates the storage cell design for the Donald C. Cook Plant.

Each cell is a square cross-section formed from an inner shroud of stainless steel, a center sheet of aluminum clad 8 C, and an outer shroud of stainless steel. This cell acts as a storage space and, in addition, provides sufficient neutron absorption by the boron carbide contained in the Boral sheet, to allow spacing of spent fuel in a 10.5 inch by 10. 5 inch array. The fuel weight is carried directly on the module base. A flared guide and transition section is provided at the top of each storage cell. This transition is designed to assure ease of entry and to preclude fuel assembly hang-up and damage.

~if' Indiana 8 Michigan Power has a contract with Allied-General Nuclear Services (AGNS) for fuel reprocessing services. Currently, however, no spent fuel can be sent to AGNS for reprocessing due to the December 23, 1977 NRC order terminating licensing proceedings for the Barnwell facility.

Presently, there are 129 spent fuel assemblies stored in the spent fuel pool. Sixty-five assemblies were discharged from Unit No. 1 in January, 1977. The remaining sixty-four assemblies were discharged in April, 1978. One hundred and twelve burnable poison clusters are contained in these assemblies and an additional thirty burnable poison clusters occupy storage locations.

The total storage capacity expected to be utilized is based on maintaining a full core discharge reserve storage capability. The estimated refueling schedules and expected number of fuel assemblies to be transferred into the spent fuel pool are given in Table 1-1.

From this table, it can be seen that the existing storage capacity would be filled by May, 1980 with FCDR.

• Trademark of Brooks and Perkins Incorporated

SPEHT fUEL CELL fEET REMOTELY AOlUSTABLE LEVELlNG COOLANT fLOW HOLES MOOULE 8ASE SPEHT fUEL POOL LlNER Figure T.2 TYPICAL HIGH-DENSITY POISON SPENT FIIEL STORAGE MODULE'O>1O ARRAY

ATTACHMENT 3 Insert the attached page 3 and Table 3.5-1 and delete the simi-larly numbered ones from to our transmittal No: AEP:NRC:00116 dated January 22, 1979.

Mr. Harold R. Denton,, w3w AEP'NRC:00116 Attachment 1 3.5.1 Continued Resul ts Thermal-hydraulic analysis of, the natural convection cooling of a single fuel assembly indicates that there is adequate cooling under normal and even under hypothetical condit'ions where a loss of forced coolant circulation is assumed to occur.

This re'suit is based on the two (2) cases presented in Table 3.5-1.

,The table results are for the 'fuel storage cell located at the pool center as shown in Figure 3.5-1 and are therefore the worst case.

The first case is the normal situation where the heat generation rate is 54.1 kw per assembly and the fuel storage cell inlet temperature is taken as 150 F, the maximum expected pool operating temperature under normal'onditions. The fuel rod peak cladding temperature is 190.gop; therefore, there can be no boi11ng within Ql the fuel assembly and the flow is single phase.

The second case is similar to the first except for the assumed inlet temperature, 240oF. This is the saturation temperature corresponding to the hydrostatic pressure at the top of the fuel storage cell. This is the maximum temperature that water flowing towards the'uel assembly inlet can attain under the hypo-thetical conditions where forced coolant cir'culation i" assumed lost and the surface of the 'pool is assumed to reach 212oF which is the saturation condition at that location. Under these assumed conditions, boiling does occur in the upper portion of, the fuel.

assembly. Maximum cladding temperature under this case is cal-culated to be less than 250oF.

In summary, the analysis indicates that even under hypothetical extreme conditions, peak clad temperatures are well below conditions .

where any degradation of the clad would occur.

3.5.2 S ent Fuel Coolin Ca abilit An evaluation has been performed to determine the capability of the spent fuel pool cooling system (SFPCS) for providing the cooling capacity required for both the annual discharge of 65 fuel'ssemblies from Unit 1 and 88 fuel assemblies from Unit 2 on a 1-> year cycle. It has been determined that the existing SFPCS, with both cooling loops in operation, can provide all necessary cooling for the normal discharge of fuel in the modi-fied storage capacity condition. However, the design criterion for the cooling system, as stated in the FSAR, is that each of the two independent cooling loops be capable of providing

,adequate heat removal capacity in the event one loop is out of service. That design criterion states that with a Revision 1

TABLE 3.5-1 Thermal Hydraulic Parameters For 54. 1 kW Fuel Assembly Located at Pool Center in Width Direction Flow Tyle Sin le Phase Two Phase S stem Parameter Case 1 Case 2 Cooling Loop Operational Yes No Fuel Assembly Heat Generation Rate, kW 54. 1 54.1 Fuel Assembly Coolant Bulk Inlet Temperature 150 240 Fuel Assembly Coolant Bulk Discharge Temperature, 'F* 180.2 240, Bundle Coolant Bulk Maximum Temperature, 'F 180.2 243 Fuel Rod Film Temperature Drop 'F, Max. 9.8 4.4 Fuel Rod Peak Cladding Temperature oF 190 247. 4 Equilibrium guality* 0 .005 Void Fraction* 0 .526 At top of assembly.