ML17317B091
| ML17317B091 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/29/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17317B089 | List: |
| References | |
| SER-790329, NUDOCS 7904160388 | |
| Download: ML17317B091 (7) | |
Text
Ol SAFETY EVALUATION BY 'THE DIVISION OF OPERATING REACTORS CONCERNING INDIANA 5 MICHIGAN POWER COMPANY'S APPLICATION FOR AMENDMENT TO ITS OPERATING LICENSE TO INCREASE THE AUTHORIZED CAPACITY OF THE SPENT FUEL POOL AT THE DONALD C.
CuOK-NUCLEAR PLANT UNIT NOS.
1 AND 2 DOCKET, NOS.
50-315 AND 50-316
- 1. 0 INTRODUCTION By its letter dated November 22, '1978, as supplemented by letters dated January 22, and January 24,
( ISMPC) applied for license amendments to increase the authorized storage capacity for spent fuel at the D.
C.
Cook Nuclear Plant, Units 1
and, 2, from 500 to 2050 fuel assemblies.
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0 DISCUSS ION The spent fuel from both units at the D. C. Cook Nuclear Plant are stored in a common pool.
The proposed spent fuel storage racks are to be made up of double-walled stainless steel containers These will be about 14 feet-long and will have a square cross section with an inner dimension of 8.97 inches.
The nominal distance between the centers of the stored fuel assemblies, i.e., the lattice pitch, is 10.5 inches.
The effective dimension of the square fuel assembly, which was used in the criticality calculations, is 8.432 inches.
This results in an overall fuel region volume fraction of.645 in the nominal storage lattice cell.
Boral plates are to be press fit and seal welded in the cavities between the double stainless steel walls.
In its January
'4, 1979 submittal ISMPC states that stringent in-process inspection and process controls are imposed during manufacturing of the Boral to
" assure that the Boral plates contain at least 0.020 grams of the boron 7 904 160 9S S'
2-2.1 ten isotope per square centimeter of plate.
In this full array of storage containers there will be two Boral plates between adjacent fuel assemblies.
This makes thy minimum areal.
density of boron between fuel assemblies 2.41 x 102'oron ten atoms per square centimeter.
CRITICALITY ANALYSES As stated in ISMPC's November 22, 1978 submittal, the fuel pool criticality calculations are based on unirradiated fuel assemblies with no burriable poision and a fuel 'loading of 44.2 grams of uranium-235 per axial cen-timeter of fuel assembly.
The Exxon Nuclear Company performed the criticality analyses for ISMPC.
The basic method was to use the NITAWL and XSDRNPM computer programs with the XSDRN 123 group microscopic cross section data to generate multigroup cross sections for the KENO IV Monte Carlo program.
These programs were used to calculate the neutron multiplication factors (ke fs) for the fuel pool.
- Also, CCELL, which is a fuel pin cell
- program, was used to determine-the effects of U02 pellet density, moderator temperature, fuel temperature, and U-235 enrichment on the neutron mul tiplication factor.
Exxon calcula'ted the worst case keff by assuming a minimum water gap thickness between adjacent storage cells and a maximum water gap temperature of 100oC.
I8MPC's November 22, 1978 submittal states that the assumption of a minimized water gap thickness of 0.953 inches;.
rather than the nominal thickness of 1.118 inches, accounts for all of the toledo ances and the possible deformations due i;o the design and seismic loads.
Exxon's calculated value for'his worst case keff is 0.923.
Exxon checked the accuracy of this KENO IV method by calculating several sets of critical experiments.
Two of these had Boral plates in them.
- These were the experiment done by E.
B. Johnson 8
G.
E. Whitesides at the Oak Ridge National Laboratory and the one done by S.
R. Bierman et al at the Pacific Northwest Laboratories (PNL-2438).
Exxon's calculated values of the keff s of these two experiements agreed with the experimental values of 1.00 within the limits of the, statistical uncertainty in the Mowte Carlo program.
In its January 24, 1979 response to our request for additional information I8MPC stated the following:
1..There are steel structural members on t'e periphery of the rack modules that will provide ar additional two inches of water between any fuel assembly outside of the rack and those in the racks.
This two inches. of water, will prevent an increase in the keff.
- 2. 'eutron attenuation tests will be performed on the rack modules at the Cook site to verify the presence of the boron.
3.
Sufficient prototypical surveillance specimens will be provided which will permit inspection of both leaking and leak tight
.Boral cells.
2.1.1 EVALUATION The above results compare favorably with the results of calculations m'ade with other methods for similar fuel pool storage lattices.
By assuming new, unirradiated fuel with no burnable poison or control rods, these calculations yield the maximum neutron mul tiplication factor that could be obtained throughout the life of the fuel assemblies.
This includes the effect of the plutonium which is generated during the fuel. cycle.
The NRC acceptance criteria for the criticality aspects of high density fuel'storage racks is that 'the neutron multiplication factor in spent fuel pools shall be less than or equal to 0 95, including all uncertainties, under all conditions throughout the life of the racks.
This 0.95 acceptance criterion is based on the overall uncertainties associated with the calculational
- methods, and it is our judgment that this provides sufficient margin to preclude criticality in fuel pools.
Accordingly, there is a technical specification which limits the neutron multiplication factor, keff in spen't fuel pools to 0.95.
Since the neutron multiplication factor in spent fuel pools is not a quantity which is measured with good accuracy, the only available value is a calculated one.
To preclude any unreviewed increase,.or increased uncertainty, in the calculated value of the neutron multi-plication factor which could raise the actual keff in the fuel pool above 0.95 without being detected, a'imit on the maximum fuel loading is also required.
Accordingly, we find that the proposed high density storage racks will meet the NRC criteria when the fuel loading in the assemblies described in these submi ttals is limited to 44.2 grams or less of uranium-235 per axial centimeter of fuel assembly.
We find that ISMPC's proposed neutron attenuation test for the verifi-cation of the boron in the racks at the Cook site is satisfactory.
Ilowever, in this test, if any Boral plates are found to be missing the NRC shall be notified and a complete test on every storage location shall be performed.
We find that ISMPC's proposed boron surveillance program,is satis-factory for monitoring,'the condition of the Boral plates.
2.1. 2 CONCLUSION We find that when any number of the fuel assemblies, which I8MPC described in these submittals, which have no more than 44.2 grams of uranium-235 per axial centimeter of fuel assembly, are loaded into the proposed
- 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, it is necessary, pending NRC review, to prohibit the
'use of these high density storage racks for fuel assemblies that contain more than 44.2 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 COOL ING The licensed thermal power for D.
C.
Cook unit one is 3250 HWth and for unit 2 it is 3391 HWth.
A full core in each of the D. C.
Cook units consists of 193 fuel assemblies.
In its January 22, 1978 submittal ISHPC states that the evaluation of the'spent fuel pool cooling system assumes that 65 fuel assemblies will be discharged annually from Unit 1
and 88 fuel assemblies will be discharged every 18 months from Unit 2.
I&MPC assumed a cooling time of 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> after the reactor is shut down following 1080 full power days of operation to calculate the maximum heat loa'd by the method given in NRC Branch Technical Position APCSB 9-2.
This gave a maximum rin-pool heat generation rate of 54.1
, kw per fuel assembly.
The spent fuel pool cooling system as described in Chapter 9 of the
- FSAR, consists of two pumps and twonheat exchangers.
Each pump is designed to pump 2300 gpm (1.15 x 106 pounds per hour),
and each heat exchanger is designed to transfer 14.9 x 106 BTU/hr from 120oF fuel pool water to 95oF Component Cooling Water, which is flowing through the heat exchanger at a rate of 1.49 x 106 pounds per hour.
1 As shown in Figures 9.4-1 and 9.2-4 of the FSAR the Seismic Category I source of makeup water for the spent fuel pool is the Chemical and Volume Control System Mold-Up Tanks.
There are three pairs of these
- tanks, and each pair has a capacity of 128,000 gallons.
The Mold-Up
= Tank Recirculation
- Pump, which is rated at 500 gpm, can be used to pump water from these tanks to 'the spent'uel pool.
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 twenty eighth refueling, which would fill the pool, could be 22.6 x 10~ BTU/hr and that the maximum peak heat load for a full core offload that essentially fills the pool could be 41 x 10 BTU/hr.- This full core offload was assumed to take place one year after the twenty fifth refueling.
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 500 to 2050 is 6.9 x 106 BTU/hr.
'This is the difference in peak heat loads for full core offloads that essentially fil.l the present and the modified pools.
t We calculate that with both pumps operating, the spent fuel pool cooling system can maintain the fuel pool outlet water temperature below 120 F
for the normal refueling offload that fills the pool and below 130 F
for the full core offload that fills the poo'l.
In the highly unlikely event that both spent fuel pool cooling systems were to fail at the time when there was a peak heat load from a full core in the pool and the water was at its maximum temperature; we calculate that boiling could commence in about eleven hours.
We also calculate that after boiling commences the 'required Hater make up rate will be 85 gpm.
We find that eleven hours will be sufficient time to establish a 85 gpm make up rate.
2.2. 2 CONCLUS IGN We find that the present cooling capacity for the spent fuel pools at D.
C.
Cook Nuclear Plant Units 1
and 2 will be sufficient to handle the incremental heat load that will be added by the proposed modifi-cations.
We also find that this incremental 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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' 2.3 INSTALLATION OF RACKS AND FUEL HANDLING less than
'2. 3.1 EVALUATION I&MPC stated in their proposal that the existing Technical Specifications (T.S. 3.9.7) prohibit travel of loads in excess of 2500 lbs over the fuel assemblies while they are stored in the spent fuel pool.
Thus the movement of storage racks over spent fuel assemblies will be prohibited.
ISMPC plans: to change the spent fuel storage racks in the summer of 1979.
At that time there will be 193 spent fuel assemblies, in racks which have a capacity for 500 assemblies.
Thus the racks will be one half full when the new racks are installed.
Since over one:half of the pool will have vacant racks when they are
- changed, 18MPC should have no difficulty in keeping the racks away from the spent fuel that is in the pool.
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 o'f the proposed racks.
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