NRC 2007-0024, License Amendment Request 248, Technical Specification 5.5.8 Steam Generator Program Supplemental Response to Request for Additional Information
| ML070860501 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/26/2007 |
| From: | Koehl D Nuclear Management Co |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| NRC 2007-0024 | |
| Download: ML070860501 (5) | |
Text
Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC March 26,2007 NRC 2007-0024 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Unit 1 Docket 50-266 Renewed License No. DPR-24 License Amendment Reauest 248 Technical S~ecification 5.5.8, Steam Generator Proaram Supplemental Response to Reauest for Additional Information
References:
(1) NMC to NRC Letter Dated July 11, 2006, License Amendment Request 248, Technical Specification 5.5.8, Steam Generator Program, (ML062050338)
(2) NMC to NRC Letter Dated January 19,2007, Supplement 1 to License Amendment Request 248, Technical Specification 5.5.8, Steam Generator Program (ML070220084)
(3) NRC to NMC Letter Dated February 27,2007, Request for Additional Information, License Amendment Request, Steam Generator Program (ML07052461)
(4) NMC to NRC Letter Dated March 9,2007, License Amendment Request 248, Technical Specification 5.5.8, Steam Generator Program Response to Request for Additional lnformation (ML070680413)
By References (1) and (2) above Nuclear Management Company, LLC (NMC) submitted a proposed one-time amendment to the Technical Specifications (TSs) for Point Beach Nuclear Plant, Unit 1. The proposed changes would revise TS 5.5.8, "Steam Generator (SG) Program," to change the repair criteria for the portion of the tubes within the hot-leg region of the tubesheet for a single operating cycle following Refueling Outage 30. The proposed amendment defines a distance downward into the hot leg tubesheet, below which flaws may remain in service regardless of size. As a result, tube inspection within the hot leg region would be required only within 17 inches of the top of the tubesheet.
661 0 Nuclear Road Two Rivers, Wisconsin 54241 -951 6 Telephone: 920.755.2321
Document Control Desk Page 2 Reference (3) transmitted a request to NMC for additional information regarding the proposed amendment. Reference (4) responded to the NRC Request for Additional Information.
On March 21,2007, a telephone conference was held between members of the NRC staff, NMC and nuclear steam supply system (NSSS) vendor. During that meeting, the NRC staff requested verbal responses to additional questions posed during the staff's review of the Reference (4) Question 3 response. Specifically, the NRC staff requested that PBNP-specific information be provided to support the conclusion that the PBNP analysis results are comparable to the results obtained for another plant with Model 44F steam generators, given that two different analysis methodologies were employed. In addition, the NRC staff requested that there be confirmation that design conditions used in the analysis are similar and that the postulated steam line break event remains the limiting condition for the analysis.
The enclosure of this letter provides the requested information. In accordance with 10 CFR 50.91, a copy of this response is being provided to the designated Wisconsin Official.
This letter contains no new commitments and no revisions to existing commitments.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 26,2007.
Dennis L. Koehl Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC
ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 248 TS 5.5.8, STEAM GENERATOR PROGRAM Backaround By References (1) and (2), Nuclear Management Company, LLC (NMC) submitted a proposed amendment to the Technical Specifications (TSs) for the Point Beach Nuclear Plant (PBNP), Unit 1. The proposed changes would revise TS 5.5.8, "Steam Generator (SG) Program," to change the repair criteria for the portion of the tubes within the hot-leg region of the tubesheet for a single operating cycle following Refueling Outage 30. The proposed amendment defines a distance downward into the hot-leg tubesheet, below which flaws may remain in service regardless of size. As a result, tube inspection within the hot-leg region would be required only within 17 inches of the top of the tubesheet.
Reference (3) transmitted a request for additional information to NMC that provides the results of studies performed by Argonne National Laboratory (ANL) relating to the leakage behavior of steam generator tube-to-tubesheet joints under postulated severe accident conditions. As part of this work ANL benchmarked its finite element model of the tube-to-tubesheet joint against pullout and leakage tests carried out by Westinghouse on tube-to-collar joint specimens for the Callaway Plant.
Reference (4) provided the NMC response to the NRC request for additional information described via Reference (3).
On March 21,2007, a telephone conference was held between representatives of the NRC staff, NMC and the nuclear steam supply system (NSSS) vendor. During the telephone conference, the NRC staff requested that PBNP-specific information be provided to support the conclusion provided in response to Question 3 that the PBNP analysis results are comparable to the results obtained for another plant with Model 44F steam generators, given that two different analysis methodologies were employed. The following discussion confirms information verbally provided to the NRC staff during the telephone conference.
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NRC Reauest Confirm that the methodology used to determine the inspection length criteria for the plant used for comparison in Reference (4) applies directly to PBNP Unit 1. In addition, demonstrate that the design conditions used in the analysis are similar and that the postulated steam line break event remains the limiting condition for the analysis.
NMC Res~onse H* distances were determined specifically for PBNP Unit 1 using the Reference (4) plant analysis methodology and bounding PBNP Unit 1 operating conditions. The differences in the methodology used in generating H* distances in LTR-CDME-05-201 -P and the Reference (4) plant analysis were:
Tube pullout test results are used to establish an effective residual contact pressure due to the hydraulic expansion process.
The divider plate is assumed to be "non-functional."
Crevice pressure ratios of 0.3686 and 0.6977 are used for steam line break (SLB) and normal operating conditions.
The same assumptions can be directly applied to the analysis of the PBNP Unit 1 steam generator tube joints.
The loadings considered in the PBNP Unit 1 analysis included in LTR-CDME-05-201 -P, Revision 1 are based on an umbrella set of conditions. The temperatures and pressures for normal operating conditions are compared with the values used in the cited Reference (4) plant as follows:
Loadinq PBNP Unit 1 Reference (4) Plant Primary Pressure 2235 psig 2235 psig Secondary Pressure 730.7 psig 806.2 psig Primary Fluid Temperature (Thot) 590.2"F 604.1 OF Secondary Fluid Temperature 5257°F 521 -4°F The postulated SLB is the limiting faulted condition for both PBNP Unit 1 and the Reference (4) plant. The primary pressure was assumed to be 2560 psig and the pressure in the crevice was assumed to be 944 psig for both plants. The SLB primary and secondary side temperature drop was -355°F.
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Using the benefit of the tube pull data that was not included in LTR-CDME-05-201 -P, assuming a "non-functional" divider plate, the crevice pressure modifications, and a lower bound thermal expansion coefficient for the tubesheet, the nuclear steam supply system vendor (NSSS) determined that the H* distance for a postulated SLB for the PBNP Unit 1 steam generators occurs near the center of the tube bundle and is 11.81 inches, which is the same that was determined for the Reference (4) steam generators.
The limiting condition for establishing the H* distance for PBNP Unit 1 was determined to be the normal operating condition. The limiting H* distance that was determined for normal operating conditions occurs near the center of the tube bundle in the cold leg and was determined to be 12.09 inches. This distance is slightly greater than the H*
distance calculated for the Reference (4) plant, which was 11.60 inches. However, the limiting H* distance remains essentially the same as that reported in LTR-CDME-05-201 -P, which is 12.1 inches.
For both PBNP Unit 1 and the Reference (4) plant, the NSSS vendor concluded that the leak rate during a postulated SLB from indications below the depth of 17 inches from the top of the tubesheet would be bounded by twice the leak rate that is present during normal operating conditions.
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