ML070330455
| ML070330455 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/20/2007 |
| From: | Mahesh Chawla NRC/NRR/ADRO/DORL/LPLIII-1 |
| To: | Thomas J. Palmisano Nuclear Management Co |
| M. Chawla LPL3-1 | |
| Shared Package | |
| ML070330441 | List: |
| References | |
| TAC MD0209, TAC MD0210 | |
| Download: ML070330455 (20) | |
Text
March 20, 2007 Mr. Thomas J. Palmisano Site Vice President Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, MN 55089
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -
ISSUANCE OF AMENDMENTS RE: (TAC NOS. MD0209 AND MD0210)
Dear Mr. Palmisano:
The Commission has issued the enclosed Amendment No. 177 to Facility Operating License No. DPR-42 and Amendment No. 167 to Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively.
By letters dated February 16, 2006, supplemented by letters dated July 21, and December 27, 2006, Nuclear Management Company (the licensee), submitted a license amendment request regarding Prairie Island Nuclear Generating Plant Units 1 and 2 steam generator (SG) tube integrity technical specifications (TS).
The amendments consist of changes to the TSs related to SG tube integrity. The amendments are modeled after the U.S. Nuclear Regulatory Commission approved TS Task Force (TSTF)
Standard TS Change Traveler, TSTF-449, Steam Generator Tube Integrity, Revision 4 (ML0510902003).
A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
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
Enclosures:
- 1. Amendment No. 177 to DPR-42
- 2. Amendment No. 167 to DPR-60
- 3. Safety Evaluation cc w/encls: See next page
Pkg: ML070330441 Amd: ML070330455 TS:ML070810131 OFFICE LPL3-1/PM LPL3-1/LA DCI/CSGB/BC OGC LPL3-1/BC NAME MChawla THarris A Hiser MBarkman LRaghavan DATE 2 /13/07 2/28/07 2/28/07 3/19/07 3/20/07 Prairie Island Nuclear Generating Plant, Units 1 and 2 cc:
Jonathan Rogoff, Esquire Tribal Council Vice President, Counsel & Secretary Prairie Island Indian Community Nuclear Management Company, LLC ATTN: Environmental Department 700 First Street 5636 Sturgeon Lake Road Hudson, WI 54016 Welch, MN 55089 Manager, Regulatory Affairs Nuclear Asset Manager Prairie Island Nuclear Generating Plant Xcel Energy, Inc.
Nuclear Management Company, LLC 414 Nicollet Mall, R.S. 8 1717 Wakonade Drive East Minneapolis, MN 55401 Welch, MN 55089 Michael B. Sellman Manager - Environmental Protection Division President and Chief Executive Officer Minnesota Attorney Generals Office Nuclear Management Company, LLC 445 Minnesota St., Suite 900 700 First Street St. Paul, MN 55101-2127 Hudson, MI 54016 U.S. Nuclear Regulatory Commission Resident Inspector's Office 1719 Wakonade Drive East Welch, MN 55089-9642 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351 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 July 2006
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 177 License No. DPR-42
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Nuclear Management Company, LLC (the licensee), dated February 16, 2006, supplemented by letters dated July 21, and December 27, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 177 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
L. Raghavan, Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: March 20, 2007
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 167 License No. DPR-60
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Nuclear Management Company, LLC (the licensee), dated February 16, 2006, supplemented by letters dated July 21, and December 27, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 167, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
L. Raghavan, Chief Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: March 20, 2007
ATTACHMENT TO LICENSE AMENDMENT NOS. 177 AND 167 FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Facility Operating License No. DPR-42 and DPR-60 with the attached revised pages. The changed areas are identified by a marginal line.
REMOVE INSERT DPR-42, License Page 3 DPR-42, License Page 3 DPR-60, License Page 3 DPR-60, License Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 1.1-3 1.1-3 3.4.14-2 3.4.14-2 3.4.14-3 3.4.14-3 5.0-13 5.0-13 5.0-14 5.0-14 5.0-15 5.0-15 5.0-16 5.0-16 5.0-17 5.0-17 5.0-18 5.0-18 5.0-19 5.0-19 5.0-20 5.0-20 5.0-21 5.0-21 5.0-22 5.0-22 5.0-30 5.0-30 5.0-31 5.0-31 5.0-38 5.0-38 5.0-39 5.0-39 5.0-40 5.0-40
(4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NMC to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to transfer byproduct materials from other job sites owned by Northern States Power Company for the purpose of volume reduction and decontamination.
C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l: Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level NMC is authorized to operate the facility at steady state reactor core power levels not in excess of 1650 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 177, are hereby incorporated in the license. NMC shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection NMC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Prairie Island Nuclear Generating Plant Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program," Revision 0, submitted by letter dated October 18, 2004.
Unit 1 Amendment No. 177
(5) Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NMC to transfer byproduct materials from other job sites owned by Northern States Power Company for the purposes of volume reduction and decontamination.
C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l: Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level NMC is authorized to operate the facility at steady state reactor core power levels not in excess of 1650 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 167, are hereby incorporated in the license. NMC shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection NMC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled:
'Prairie Island Nuclear Generating Plant Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program," Revision 0, submitted by letter dated October 18, 2004.
Unit 2 Amendment No.167
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 177 TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 167 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
By application dated February 16, 2006 (Agency Documents Access and Management System (ADAMS) Accession No. ML060480440), as supplemented by letters dated July 21, (ADAMS Accession No. ML062370052), and December 27, 2006 (ADAMS Accession No. ML063620460), Nuclear Management Company (the licensee), requested changes to the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plant Units 1 and 2. The supplements dated July 21, and December 27, 2006 provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 11, 2006 (71 FR 18376).
The proposed changes would revise the existing steam generator (SG) tube surveillance program. The changes are modeled after TS Task Force (TSTF) traveler TSTF-449, Revision 4, Steam Generator Tube Integrity, and the model safety evaluation prepared by the NRC and published in the Federal Register on March 2, 2005 (70 FR 10298). In this regard, the scope of the application includes changes to the definition of leakage, changes to the primary-to-secondary leakage requirements, changes to the SG tube surveillance program (SG tube integrity), changes to the SG reporting requirements, and associated changes to the TS Bases.
2.0 REGULATORY EVALUATION
The background, description, and applicability of the proposed changes associated with the SG tube integrity issue and the applicable regulatory requirements, were included in the NRC staffs model safety evaluation (SE) published in the Federal Register on March 2, 2005 (70 FR 10298). The Notice of Availability of Model Application Concerning Technical Specification Improvement To Modify Requirements Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process, was published in the Federal Register on May 6, 2005 (70 FR 24126), and made the model SE available for licensees to reference.
3.0 TECHNICAL EVALUATION
3.1 Overview In its February 16, 2006, application, and July 21, and December 27, 2006, supplements, the licensee proposed changes to the TSs that are modeled after TSTF Standard TS Change Traveler, TSTF-449, Steam Generator Tube Integrity. Consistent with TSTF-449, the proposed TS changes include: (1) a revised definition of LEAKAGE in TS 1.1, (2) a revised TS 3.4.14, "RCS (Reactor Coolant System) Operational LEAKAGE," (3) a new TS 3.4.19, "Steam Generator (SG) Tube Integrity," (4) a revised TS 5.5.8, "Steam Generator (SG) Program," (5) a revised TS 5.6.7, "Steam Generator Tube Inspection Report," and (6) revised Table of Content pages to reflect the proposed changes. There were minor differences between TSTF-449 and the licensee's application. These included differences in the facility licensing basis and differences in TS numbering. These differences are discussed below.
With respect to the differences in the facility licensing basis, the differences did not invalidate the technical evaluation of TSTF-449; rather they resulted in the licensee having to slightly deviate from some of the modifications discussed in TSTF-449. For example, in the Bases section for Steam Generator Tube Integrity, TSTF-449 indicated that the accident analysis for a SG tube rupture, assumed the contaminated secondary fluid was only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser. Since the licensee has a different licensing basis than the one described in the standard TSs (i.e., TSTF-449), they did not include this sentence, rather they indicated that the analysis assumes the contaminated fluid is released to the atmosphere via atmospheric steam dumps.
Another example (in the Bases section for Reactor Coolant System Operational Leakage and SG Tube Integrity), is that the licensee provided additional text concerning the appropriate accident induced primary-to-secondary leakage limit and structural integrity performance criteria when specific alternate repair criteria are implemented. Another example is that the definition of a SG tube (as defined in the Bases section for SG Tube Integrity), is slightly different than the definition of a tube in TSTF-449. This difference in definition is a result of the licensee being authorized to implement alternate tube repair criteria and to repair tubes by sleeving.
With respect to the definition of a tube in the Bases, the licensee clarified in a conference call, that the original tube wall at the location where sleeve joints are established is considered part of the tube.
Since these differences were minor in nature, they were consistent with the plant's licensing basis (e.g., in the level of detail incorporated into the TS Bases), or they were consistent with the intent of TSTF-449, the NRC staff determined they were acceptable.
With respect to the differences in the numbering of the TSs, these differences were administrative in nature and did not affect the technical adequacy of the submittal. As a result, the NRC staff determined they were acceptable.
The licensee proposed a few changes that went beyond TSTF-449. For example, the licensee proposed to incorporate a new requirement in TS 5.5.8.c.2(a)(3) to specify the repair criteria for flaws in the original tube wall at the sleeve to tube joint. Since the proposal was to plug tubes
upon detection of flaws at this location (consistent with the design basis of the joint), the staff found the proposal acceptable. Since these differences were administrative in nature and did not affect the technical adequacy of the submittal, the NRC staff determined they were acceptable.
In addition to these minor changes, the licensee proposed to include previously approved alternate repair criteria and repair methods into their proposed new TSs. The structure of the TSTF-449 allows licensees to incorporate alternate repair criteria and methods into the TSTF-449 format. By incorporating the previously approved repair criteria and repair methods into the TSTF-449 format, there were several additions, deletions, and changes to the requirements.
These changes (including additions and deletions), were made as a result of the format, content, and performance-based approach of TSTF-449.
The staff verified that (a) the inspection criteria associated with these repair criteria and methods were moved, as appropriate, to the inspection section of the proposed SG TSs, (b) the repair criteria were moved, as appropriate, to the repair criteria section of the proposed SG TSs, (c) the repair methods were moved to the repair method section of the proposed SG TSs, and (d) the reporting requirements were moved to the reporting section of the proposed SG TSs. There were some pre-existing reporting requirements associated with these previously approved repair criteria deleted, since the reporting requirements were no longer necessary.
These requirements were no longer necessary because the licensee incorporated the limits that would require the report to be submitted into the definition of tube integrity (and the plant can not operate when tube integrity is not maintained under the proposed new SG TSs). In addition, the timing for one of the reports was changed from within 15-days following the inspection to 180-days after the initial entry into Mode 4. This change in the timing of the report was considered acceptable since 180-days is consistent with the TSTF-449 reporting time frame and operating experience does not warrant a report (with this specific information) within 15-days following the inspection.
The licensee also deleted several definitions associated with a tube repair criteria. Since these definitions were no longer necessary to clarify the repair criteria, the staff found the deletion of these definitions acceptable. In addition, the licensee clarified that a region of the tube within the tubesheet (i.e., below a portion of the tube referred to as F* or EF* region), does not require an inspection unless there is a sleeve within this region. Since this proposal was consistent with the original approval of the F* and EF* repair criteria (i.e., degradation below this region is acceptable), the staff found this clarification acceptable. The staff notes, however, that as a practical matter, this region of the tube is still typically inspected, and indications found in this region of the tube are reported to the NRC. The licensee also clarified in their proposed TS that re-rolling of a tube in the tubesheet was an acceptable tube repair method. This clarification was considered acceptable since re-rolling was implemented in combination with application of the NRC approved F* and EF* repair criteria. In summary, the NRC staff determined that the previously approved repair criteria and repair methods were appropriately incorporated into the plants TSs.
In their TS Bases, the licensee described various assumptions regarding primary-to-secondary leakage in their accident analyses. This description focused only on accidents that have a faulted SG. Since some design-basis accidents do not have a faulted SG (and yet assume primary-to-secondary leakage exists), the licensee clarified in a conference call (per an NRC
request), that it was not implying in their Bases that the accident induced leakage performance criteria (discussed in TS 5.5.8.b.2) only addresses those accidents with a faulted SG, but rather it addresses all design-basis accidents (other than a SG tube rupture) which assume primary-to-secondary leakage exists.
The remainder of the application was consistent with, or more limiting than, TSTF-449.
In summary, the staff determined that the model SE is applicable to this review and finds the proposed changes acceptable.
4.0
SUMMARY
The proposed TS changes establish a programmatic, largely performance-based regulatory framework for ensuring SG tube integrity is maintained. The NRC staff finds that it addresses key shortcomings of the current framework by ensuring that SG programs are focused on accomplishing the overall objective of maintaining tube integrity. It incorporates performance criteria for evaluating tube integrity that the NRC staff finds consistent with the structural margins and the degree of leak tightness assumed in the current plant licensing basis. The NRC staff finds that maintaining these performance criteria provides reasonable assurance that the SGs can be operated safely without an increase in risk.
The revised TSs will contain limited specific details concerning how the SG Program is to achieve the required objective of maintaining tube integrity; the intent being that the licensee will have the flexibility to determine the specific strategy for meeting this objective. However, the NRC staff finds that the revised TSs include sufficient regulatory constraints on the establishment and implementation of the SG Program such as to provide reasonable assurance that tube integrity will be maintained.
Failure to meet the performance criteria will be reportable pursuant to the requirements in 10 CFR Parts 50.72 and 50.73. The NRC reactor oversight process provides a process by which the NRC staff can verify that the licensee has identified any SG Program deficiencies that may have contributed to such an occurrence and that appropriate corrective actions have been implemented.
In conclusion, the NRC staff finds that the TS changes proposed by the licensee in its February 16, 2006, application and July 21, and December 27, 2006, supplements conform to the requirements of 10 CFR 50.36 and establish a TS framework that will provide reasonable assurance that SG tube integrity is maintained without undue risk to public health and safety.
The licensee included in its application the revised TS Bases to be implemented with the TS change. The NRC staff finds that the TS Bases Control Program is the appropriate process for updating the affected TS Bases pages and has, therefore, not included the affected Bases pages with this amendment.
5.0 REFERENCES
A complete list of references used to complete this review can be found in the NRCs model SE published in the Federal Register on March 2, 2005 (70 FR 10298).
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 and changes the surveillance requirements. The 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 18376). 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: YDiaz-Castillo, KKarwoski Date: March 20, 2007