ML070570506
| ML070570506 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/01/2007 |
| From: | Farideh Saba NRC/NRR/ADRO/DORL/LPLIV |
| To: | Mitchell T Entergy Operations |
| Saba F, NRR/DORL/LPL4, 301-415-1447 | |
| Shared Package | |
| ML070570502 | List: |
| References | |
| TAC MD0700 | |
| Download: ML070570506 (13) | |
Text
March 1, 2007 Mr. Timothy G. Mitchell Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72802
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: REMOVAL OF REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY REQUIREMENTS (TAC NO. MD0700)
Dear Mr. Mitchell:
The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 270 to Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit No. 2 (ANO-2).
The amendment consists of changes to the ANO-2 Technical Specifications (TSs) in response to the Entergy Operations, Inc. (the licensee), application dated March 20, 2006.
The amendment removes ANO-2's reactor coolant structural integrity requirements contained in TS 3/4.4.10. The TS change is consistent with NUREG-1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.1. The Bases for TS 3/4.4.10 will be deleted and performed under the ANO-2 TS Bases Control Program, and is not included with the submittal. The amendment also renumbers TS pages 3/4 4-22a, 23, 23a, and 23b as TS pages 3/4 4-23, 24, 25, and 26, respectively.
T. Mitchell A copy of the NRC staffs related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/RA/
Farideh E. Saba, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-368
Enclosures:
- 1. Amendment No. 270 to NPF-6
- 2. Safety Evaluation cc w/encls: See next page
Package ML070570502 (Amdt./License ML070570506, TS Pages ML070570519)
OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/CPNB/BC NRR/ITSB/BC OGC - NLO NRR/LPL4/BC NAME FSaba LFeizollahi TChan TKobetz APHodgdon DTerao DATE 2/27/07 2/27/07 2/27/07 2/28/07 2/28/07 3/1/07
February 2007 Arkansas Nuclear One cc:
Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995 General Manager Plant Operations Entergy Operations, Inc.
Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802 Director, Nuclear Safety Assurance Entergy Operations, Inc.
Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802 Manager, Licensing Entergy Operations, Inc.
Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802 Director, Nuclear Safety & Licensing Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213-8298 Section Chief, Division of Health Radiation Control Section Arkansas Department of Health and Human Services 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Section Chief, Division of Health Emergency Management Section Arkansas Department of Health and Human Services 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 310 London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 County Judge of Pope County 100 W. Main Street Russellville, AR 72801 Vice President, Operations Support Entergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995
ENTERGY OPERATIONS, INC.
DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. NPF-6 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee), dated March 20, 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 license 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 and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-6 as indicated in the attachment to this license amendment.
3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and the Technical Specifications Date of Issuance: March 1, 2007
ATTACHMENT TO LICENSE AMENDMENT NO. 270 RENEWED FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Replace the following pages of the Renewed Facility Operating License No. NPF-6 and 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.
Operating License REMOVE INSERT 3
3 Technical Specifications REMOVE INSERT 3/4 4-22a 3/4 4-23 3/4 4-23 3/4 4-24 3/4 4-23a 3/4 4-25 3/4 4-23b 3/4 4-26 3/4 4-26
3 facility at the designated location in Pope County, Arkansas in accordance with the procedures and limitations set forth in this renewed license; (3)
EOI, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time at the facility site and as designated solely for the facility, special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)
EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, 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 calibration or associated with radioactive apparatus or components; and (6)
EOI, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain and is subject to conditions specified in the following Commission regulations in 10 CFR Chapter 1; 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 3026 megawatts thermal. Prior to attaining this power level EOI shall comply with the conditions in Paragraph 2.C.(3).
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 270 are hereby incorporated in the renewed l
Renewed License No. NPF-6 Amendment No. 270
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT NO. 2 DOCKET NO. 50-368
1.0 INTRODUCTION
By application dated March 20, 2006 (Agencywide Document Access and Management System Accession No. ML060880332), Entergy Operations, Inc. (Entergy, the licensee) requested changes to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit No. 2 (ANO-2).
The proposed changes remove ANO-2 Reactor Coolant System (RCS) TS 3/4.4.10 (TS Page 3/4 4-26), Structural Integrity, and renumber TS pages 3/4 4-22a, 23, 23a, and 23b as TS 3.4.10.1 pages 3/4 4-23, 24, 25, and 26, respectively. The Bases for TS 3/4.4.10 will be deleted and performed under the ANO-2 TS Bases Control Program, and are not included with the submittal.
2.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act, as amended (the Act), requires applicants for nuclear power plant operating licenses to incorporate TSs as part of the license. The Commissions regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) 50.36. That regulation requires that the TSs include items in five categories, including: (1) safety limits, limiting safety system settings, and limiting control settings (2) limiting conditions for operation (LCOs), (3) surveillance requirements (SRs),
(4) design features, and (5) administrative controls.
On July 22, 1993, the Commission issued its Final Policy Statement, expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a of the Act and 10 CFR 50.36. The Final Policy Statement gave guidance for evaluating the required scope of the TSs and defined the guidance criteria to be used in determining which of the LCOs and associated SRs should remain in the TSs. The Commission noted that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TSs, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979). There, the Appeal Board observed:
[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.
By this approach, existing LCO requirements that fall within or satisfy any of the criteria in the Final Policy Statement should be retained in the TSs; those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents.
The Commission codified the four criteria in 10 CFR 50.36 (60 FR 36953, July 19, 1995). The four criteria are stated as follows:
(1)
Installed instrumentation that is used to detect, and indicate in a control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2)
A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (3)
A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of, or represent a challenge to the integrity of a fission product barrier; and (4)
A structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
As a result, existing LCO requirements that fall within or satisfy any of the criteria in 10 CFR 50.36(c)(2)(ii) must be retained in the TSs while those LCO requirements that do not fall within or satisfy these criteria may be relocated to other licensee-controlled documents.
3.0 TECHNICAL EVALUATION
3.1 Entergys Proposed TS Changes The licensee, in its application, stated that the purpose of TS 3.4.10.1, RCS structural integrity LCO, is to specify the requirements of maintaining the structural integrity of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Class 1, 2, and 3 components. This specification was originally intended to support assurance that structural integrity and operational readiness of these components are maintained at an acceptable level throughout the life of the facility. The specification is applicable in all operational modes. However, the specification does not provide actions for a plant shutdown if its LCO is not met. In addition, the specification contains no SRs as stated by the licensee.
This is because the specification addresses the passive pressure boundary function of ASME Code Class 1, 2, and 3 components as established under the inservice inspection (ISI) program. In addition to the above, the ISI program is required pursuant to 10 CFR 50.55a, thereby addressing the inspections necessary to maintain structural integrity.
Furthermore, the specification wording could be misconstrued to conflict with normal outage-related activities, including removal of RCS manways and the reactor vessel head in preparation for refueling, which would make the pressure boundary no longer structurally intact.
The licensee states that maintaining a program-type requirement within an LCO creates significant interpretation issues for Operations personnel. The RCS structural integrity TS was part of the original TS and, the TS basis history regarding its intent is not documented.
TS 3.4.10.1 appears to have been included to help ensure that plant heatup and startup would not occur until all required portions of the RCS were verified to meet ISI acceptance criteria following inspections performed during a plant outage. Meeting these acceptance criteria helps ensure the integrity of the RCS pressure boundary during all modes of operation, including accident events. Furthermore, TS 3.4.10.1 contains no action suggesting it was designed to accommodate integrity concerns once plant heatup has commenced. RCS structural integrity ISI activities are performed only during plant outages when conditions exist that permit access to the RCS pressure boundary and are not monitored or controlled through application of the ISI program during the operational cycles.
The licensee stated that other TSs are designed to monitor the structural integrity of the RCS during operation and provide actions to shut down the unit if compliance is not maintained. For example, reactor coolant heatup and cooldown rates protect against applying undue stresses as a result of pressure/temperature transients on RCS components and piping. The RCS leakage TSs provide a means of evaluating the RCS structural integrity by detecting and monitoring leakage. Therefore, the licensee stated it is not necessary to apply TS 3.4.10.1 when integrity issues become evident during plant operation above cold shutdown. Because TS 3.4.10.1 is redundant to other regulation, it is acceptable to remove TS 3.4.10.1 from the TSs. Finally, removal of this specification does not reduce the controls that are necessary to ensure compliance with the ASME Code or the need to maintain the RCS pressure boundaries.
Structural integrity is maintained by compliance with 10 CFR 50.55a, as implemented through the ANO-2 ISI Program.
3.2 Nuclear Regulatory Commission (NRC) Staffs Evaluation of TS Changes The TS changes proposed by Entergy in this license amendment request are required to be evaluated to confirm compliance with the regulatory requirements in Section 2.0 of this safety evaluation.
For Criterion 1, above, the RCS ASME Code Class 1, 2, and 3 components do not include any instrumentation. Therefore, the NRC staff finds that this TS does not meet Criterion 1.
For Criterion 2, structural integrity is neither a process variable, design feature, or operating restriction that is an initial condition of a design-basis analysis (DBA) or transient analysis.
Structural integrity is not monitored or controlled during plant operation; it is verified during periodic inspections. Therefore, the NRC staff finds that this TS does not meet Criterion 2.
For Criterion 3, ASME Code Class 1, 2, and 3 components that are part of the primary success path and function to mitigate DBAs or transients that either assume the failure of, or present a challenge to, the integrity/operability of these components are included in the individual specification that covers these components. The portion of this TS that is proposed to be removed addresses only the passive pressure boundary function of these components.
Therefore, the NRC staff finds that this TS does not meet Criterion 3.
For Criterion 4, the requirements covered by this TS that are being removed have not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. In addition, failure modes of applicable structures, systems, or components (SSCs) would not be identified from the requirements of this TS. Furthermore, the requirements of this TS do not affect the risk review/unavailability monitoring of applicable SSCs. Therefore, the NRC staff finds that this specification does not meet Criterion 4.
The review for the structural integrity LCO relocation was actually performed and presented in a split report from the Director of Nuclear Reactor Regulation, NRC, Thomas Murley, on May 8, 1988. Originally, NUREG-0212, Standard Technical Specifications for Combustion Engineering Plants, contained provisions for the LCOs and SRs in reference to the structural integrity of ASME Code Class 1, 2, and 3 components. This split report identified Section 3/4.4.10, Structural Integrity, as not meeting the criterion for 10 CFR 50.36 and was therefore removed from subsequent revisions of the Standard Technical Specifications.
Therefore, since this TS does not fulfill any of the 10 CFR 50.36(c)(2)(ii) criteria for items for which TSs must be established, the NRC staff finds that removing TS 3/4.4.10 and the associated bases is acceptable. Finally, the removal of TS 3/4.4.10 and its associated references to structural integrity eliminates from the TSs the redundancy of structural integrity requirements that are already covered under 10 CFR 50.55a.
Normally in applying the Commission Final Policy Statement on Technical Specifications for Nuclear Power Reactors, the NRC staff would require that a licensee identify both the licensee-controlled document receiving a relocated TS and the change control mechanism that governs that document. However, in this instance to achieve efficiency in the issuance of this license amendment to relocate TS 3/4 4.10, elimination of a duplicate regulatory requirement, the NRC staff will permit deletion without relocation of the TS. Therefore, the NRC staff finds this proposed change to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.
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 SRs.
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 published on May 9, 2006 (71 FR 26999). 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.
6.0 CONCLUSION
The NRC staff concludes, 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 Commissions 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. Therefore, the NRC staff approves the changes to the ANO-2 TSs, which effectively remove references to structural integrity under TS 3.4.10.1.
Principal Contributor: Tim Steingass Date: March 1, 2007