NRC 2004-0042, Response to Request for Additional Information Regarding Request for Review of Dynamic Effects Design Basis and Leak-Before-Break Analysis
| ML041250404 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/22/2004 |
| From: | Vanmiddlesworth G Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2004-0042 | |
| Download: ML041250404 (3) | |
Text
NMC Committed to Nuclear Excellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC April 22, 2004 NRC 2004-0042 10 CFR 50, Appendix A, GDC 4 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 License Nos. DPR-24 and DPR-27 Response to Request for Additional Information Regarding Request for Review of Dynamic Effects Design Basis and Leak-Before-Break Analysis
Reference:
- 1) Letter from NMC to NRC dated November 5, 2003 (NRC 2003-0105).
In Reference 1, Nuclear Management Company, LLC (NMC), submitted a request for review and approval of evaluations, which demonstrate that dynamic effects of reactor coolant system (RCS) primary loop pipe breaks need not be considered in the structural design basis of the Point Beach Nuclear Plant (PBNP), Units 1 and 2.
During a conference call between NMC personnel and NRC staff on April 5, 2004 the NRC requested additional information in support of their review of Reference 1. The NMC response to the staff's questions are provided in Enclosure I.
This letter contains no new commitments and no revision to existing commitments.
Gary D. Van Middlesworth Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC
Enclosure:
I Response to Request for Additional Information cc:
Project Manager, Point Beach Nuclear Plant, NRR, USNRC Regional Administrator, Region 1II, USNRC NRC Resident Inspector - Point Beach Nuclear Plant PSCW 6590 Nuclear Road
- Two Rivers, Wisconsin 54241 4
0 Telephone: 920.755.2321
ENCLOSURE I RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST FOR REVIEW OF DYNAMIC EFFECTS DESIGN BASIS AND LEAK-BEFORE-BREAK ANALYSIS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 The following information is provided in response to the Nuclear Regulatory Commission staffs request for additional information (RAI) regarding Reference 1.
The NRC staff's questions are restated below, with the NMC response following.
NRC Question 1:
How does Point Beach preclude primary water stress corrosion cracking (PWSCC) on reactor vessel (RV) nozzle safe-end welds which have Inconel 82/182 weld metal?
NMC Response:
With the exception of the PBNP Unit 2 Steam Generators' primary nozzle-to-safe-end welds, Inconel welds are not used in the PBNP primary loop piping. The Units 1 and 2 RV nozzle safe ends are constructed solely from a stainless steel weld butter.
NRC Question 2:
Please provide T(hot) and T(cold) temperatures for rated power and stretch power (10.4% uprate) operation.
NMC Response:
The analyses for current licensed rated power conditions (1540 MWt) include a range of T(avg) from 558.1 to 574 degrees Fahrenheit. The resulting T(hot) and T(cold) ranges are 588.1 to 603.5, and 528 to 544.5 degrees Fahrenheit, respectively. PBNP currently uses a T(avg) program of 547 to 570 degrees Fahrenheit (no load to full load), resulting in a T(hot) and T(cold) of approximately 597 and 542 degrees Fahrenheit, respectively.
The analyses for the 10.5 percent uprated power condition (1678 MWt) include a range of T(avg) from 558.6 to 573.4 degrees Fahrenheit. The resulting T(hot) and T(cold) ranges are 591.2 to 605.5, and 526 to 541.4 degrees Fahrenheit, respectively.
The LBB analysis used a T(hot) of 605.5 degrees Fahrenheit, and a T(cold) of 541.4 degrees Fahrenheit.
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NRC Question 3:
In the leak rate calculation for a 10 gpm leakage flaw for the case of two-phase choked flow, please justify conservatism of leak rate prediction with the use of higher T(hot) temperature.
NMC Response:
For the leak-before-break (LBB) analysis leak rate calculation for a 10 gpm leakage flaw, Westinghouse used an operating hot leg temperature of 605.5 degrees Fahrenheit (T hot) for the hot leg. Hot leg temperatures, versus cold leg temperatures, are used for the LBB analysis leak rate calculations for the 10 gpm leakage flaw since the piping material properties are more limiting at the higher temperatures. Using the lower full power hot leg temperature of 588.1 degrees Fahrenheit would have an insignificant impact on the 10 gpm leakage flaw sizes (in the case of two phase choked flow).
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