ML080040029
| ML080040029 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/15/2008 |
| From: | Mahesh Chawla NRC/NRR/ADRO/DORL/LPLIII-1 |
| To: | Wadley M Nuclear Management Co |
| Chawla M, NRR/DORL/LPL3-1, 301-415-8371 | |
| References | |
| TAC MD2567, TAC MD2568 | |
| Download: ML080040029 (8) | |
Text
January 15, 2008 Mr. Michael D. Wadley Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
CORRECTION LETTER FOR THE LICENSE AMENDMENTS TO INCORPORATE LARGE-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS USING ASTRUM FOR PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 (TAC NOS. MD2567 AND MD2568)
Dear Mr. Wadley:
On June 28, 2007 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML071800155), the Nuclear Regulatory Commission (NRC) issued Amendment No. 179 to Facility Operating License No. DPR-42 and Amendment No. 169 to Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2 respectively.
In a letter dated September 17, 2007 (ADAMS Accession No. ML072620102), the Nuclear Management Company (the licensee) provided comments to the above license amendments and the supporting safety evaluation (SE). These comments were generated as a result of certain administrative errors in the safety evaluation. These changes however do not affect any technical specifications pages issued with the amendments. The above comments and the required corrections to the SE were discussed and agreed upon between your representative and the NRC staff in a teleconference dated December 19, 2007.
A copy of our revised SE is enclosed. Please replace the originally issued SE with the enclosed SE. We regret any inconvenience this may have caused. If you have any further concerns, please contact me at 301-415-8371.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
Revised Safety Evaluation cc w/encls: See next page
ML072620102), the Nuclear Management Company (the licensee) provided comments to the above license amendments and the supporting safety evaluation (SE). These comments were generated as a result of certain administrative errors in the safety evaluation. These changes however do not affect any technical specifications pages issued with the amendments. The above comments and the required corrections to the SE were discussed and agreed upon between your representatives and the NRC staff in a teleconference dated December 20, 2007.
A copy of our revised SE is enclosed. Please replace the originally issued SE with the enclosed SE. We regret any inconvenience this may have caused. If you have any further concerns, please contact me at 301-415-8371.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosure:
Revised Safety Evaluation cc w/encls: See next page DISTRIBUTION PUBLIC LPL3-1 R/F RidsNrrDorlLpl3-1 RidsNrrPMMChawla RidsNrrLATHarris RidsOgcRp RidsAcrsAcnwMailCenter RidsNrrDirsltsb G. Hill, OIS RidsRgn3MailCenter RidsNrrDorlDpr FOrr, NRR ADAMS Accession No: ML080040029 OFFICE NRR/LPL3-1/PM NRR/LPL3-1/LA NRR/LPL3-1/LA NRR/SRXB/BC NRR/LPL3-1/(A)BC NAME MChawla BTully THarris GCranston CMunson DATE 1/14/08 1/11/08 1/11/08 1/14/08 1/15/08
Prairie Island Nuclear Generating Plant, Units 1 and 2 cc:
Jonathan Rogoff, Esquire Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Manager, Regulatory Affairs Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089 Manager - Environmental Protection Division Minnesota Attorney General=s Office 445 Minnesota St., Suite 900 St. Paul, MN 55101-2127 U.S. Nuclear Regulatory Commission Resident Inspector's Office 1719 Wakonade Drive East Welch, MN 55089-9642 Administrator Goodhue County Courthouse Box 408 Red Wing, MN 55066-0408 Commissioner Minnesota Department of Commerce 85 7th Place East, Suite 500 St. Paul, MN 55101-2198 Tribal Council Prairie Island Indian Community ATTN: Environmental Department 5636 Sturgeon Lake Road Welch, MN 55089 Nuclear Asset Manager Xcel Energy, Inc.
414 Nicollet Mall, R.S. 8 Minneapolis, MN 55401 Michael B. Sellman President and Chief Executive Officer Nuclear Management Company, LLC 700 First Street Hudson, MI 54016 Dennis L. Koehl Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Joel P. Sorenson Director, Site Operations Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089 July 2006
ENCLOSURE REVISED SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 179 TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 169 TO FACILITY OPERATION LICENSE NO. DPR-60 NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
On July 6, 2006 (Agencywide Documents Access and Management System (ADAMS),
Accession No. ML061880026), the Nuclear Management Company, LLC (NMC, or licensee) submitted a license amendment request (LAR) to apply the Nuclear Regulatory Commission (NRC)-approved Westinghouse best-estimate (BE) large-break loss-of coolant accident (LBLOCA) methodology, as described in WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005 (ADAMS Accession Nos. ML050910159/61(NP) and ML050910162/63(P)), to its Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP 1 and 2). The licensee also requested a license amendment to include the ASTRUM LBLOCA methodology in the core operating limits reports (COLRs) for both plants. In response to the NRC staffs request for additional information (RAI), the licensee supplemented its requests in letters dated September 15, 2006 (ADAMS Accession No. ML062610088), and December 26, 2006 (ADAMS Accession No. ML063600313). The supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on September 12, 2006 (71 FR 53718).
2.0 REGULATORY EVALUATION
The BE LBLOCA analyses were performed to demonstrate that the emergency core cooling system (ECCS) design would provide sufficient ECCS flow to transfer the heat from the reactor core following a LBLOCA and at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling would be prevented, and (2) the clad metal-water reaction would be limited to less than the amounts that would compromise cladding ductility and result in excessive hydrogen generation.
The NRC staff reviewed the analyses to assure that the PINGP 1 and 2 reflect suitable redundancy in components and features; and that suitable interconnections, leak detection, isolation, and containment capabilities are available such that the safety functions could be accomplished. The analyses assumed a single-failure for LBLOCAs and considered the availability of only onsite or offsite electric power (i.e., assuming offsite electric power is not available, with onsite electric power available; or assuming onsite electric power is not available, with offsite electric power available), consistent with the requirements of General Design Criterion 35 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR).
The NRC staff used the acceptance criteria for ECCS performance provided in Section 50.46 of 10 CFR Part 50 (10 CFR 50.46), in assessing the acceptability of the Westinghouse ASTRUM methodology for PINGP 1 and 2.
In its assessment of the acceptability of the methodology for PINGP 1 and 2, the NRC staff also reviewed the limitations and conditions stated in its safety evaluation (SE) supporting general approval of the Westinghouse ASTRUM methodology and the range of parameters described in the ASTRUM topical report.
3.0 TECHNICAL EVALUATION
The NRC staff reviewed the licensees demonstration evaluations of the ECCS performance analyses, conducted in accordance with the ASTRUM methodology, at the core power of 1683 MWt which includes measurement uncertainty. These specific analyses were performed to demonstrate the suitability of the ASTRUM methodology for application at PINGP 1 and 2. Upon approval, the specific analyses would be acceptable and specifically applicable to PINGP 1 and 2, when operated with the fuel(s) identified in Table 1 of this SE. For PINGP 1 and 2, the BE LBLOCA analyses were conducted assuming that the plants use cores containing Vantage+
(Zirlo - clad fuel) assemblies.
In its application, the licensee provided the results for the PINGP 1 and 2 BE LBLOCA analyses that were performed in accordance with the ASTRUM methodology with each unit operating at rated power of 1683 MWt which included measurement uncertainty. The licensees results for the calculated peak cladding temperatures (PCTs), the maximum cladding oxidation (local), and the maximum core-wide cladding oxidation are provided in the following table along with the acceptance criteria of 10 CFR 50.46(b).
TABLE 1 LARGE-BREAK LOCA ANALYSIS RESULTS - PINGP Units 1 and 2 Parameter Unit 1 ASTRUM Vantage+
Results Unit 2 ASTRUM Vantage+
Results 10 CFR 50.46 Limits Limiting Break /Location See Table 2 N/A Cladding Material Zirlo Zirlo (Cylindrical) Zircaloy or Zirlo Peak Clad Temperature 1594 oF 1546 oF 2200 oF (10 CFR 50.46(b)(1))
Maximum Local Oxidation 0.2%
0.5%
17.0% (10 CFR 50.46(b)(2))
Maximum Total Core-Wide Oxidation (All Fuel)
< 0.01%
< 0.01%
1.0% (10 CFR 50.46(b)(3))
TABLE 2 Unit Criteria Limiting Break Size 1
Peak Clad Temperature (PCT)
Split Break (1.0492
- Cold Leg Area) 1 Local Maximum Oxidation (LMO)
Split Break (0.7644
- Cold Leg Area) 1 Core Wide Oxidation (CWO)
(1) 2 Peak Clad Temperature (PCT)
Split Break (1.2395
- Cold Leg Area) 2 Local Maximum Oxidation (LMO)
Split Break (0.8037
- Cold Leg Area) 2 Core Wide Oxidation (CWO)
(1)
(1) Due to the low calculated peak clad temperatures, Core Wide Oxidation was estimated and no limiting break size applies In its analyses, the licensee also addressed the concern that the Vantage+ fuel cladding may have pre-existing oxidation that must be considered in its LOCA analyses. In the supplemental letter dated September 15, 2006, responding to an NRC staff RAI, the licensee indicated that it considered whether the fuel cladding has both pre-existing oxidation and oxidation resulting from the LOCA (pre-and post-LOCA oxidation both on the inside and the outside cladding surfaces).
In the supplemental letter dated December 26, 2006, the licensee noted that both PINGP units will be fueled only with Vantage+ fuel (with Zirlo cladding) and clarified that the optimized fuel assembly (OFA) fuel mentioned in its September 15, 2006, letter, referred to PINGP Optimized Fuel Assembly (OFA) fuel. The OFA is an assembly design (geometric, structural, etc.) that was customized for PINGP 1 and 2 to enhance the performance of the fuel (in this case, 14x14, 0.400" outside rod diameter Vantage+ fuel). The term Vantage+ fuel is also commonly used to designate fuel assemblies with Zirlo-clad fuel.
In the September 15, 2006, letter, the licensee indicated that the calculated pre-LOCA cladding oxidation was factored into the licensees BE LBLOCA analyses for the Zirlo clad fuel, consistent with the Westinghouse ASTRUM methodology. Both plants will be operated as governed by the Westinghouse-recommended program to limit operational duty within fuel duty limits, even during a fuel pins final cycle in the core, such that the sum of the calculated pre-and post-LOCA oxidation will be sufficiently small so that the total local oxidation will remain less than the 17 percent acceptance criterion of 10 CFR 50.46(b)(2) as noted above.
The concern with core-wide oxidation relates to the amount of hydrogen generated during a LOCA. Because hydrogen that may have been generated pre-LOCA (during normal operation) will be removed from the reactor coolant system throughout the operating cycle, the NRC staff noted that pre-existing oxidation does not contribute to the amount of hydrogen generated post-LOCA, and therefore it does not need to be addressed when determining whether the calculated total core-wide oxidation meets the 1.0 percent criterion of 10 CFR 50.46(b)(3).
As discussed previously, NMC had Westinghouse conduct BE LBLOCA analyses for PINGP 1 and 2 operating at about 102 percent of the current licensed power level of 1650 MWt using an NRC-approved Westinghouse methodology (ASTRUM). The NRC staff concluded that the results of these analyses demonstrated compliance with 10 CFR 50.46(b)(1) through (b)(3) for licensed power levels of up to 1650 MWt. Meeting these criteria provides reasonable assurance that, at the current licensed power level, the PINGP 1 and 2 cores will be amenable to cooling as required by 10 CFR 50.46(b)(4). The capability of PINGP 1 and 2 to satisfy the long-term cooling requirements of 10 CFR 50.46(b)(5) is unaffected by this amendment, and will be addressed, if needed, in a future NRC SE.
In its September 15, 2006, letter, the licensee stated, NMC and its vendor (Westinghouse) have ongoing processes that assure that the ranges and values of the input parameters for the Prairie Island Nuclear Generating Plant (PINGP) Large Break Loss of Coolant Analyses conservatively bound the ranges and values of the as-operated PINGP Unit 1 parameters. Likewise, NMC and its vendor (Westinghouse) have ongoing processes that assure that the ranges and values of the input parameters for the PINGP, Unit 2 LBLOCA analyses conservatively bound the ranges and values of the as-operated PINGP Unit 2 parameters. The NRC staff finds that this statement, along with the generic acceptance of ASTRUM, provides assurance that ASTRUM and its LBLOCA analyses apply to PINGP 1and 2, respectively, operated at their current licensed power levels.
4.0 PINGP 1 & 2 TECHNICAL SPECIFICATIONS COLR REFERENCES AND BASES CHANGES In support of the LAR, the licensee proposed to make changes to TS 5.6.5, Core Operating Limits Report, for each plant to reflect use of a new LBLOCA analysis methodology to perform LBLOCA analyses in support of PINGP 1 and 2 operation. The licensee also provided new TS Bases sections to describe the rationales for the new TS. The NRC staff reviewed these TS provisions, assessed them for consistency against NUREG-1431,Standard Technical Specifications, Westinghouse Plants, Revision 3, TS 5.6.5 (quoted on Page 14 of 59 in the LAR) as stated below, and found their content acceptable.
TS 5.6.5 COLR:
- 27. WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM).
This methodology was found to apply to all Westinghouse and Combustion Engineering pressurized-water reactor (PWR) designs in the NRC generic safety evaluation of the ASTRUM methodology. Therefore, ASTRUM is acceptable for application to PINGP 1 and 2, which are PWRs of Westinghouse design, and for inclusion in PINGP 1 and 2 TS for each plant. The above listed TS 5.6.5 Reference 27 was presented in the licensees submittal as a TS addition.
This reference does not include the WCAP-16009-P-A revision number (i.e., 0); nor does it include the date of approval for the methodology. The licensee will list the topical report, including the latest revision number, and date of approval in the COLR for each of the PINGP units consistent with guidance provided in NUREG-1431.
The NRC staff finds that ASTRUM is applicable to PINGP 1 and 2 and that the limitations and conditions of the NRCs SE approving ASTRUM were satisfied. Thus, the NRC staff concludes that the proposed addition of WCAP-16009-P-A to TS 5.6.5 is acceptable.
5.0
SUMMARY
Based on its review as discussed above, the NRC staff concluded that the Westinghouse ASTRUM methodology, as described in WCAP-16009-P-A, is acceptable for use for PINGP 1 and 2 in demonstrating compliance with the requirements of 10 CFR 50.46(b). The NRC staffs conclusion is based on the staffs verification that the PINGP design is among the designs for which ASTRUM application was approved.
The NRC staffs review of the acceptability of the ASTRUM methodology for PINGP 1 and 2 focused on assuring that the licensee and its vendor have a processes to assure that specific input parameters or bounding values and ranges (where appropriate) are used to conduct the PINGP 1 and 2 LBLOCA analyses, that the analyses will be conducted within the conditions and limitations of the NRC-approved Westinghouse ASTRUM methodology, and that the results will satisfy the requirements of 10 CFR 50.46(b) for PINGP 1 and 2.
The NRC staff finds the Westinghouse ASTRUM BE LBLOCA analysis methodology acceptable for application to PINGP 1 and 2 and for inclusion in the PINGP 1 and 2 TS 5.6.5 and COLRs.
As discussed above, the staff also finds the specific LBLOCA analyses acceptable that were performed with the ASTRUM methodology for PINGP 1 and 2 operating at powers up to their licensed power level of 1650 MWt.
6.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment. The State official had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. The NRC staff has determined that the amendment involves 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 a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (71 FR 53718). Accordingly, the amendment meets 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 amendment.
8.0 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: F. Orr, NRR Date: January 15, 2008