ML15112A498
| ML15112A498 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/08/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML15112A497 | List: |
| References | |
| NUDOCS 9906110144 | |
| Download: ML15112A498 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION LICENSING TOPICAL REPORT DPC-NE-2005P DUKE ENERGY CORPORATION OCONEE NUCLEAR STATION, UNITS 1. 2. AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
By letter dated April 22, 1997 (Reference 1), as supplemented September 21, 1998, and May 13, 1999, (References 2 and 3 respectively), Duke Energy Corporation (DEC), licensee for the Oconee Nuclear Station, Units 1, 2, and 3, requested NRC staff review and approval of Appendix D, "Oconee Plant Specific Data, Mark-B1 1 Fuel, Application of BWU-Z CHF Correlation to Mark-B1 1 Mixing Vane Spacer Grid Fuel Design" (Reference 1), to DPC-NE 2005P, "Duke Power Company Thermal-Hydraulic Statistical Core Design Methodology" (Reference 4). The submittal contains analyses of the Mark-B1 1 fuel assemblies, analyzed using the BWU-Z critical heat flux correlation, and provides the required justifications for the specific uncertainties and distributions used in the application, and for the selected statepoints used to generate the statistical design limit.
The staff was assisted in this review by its consultant, Pacific Northwest National Laboratory (PNNL). The staff's evaluation includes the licensee's submittal (Reference 1), the licensee's response to the staff's request for additional information (RAI) dated September 21, 1998 (Reference 2), and the licensee's clarification dated May 13, 1999 (Reference 3). The staff has adopted the findings recommended in our consultant's attached technical evaluation report.
2.0 EVALUATION This review considered Appendix D "Oconee Plant Specific Data, Mark-B1 1 Fuel, Application of BWU-Z CHF Correlation to Mark-B1 1 Mixing Vane Spacer Grid Fuel Design" to DPC-NE 2005(P) "Duke Power Company Thermal-Hydraulic Statistical Core Design Methodology". The details of the evaluation are provided in the attachment.
This appendix contains plant-specific data for two-loop Babcock and Wilcox pressurized water reactors and specific limits for the Oconee Nuclear Station with Mark-B1 1 fuel using the BWU-Z Enclosure 9906110144 990608 PDR ADOCK 05000269 P
_jay
-2 form of the BWU critical heat flux (CHF) correlation. Approved methodologies, including the VIPRE-01 thermal-hydraulic computer code (EPRI NP-2511-CCM-A, Vol. 1-4) and the Oconee eight and nine channel models (DPC-NE-2003P-A), are used in this analysis.
The statistical core design (SCD) analysis includes: (1) statepoints that represent the range of conditions to which the statistical DNB analyses limit will be applied; (2) uncertainties that were selected to bound the values calculated for each parameter at Oconee and have not changed except for the rod power hot channel factor (Fq), core flow measurement, and departure from nuclear boiling ratio (DNBR) correlation; (3) the statistical DNBR design limit for each Jate point evaluated that was determined based on the 500 and 5000 case runs; and (4) the transition core model that determines the impact of the geometric and hydraulic difference between the resident Mark-B1i0 series fuel and the new Mark-B 1 design. The staff's concerns with respect to the statistical core design analysis in the areas of the applicable range of conditions, the uncertainties for core flow, the hot channel factor Fq and DNBR correlation, and the mixed core penalty were clarified in the licensee's response to the staff RAI (Reference 2).
Based on our review of Appendix D to Topical Report DPC-NE-2005P and the response to the staff's RAI (Reference 2), the staff finds Appendix D, "Application of BWU-Z CHF Correlation to Mark-B1 1 Mixing Vane Spacer Grid Fuel Design" to be acceptable. However, this approval is subject to the following conditions that were committed to by DEC in Reference 3:
(1) Omission of the parameter "Fq" from the SCD analysis of the Oconee plant with a new fuel design must be justified for each particular case. Acceptance of its omission in the case of Mark-B1 1 fuel does not constitute a general approval of its removal from the parameters to be considered in this methodology.
(2) The applicability of a CHF correlation to mixed core geometries is an issue that must be examined for each transition to new fuel to determine if the mixed core non-uniformities take the local hot channel conditions outside the range of applicability of CHF correlation.
(3) The SCD analysis shall be reviewed and revised as needed if the Mark-B1 1 CHF correlation range of applicability is changed.
3.0 CONCLUSION
Based on our review of Appendix D to the topical report DPC-NE-2005P and supplemental information supplied by DEC, the staff concludes that Appendix D, "Application of BWU-Z CHF Correlation to Mark-B1 1 Mixing Vane Spacer Grid Fuel Design" is acceptable. However, actions should be taken whenever a new fuel design is introduced, as follows:
- 1. Omission of the parameter "Fq" from the SCD analysis of the Oconee plant with a new fuel design must be justified for each particular case. Acceptance of its omission in the case of Mark-B1 1 fuel does not constitute a general approval of its removal from the parameters to be considered in this methodology.
-3
- 2. The applicability of a CHF correlation to mixed core geometries is an issue that must be examined for each transition to new fuel to determine if the mixed core non-uniformities take the local hot channel conditions outside the range of applicability of CHF correlation.
- 3. The SCD analysis shall be reviewed and revised as needed if the Mark-B1 1 CHF correlation range of applicability is changed.
Attachment:
Technical Report Principal Contributor: Tai Huang Date: June 8, 1999
REFERENCES
- 1. Letter from M. S. Tuckman to USNRC, Oconee Nuclear Station Docket Nos. 50-269, 50-270, and 50-287, Use of BWU-Z Critical Heat Flux Correlation for Mark-B1 1 Fuel, April 22, 1997 (Proprietary and Non-Proprietary Information Available).
- 2. Letter from M. S. Tuckman to USNRC, Response to NRC Request for Additional Information on Appendix D to Topical Report DPC-NE-2005-P, "Duke Power Company Thermal-Hydraulic Statistical Core Design Methodology," September 21, 1998 (Proprietary and Non-Proprietary Information Available).
- 3. Letter from M. S. Tuckman to USNRC, Oconee Nuclear Station Docket Nos. 50-269, 50-270, and 50-287, Duke Commitment to Conditions of SER and Clarification of Topical Report DPC-NE-2005 Revision Level, May 13, 1999.
- 4. DPC-NE-2005P-A, Duke Power Company Thermal-Hydraulic Statistical Core Design Methodology, February 1995.
- 5. BAW-10199P-A, The BWU Critical Heat Flux Correlations, August 1996.