ML20086B696
| ML20086B696 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/14/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20086B692 | List: |
| References | |
| NUDOCS 9507060053 | |
| Download: ML20086B696 (5) | |
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UNITED STATES 7,
. NUCLEAR REGULATORY COMMISSION y
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
RELATED TO AMENDMENT N0.108 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS. INC.
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WATERFORD STEAM ELECTRIC STATION. UNIT 3 DOCKET NO. 50-382
1.0 INTRODUCTION
By application dated January 27, 1995, Entergy Operations, Inc. (the licensee), submitted a request for changes to the Waterford Steam Electric Station, Unit 3 (WSES3), Technical Specifications (TSs).
The requested changes would change the TSs to increase the maximum enrichment for the spent fuel pool and containment temporary storage rack from 4.1 to 4.9 weight percent (w/o) U-235 when fuel assemblies contain fixed poisons.
The staff's evaluation of the criticality cspects of the proposed changes is provided below.
2.0 EVALUATION The WSES3 spent fuel rack is composed of cells containing stainless steel partitions which provide a fuel assembly storage area and two Boraflex insert areas per cell. The cells are oriented so that face adjacent assemblies are separated by a Boraflex insert area.
The panels are arranged in a rectangular configuration which maintains a 1-inch flux trap between panels and a 10.38-inch center-to-center spacing between cells.
The NRC acceptance criterion for conforming to General Design Criterion 62 for the prevention of criticality in fuel handling and storage is that the effective multiplication (k,,,) of the storage racks, fully loaded with fuel of the highest anticipated enrichment and fully moderated by unborated water, shall not exceed 0.95.
This value shall include all known uncertainties at the 95% probability, 95% confidence level (95/95).
The analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed with the SCALE 4 system of computer codes which includes the three-dimensional. multi-group Monte Carlo computer code, KENO Va.
Neutron cross sections were generated by the NITAWL and BONAMI codes. The CASMO-3 integral transport theory code was used to determine the reactivity effects of uncertainties or tolerance factors in the rack and fuel design parameters.
These codes are widely used for the analysis of fuel rack 9507060053 950614 DR ADOCK o 32 reactivity and have been benchmarked against results from numerous critical experiments. These experiments simulate the WSES3 fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment, absorber and assembly spacing.
The intercomparison between two independent methods of analysis (KEN 0 Va and CASMO-3) also provides an acceptable technique for validating calculational methods for nuclear criticality safety.
To minimize the statistical uncertainty of the KEN 0 Va reactivity calculations, a minimum of 600,000 neutron histories were accumulated in each calculation.
Experience has shown that this number of histories is quite sufficient to assure convergence of KENO Va reactivity calculations.
The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the WSES3 storage racks with a high degree of confidence.
WSES3 uses the ABB Combustion Engineering 16x16 fuel rod array assembly design which contains five large water holes for control rod insertion. A typical reload fuel assembly of this design contains two separate fuel rod initial enrichments, 4.1 w/o U-235 around the water holes and assembly corners and 4.5 w/o U-235 in the remaining locations.
Because of these higher fuel enrichments, reload batches usually contain burnable absorbers.
The base fuel assembly used in the WSES3 fuel pool criticality analysis contained U-235 enrichments of 4.1 and 4.5 w/o and eight absorber rods (shims), each with 0.016 grams of boron-10 (B-10) per inch.
These absorber rods replace fuel rods and are not susceptible to inadvertent removal. Therefore, the NRC considers them to be fixed absorbers and credit may be taken for their reactivity control effects.
Various other combinations of fuel enrichments and burnable absorber loadings which meet the NRC acceptance criterion of km no greater than 0.95 were also analyzed and the results are presented in Table 1 attached to this safety evaluation.
Uncertainties or tolerance factors in the rack and fuel design parameters were evaluated by either setting the parameter to its most adverse value or performing sensitivity studies with CASMO 3 to determine the reactivity impact of the tolerance factor.
The tolerance factors included uncertainties in U-235 enrichment, fuel pellet density, fuel pellet diameter, clad I.D., guide tube thickness and burnable absorber loading.
In addition, a method bias and uncertainty and an enrichment bias, determined from the benchmarking, were included, as well as the effects of Boraflex gaps as discussed below.
The staff has reviewed the assumptions made in determining these biases and uncertainties and concludes that they are appropriately conservative and meet the 95/95 probability / confidence requirement.
The WSES3 Boraflex surveillance program includes periodic blackness testing, using neutron attenuation, of selected Boraflex panels in the spent fuel racks at a maximum 4-year time interval. The panels selected are those expected to receive the highest cumulative gamma dose and therefore, result in the largest gap formation.
Periodic destructive testing on selected Boraflex panels will also be performed if engineering assessment determines it is i
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necessary.
Blackness testing of 697 Boraflex panels was November 1992 and 538 panels were found to have no gaps. performed in The largest gap size observed in the remaining panels was 3.6 inches, corresponding to an axial shrinkage of about 2.6%.
1 Since there are two Boraflex panels in the flux trap between each storage cell, it is highly unlikely that gaps would form at the same axial location in each panel.
If both panels in a flux trap form gaps, but the gaps occur at different axial locations, the reactivity impact will be much lower than if the gaps occur at the same elevation.
The assumption used in the WSES3 spent fuel pool criticality analysis was that all Boraflex panels contain 4.5-inch coplanar gaps at the top of each panel. This gap size bounds the WSES3 blackness measurements mentioned previously and is conservative relative to more realistic analyses based on these measurements, which would include the
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variations in gap size and location based on a probabilistic distribution. A licensee reactivity evaluation of placing coplanar gaps at various positions confirmed that the most reactive axial location for the placement of gaps was at the top of the panel.
In addition, a 4.1% shrinkage in the width of each Boraflex panel was assumed. Therefore, the NRC staff finds these Boraflex gap i
assumptions acceptable.
i Most abnormal storage conditions will not result in an increase in the k,n of the spent fuel racks. However, it is possible to postulate events, such as the misloading of an assembly with an enrichment and burnable absorber combination outside of the acceptable requirement, which could lead to an increase in reactivity. However, for such events credit may be taken for the presence of at least 1720 ppm of baron in the pool water required by plant procedures, since the staff does not require the assumption of two unlikely, j
independent, concurrent events to ensure protection against a criticality accident (double contingency principle). The reduction in k caused by the boron more than offsets the reactivity addition caused by cr,e,dible accidents.
Therefore, the staff criterion of k,u no greater than 0.95 for any postulated accident is met Containment temporary storage racks, which rely on fuel assembly spacing to maintain k,,f no greater than 0.95, are also provided in the WSES3 fuel storage facility. During normal storage conditions, assemblies in these racks are essentially neutronically decoupled due to the nominal assembly spacing of 18 inches, and k is calculated to be less than 0.90.
A dropped assembly n
accident was als,o evaluated assuming a minimum spacing of 1.762 inches between an assembly in the center rack location and the dropped assembly.
In this case, credit was taken for the minimum required boron concentration during refueling (double contingency principle) and the resulting k,n was also less than 0.90.
These values included uncertainties and biases at the 95/95 probability / confidence level, thereby meeting the NRC acceptance criterion.
The following changes to TS 5.3.1 have been proposed as a result of the requested enrichment increase. The staff finds these changes acceptable, for j
the reasons stated above.
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5 (1) The nominal total weight of uranium in a fuel rod has been increased from 1807 grams to 1830 grams.
(2) The maximum enrichment of reload fuel assemblies has been increased from 4.1 to 4.9 weight percent U-235 with the provision that the assemblies contain sufficient fixed poisons to ineet the final storage requirements described in TS 5.6, i.e., the storage rack k,, equivalent to less than or equal to 0.95 when flooded with unborated water,, which includes a conservative allowance for uncertainties.
CONCLUSION Based on the review described above, the staff finds the criticality aspects of the proposed enrichment increase to the WSES3 spent fuel pool storage racks are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.
Although the WSES3 TSs have been modified to specify the above-mentioned fuel as acceptable for storage in the spent fuel racks, evaluations of reload core designs (using any enrichment) will, of course, be performed on a cycle by cycle basis as part of the reload safety evaluation process.
Each reload design is evaluated to confirm that the cycle core design adheres to the limits that exist in the accident analyses and TSs to ensure that reactor operation is acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21. 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Reaister on June 13, 1995 (60 FR 31171). Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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Attachment:
Table 1 Principal Contributor:
L. Kopp Date: June 14, 1995
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TABLE 1 NO. OF SHIMS GM B-10/IN.
ENRICHMENT L,, --
4 0.012 4.11/3.71 0.94490 4
0.016 4.20/3.80 0.94551 4
0.02 4.24/3.84 0.94757 4
0.024 4.33/3.93 0.94904 8
0.012 4.37/3.97 0.94640 8
0.016 4.50/4.10 0.94956 8
0.02 4.55/4.15 0.94705
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0.024 4.61/4.21 0.94787 8
0.028 4.65/4.25 0.94754 12 0.02 4.85/4.45 0.94949 16 0.012 4.90/4.50 0.94598-i I
Attachment