ML18106A872
| ML18106A872 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/15/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18106A871 | List: |
| References | |
| NUDOCS 9809210178 | |
| Download: ML18106A872 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C.. 20555--0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO REQUEST FOR APPROVAL OF INCORPORATION OF WRB-2 CHF CORRELATION INTO LICENSING BASIS PUBLIC SERVICE ELECTRIC & GAS COMPANY SALEM NUCLEAR GENERATING.STATION. UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311
1.0 INTRODUCTION
By [[letter::05000311/LER-1998-008-01, :on 970814,failure to Test 21 & 22 AF 40 Valves in Closed Direction as Required by TS 4.0.5 Was Noted.Caused by Inadequate Design Mod Process.Motor Driven 21/22 AF 40 Valves Were Tested IAW Revised Procedure|letter dated May 18, 1998]], (Ref.1) thePublic Service Electric and Gas Company (PSE&G) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for approval to
- incorporate the Westinghouse Electric Company WRB-2 correlation into the licensing basis for the Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem). The Westinghouse WRB-2 correlation is a critical heat flux (CHF) correlation that is dependent on the fuel design features incorporated with the reload batch and is required for fuels with intermediate flow mixing (IFM) grids.
The currently licensed CHF correlation used by PSE&G for thermal-hydraulic analysis at Salem is WRB-1, which is not applicable to fuel with IFM grids.
2.0 BACKGROUND
The Margin Recovery Program (MRP) was a joint venture between PSE&G and Westinghouse to provide a more efficient core design and to also increase the fuel reliability, decrease reactor coolant system activity level, and improve operating margins by revising the safety analyses for Salem. The Fuel Upgrade Margin Recovery Program (FUMRP) - (MRP is a subset of FUMRP) -
consists of two parts: (1) the fuel upgrade portion consisting of utilizing VANTAGE+ fuel and ZIRLO cladding approved via Amendment 154/13S, dated August 22, 199S, and (2) the margin recovery program which forms the second part of FUMRP.
The analyses in support of the MRP license change submittal (Ref. 2) were originally performed to support the use of IFM grids with the WRB-2 CHF correlation (Ref. 3). However, because of vibration problems with the VANTAGE-SH (VSH) fuel at the time of licensing, PSE&G opted for not using fuel with IFM grids. This enabled PSE&G to continue to use the WRB-1 correlation.
As a consequence of this, the MRP submittal by the licensee was modified fo remove all descriptions of any analyses supporting fuel with IFM grids and the WRB-2 correlation.
3.0 EVALUATION Westinghouse resolved the flow-induced vibrational issues associated with the VANTAGE-SH fuel by developing an improved fuel design and with no loss in the "departure from nucleate boiling" ratio (DNBR) margin. Westinghouse refers to this fuel as the "Robust Fuel Assembly" (RFA). It maintains most of the VANTAGE+ and PERFORMANCE+ design features and also
-- *9ao92I 0178-- 98091°5 -
PDR ADOCK 05000272 p
.* incorporates a modified low-pressure drop (LPD) grid and a modified IFM grid as well as a thicker guide thimble. Concurrently, these design features address the flow-induced vibration (LPD and IFM grid modifications) and the incomplete rod insertion problems in high-bumup fuel (thicker guide tube thimble). The RFA has undergone extensive testing and design review, and is currently loaded in such operating reactors as Wolf Creek (Ref. 4). PSE&G intends to use RFA fuel in the next (upcoming) cycles; cycle 14 for Unit 1 and cycle 11 for Unit 2.
The NRC safety evaluation report (SER) for the Salem VSH fuel change (Ref. 6) was based on prior NRC approval of WCAP-10444-P-A and WCAP-10444-P-A Addendum 2. The SER for references 3 and S identified 13 conditions to be addressed by licensees using this fuel design.
Only those conditions that affected the Salem VSH without IFMs application were previously addressed in the SER. Those conditions that pertained to the VSH and the WRB-2 correlation were not considered until this submittal (Ref. 1). In support of this effqrt and for completeness, PSE&G included in 'his submittal those conditions pertaining to VANTAGE-SH fuel with IFM grids that were part of the MRP license change submittal (Ref. 2) for the staff to review. The NRC staff has reviewed these conditions and finds them to be acceptable.
The RFA design utilizes the modified 17x17 VSH LPD grid and the modified 17x17 VSH/IFM grid designs with the 0.482 inch outer diameter thimble and instrument tube. This design is identical to the one that the NRC recently approved and utilized at Wolf Creek (Refs. 4 and 9).
As documented in reference 4, extensive DNB testing was performed on the modified VSH/LPD and the VSH/IFM grids in order to demonstrate the applicability and the conservatism of the WRB-2 correlation to predicting the DNB performance of the modified new design. Analysis of the geometric effects of the thicker thimble tube was found to be bracketed by the geometric parameter range of the WRB-2 correlation. Therefore, the 95/95 correlation limit of 1.17 remains applicable to the RFA design with IFM grids intended for the Salem units (Ref. 10).
PSE&G has analyzed the WRB-2 range of applicability for the Salem RFA fuel. Some of the parameters can only be verified on a reload basis such as pressure, local mass velocity, and local quality. These parameters (variables) are functions of the core nominal conditions and thermal-hydraulic design parameters, and thus are reload dependent. Consequently, these parameters are verified for every reload design to ensure that CHF correlation is being* used within the applicable NRG-approved range.
Other correlation variables, such as fuel active length for RFA is the same as that of the VSH, and thus is bounded by the correlation heated length range. The grid spacing for the RFA is the same as that for the VS with IFMs, and therefore the grid spacing for the RFA remains applicable to the WRB-2 correlation. All the RFA design changes to the VSH/LPD grid and thimble/
instrument tube were performed within the guidelines of the Westinghouse Fuel Criteria Evaluation Process (FCEP) (Ref. 10). Analysis by the licensee shows that these changes remain bracketed by the WRB-2 correlation parameters for hydraulic and heated diameters.
The application of the WRB-2 critical heat flux correlation to the Salem RFA design within the context of the Westinghouse FCEP demonstrates that the WRB-2 correlation with a limit of 1. 17 can be conservatively applied to the RFA with IFMs. The incorporation of WRB-2 CHF correlation into the Salem licensing basis r(Jquires no changes to the current Technical Specifications because WRB-2 is already in the Salem Unit 1 Bases (Section 2.1. 1 ). The incorporation of WRB-2 into the Salem Unit 2 Technical Specifications Bases will become active with the issuance and implementation of the MRP amendment in the early part of 1999.
3.0 CONCLUSION
On the basis of the preceding review, the NRC staff finds the proposed request for the use of the Westinghouse WRB-2 correlation at Salem Units 1 and 2 to be acceptable, subject to the following commitments agreed upon by the licensee in its [[letter::LR-N980337, Forwards 90-day Rept for ISI Activities Conducted at Salem Generating Station,Unit 1,during Twelfth Refueling Outage. Rept Submitted IAW Plant TS Requirements & Article IWA-6210(c),6220(c) & 6230 of Section XI of ASME BPV Code|letter dated July 13, 1998]] (Ref. 11 ):
(1)
Transition core effects (such as large break LOCA transients effects, and transient core penalty, PCT penalty etc.,) and safety analysis impact assessment will be performed as part of the normal reload design and safety evaluation process under 10 CFR 50.59 and in accordance with 10 CFR 50.46.
(2)
Necessary changes to the Updated Final Safety Analysis Report (UFSAR) with respect to Chapters 4 and 15 as a result of incorporation of the WRB-2 correlation, will be made under 10 CFR 50.59 and upon approval of this submittal.
Principal Contributor. A. Attard Date: September 15, 1998
- REFERENCES
- 1.
Letter from E. C. Simpson, Public Service Electric and Gas Company, to the U.S. Nuclear Regulatory Commission,*Request for approval of WRB-2 CFH correlation into the Salem Licensing Basis,* May 18, 1998.
- 2.
Letter from L. Storz (PSE&G) to NRC, "Request for change to Technical Specifications Margin Recovery Program*, LCR 94-41, May 10, 1996
- 3.
WCAP-10444-P-A, Reference Core Report VANTAGE 5 Fuel Assembly, September 1985.
- 4.
Letter from N.J. Liparulo (Westinghou_se) to J. E. Lyons (NRC), "Transmittal of Response to NRC Request for Information on Wolf Creek Fuel Design Modifications,* NSD-NRC 5189, dated June 30, 1997.
- 5.
WCAP-10444-P-A, Addendum 2, VANTAGE 5 Fuel Assembly, April 1988.
- 6.
J. Stone (NRC) to S. Miltenberger (PSE&G), "Use of VANTAGE 5 Hybrid Fuel (TAC Nos.
71836/71837)", Amendment Nos. 96 and 72, May 1989.
- 7.
WCAP-8762-P-A, New Westinghouse Correlation WRB-1 for predicting critical heat flux in rod bundles with mixing vane grids, July 1984.
- 8.
Letter from L. Olshan (NRC) to L. Eliason (PSE&G), "Salem Nuclear Generating Station, Unit No. 1 (TAC No. M95383),* Amendment 201, November 26, 1997.
- 9.
Letter from N.J. Liparulo (Westinghouse) to R. J. Jones (NRC), "Transmittal of Presentation material from Westinghouse fuel design change meeting on April 15, 1996,"
NSD-NRC-96-4964, dated April 22, 1996.
- 10.
WCAP-12488-A, Westinghouse Fuel Criteria Evaluation Process, October 1994.
- 11.
Letter (LR-N980343) from E. C. Simpson, Public Service Electric and Gas Company, to the U.S. Nuclear Regulatory Commission, July 13, 1998.