ML16138A789
| ML16138A789 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/03/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A788 | List: |
| References | |
| NUDOCS 9505100199 | |
| Download: ML16138A789 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 209 TO FACILITY OPERATING LICENSE DPR-38 AMENDMENT NO.
209 TO FACILITY OPERATING LICENSE DPR-47 AND AMENDMENT NO.
206 TO FACILITY OPERATING LICENSE DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1. 2. AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
By letter dated November 22, 1994, as supplemented by letters dated January 30, March 2, March 13, and May 2, 1995, Duke Power Company, et al.
(the licensee), submitted a request for changes to the Oconee Nuclear Station, Units 1, 2, and 3, Technical Specifications (TS). The changes will allow an increased limit for fuel enrichment. The May 2, 1995, letter did not change the scope of the November 22, 1994, application and the initial proposed no significant hazards consideration determination.
The Oconee Nuclear Station has two separate spent fuel storage pools. One pool is shared by the Unit 1 and 2 reactors and has a current maximum nominal enrichment of 4.3 weight percent (w/o) U-235. The Unit 3 pool has a current maximum nominal enrichment of 4.0 w/o U-235. The proposed changes would allow for the storage of fuel with an enrichment not to exceed a nominal 5.00 w/o U 235 in the spent fuel storage racks. As-built manufacturing variations of up to 0.05 w/o U-235 are accounted for in the reactivity analyses.
The increased fuel enrichment limits for fuel storage in the Oconee spent fuel pools were evaluated against the requirements of General Design Criteria 62 of 10 CFR Part 50, Appendix A. The staff's evaluation of the criticality aspects of the proposed changes follows.
2.0 EVALUATION Each of the two independent spent fuel pools is designed for storage of either fresh or irradiated fuel.
The stainless steel cells for the Unit 1 and Unit 2 storage racks are spaced on a 10.65-inch center-to-center distance and have a storage capacity of 1312 fuel assemblies. The Unit 3 racks contain 825 available storage cells with a 10.60-inch center-to-center spacing.
9505100199 950503 PDR ADOCK 05000269 P
-2 The analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed with the SCALE system of computer codes using the three-dimensional multi-group Monte Carlo computer code, KENO Va. Neutron cross sections were generated by the NITAWL and BNAMI codes using the 27 Group NDF4 library. Since the KENO Va code package does not have depletion capability, burnup analyses were performed with the CASMO-3/SIMULATE-3 methodology. CASMO-3 is an integral transport theory code and SIMULATE-3 is a nodal diffusion theory code. These codes are widely used for the analysis of fuel rack reactivity and have been benchmarked against results fromnumerous critical experiments. These experiments simulate the Oconee fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment and assembly spacing. The intercomparison between two independent methods of analysis (KENO Va and CASMO-3/SIMULATE-3) also provides an acceptable technique for validating calculational methods for nuclear criticality safety. To minimize the statistical uncertainty of the KENO Va reactivity calculations, a nominal 90,000 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 has reviewed the licensee's analysis as described above and concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the Oconee storage racks with a high degree of confidence.
Duke Power has indicated a desire to incorporate two new fuel assembly designs, designated MkB1OT and MkB11, at Oconee in the near future. These new fuel assembly designs are more reactive than the designs previously analyzed and would violate the spent fuel storage design criteria under the current TS.
Therefore, a reanalysis was performed by DPC to allow for an increase in the maximum allowable initial enrichment of the stored fuel.
The results indicate that fuel with nominal enrichments up to 3.73 w/o U-235 for MkB11 fuel and 3.93 w/o U-235 for all other Oconee fuel can be stored in every cell of the Unit 1 and 2 spent fuel storage racks. For the Unit 3 storage racks, MkB11 fuel having an initial enrichment of up to 3.66 w/o U-235 or all other Oconee fuel having a maximum enrichment of 3.86 w/o U-235 can be stored in every cell.
Since the reanalysis was performed using the methodology described above, the results stated are acceptable.
To enable the storage of depleted fuel assemblies initially enriched to greater than the 3.73 w/o and 3.93 w/o limits stated above, the concept of burnup credit reactivity equivalencing was used. This is predicated upon the reactivity decrease associated with fuel depletion and has been previously accepted by the staff for spent fuel storage analysis. For burnup credit, a series of reactivity calculations are performed to generate a set of initial enrichment-fuel assembly discharge burnup ordered pairs which all yield an equivalent keff less than 0.95 when stored in the spent fuel storage racks.
This is shown in Table 3.8-1 in which a fresh 3.73 w/o MkB11 enriched fuel assembly yields the same rack reactivity as an initially enriched 5.00 w/o MkB11 assembly depleted to 7.95 GWD/MTU. A similar result is shown for other Oconee fuel assemblies where a fresh 3.93 w/o enriched assembly yields the same storage rack reactivity as an initially enriched 5.00 w/o assembly depleted to 6.03 GWD/MTU. The curves shown in the Table include biases due to methodology, Boraflex shrinkage, and boron self-shielding, a 95/95 methodology
-3 uncertainty, and a mechanical uncertainty due to manufacturing tolerances. In addition, a bias and uncertainty associated with fuel burnup was also included. The staff has reviewed the assumptions made in determining these biases and uncertainties and concludes that they are appropriately conservative and the burnup limits are acceptable.
New or irradiated assemblies with initial enrichments up to 5.00 w/o U-235 which do not meet the requirements for unrestricted storage must be placed in a restricted loading pattern. Reactivity analyses for these assemblies, stored in every other row of the spent fuel pool, were performed using the previously discussed methods. Acceptable fuel assemblies which qualify for storage in the alternating rows between adequately depleted assemblies are shown in Table 3.8-2 and are referred to as filler assemblies. These filler assemblies were also determined from minimum burnup versus initial enrichment calculations as described above. These special configurations have been analyzed using the acceptable reactivity methods described previously and meet the NRC acceptance criterion of k no greater than 0.95, including all appropriate uncertainties at the Vg/ 95 probability/confidence level.
The results are, therefore, acceptable.
Similar analyses were performed for the Unit 3 spent fuel pool and the resulting minimum qualifying burnups are shown in Tables 3.8-3 and 3.8-4.
Since fuel will be stored in the pools according to two different loading configurations to accommodate both unrestricted and restricted storage, the boundary conditions between these configurations were analyzed to determine the effects of neutronic coupling. The results show that, in order to satisfy the keff criterion, a row of restricted assemblies must not be directly adjacent to a row of unrestricted fuel.
This additional restriction has been incorporated into the proposed Oconee fuel storage TS.
A statement is included in Tables 3.8-1 through 3.8-4 to allow for specific criticality analyses for fuel which differs from those designs used to determine the requirements for storage defined in these tables. This would allow storage of fuel from another facility or storage of individual fuel rods as a result of fuel assembly reconstitution. A similar specification was previously approved for the McGuire Nuclear Station and has been implemented to accommodate storage of Oconee spent fuel shipped to McGuire for storage.
These analyses would require using the NRC approved methodology described above to ensure that keff does not exceed 0.95 at a 95/95 probability /
confidence level and fuel storage would still be limited to the configurations defined in TS 3.8-16. At the staff's request, the Bases was revised to include additional discussion which reflects the intended use of this provision. The staff finds this proposed specification acceptable.
Most abnormal storage conditions will not result in an increase in the keff of the spent fuel racks. However, it is possible to postulate events, such as the misloading of an assembly with a burnup and enrichment 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 boron in the pool water required by TS 3.8.15 when fuel is stored in the spent fuel pool since the staff does not require the assumption of two unlikely, independent,
-4 concurrent events to ensure protection against a criticality accident (Double Contingency Principle). The reduction in keff caused by the boron more than offsets the reactivity addition caused by credible accidents. Therefore, the staff criterion of keff no greater than 0.95 for any postulated accident is met.
The following TS changes have been proposed as a result of the requested enrichment increase. The staff finds these changes, and the associated Bases changes, acceptable.
(1) TS 3.8.15 is being added to establish limits for the required spent fuel pool boron concentration in the Core Operating Limits Report (COLR). The relocation of the minimum spent fuel pool boron concentration to the COLR has previously been approved by the NRC in other licensing actions.
(2) TS 3.8.16 is being added to specify the new fuel storage requirements given in Tables 3.8-1 through 3.8-4 and Figures 3.8-1 and 3.8-2 based on the reactivity analyses evaluated above.
(3) TS 3.8.17 is being added to state the required actions if the limiting conditions stated in TS 3.8.15 or 3.8.16 are not met.
(4) The Bases is being modified to allow for specific criticality analyses for special situations without requiring additional TS changes, as described above. In addition, the Bases are being changed to address new Loss of Coolant Accident (LOCA) limits for the new MkB10T fuel assemblies.
(5) TS 5.3.1 is being revised to accommodate changes in the fuel assembly design evaluated above. The proposed changes are consistent with the standard TS.
(6) TS 5.4.1.1 is being revised to remove references to maximum fuel enrichments since this is now specified in TS Tables 3.8-1 through 3.8-4.
In addition, TS 5.4.1.1 and TS 5.4.1.2 are being combined into TS 5.4.1.
(7) TS 5.4.2.1 is being modified to delete extraneous information.
(8) TS 5.4.3, which specifies the spent fuel pool boron concentration, is being relocated to TS 3.8.15.
(9) TS 6.9.1 is being changed to include the spent fuel pool boron concentration in the list of COLR parameters.
Based on the review described above, the staff finds the criticality aspects of the proposed enrichment increase to the Oconee 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 Oconee TS have been modified to specify the above-mentioned fuel as acceptable for storage in the fresh or spent fuel racks, evaluations of reload core designs (using any enrichment) will be performed on a cycle-by-
Sg
-5 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 TS to ensure that reactor operation is acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued proposed findings that the amendments involve no significant hazards consideration, and there has been no public comment on such findings (60 FR 8746 dated February 15, 1995; 60 FR 16185 dated March 29, 1995).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Attachments:
TS Tables Principal Contributor: L. Kopp Date:
May 3, 1995
able 3.8-1 Minimum Qualifying Burnue Versus Initial Enrichment for Unrestricted Storage in the Unit 1 and 2 Soent Fuel Pool MkB1 OT and Earlier MkB11 Fuel Assembly Designs Fuel Assembly Design Initial Nominal Enrichment Assembly Burnup initial Nominal Enrichment Assembly Burnup (Weight% U-235)
(GWD/MTL (Weight% U-235)
(GWD/MTU 3.93 (or less) 0 3.73 (or less) 0 4.00 0.43 4.00 1.83 4.50 3.30 4.50 4.80 5.00 6.03 5.00 7.95 10.00 MkB1 OT and Earlier Fuel Assembly Designs eg 8.00 ACCEPTABLE 6.0For Unrestricted Storage 6.00 c
4.00 25 --
Mk31 1 E 0 Fuel Assembly Design n2.00 --
UNACCEPTABLE For Unrestricted Storage 0.00 3.50 3.75 4.00 4.25 4.50 4.75 5.00 Initial Nominal Enrichment (Weight/a U-235)
Fuel which differs from those designs used to determine the requirements of Table 3.8-1 may be qualified for Unrestricted storage by means of an analysis using NRC approved methodology to assure that k. is less than or equal to 0.95.
Likewise, previously unanalyzed fuel up to 5.0 weight/
U-235 may be qualified for Restricted storage by means of an analysis using NRC approved methodology to assure that k., is less than or equal to 0.95.
3.8-6
4, 0
Table 3.8-2 Minimum Qualifying Burnue Versus Initial Enrichment for Filler Assemblies in the Unit 1 and 2 Spent Fuel Pool All Fuel Assembly Designs Initial Nominal Enrichment Assembly Burnup (Weight% U-235)
(GWD/MTUI 2.72 (or less) 0 3.00 3.25 3.50 8.22 4.00 13.13 4.50 18.10 5.00 22.69 30.00 ACCEPTABLE 3:20.00 For Use As Filler Assembly CL cc 10.00 UNACCEPTABLE For Use As Filler Assembly 0.00 2.00 2.50 3.00 3.50 4.00 4.50 5.00 Initial Nominal Enrichment (Weight%!
0 U-235)
Fuel which differs from those designs used to determine the requirements of Table 3.8-2 may be qualified for use as a Filler Assembly by means of an analysis using NRC approved methodology to assure that kg is less tian or equal to 0.95.
3.8-7
Table 3.8-3 Minimum Qualifying Burnue Versus initial Enrichment for Unrestricted Storage in the Unit 3 Soent Fuel Pool MkB10T and Earlier MkB11 Fuel Assembly Designs Fuel Assembly Design Initial Nominal Enrichment Assembly Burnup Initial Nominal Enrichment Assembly Burnup (Weight% U-235)'
(CWD/MLU)
(Weight% U-235)
LGWD/MIU1 3.86 (or less) 0 3.66 (or less) 0 4.00 0.91 4.00 2.31 4.50 3.73 4.50 5.34 5.00 6.60 5.00 8.49 MkB 1OT and Earlier 10.00 Fuel Assembly Designs 8.00 ACCEPTABLE For Unrestricted Storage 6.00 4.00 MkBI 1 Fuel Assembly Desiqn E
Cn2.00 UNACCEPTABLE For Unrestricted Storage 0.00 I
3.50 3.75 4.00 4.25 4.50.
4.75 5.00 Initial Nominal Enrichment (Weight% U-235)
Fuel which differs from those designs used to determine the requirements of Table 3.8-3 may be qualified for Unrestricted storage by means of an analysis using NRC approved methodology to assure that k,, is less than or equal to 0.95.
Likewise, previously unanalyzed fuel up to 5.0 weight/o U-235 may be qualified for Restricted storage by means of an analysis using NRC approved methodology to assure that km, is less than or equal to 0.95.
3.8-8
7abe 3.8-4*
Minimum Qualifying Burnue Versus Initial Enrichment for Filler Assemblies in tne Unit 3 Spent Fuel Pool All Fuel Assembly Designs Initial Nominal Enrichment Assembly Burnup (Weight% U-235)
(GWD/MTU) 2.61 (or less) 0 3.00 4.49 3.50 9.62 4.00 14.68 4.50 19.96 5.00 24.37 30.00 ACCEPTABLE 3=20.00 2.0For Use As Filler Assembly 10.00 UNACCEPTABLE E
<W
-For Use As Filler Assembly Ct) 0.00 2.00 2.50 3.00 3.50 4.00 4.50 5.00 Initial Nominal Enrichment (Weight/ U-235)
Fuel which differs from those designs used to determine the requirements of Table 3.8-4 may be qualified for use as a Filler Assembly by means of an analysis using NRC approved methodology to assure that k. is less than or equal to 0.95.
3.8-9