ML19249E014
| ML19249E014 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 7909250680 | |
| Download: ML19249E014 (5) | |
Text
SAFETY EVALUATION BY THE DIVISION OF OPERATING REACTORS CONCERNING FLORIDA POWER CORPORATION'S APPLICATION FOR AMENDMENT TO ITS OPERATING LICENSE TO INCREASE THE AUTHORIZED CAPACITY OF THE SPENT FUEL POOL AT CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302
1.0 INTRODUCTION
By its letter dated January 9,1978, as supp'emented by letters dated 2/3/78; 3/22/78; 8/30/78; 1/18/79; 3/16/79; and 6/29/79; the Florida Power Corporation (FPC) applied for a license amendment to increase the authorized storage capacity for spent fuei at Crystal River 3 from 256 to 1159 fuel assemblies.
2.0 DISCUSSION 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 steel 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.
FPC states that the highest anticipated U-235 enrichment is 3.3 weight percent.
This enrichment along with the technical specificaticn 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 U-235 per axial centimeter of fuel assembly.
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. 2.1 CRITICALITY ANALYSES As FPC stated in its March 22, 1978 and March 16, 1979 submittals, the fuel pool criticality calculations are based on unirradiated fuel assemblies with no burnable poisons which have a fuel enrich-ment of 3.3 weight percent U-235 and pure, i.e., unborated, water in the ' pool. 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 ten per sqo e centimeter of plate and that this minimum amount of boron would
, used to calculate the neutron multiplication factors.
Nuclear Energy Services, Incorporated (NES) perfomed 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 bicckness 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 temperatunt, voids in the pool water, and abnormal dislocations of fuel assemblies in the racks.
NES also did verification calculations with the KENO Monte Carlo program with sixteen group Hansen-Roach cross sections.
In i ts 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, 1979 response to our request for additional infomation, FPC stated that it will perform a surveillance test on coupons of the 84 / Polymer Composite plates to verify the continued presenta of the C
boron in the plates in the pool over the complete life of the storage racks.
2.1.1 EVALUATION The results of the neutron multiplication factor calculations submitted by FPC are generally lower than the results from other methods for similar fuel pool storage lattices.
By comparing FPC's results with those from other methods we have determined that an additional uncer-tainty of +0.01 needs 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 boron 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 includes the effect of the plutonium which is generated during the fuel cycle.
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. Since the neutron multiplication factor could increase with the fuel loading we will require that the use of these storage racks be prohibited for fuel assemblies that contain more than 42.7 grams of uranium-235 per axial centimeter of fuel assembly.
Since this neutron multiplication factor will increase if the boren loading in the plates is decreased below the stated minimum, we.equire that FPC make two tests. The first is an onsite neutron attentuation test to verify that there are no missing boron plates in the installed racks. The second is a surveillance test on coupons to continuously verify that the boron loading in any plate will not decrease below.012 grams of boron ten per square centimeter of plate. We also require that if the surveillance tests indicate that this boron loading is likely to go below.012 grams of boron ten per square centimeter of plate the NRC shall be notified and FPC shall recalculate the neutron multiplication factor in tne pool and send a report on this calculation to the NRC.
With these two tests, we find that all factors that could affect the neutron multiplication factor in this pool have been conservatively accounted for and that the maximum neutron multiplication factor in this pool 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 spent fuel pool.
This 0.95 acceptance criter-ion is based on the uncertainties associated with the calculational 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 the spent fuel pool to 0.95.
2.
1.2 CONCLUSION
We find that when any number of the fuel assemblies, which FPC described in these submittals, having no more than 42.7 grams of uranium-235 per axial cantimeter of fuel assembly or equivalent are loaded into the pro-posed racks, the keff in the fuel pool will be less than the 0.95 limit.
We also find that in order to preclude the possibility of the keff in the fuel pool 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 equivalent, 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 health and safety of the public will not be endangered by the use of the proposed racks.
b 2.2 SPENT FULL COOLING The licensed thermal power for Crystal River 3 is 2452 Nth.
FPC plans to refuel this reactor annually at which times about 59 of the 177 fuel assemblies in the core will be replaced.
To calculate the maximum heat loads in the spent fuel pools PFC assumed a 150 hour0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> time interval betweea reactor shutdown and the time when either the 59 fuel assemblies in the normal refueling or the 177 fuel assemblies in a full core officad are placed in the spent fuel pools.
For this cooling time, FPC used the method given in the NRC Standard Review Plan 9.2.5 to calculate
. 6 maximum heat joads 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 changers.
Each pump is designed to pump 1500 gpm (7.5 x 10 pounds 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 flgwing through the shell side of the heat exchanger at a rate of 7.5 x 103 pounds per hour.
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 offloaded full core FPC, in its March 16, 1977 response to our request for additional information, stated that the Decay Heat Removal system could be aligned to cool the spent fuel pool by closing six valves in the spent fuel cooling system and opening seven valves in the Decay Heat Removal system $ This system has two loops each of which is designed to remove 30 x 10 BTU /hr at a 140*F outlet temperature.
In regard to emergency make up water for the spent fuel pool, Section 9.3.2.8 of the FSAR FPC stated that the eight inch diameter pipe to the Decay Heat Removal System is designed to Seismic Class I criteria and that it connects the spent fuel pool to the 420,000 gallon borated water storage tank.
2.2.1 EVALUATION 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 the uncertainty 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 spent fuel pool are conservatively high.
We also find that the maximum incremental heat load that could be added by increasing the number of spent fuel assemblies in the pool from 256 to 1159 is 3.5 x 106 BTV/h r.
This is the difference in peak heat loads for the present and the modified pools We find that with two pumps operating the spent fuel pool cooling system can maintain the fuel pool outlet water temperature below 128 F for the normal refueling orfload 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 pool.
Since both the Spent Fuel Fool 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 spent fuel pool cooling.
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However, if this did occur just after a full core offload, the maximum heat up rate of the spent fuel pool water would be about 9 F/hr. Thus assuming that the spent fuel pool water temperature was at its maximum 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 make up rate will be less than 70 gpm.
We find that eight hours will be sufficient time to establish a 70 gpm make up rate.
2.
2.2 CONCLUSION
We find that the pr'esent cooling capacity for the Crystal River 3 spent fuel pool will be sufficient to handle the incremental heat load that will be added by the proposad modification. We also find that this incremental heat load will not alter the safety considerations of spent fuel pool cooling frc;n that wwhich we previously reviewed and found to be acceptable.
We conclude that there is reasonable assurance that the health and safety of the public will not be endangered by the use of the proposed design.
2.3 INSTALLATION OF RACKS AND FUEL HANDLING Because of the first refueling at Crystal River 3, which was performed in May 1979, FPC stated in its March 16, 1979 submittal that there will be sixty spent fuel assemblies in the poolswhen the new racks are installed. However, FPC stated that these sixty fuel assemblies will be located in Pool 8 during the modification of Pool A and that 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.
2.3.1 EVALUATION We find that, because Crystal River 3 has two separated spent fuel pools and has a missle shield over the pools, FPC can adequately pro-tect the spent fuel assemblies stored in the pools during the change of racks.
After the racks are installed in the pool, the fuel handling procedures in and around the pool will be the same as those procedures that were in effect prior to the proposed modifications.
2.
3.2 CONCLUSION
We conclude that there is reasonable assurance that the health and safety of the public will not be endangered by the installation and use of the proposed racks.
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