ML19345C882
| ML19345C882 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/17/1980 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19345C883 | List: |
| References | |
| NUDOCS 8012080538 | |
| Download: ML19345C882 (9) | |
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UNIVED STATES NUCLEAR REGULATORY COMMISSION 3
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,t WASHINGTON, D. C. 20555 ty.....f
--F'.0P.ivA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSIMMEE CITY OF LEESBURG CITY OF HEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF_NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO SEBRING UTILITIES COMMISSION SEMINOLE ELECTRIC CDOTERATIVE, INC.
CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 License No. DPR-72 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The. application for amendment by Florida Power Corporation, et al (the licensees) dated March 17, 1978, as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regu-lations set forth in 10 CFR Chapter I; B.
The facility will operate in. conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities' authorized by this amendment can be conducted without endangering the health and safety i
of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is ir accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied, l
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. 2.
Accordingly, the license is amend'ed by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-72 is hereby amended to read as follows:
4 (2). Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 36, are hereby incorporated in the license.
Florida Power Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMM?SSION
<E hY, obert W. Reid, Chief Operating Reactors Brar -h #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 17, 1980 O
i.
ATTACHMENT TO LICENSE AMENDMENT N0.
36 FACILITY OPERATIrlG LICENSE NO. DPR-72 DOCKE1,:j0. 50-302 Replace the following pages of Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amend-l ment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain docu-ment completeness.
3/4 9-7 5-4 5-5 5-6 i
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REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2750 pounds, except for movement of the missile shield and pool divider gate as necessary for access to the fuel assemblies, shall be prohibited from travel over fuel assemblies in the storage pool.*
l APPLICABILITY: With f.el assemblies and water in the storage pool.
ACTION:
With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.7.1 Crane interlocks and/or physical stops which prevent crane travel
~
with loads in excess of 2750 pounds-over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane operation and at least once per 7 days during crane operation.
4.9.7.2 Prior to operating tne crane in the cask handling mode, verify that:
a.
No fuel assemblies are in the storage pool adjacent to the cask loading area, and b.
The watertight gate between storage pools is in place and sealed.
'*except for the removal of old spent fuel racks and the installation of the high density spent fuel storage racks in the spent fuel storage pool.
The missile shield shall cover the spent fuel in the alternate pool during rack handling.
CRYSTAL RIVER - UNIT 3 3/49-7 Amendment No. 36 i
REFUELING OPERATIONS COOLANT CIRCULATION i
LIMITING CONDITION FOR OPERATION
+
3.9.8 AtLleast one decay heat removal loop shall be in operation.
APPLICABILITY: MODE 6.
ACTION:
4 a.
With less than one decay heat removal loop in ope ation, except as provided in b. below, suspend all operr. ions involving an increase in the reactor decay heat load or a reduction in n.
i boron concentration of'the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
The decay heat removal loop may be removed from operation for
+
up to I hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of CORE ALTERATIONS to prevent water turbulence problems.
c.
The provisiont of Specification 3.0.3 are not applicable.
4 SURVEILLANCE REQUIREMENTS 4.9.8 A decay heat removal loop shall be determined to be operating and circulating reactor coolant at a flow rate of 3,2700 gpm at least onde per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
CRYSTAL RIVER - UNIT 3 3/4 9-8
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LOW POPULATION ZONE FIGURE 5.1-2 CRYSTAL RIVER - UNIT 3 5-3 D"Ild A3'9'[ L)
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~ DESIGN FEATURES ~
DESIGN PRESSURE AND TEMPERATURE _
The reactor containment. building is designed and shall be maintained 5.2.2
'for a maximum internal pressure of 55 psig and a temperature of 281*F.
REACTOR CORE 5.3
- FUEL ASSEM8 LIE 5_
The rector core shall contain 177 fuel assemblies with each fuel'
+
5.3.1 Each fuel assembly containing 208 fue1~ rods clad with Zircaloy -4.
rod shall have a nominal active fuel length of 144 inches and contain a The initial core loading shall l
maximum total weight of 2253 grams uranium.
have a maximum enrichment of 2.83 weight percent U-235.
Reload fuel shall be similar in~ physical design to the initial core loading and shall have a maximum enrichment of 3.30 weight percent U-235.
' CONTROL RODS The reactor core shall contain 61 safety and regulating and 8 axial 5.3.2 power shaping (APSR) control rods. The safety and regulating control rods shall contain a nominal.134 inches of absorber material.
The APSR's shall The contain a nominal 36 inches of absorbe'r material at their lower ends.
nominal values of absorber material shall be 80 percent silver,15' percent indium and 5 percent cadmium. :All control. rods shall be clad with stainless steel tubing.
CRYSTAL. RIVER - CIT 3 5-4 AmendmentNo.M36
DESIGN FEATURES 5.t. REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
In accordance with the code requirements specified in a.
Section 4.1.2 of the FSAR, with allowance for nonnal degradation pursuant to applicable Surveillance Requirements.
b.
For a pressure of 2500 psig, and For a temperature of 550*F, except for the pressurizer c.
and pressurizer surge line which is 670*F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 12,180 1 200 cubic feet at a nominal T of 525'F.
avg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1 The new fuel storage racks are designed and shall be maintained with a nominal 21-1/8 inch center-to-center distance between fuel assemblies placed in the storage racks. The high density spent fuel storage racks are designed and shall be maintained with a nominal 10.5 inch center-to-center distance between fuel assemblies placed in the storage racks.
Both of these rack designs ensure a keff equivalent to < 0.95 with the storage pool filled with unborated water. The k 1 6.95 includes a conservative allowance of > 1% a k/k for uncertainties.ff of e
In addition, fuel in the new and spent fuel storage racks shall have a U-235 loading of < 42.7 grams of U-235 per axial centimeter of fuel assembly (<an enrTchment of 3.3 3.3 weight percent U-235).
i ORAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent drair.'ag of the pool below elevation 138 feet 4 inches.
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, CRYSTAL RIVER - UNIT 3 5-5 Amendment No. 36 I
DESIGN FEATURES CAPACITY 5.6.3 -The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1153 fuel assemblies and 6 failed fuel containers.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limit of Table 5.7-1.
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. CRYSTAL RIVER - Unit 3 5-6 Amendment No. 36 g.-,.a
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QJ pm Es UNITED STATES
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j NUCLEAR REGULATORY COMMISSION e E WASHINGTON. D. C. 20655 '
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDl1ENT NO. 36 TO FACILITY OPERATING LICEllSE NO. DPR-72 FLORIDA POWER CORPORATION, ET AL.
CRYSTAL RIVER UNIT NO. 3 NUCLEAR GENERATING PLANT DOCKET N0. 50-302 i
1.0 Introduction By application-dated March 17, 1978, as supplemented January 9, 1978, March 3 and 22,1970 August 30, 1978, January 18,1979, March 16,1970, June 29, 1979, September 5,1979, October 1 and 10,1979, and December 5, 1979, the Florida Power Corporation (FPC) proposed to install high density fuel storage racks at the Crystal River Unit No. 3 Nuclear Generating Plant (CR-3).
The proposed modification would increase the storage capacity in the spent fuel pools (SFPs) for up to 1153 fuel assemblies and six failed fuel contain-ers. The proposed modification consists of replacing existing fuel assembly racks with high density, free standing storage racks without changing the basic structural geometry of the SFPs.
(Two individual spent fuel storage pools are located within the fuel handling area of the Auxiliary Building.
Both' pools are rectangular:
Pool A, ?:' 2" by 24' 0" and Pool B, 32' 7" by 24' 0", with a depth of 43' 8". )
2.0 Discussion Each storage rack consists of an assembly of fue' storage cells spaced 10.5 inches on center and welded to a base grid struuure.
Each storage cell is a double wall Type 304 stainless steel box with an inside square dimension of 8.9375 inches which allows sufficient clearance (0.2005 inch each side of the fuel assembly) to avoid interferences during fuel storage and removal operations.
The double wall construction provides four compartments in which poison elements (8 C poison sheets) can be placed.
The top opening of thestoragecellisflarhdtofacilitateinsertionofthefuelassembly;the bottom member of the storage cell provides the level support surface required for the fuel assembly and contains the cooling flow orifice.
The bottom member of each storage cell sits on and is welded to the rack based unit which is basically a grid structure constructed from Type 304 stainless steel wide flange and box beam members.
Continuous spacer bars are provided at the middle and top of the storage cells to ensure that the required pitch (10.5 inches) is maintained between storage cells in both directions (north /
south and east / west).
The spacer bars which are intermittently welded to the storage cells also maintain the vertical alignment of the cells.
Support feet attached to the bottom of the rack base raise the rack above the pool floor to the height required to provide an adequately' sized cooling water supply plenum (for natural circulation).
Each support foot contains a remotely adjustable
.iackscrew to permi the rack to be leveled following installation.
8012080 y37
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Because of the pool configurations (Pool A and Pool B) and requirements for-failed fuel storage, it is necessary to supply different rack. sizes.
In addition to the basic 6 x 6 storage rack, racks are provided with 6--x'5, 5 x 5, 4 x 5, and 4 x 6 arrays of storage locations.
The storage racks which are free to slide are positioned on the pool floor so that adequate clearances are provided between racks and between the racks and g
pool structure to avoid impacting during seisrcic events.
The horizontal seismic loads transmitted from the rack structure to the pool floor are only those associated with friction between the rack structure and the pool liner.
The vertical dead-weight and seismic loads are transmitted directly to the pool floor by the support feet.
3.0 Evaluation-3.1 Structural'and Mechanical The supporting arrangemcnts for the spent fuel modules, including their restraint, design, fabrication, and installation procedures; the structural design and analysis procedures for all loadings, including seismic and impact loadings; the load combinations; the structural acceptance criteria; the quality assurance requirements for design, fabrication, and installation; and applicable industry codes were all reviewed in'accordance with the applicable portions of the NRC Position for Review and Acceptance of SFP Storage and Handling Applications of April 1978, as revised January 1979.
Seismic analysis was performed using pool floor response time histories which conform to those approved in the original plant design.
The pool floor response time histories were determined in the seismic analysis of the Auxiliary Building using a base acceleration time-history compatible with smoothed response spectra which conform to the positions in Regulatory Guide 1.60, " Design Response Spectra for Seismic Design of Nuclear Power Plants," and structural damping values wnich conform to the positions in Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants." The pool floor horizontal time histories were then used as input to perform non-linear time-history analyses of the lateral motion of the fuel racks.
The use of non-linear time-history analyses in the horizontal direction was necessitated by the non-linear characteristics of the fuel racks in the lateral direction.
The combination of modes and spatial earthquake components in the seismic response analysis is in accordance with Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis".
In the Spent Fuel Rack. Structural Analysis, the effects of a. gap between a l
storage cavity and a fuel assembly, and the effects of submergence in water of the fuel racks were accounted for.
The rack has been mathematically modeled as a three-dimensional finite-element structure consisting of discrete elastic beam and plate elements.
Two representative fuel assembly load conditions t
(partially and fully loaded), were used in the analyses.
The static analysis of the finite-element model has been performed using the direct stiffness methods of structural analysis to determine the internal forces and stresses in each element.
For dynamic analyses, the natural frequencies and the moda shapes of the finite-element model were detennined first, then the analyses were perfomed to determine the dynamic responses due to the seismic effects including the fuel impacting and hydrodynamic action. The total system response is obtained by combining the individual model response values in accordance with Regulatory Guide 1.92.
n 1
9 4 Non-linear time-history analyses were also performed to determine ar1y potential impacting between adjacent fuel racks and between the fuel racks and the spent fuel pool structure.
The storage rack and the stored fuel assemblies are represented by a.two-dimensional lumped mass finite element model consisting of two finite-element cantilever beams, representing the storage cells and the stored fuel assemblies, attached to a floor mass by means of a non-linear sliding element. The range of friction coefficient used in the analyses, between the rack and two pool floor, was selected based on published test data.
These analyses resulted in conservative values for the rack sliding and the shear forces transmitted through the rack and pool interfaces.
Rack material properties used in the analysis of the spent fuel racks are in accordance with the requirements of Subsection NF and Appendix I of Section III of tne ASME Boiler and Pressure Vessel Code.
Results of the seismic analysis show that the racks are capable of withstanding the loads associated with all the design loading conditions without exceeding allowable stresses.
Analyses were performed to assess the effects of the fuel drcp accidents.
The postulated drop accidents include a straight drop on the top of a rack, a straight drop through an individual cell all the way to the bottom of the rack, and an inclined drop on the top of a rack.
The drop height used in the analyses is 34 inches which is the maximum ' height that the crane can lift the fuel.
The 'SFPs are constructed of concrete walls and floors lined with stainless steel plates. Fuel pool structures have been analyzed for the additional loads resulting from the proposed increase in pool storage capacity and the most severe load combination conditions with the results indicating that the maximum loads are within the allowable stresses and the fuel pool floord are, adequate to withstand the effects of the new racks and additional fuel.
3.2 flaterials The Type 304 stainless steel (ASTM Specification A-240)'used in the new spent fuel storage racks is compatible with the storage pool environment, which is demineralized borated water controlled to a maximum 150 F temps.rature.
To prevent any adverse effect from gas generated by B C material exposure to the g
fuel pool environment, the poison material compartrhents are open at the top.
Based on our review of previous operating experience with similar materials approved and in use, we have concluded that there is reasonable assurance that no significant corrosion of the racks, the fuel cladding, or the pool liner will occur over the lifetime of the plant.
3.3 Analysis, Design, Fabrication and Installation e
The analysis, design, fabrication, and installation' of the proposed new spent fuel rack storage system are in conformance with accepted codes and criteria.
The analysis of the structural loads imposed by dynamic, static, seismic and thermal loadings;..and the acceptance criteria for the appropriate loading conditions are in accordance with the.ppropriate portions of the NRC Position for Review and Acceptance of Spent 'sel Pool Storage and Handling Applications, April 1978, including errata, January 1979.
~ The mechanical properties for the materials used in the rack design are consistent with the normal and accident pool conditions.
The quality assurance procedures for the materials, fabrication, installation, and examination of the new racks are in accordance with the accepted requirements of ASME Code,Section III, Subsection NF, Articles NF-2000, NF-4000, and NF-5000.
+
In addition. the design, procuremei.', and fabrication of the spent Del racks comply with the pertinent requirements of Appe.Six B to 10 CFR 50, and delineated in Regulatory Guide 1.29, " Seismic Design Classification".
The effects of the additional loads on the existing pool structurcs due to the high capacity storage racks have been examined.
The pool structural integrity is assured by conformance with the original Final Safety Andlysis Report (FSAR) acceptance criteria.
There is na evidence at this time to indicate that corrosion of the fuel assemblies, the stainless steel rack structures, or the fuel pool liner will occur over the lifetime of the plant, at the temperatures and quality of the demineralized borated water to be mintained in the pools.
Installation procedures for the new racks have also been reviewed. Missile shields that are normally in place over the SFPs will remain in place over Pool B while old racks are being removed and new racks installed in Pool A.
A similar procedure will be used for installation of new racks in Pool B.
The licensee's ---lysis has shown that the missile shields will withstand the force of dropping a fuel rack from the maximum height that would be used in rack transfer operations. Based on handling procedures described to prevent damage to the stored fuel and to prevent interaction between old and new racks, the installation procedures have been found to be acceptable.
We found that the subject modification proposed by FPC is acceptable and satisfies the applicable requirements of the General Design Criteria 2, 4, 61, and 62 of 10 CFR, Part 50, Appendix A.
3.4 Criticality Consideratinn The proposed spent fuel racks are to be made up of individual containers which are approximately 9 inches square by 14 feet long.
These containers are to be fabricated from Type 304 stainless ste.al by using 1-1/4" x 1/8" angle stock for the corners which are welded to sides which consist of double sheets of.060" thick stock.
Sheets of the Carborundum Company's Boron Carbide Composite Material, which are approximately 6.7 inches wide by 0.075 inches thick will be placed between these double sheets of stainless steel prior to weld-ing.
Since there will be a sheet of boron material in each of the double container walls, and since there will be one container for every fuel assembly, there will be two sheets of this boron material between every two fuel assemblies.
Spacer grids and clips will be used to separate these containers in the modules to obtain a design lattice pitch of 10.5 inches.
This will result in their being about one inch of water between the containers. This 10.5 inch pitch and the overall dimension of the fuel assembly, which is 8.52 inches, gives a fuel region volume fraction of 0.658 for the storage lattice.
-~
. i FPC states that the highest anticipated Uran.ium-235 enrichment is 3.3 welqht percent.
This enrichment along with the technical specification limit on the loading of uranium dioxide in a fuel assembly, which is 536.94 kilograms, results in a maximum loading of 42.7 grams of Uranium-235 per axial centimeter of fuel assembly.
The FPC fuel pool criticality calculations a're based cn unirradiated fuel assemblies with no burnable poisons which have a fuel enrichment of 3.3 weight percent Uranium-235 and pure, i.e., unborated, water in the pool.
3.3 weight percent Uranium-235 corresponds to 42.7 grams of Uranium-235 per axial centimeter of fuel assembly with the present fuel.
FPC also stated in its March 22, 1978 submittal that the areal density of the boron in each of the plates would be a minimum of 0.012 grams of boron-10 per square centimeter of plate and that this minimum amount of boron is used to calculate the neutron multiplication factors.
Nuclear Energy Sers ices, Incorporated (NES) performed the criticality analyses for FPC.
NES made parametric calculations by using the HAMMER computer program to obtain four-group cross sections for EXTERMINATOR diffusion theory calculations.
The blackness theory program, BRM, was used to calculate the thermal and epithermal group cross sections for the boron region.
This calculational method was used to determine the nominal koo and then the effects of design and fabrication tolerances, changes in temperatum, voids in the pool water, and abnormal dislocations of fuel assemblies in the racks.
NES also did verification calculations with the KEN 0 Monte Carlo '
program with sixteen group Hansen-Roach cross sections.
In its March 22, 1978 submittal, FPC stated that the overall result of all of these calculations is that, with an assumed calculational uncertainty of +0.01, the maximum, " worst case", abnormal neutron multiplication factor is 0.9356.
In its March 16,197, response to our request for additional information.
FPC sf ted that it will perform a surveillance test on coupons of the B4 / Polymer Composite plates to verify the continued presence of the C
boron in the pfates in tne pools over the complete life of the storage racks.
In addition, FPC will perform an on-site neutron attenuation test to verify that there are no missing boron plates in the racks.
The results of the neutron multiplication factor calculations submitted by FPC are generally lower than the results from other methods for l
similar fuel pool storage lattices.
By comparing FPC's results with I
those from other methods we have determined that an additional uncer-tainty of +0.01 need: to be added to FPC's maximum, " worst case", abnormal neutron multiplication factor of 0.9356; so that for practical purposes the maximum neutron multiplication factor in these racks for the specified fuel loading and baron plate loading is 0.95.
By assuming new, unirradia-ted fuel with no burnable poison or control rods, these calculations yield the maximum neutron multiplication factor that could be obtained~ through-out the life of the fuel assemblies.
This incluces the effect of the plutonium which is generated during the fuel cycle.
i
Since the neutron multiplication factor could increase with the fuel loading, we have determined that the'dse of these storage racks-should be prohibited for fuel assemblies that contain more than 42.7 grams of Uranium-235 per axial centimeter of fuel assembly. An appropriate Technical Speci-fication has been established.
We find that all factors that could affect the neutron multiplication factor in the pools have been conservatively accounted for and that the nnximum neutron multiplication factor in the pools with the proposed racks will not exceed 0.95.
This is NRC's acceptance criterion for the maximum (worst case) calculated neutron multiplication factor in a SFP.
This 0.95 acceptance criterion is based on the uncertainties associated with the cal-culational methods and provides sufficient margins to preclude criticality in the fuel. Accordingly, there is a Technical Specification which limits the effective neutron multiplication factor in each SFP to 0.95.
We find that when aay number of the feel assemblies, which FPC described in these submittals, having no more than 42.7 grams of Uranium-235 per axial centimeter of fuel assembly or equivalent are loaded into the pro-posed racks, the keff in the fuel pools will be less than the 0.95 limit.
We also find that in order to preclude the possibility of the keff in the fuel pools from exceeding this 0.95 limit without being detected, the use of these high density storage racks will be prohibited for fuel assemblies that contain more than 42.7 grams of Uranium-235, or eqt.ivalent, per axial centimeter of fuel assembly. On the basis of the information submitted, and the keff and fuel loading limits stated above, we conclude that the
' design is acceptable from~ criticality consideration.
3.5 Spent Fuel Cooling The licensed thermal power for CR-3 is 2452 MWt.
FPC currently plans to refuel this reactor annually at which time.s about 59 of the 177 fuel assemblies in the core will be replaced.
To calculate the maximum heat 1 ads in the SFPs, FPC assumed a 150-hour time interval betwun reactor shutdown and the time when either the 59 fuel assemblies in the normal refueling or the 177 fuel assemblies in a full core offload are placed in the spent fuel pools.
For this cooling time, FPC used the method given in the NRg Standard Review Plan 9.2.5 to calculate maximum heat goads of 16.7 x 10 BTU /hr for sixteen successive refuelings and 33.4 x 10 BTU /hr for the full core offload which fills the pool after sixteen refuelings have been performed.
The spent fuel cooling system consists of two pumps and two heat ex-5 Each pump is designed to pump 1500 gpm (7.5 x 10 pounds changers.
per hour), and each heat exchanger is designed to transfer 8.75 x 106 BTU /hr from 129'F fuel pool water to 95*F closed cycle cooling water, which is f1pwing through the shell side of the heat exchanger at a rate of 7.5 x 103 pounds per hour.
.J
. FPC states that this system, with two pumps running, will be able to keep the spent fuel pool outlet temperature below 128'F through the sixteenth annual refueling.
For cooling an of floaded full core, FPC's March 16, 1977 response to our request for additional infonnation stated that the Decay Heat Removal System could be aligned to cool the SFPs by closing six valves in the spent fuel cooling system and opening seven valves in the De::ay Heat Removal System This system 8
has two loops each of which is designed to remove 30 x 10 BTU /hr at a 140"F outlet temperature.
Ir. cgard to emergency makeup water for the SFPs, Section 9.3.2.8 of the FSAR states that the eignt inch diameter pipe to the Decay Heat Hemoval System is designed to Seismic Class I criteria and that it con-nects the 5 fps to tne 420,uuu gailon borated water storaga tank.
By using the method given on pages 9.2.5-8 through 14 of the November 24, 1975 version of the NRC Standard Review Plan, with.th n ertainty 7
factor, K, equal to 0.1 for decay times longer than 10 seconds, for a decay time of 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />, we find the FPC's maximum heat loads in the SFPs are conservatively hich. We also. find that the maximun incremental heat load that could be added by increasing the number of spent fuel assemblies in the pools from 256 to 1159 is 3.5 x 106 BTU /hr.
This is the difference in peak heat loads for the present and the modified pools.
We find that with two pumps operating the SFP cooling system can maintain the feal pool outlet water temperature belog 128*F for the normal refueling offload that fills the pools. We find that the capacity of the Decay Heat Removal System is adequate for maintaining the spent fuel water temperature below 140*F for the full core offload that fills the poois.
Since both the SFP Cooling System and the Decay Heat Removal System are seismic Class I systems, it is highly unlikely that a single failure could result in a complete loss of SFP cooling.
However, if this did occur just after a full core offload, the maximum heatup rate of SFP water would be about 9 F/hr. Thus assuming that SFP water temperature was at its maximtsa of 140 F at the time of the loss of cooling, it would be more than eight hours before the pool would start to boil. We calculate that after boiling starts the required water makeup rate will be less than 70 gpm. We find that eight hours will be sufficient time to establish a 70 gpm makeup rate.
We find that the present cooling capacity for the CR-3 SFPs will be sufficient to handle the incremental heat load that will be added by the proposed modification. We also find that this incremental heat load will j
not alter the safety considerations of SFP cooling from that which we pre-i viously reviewed and.found to be acceptable.
.l
. 3.6 Installation of Racks and Fuel Handling Because of the two refuelings at CR-3, there are 120 spent fuel assemblies
' in the pools. These fuel assentlies will be located in Pool B during the modification of Pool A and the missile shield over Pool B will prevent damage to any of the spent fuel assemblies in the unlikely event that a load is dropped during the change of racks in Pool A.
A similar routine will be used for Pool B nodifications in which all the fuel will be in Pool A.
We find that, because CR-3 has two separated SFPs and has a missile shield over the pools. FPC can adequately protect the spent fuel assemblies stored in the pools during the change of racks. After the rack., are installed in the pools, the fuel handling procedures in and around the. pools will be the same as those procedures that were in effect prior to the proposed modifica-tions.
3.7 Fuel Handling The CR-3 Technical Specifications prohibit loads greater than 2750 pounds (the nominal weight of a fuel assembly and handling tool) to be transported over spent fuel in the SFPs except for removal of old racks and installation of new racks in which case the missile shield must be over the fuel in the alter-nate pool. We have, therefore, concluded that the likelihood of any heavy load handling accident is sufficiently small that the proposed modification is acceptable.
The potential consequences of fuel handling accidents in the SFP area pre-sented in the CR-3 Safety Evaluation Report (SER) dated July 197-4 are not changed by the proposed modifications to the SFPs.
3.8 Occupational Radiation Exposure We have reviewed FPC's plans for the removal and disposal of the low density racks and the installation of the high density racks with respect to occupa-tional radiation exposure. The occupational radiation exposure for this operation is estimated by FPC to be about 8.5 man-rem. We consider this to be a reasonable estimate.
This estimate reoresents a small fraction of the total man-rem burden frem occupational exposure at the plant.
The estimated man-rem exposure to re-rack the pools is based on FPC's detailed breakdown of occupational exposure for each phase of the pools' modification.
FPC considered the number of individuals perfoming a specific. job, their occupancy time while perfonning this job and the average dose rate in the area where the job will be performed. The modification will be done in a dry : pool (i.e., each pool will be drained) after decontamination of the pool nas been performed by hydro-lasers followed by vacuuming and filtra-tion of the final 6 inches of water on the pool floor.
By using these techniques, we conclude that each pool modification will be done in a manner that will ensure as low as is reasonably achievable (ALARA) exposures to the occupational workars.
FPC has presented altern&tive plans for the disposal of the old racks which considered removing, decontaminating and crating intact racks vs. removing, decontaminating, cutting and packaging the small sections.
FPC is consi-dering three methods of disposal.of the old racks:
(1) shipping the racks
.g.
whole to Barnwell for burial; (2) cutting them into small sections to reduce the volume and then shipping them to Barnwell for burial; cr (3) crating the racks whole and shipping the intact racks to a vendor for further decontamina-tion and scrapping. Taking into account alternative disposal costs and ex-posures, FPC will make the final decision as to the choice of disassembly and disposal of the old racks so that exposures will be kept to levels that are ALARA.
We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies on the basis of information supplied by FPC for dase rates in the SFP area from radionuclide concentra-tions in the pool water and the spent fuel asserblies. The spent fuel as-semblies themselves will contribute a negligible fraction of the dese rates in the pool area because of the depth of water shielding the fuel. Conse-quently, the occupational radiation exposure resulting from the additional spent fuel in the pools represents a negligible burden. Based on present and projected operations in the SFP area, we estimate that the proposed modifi-cation should add lass than one percent to the total annual occupational radiation exposure burden at this facility. The small increase in radiation exposure will not affect FPC's ability to maintain individual occupational doses to ALARA and within the limits of 10 CFR Part 20. Thus, we conclude that storing additional fuel in the SFPs will not result in any significant increase in doses received by occupational workers.
3.P Radioactive Waste Treatment The plant contains waste treatment systems designed to collect and process the gaseous, liquid and solid wastes that might contain radioactive material.
The waste treatment systems were evaluated in the SER dated July 1974. There will be no change in the waste treatment systems or in the conclusions of the evaluation of these systems as described in Section 11 of the SER because of tbo proposed modification.
4.0 CONCLUSION
We have concluded, based on the considerations riiscussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in 'he proposed manner, and (2) such activities will be conducted in compliance with the Com-mission's regulations and the issu;nce of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: November 17, 1930 l