ML19282B328
| ML19282B328 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 02/21/1979 |
| From: | Lainas G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19282B327 | List: |
| References | |
| TASK-09-01, TASK-9-1, TASK-RR NUDOCS 7903130096 | |
| Download: ML19282B328 (6) | |
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SAFETY EVALUATION BY THE DIVISION OF OPERATING REACTORS CONCERNING DAIRYLAND POWER COOPERATIVE'S APPLICATION FOR AMENDMENT TO ITS OPERATING LICENSE TO INCREASE THE AUTHORIZED CAPACITY OF THE SPENT FUEL POOL AT THE LACROSSE BOILING WATER REACTOR DOCKET NUMBER 50-409
1.0 INTRODUCTION
By its letter dated April 20, 1978 as supplemented by letters dated June 7 July ll, August 7, October 16, 1978 and January 4 and 31,1979, the Dairyland Power Cooperative (DPC) applied for a license amendment to increase the authorized storage capacity for spent fuel at the Lacrosse Boiling Water Reactor from 133 to 440 fuel assemblies.
2.0 DISCUSSION The proposed spent fuel storage racks are to be made up of stainless steel containers. These will be about nine feet long and will have a square cross section with an inner clearance dimension of 6.13 inches.
These containers are to be made by welding four 0.062" sheets of type 304 stainless steel to four 1.5" x 0.187" corner angle pieces. As stated in DPC's October 16, 1978 submittal, on each of two opposite sides of each container will be a centrally located 3.94" wide by 0.1" thick sheet of 8 C/ Polymer Composite, which is mcnufactured by the 4
Carborundum Company.
In its January 4,1979 response +o our request for additional information, DPL stated that there will be 6 minimum of 0.024 grams of boron ten (i.e.,1.45 x 1021 boron ten atoms) in every square centimeter of the composite sheet.
In its January 31, 1979 response to our request for additional information, DPL stated that the size of the B C particles in the composite sheet will be less 4
than
.020 inches. This composite boron plate will be held in place by a 0.028 inch thick, stainless steel, cover sheet.
7903130096
. These containers will be oriented in the racks so that one composite boron plate will be located between each pair of fuel assemblies. The nominal pitch between fuel assemblies will be 7.0 inches within the rack and not less than 7.0 inches in contiguous racks. The size of the square fuel region in the calculational model is 5.65 inches.
This resul ts in an overall fuel region volume fraction in the storage lattice of 0.65.
2.1 CRITICALITY ANALYSES The Lacrosse fuel pool criticality calculations are based on unirradiated fuel assemblies with no burnable poison. Calculations were made for both stainless steel and Zircaloy clad fuel elements.
The fuel loadings which were used in the analyses were 22.4 grams of uranium-235 per axial centimeter of fuel assembly for the stainless steel clad fuel elements and 16.6 grams of uranium-235 per axial centimeter of fuel assembly for the Zircaloy clad fuel elements.
Nuclear Energy Services, Incorporated (NES) performed the criticality analysas for OPC.
NES made parametric calculations by using the i
HAMMER computer program t7 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 baron region. This calculational method was used to determine the nominal koo and then the effects of design and fabrication tolerances, changes in temperature, voids in the pool water, and abnormal dislocations of fuel assemblies in the racks.
To obtain its January 31, 1979 response to our request for additional information, NES used the KENO IV Monte Carlo program to calculate the increase in koo that would be obtained if it was assumed that all of the boron carbide particles were the maximum size of.020".
NES found that this increase would be 0.9 percent. Adding this increment to NES's previously calculated value for the nominal reference configu-ration gives a koo of 0.922. With all of the other effects listed above included NES's maximum value of koo is 0.925. NES found that if a fuel assembly in the shipping cask area of the pool was brocght up to the outside of a fully loaded rack, the koo could increase by 0.004 Thus NES's " worst abnormal" koo is 0.93.
Moreover, NES calculated the reference configuration with the more vigorous KENO IV Monte Carlo program. This resulted in a lower koo then was obtained with the EXTERMINATOR program; so NES assumed that no calculational bias was needed on the worst case koc of 0.93.
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.. In its January 22, 1979 response to our request for additional information, DPC stated that two inspections will be performed, on site, to verify the presence of the boron plates in the racks.
First, every location will be checked by visually observing the boron plates through a test hole which will te in everyone of the stainless steel jackets.
- Second, a blackness test shall be performed on ten percent of the neutron absorber plate locations, which will be randomly selected in each storage rack. DPC also stated that if the blackness test reveals any missing boron plates, all of the plate locations will be checked.
In its January 4,1979 response to our request for additional information, DPC stated that it will perform a surveillance test on coupons of the 8 C/ Polymer Composite plates to verify the continued presence of the 4boron in the plates in the pool over the complete life of the storage racks.
2.1.1 EVALUATION Considering the fact that the neutron absorber plates in these racks only partially cover the fuel assemblies (i.e., 3.94 inches out of 5.65 inches) the above described results compare favorably with the results of calculations made with other methods for fuel pool storage lattices with boron plates. By assuming new, unirradiated fuel with no burnable poison or control rods, these calculations yield the maximum neutron multiplication factor i. hat could be obtained throughout the life of the fuel assemblies. This includes the effect of the plutonium which is generated during the fuel cycle.
We find that DPC's on site tests for the presence of the boron is acceptable provided that if any boron plates are found to be missing in the blackness test, the NRC is promptly notified.
We find that DPC's proposed surveillance test is an acceptable way to verify the continued presence of the boron over the life of the racks.
We find that all factors that could affect the neutron multiplication factor in this pool have bean conservatively accounted for. We also find that if the fuel loading is limited to no more than 22.4 grams of uranium-235 per axial centimei.er of fuel assembly for the stainless steel clad fuel elements and to no more than 16.6 grams of uranium-235 per axial centimeter of fuel assembly for theZircaloy clad fuel elements, 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 (vorst case) calculated neutron multiplication factor in a spent fuel pool.
This 0.95 acceptance criterion is based on the uncertainties associated with the calculational methods and provides sufficient margins to preclude criticality in the fuel pool.
. 2.
1.2 CONCLUSION
We find that when any number of the fuel assemblies, which DPC described in these submittals which have no more than 22.4 grams of uranium 235 per axial centimeter of fuel assembly for the stain-less steel clad fuel elements and no more than 16.6 grams of uranium 235 per axial centimeter of fuel assembly for the Zircaloy clad fuel elements, are loaded into the proposed racks the neutron multiplica-tion factor 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, it is necessary, pending NRC review, to prohibit the use of these high density storage racks for fuel assenblies which have stainless steel clad fuel elements with more than 22.4 grams of uranium 235 per axial centimeter of fuel assembly and those which have Zircaloy clad fuel elements with more than 16.6 grams of uranium 235 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.
2.2 SPENT FUEL C00LIN3G_
The licensed thermal power for the Lacrosse Boiling Water Reactor is 165 MWT. DPC plans to refuel this plant annually. This will require the replacement o' about twenty four of the seventy two fuel assemblies in the core every year.
In its June 7,1978 submittal DPC assumed a 3 day (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) decay time for calculating the maximum heat generation rates in the fuel pool for the annual refueling and a 7 day (168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />) decay time for a full core offload. With these decay times, DPC used the method given in the NRC Standard Review Plan, NUREG-75/087, to calculate 0.93 x 106 BTU /hr as the maximum possible heat load for i
any annual refuelim The spent fuel pool cooling system consists of two pumps and one heat excha nger.
Each pump 1s designed to pump 260 gpm (1.3 x 105 pounds per hour).
Only one pump is operated at a time; so there is a redundant pump. At the design flow, the heat exchanger is designed to transfer 2.6 x 106 BTV/hr from 1200 fuel pool water to 900F Component Cooling water, which is flowing through the heat exchanger at a rate of 1.3 x 105 pounds per hour.
In its August 7,1978 submittal DPC included an analysis of the natural convection cooling of the spent fuel in the proposed two tier racks.
This showed that the maximuu increase in water temperature in the storage container with minimum natural circulation flow will be less than 240F and that the temperature of all of the fuel pool water will always be far below the saturation temperature.
4 2.2.1 EVALUATION Using the method given on pages 9.2.5-8 through 14 of the NRC Standard Review plan, with the uncertainty factor, K, equal to 0.1 for decay times longer than 107 seconds, we calculate that the maximum peak heat load during the 1990 refueling could be 1.1 x 106 BTU /hr and that the maximum peak heat load for a full core offload that essentially fills the pool could be 2.0 x 106 BTU /hr. This full core offload was assumed to take place one year after the 1987 refueling. We also find that the maximum incremental heat load that could be added by increasing the number of spent fuel assemblies u the pool from 133 to 440 will be 0.2 x 106 BTU /hr. This is the difference in peak heat loads for full core offloads that essentially fill the present and the modified pools.
Sirjce the spent fuel pool cooling system is designed to remove 2.6 x 100 BTU /hr from 1200F fuel pool water with one pump operating, the maximum fuel pool outlet water temperature will be less than 120 F 0
for the 2.0 x 106 BTV/hr which we calculate as the maximum for any full core offload.
In the unlikely event that both spent fuel pool cooling pumps fail just after a full core offload which generates 2.0 x 106 BTU /hr the fuel pool water would heat up at a rate of about 90F/hr. Thus assunting an initial outlet water temperature of 1200F there would be a time interval of ten hours before boiling would commence. During boiling a makeup rate of four gallons of water per minute (qom) would be required to keep the pcol full.
In its January 31, 1979 response to our request for additional information, DPL stated that either the overhead storage tank or the Demineralized Water Hose Station could be used for this purpose.
From this we find that ten hours would be sufficient time to pmvide a 4 gpm source of makeup water for the spent fuel pool.
2.
2.2 CONCLUSION
We find that the present cooling capacity for the spent fuel pool at the Lacrosse Boiling Water Reactor will be sufficient to handle the incremental heat load that will be added by the proposed modifications.
We also find that this increme-tal heat load will not alter the safety considerations of spent fuel cooling from that which 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.
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