ML19290D592
| ML19290D592 | |
| Person / Time | |
|---|---|
| Site: | 07002443, 07002913 |
| Issue date: | 11/09/1979 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Crow W NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 14666, NUDOCS 8002220248 | |
| Download: ML19290D592 (10) | |
Text
7s-avv3 TENNESSEE VALLEY AUTHORITY s
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CHATTANOOGA TENNESSEE 374ot g
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Mr. W. T. Crow, Section Leader k?
g Uranium Fuel Fabrication Section Fuel Processing & Fabrication Section Division of Fuel Cycle and Material Safety U.S. Nuclear Regulatory Co.amission Washington, DC 20555,
Dear Mr. Crow:
In the Matter of
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Docket Nos. 070-2443 Tennessee Valley Authority
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070-2913 The Tennessee Valley Authority (TVA) hereby requests an amendment to its Sequoyah Nuclear Plant unit 1 and 2 special nuclear material licenses SNM-1716, dated March 25, 1977, and SNM-1863, dated October 2, 1979. We request authorization to move fuel stored in the spent fuel pool racks into 10.375-inch center-to-center spacing high-density poison fuel racks that will be installed in the spent fuel pool. TVA will move the fuel after sufficient racks for safe fuel storage are installed, inspected, and adequate provisions have been made to ensure safe storage of the fuel during the installation of additional racks. and Sequoyah FSAR Amendment No. 62 which was submitted on October 31, 1979, describe the PAR high-density spent fuel storage racks and discuss criticality considerations. Enclosure 2 is a revised drawing of the cask-loading area submitted as Figure 1 in the original applications of SNM-1716 and SNM-1863. The new Figure I shows the temporary barrier removed. The cask-loading area is controlled by unit 1 Plant Physical Security Progras and this barrier is no longer needed.
TVA requests the amendment on or about January 21, 1980.
In accordance with 10 CFR 170.11(a)3, an amendment fee is not required.
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An Equal opportunity Employer d
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, Mr. W. T. Crow November 9, 1979 If you have any questions regarding this amendment, please get in touch
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with P. J. Hammons of my staff at FTS 854-2584.
Very truly yours, TENNESSEE VALLEY AUTHORITY k
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N L. M. Mills, Manager Nuclear Regulation and Safety Enclosures
ENCLOSURE 1 1.2.1 Storage Area - To describe the replacement racks add the following paragraphs to the end of this section.
There is storage space for 1386 fuel assemblies in the spent fuel storage racks.
The fuel shall be stored in an array such that K gg will be less than 0.95 even if e
immersed in unborated water including all mehcanical tolerances.
The high density poison PWR Spent Fuel Racks are of
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stainleas steel welded construction.
The racks are designed to store fuel safely on 10.375 inch center to center spacing.
They consist of four basic components:
(1) top grid casting (2) bottom grid casting.
(3) neutron absorber canister (poison cans)
(4) adjustable foot assembly The top and bottom grids type CF-35 SS locate and support the poison cans and fuel elements.
The grid castings have pockets cast in every cavity opening into which the inner tube of the neutron absorber canistors are welded.
The neutron absorber canister, hereafter called poison cans, consists of two concentric type 304 SS tubes with the neutron absorber plates located in the annular gap.
The neutron absorber plates "Boral" Tu consist of Boron carbide in an aluminum composite matrix which is clad with type 1100 aluminum. sheets.
The outer tube of the poison can is folded into the inner tube at the ends and totally seal welded to isolate the neutron poison from the pool water.
The inner tubes of the poison cans act as structural elements between the top and bottom grids since they are welded onto them.
In the unlikely event of swelling, the poison can will not impact the fuel rods.
Large leveling screws type 17-4 PH H1100 SS are located at the bottom grid corners to adjust for variations in pool floor level.
The adjustable feet of the spent fuel rack nest into cups which are welded onto a seismic support grid work.
This grid rests on but is not attached to existing 1-1/2-inch floor plates.
The floor plates are anchored to the pool floor and were used on the previous rack design.
The grid is mainly comprised of 3 x 5 and 5 x 5 bar stock, and is compression supported to the base of all the pool walls.
Snubber restraints are used on the south and west wall to eliminate thermal loads, and fixed wedge restraints are used on the north and east wall.
No seismic restraints bear against the 18 inch thick portion of the north wall adjacent to the fuel cask loading pit.
Grid diagonal members are used to transfer this load to the middle north wall restraint.
The north wall restraints are located under the existing sparger pipe.
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The racks are horizontally restrained by the grid but are rotationally unrestrained.
I!ence the rack feet vertically bear in cups on the grid but do not take any uplift forces.
The vertical impact forces generated by this rocking motion of the rack under seismic excitation are elastically, absorbed by the grid structure and pool floor.
All the seismic supports and associated grid work are constructed from type 304 stainless steel.
Figure 1 shows the design of a (7x8) module spent fuel rack assembly.
Figure 2 shows the sequoyah spent fuel rack layout.
Figure 3 shows the Sequoyah spent rack seismic support layout.
2.2 Nuclear Criticality Safety - To describe the replacement racks add the following paragraphs af ter the third paragraph:
The spent fuel storage racks provide a nominal spacing of 10. 37 5 in ches.
The nominal gap between racks arrays is 1.0 inch N-S and 1.5 inches E-W.
The spent fuel storage racks are individual vertical cells welded together in either 7 x 8 or 7 x 9 arrays forming modules that are supported to the pool walls via a floor gridwork.
Criticality of fuel assemblies outside of the reactor is precluded by adequate design of fuel transfer and fuel storage facilities and by administrative control procedures.
This section identifies those criteria important to criticality safety analyses.
New fuel stored in 10.375 inch center-to-center racks in the spent fuel storage pit may be stored wet or in a dry condition.
The racks use Boral with 0.232 g/cm2 B-10 loading.
For the flooded condition with unborated water assuming new fuel in the 10.375 inch racks, the K,gf does not exceed 0.95 with fuel of the highest anticipated enrichment (3.5 wt % U-235) in place.
The calculations were performed by using the Monte Carlo transport code, KENO-IV.
In the analysis for the storage facilities, the fuel assemblies are assumed to be in their most reactive condition, namely fresh or undepleted and with no control rods or removable neutron absorbers present.
Credit is taken for the inherent neutron-absorbing ef fect of materials of construction of the racks.
Assemblies can not be closer together than the design separation provided by the storage facility.
The mechanical integrity of the fuel assembly is assumed.
The infinite array reactivity of the new rack design, with fuel enrichment of 3.5 weight percent of U-23S, was calculated by both KENO-IV Honte Caro transport model and by the PDC diffusion theory model.
The latter was further used in the calculations of various biases and
sensitivities related to the tolerances and variances of the rack design, cuch as pitch, position, geometry, dimension, material, enrichment, temperature, voids, and boron density.
The enrichment variation is restricted by the Fuel Specifications:
the boron density has a minimum B-10 loading of 0.0232 g/cm2, with a minimum boral core thickness of 80 mils.
The biases due to other uncertainties, such geometrical, positional -
dimensional, and material variations, have all been taken into accourr in the process of assessing the final reactivity for the racks.
The biases due to temperature variation from 680 F to 2120 F, and due to voiding in the racks, are all found to be negative.
The total bias due to these unc ertainties, including the worst geometry tolerances, is found to be worth 1.78% in Ak.
The criticality calculatiens described are based on the assumption that the core of the Boral slabs contains a homogeneous mixture of fine B C and aluminum powder.
4 The computer cedes used for these calculations contain no provision tc take into account directly the self-shielding effec: due to the random distribution of B C 4
grains of finite sizes in the mixture.
Based on a homogeneous mixture, the neutron attenuation factor for a Boral slab of 80 mil thickness with 0.0232 g/cm2 B-10 loading was calcui'.ated to be 0.969.
The vendor has proposed the value 0.963 as the acceptance ceiterion for their measured neutron attenuation factor for the above Boral slabs, which takes into account the heterogeneous mixture.
Based on analysis the difference between an attenuation factor of 0.969 and 0.963 amounts to about 0.17% Ak for Sequoyah Spent Fuel Pool.
With the inclusion of this bias, the final average k value including all the uncertainties becomes 0.90gg0 with a 95% confidence interval ranging from 0.932 to 0.948.
The design basis for wet fuel storage criticality analysis is that, considering possible variations, there is a 95 percent confidence level that effective multiplication factor (K of the fuel storage array will be less than 0.95 pe,f g )
r ANSI Standard N18.2-1973.
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