ML120900189
| ML120900189 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 05/02/2012 |
| From: | Boska J Plant Licensing Branch 1 |
| To: | Entergy Nuclear Operations |
| Boska J, NRR/DORL/LPLI-1, 301-415-290 | |
| References | |
| TAC ME7230, TAC ME7231, FOIA/PA-2016-0148 | |
| Download: ML120900189 (19) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 May 2,2012 Vice President, Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 -ISSUANCE OF AMENDMENTS RE: REVISIONS TO INSERVICE TEST PROGRAM TECHNICAL SPECIFICATIONS (TAC NOS. ME7230 AND ME7231)
Dear Sir or Madam:
The Commission has issued the enclosed Amendment No. 267 to Facility Operating License No.
DPR-26 for the Indian Point Nuclear Generating Unit No.2 (IP2) and Amendment No. 245 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit NO.3 (IP3).
The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated September 16, 2011.
The amendments revise the Inservice Testing Program, TS 5.5.6 for IP2 and TS 5.5.7 for IP3, by incorporating TS Task Force Traveler (TSTF) 479, "Changes to Reflect Revision of 10 CFR 50.55a," and TSTF-497, "Limit Inservice Testing Program SR [Surveillance Requirement] 3.0.2 Application to Frequencies of 2 Years or Less." Specifically, the amendments (1) replace references to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, with the ASME Code for Operation and Maintenance of Nuclear Power Plants for inservice testing activities, and (2) applies the extension allowance of SR 3.0.2 to other normal and accelerated inservice testing frequencies of 2 years or less that were not included in the frequencies of the table listed in TS 5.5.6.a for IP2 and in TS 5.5.7.a for IP3.
-2 A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely, o n P. Boska, Senior Project Manager ant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286
Enclosures:
- 1. Amendment No. 267 to DPR-26
- 2. Amendment No. 245 to DPR-64
- 3. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 2, LLC ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 267 License No. DPR-26
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated September 16, 2011, 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-26 is hereby amended to read as follows:
- 2 (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 267, are hereby incorporated in the license. ENO 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 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION George Wilson, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the License and Technical Specifications Date of Issuance: May 2, 2012
ATTACHMENT TO LICENSE AMENDMENT NO. 267 FACILITY OPERATING LICENSE NO. DPR-26 DOCKET NO. 50-247 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page 3 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 Pages Insert Pages 5.5-5 5.5-5 5.5-6 5.5-6
-3 instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) ENO pursuant to the Act and 10 CFR Parts 30,40 and 70, Amdt. 42 to receive, possess, and use in amounts as required any 10-17-78 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Amdt. 220 possess, but not separate, such byproduct and special 09-06-01 nuclear materials as may be produced by the operation of the facility.
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 I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, 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 ENO is authorized to operate the facility at steady state Amdt. 241 reactor core power levels not in excess of 3216 10-27-04 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 267, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.
(3) The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:
- 1. This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documents as described in Table R, "Relocated Technical Specifications from the CTS," and Table LA, "Removed Details and Less Restrictive Administrative Changes to the CTS" attached to the NRC staff's Safety Evaluation enclosed with this amendment. The relocation of requirements and detailed information shall be completed on or before the implementation of this amendment.
Amendment No. 267
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Radioactive Effluent Controls Program (continued)
- j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.
5.5.4 Component Cyclic or Transient Limit This program provides controls to track the UFSAR. Section 4.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.
5.5.5 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel using ultrasonic methods. The program shall include inspection frequencies and acceptance criteria. The inspection frequency will ensure that each reactor coolant pump flywheel is inspected at 20-year intervals.
5.5.6 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1,2, and 3 components. The program shall include the following:
- a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code and applicable Addenda as follows:
ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice testing inservice testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days INDIAN POINT 2 5.5 - 5 Amendment No. 267
Programs and Manuals 5.5 5.6 Programs and Manuals 5.5.6 Inservice Testing Program (continued)
- b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities,
- c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and
5.5.7 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
- a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tu bing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.
- b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads INDIAN POINT 2 5.5 - 6 Amendment No. 267
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 3, LLC ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO.3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 245 License No. DPR-64
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated September 16, 2011, 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-64 is hereby amended to read as follows:
-2 (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 245, are hereby incorporated in the license. ENO 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 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
~ oJll.-~.
George Wilson, Chief Plant Licensing Branch 1-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the License and Technical Specifications Date of Issuance: May 2, 2012
ATTACHMENT TO LICENSE AMENDMENT NO. 245 FACILITY OPERATING LICENSE NO. DPR~64 DOCKET NO. 50-286 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page 3 3 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Page Insert Page 5.0-12 5.0-12
- 3 (4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Amdt. 203 to receive, possess, and use in amounts as required any 11/27/00 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Arndt. 203 possess, but not separate, such byproduct and special 11/27/00 nuclear materials as may be produced by the operation of the facility.
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 I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and 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 ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal (100% of rated power).
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 245, are hereby incorporated in the License. ENO shall operate the facility in accordance with the Technical Specifications.
(3) (DELETED) Amdt. 205 2-27-01 (4) (DELETED) Arndt. 205 2-27-01 D. (DELETED) Amdt.46 2-16-83 E. (DELETED) Amdt.37 5-14-81 F. This amended license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter of May 2, 1975, to Consolidated Edison Company of New York, Inc., granting a Section 401 certification under the Federal Water Pollution Control Act Amendments of 1972.
Amendment No. 245
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class I, 2, and 3 components including applicable supports. The program shall include the following:
- a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:
ASME OM Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At 1east once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At 1east once per 366 days Biennially or every 2 years At 1east once per 731 days
- b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservic~ testing activities;
- c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and
(continued)
INDIAN POINT 3 5.0 -12 Amendment 245
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INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286
1.0 INTRODUCTION
By letter dated September 16,2011, Agencywide Documents Access and Management System (ADAMS) Accession No. ML11273A036, Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted a request to the Nuclear Regulatory Commission (NRC) for changes to the Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3) Technical Specifications (TSs).
The proposed changes revise the Inservice Testing Program, TS 5.5.6 for IP2 and TS 5.5.7 for IP3, by incorporating TS Task Force Traveler (TSTF)-479, "Changes to Reflect Revision of 10 CFR 50.55a," and TSTF-497, "Limit Inservice Testing Program SR [Surveillance Requirement]
3.0.2 Application to Frequencies of 2 Years or Less," Specifically, the changes (1) replace references to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, with the ASME Code for Operation and Maintenance of Nuclear Power Plants for inservice testing (1ST) activities, and (2) applies the extension allowance of SR 3.0.2 to other normal and accelerated inservice testing frequencies of 2 years or less that were not included in the frequencies of the table listed in TS 5.5.6.a for IP2 and in TS 5.5.7.a for IP3.
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. In Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36 (10 CFR 50.36), the NRC established regulatory requirements related to the content of TSs. That regulation requires that the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings, (2) limiting conditions for operations (LCOs), (3) SRs, (4) design features, (5) administrative controls, (6) decommissioning, (7) initial notification, and (8) written reports.
The NRC staff reviewed the proposed changes for compliance with 10 CFR 50.36, "Technical specifications," and consistency with the precedent as established in NUREG-1431, "Standard Technical Specifications - Westinghouse Plants," Revision 3, including changes incorporated via TSTF-479 and TSTF-497.
-2 By letter dated December 6, 200S, the NRC approved Revision 0 of TSTF-479. TSTF-479, Revision 0, revises references in the Standard Technical Specifications (STS) Administrative Controls 1ST Program and STS Bases to reflect the current edition of the ASME Code specified in 10 CFR SO.SSa(b). The NRC concluded that the revision was acceptable because the requirements of 10 CFR SO.SSa adequately provide for 1ST.
By letter dated October 2, 2006, the NRC approved Revision 0 of TSTF-497. TSTF-497, Revision 0, revises the STS 1ST program by clarifying that the application of the 2S percent (%)
1ST interval extension allowed by SR 3.0.2 was for 1ST frequencies of 2 years or less. The NRC concluded that the revision was acceptable because it was an administrative change that clarified that the provisions of SR 3.0.2 (i.e., the 2S% interval extension) are applicable to 1ST intervals of 2 years or less.
The changes were also reviewed for compliance with the requirements for 1ST as contained in 10 CFR SO.SSa(f)(4) for ASME Code Class 1, 2, and 3 pumps and valves. They are also consistent with the guidance in NUREG-1482, "Guidelines for In-service Testing at Nuclear Power Plants."
3.0 TECHNICAL EVALUATION
This license amendment request proposed changes to revise IP2 TS S.S.6, "Inservice Testing Program" and IP3 TS S.S.7, "Inservice Testing Program." Both referenced TSs would be revised to (1) specify the ASME Operations and Maintenance (OM) Code for 1ST for consistency with the requirements of 10 CFR SO.SSa(f)(4) for pumps and valves which are classified as ASME Code Class 1, 2, and 3, and (2) address the applicability of SR 3.0.2 to other normal and accelerated testing frequencies not listed. The licensee has stated that the proposed changes are consistent with TSTF-479, Revision 0, "Changes to Reflect Revision of 10 CFR SO.SSa,"
TSTF-497, Revision 0, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less," and NUREG-1431, "Standard Technical Specifications - Westinghouse Plants," Revision 3.0.
3.1 TSTF-479. Revision O. "Changes to Reflect Revision of 10 CFR SO.SSa" The purposes of the 1ST programs are to assess the operational readiness of pumps and valves, to detect degradation that might affect component operability, and to maintain safety margins with provisions for increased surveillance and corrective action. The NRC regulations, in 10 CFR SO.SSa, define the requirements for applying industry codes to each licensed nuclear power facility. Licensees are required by 10 CFR SO.SSa(f)(4)(i) to prepare programs to perform 1ST of certain ASME Section III, Code Class 1, 2, and 3 pumps and valves during the initial 120 month interval. The regulations require that 1ST be revised during the successive 120-month intervals to comply with the latest edition and addenda of the Code incorporated by reference in 10 CFR SO.SSa(b) 12 months prior to the start of the interval.
Section XI of the ASME Code has been revised on a continuing basis over the years to provide updated requirements for the inservice inspection and 1ST of components. Until 1990, the ASME Code requirements addressing the 1ST of pumps and valves were contained in Section XI, Subsections IWP (pumps) and IWV (valves). In 1990, the ASME published the initial edition of the OM Code that provides the rules for the 1ST of pumps and valves. Since the
-3 establishment of the 1990 Edition of the OM Code, the rules for the 1ST of pumps are no longer being updated in Section XI.
The TS change does not eliminate any tests and does not relieve the licensee of its responsibility to seek relief from ASME Code test requirements when they are impractical. The proposed change of the ASME Code reference from "ASME Section XI" to "ASME OM Code" eliminates the ASME Code inconsistency between the 1ST program and the TS, as required by 10 CFR 50.55a(f)(4)(ii); therefore, the NRC staff finds these proposed changes to be acceptable.
This change is also consistent with the NRC's basis for approval of TSTF-479 in that the requirements of 10 CFR 50.55a adequately provide for inservice testing.
3.2 TSTF-497. Revision O. "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less" TSTF-479, Rev. 0, "Changes to Reflect Revision of 10 CFR 50.55a," revised the Inservice Testing Program located in Chapter 5 of the STS to reflect the ASME OM Code rather than ASME Code,Section XI. TSTF-479 also revised paragraph b of the STS Chapter 5 Inservice Testing Program by adding "The provisions of SR 3.0.2 are applicable to the above required Frequencies and other normal and accelerated Frequencies specified in the Inservice Testing Program for performing inservice testing activities." However, TSTF-479 does not provide adequate justification for applying SR 3.0.2 to Frequencies specified in the Inservice Testing Program as greater than 2 years, and the NRC staff does not approve plant-specific amendments without further justification.
In order to enhance the 2 years or less test frequency requirement, TSTF-497 revised paragraph b of the STS Inservice Testing Program to state, "The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing 1ST activities." Without this limitation, some components, such as safety and relief valves that may be tested at surveillance intervals significantly greater than 2 years, could have extensions applied which would be much greater than needed for operational flexibility. This is an administrative change to clarify that the extensions allowed by TS SR 3.0.2 are applicable to 1ST activities with intervals of 2 years or less. The licensee's proposed changes are consistent with the bases for the NRC's approval of TSTF-479 and TSTF-497.
3.3 Conclusion Based on the above evaluation, the NRC staff finds the proposed revisions to the IP2 and IP3 TSs meet the requirements of 10 CFR 50.55a and 10 CFR 50.36. In addition, the amendment is consistent with the NRC staff position contained in Revision 1 of NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants." Therefore, the NRC staff finds the proposed changes acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official provided the following comment:
-4 The second portion of the requested operating license amendment would change the scope and purpose of Surveillance Requirement 3.0.2 and would permit intervals not exceeding 2 years to be extended by 25%. This could result in a delay of up to 6 months for certain surveillances.
This proposed amendment and proposed standard is not written in clear, enforceable text. Instead, its wording and jargon could impede public understanding of the actual terms of the commitment and whether or not the commitment is measurable and enforceable. This lack of clarity could make enforcement by a federal regulator and the Department of Justice difficult. Stated differently, could this new proposed requirement be explained to a federal court jury and provide an unambiguous commitment for which there would be no doubt as to whether the operator was in compliance or not in compliance? Lack of clarity could frustrate and prevent enforcement.
New York State looks to the NRC, in its response to this application, to clarify what is understood by the Entergy proposal and provide the rationale for why deviating from the current license requirement will not result in degraded oversight or increased risks.
The NRC staff finds that the amended license is clear and enforceable. SR 3.0.2 allows a 25%
extension of the time to perform the surveillances required by the plant TS, unless the use of SR 3.0.2 is prohibited in a specific surveillance. This extension facilitates scheduling and considers plant operating conditions that may not be suitable for conducting the surveillance (e.g., transient conditions or other ongoing surveillance or maintenance activities). This license amendment does not change SR 3.0.2. This license amendment revises TS Section 5.5.6 for IP2 and TS Section 5.5.7 for IP3, which is the Inservice Testing Program. The Inservice Testing Program applies the requirements of the ASME OM Code, which requires additional tests of plant equipment beyond those already stated in the plant's TSs. The amended license specifies that for frequencies listed in the ASME OM Code as 2 years or less, the licensee may apply a 25%
extension as specified in SR 3.0.2. This same change has been approved for many of the nation's nuclear plants, and NRC inspectors have inspected and enforced this provision.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes 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 (76 FR 80976). 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.
-5
6.0 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 wi" not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: S. Anderson Date: May 2, 2012
-2 A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely, Ira/
John P. Boska, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286
Enclosures:
- 1. Amendment No. 267 to DPR-26
- 2. Amendment No. 245 to DPR-64
- 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC LPL 1-1 R/F RidsNrrDorl RidsN rrDorl Lp11-1 RidsN rrDorl Dpr RidsNrrPMlndianPoint Ridsl\lrrLASLittie RidsN rrDssStsb RidsRgn11\/iailCenter RidsOGCRp RidsAcrsAcnw_MailCTR SAnderson RidsNrrDeEptb ADAMS ACCESSION NO.: ML120900189 *Via email **See memo dated 2/23/12 OFFICE LPL 1-1/PM LPL 1 STSB/BC** EPTB/BC OGC LPL 1-1/BC 1/LA*
NAME JBoska SLittie RElliott AMcMurtray STurk GWilson DATE 4/6/12 4/5/12 2/23/12 4/9112 4/27/12 5/1112 OFFICIAL RECORD COPY