Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 9821-9832 [2011-3721]

Download as PDF Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices Dated: February 5, 2011. Richard W. Sherman, Deputy General Counsel. Estimated No. of Respondents/ Recordkeepers: 650. Estimated Burden Hours per Response: 2.0 hours. Frequency of Response: Reporting and on occasion. Estimated Total Annual Burden Hours: 1300. Estimated Total Annual Cost: $ 0. [FR Doc. 2011–3781 Filed 2–18–11; 8:45 am] BILLING CODE P NATIONAL CREDIT UNION ADMINISTRATION By the National Credit Union Administration Board on February 15, 2011. Mary Rupp, Secretary of the Board. Agency Information Collection Activities: Submission to OMB for Review; Comment Request [FR Doc. 2011–3792 Filed 2–18–11; 8:45 am] National Credit Union Administration (NCUA). ACTION: Request for comment. AGENCY: The NCUA intends to submit the following information collection to the Office of Management and Budget (OMB) for review and clearance under the Paperwork Reduction Act of 1995 (Pub. L. 104–13, 44 U.S.C. Chapter 35). This information collection is published to obtain comments from the public. DATES: Comments will be accepted until April 25, 2011. ADDRESSES: Interested parties are invited to submit written comments to NCUA Clearance Officer listed below: Clearance Officer: Tracy Sumpter, National Credit Union Administration, 1775 Duke Street, Alexandria, Virginia 22314–3428. Fax No. 703–837–2861, Email: OCIOmail@ncua.gov. FOR FURTHER INFORMATION CONTACT: Requests for additional information or a copy of the information collection request should be directed to Tracy Sumpter at the National Credit Union Administration, 1775 Duke Street, Alexandria, VA 22314–3428, or at (703) 518–6440. SUPPLEMENTARY INFORMATION: Proposal for the following collection of information: OMB Number: 3133–0121. Form Number: 4063 and 4063a. Type of Review: Reinstatement, without change, of a previously approved collection. Title: Notice of Change of Official or Senior Executive Officer and Individual Application for Approval of Official or Senior Executive Officer. Description: In order to comply with statutory requirements, the agency must obtain sufficient information from new officials or senior executives officers of troubled or newly chartered credit unions to determine their fitness for the position. These forms standardize the information gathered to evaluate the individual’s fitness for the position. The format is similar to the one used by the FFIEC agencies and the FRB. 12 CFR 701.14 and 741.205. SUMMARY: mstockstill on DSKH9S0YB1PROD with NOTICES BILLING CODE 7535–01–P VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 NUCLEAR REGULATORY COMMISSION [NRC–2011–0040] Biweekly Notice Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. Background Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from January 27, 2011, to February 10, 2011. The last biweekly notice was published on February 8, 2011 (76 FR 6830). Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission’s regulations in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously PO 00000 Frm 00082 Fmt 4703 Sfmt 4703 9821 evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. Written comments may be submitted by mail to the Chief, Rules, Announcements and Directives Branch (RADB), TWB–05–B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be faxed to the RADB at 301–492– 3446. Documents may be examined, and/or copied for a fee, at the NRC’s Public Document Room (PDR), located at One White Flint North, Room O1– F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission’s E:\FR\FM\22FEN1.SGM 22FEN1 mstockstill on DSKH9S0YB1PROD with NOTICES 9822 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices ‘‘Rules of Practice for Domestic Licensing Proceedings’’ in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the Commission’s PDR, located at One White Flint North, Room O1–F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System’s (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, https:// www.nrc.gov/reading-rm/doccollections/cfr/. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor’s/petitioner’s right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor’s/petitioner’s property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor’s/petitioner’s interest. The petition must also identify the specific contentions which the requestor/ petitioner seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/ petitioner to relief. A requestor/ petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment. All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The EFiling process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by telephone at (301) 415–1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign PO 00000 Frm 00083 Fmt 4703 Sfmt 4703 documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRCissued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket. Information about applying for a digital ID certificate is available on NRC’s public Web site at https:// www.nrc.gov/site-help/e-submittals/ apply-certificates.html. System requirements for accessing the ESubmittal server are detailed in NRC’s ‘‘Guidance for Electronic Submission,’’ which is available on the agency’s public Web site at https://www.nrc.gov/ site-help/e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC’s E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software. If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC’s online, Web-based submission form. In order to serve documents through EIE, users will be required to install a Web browser plugin from the NRC Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC’s public Web site at https://www.nrc.gov/site-help/esubmittals.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at https://www.nrc.gov/site-help/esubmittals.html. A filing is considered complete at the time the documents are submitted through the NRC’s E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the E:\FR\FM\22FEN1.SGM 22FEN1 mstockstill on DSKH9S0YB1PROD with NOTICES Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/ petition to intervene is filed so that they can obtain access to the document via the E-Filing system. A person filing electronically using the agency’s adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the ‘‘Contact Us’’ link located on the NRC Web site at https:// www.nrc.gov/site-help/esubmittals.html, by e-mail at MSHD.Resource@nrc.gov, or by a tollfree call at 1–866–672–7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays. Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by firstclass mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory proceedings will appear in NRC’s electronic hearing docket which is available to the public at https:// ehd1.nrc.gov/EHD/, unless excluded VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Nontimely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)–(viii). For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the Commission’s PDR, located at One White Flint North, Room O1–F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, https:// www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC PDR Reference staff at 1–800–397–4209, 301– 415–4737, or by e-mail to pdr.resource@nrc.gov. Entergy Nuclear Operations, Inc., Docket No. 50–333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York Date of amendment request: January 13, 2011. Description of amendment request: The proposed amendment would modify the Facility Operating License (FOL) by deleting references to specific Safety Evaluation Reports (SER), Technical Specification (TS) Amendments, and Exemptions from License Condition 2.C(3), Fire Protection, and replacing them with the words ‘‘as supplemented.’’ This is an administrative amendment to the FOL. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards PO 00000 Frm 00084 Fmt 4703 Sfmt 4703 9823 consideration, which is presented below: 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed FOL change is administrative and does not involve a plant or design function change. It has no effect on reactor operation or accident analyses, and thus, the proposed FOL change does not increase the probability or consequence of an accident previously evaluated. 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed FOL change is administrative and does not involve a plant or design function change. Because the proposed amendment would not change the design, configuration, or method of operation of the plant, it would not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety? Response: No. The proposed FOL change is administrative and does not involve a plant or design function change. No design or safety margin is involved. Therefore, the proposed change does not involve a reduction in any margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: Nancy L. Salgado. Entergy Nuclear Vermont Yankee (VY), LLC and Entergy Nuclear Operations, Inc., Docket No. 50–271, Vermont Yankee Nuclear Power Station, Vernon, Vermont Date of amendment request: December 21, 2010. Description of amendment request: The proposed amendment would revise Technical Specifications (TS) Section 3.6.A ‘‘Pressure and Temperature Limitation’’ to reflect the pressure and temperature (P–T) limits for the reactor coolant system through, approximately the end of the prospective 20-year E:\FR\FM\22FEN1.SGM 22FEN1 9824 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices mstockstill on DSKH9S0YB1PROD with NOTICES renewed license period, depending on the plant capacity factor. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change revises the period of applicability of the P–T limits. The technical bases for the new period of applicability have been previously reviewed and approved by the NRC as discussed in the submittal. Because the applicable regulatory requirements continue to be met, the change does not significantly increase the probability of any accident previously evaluated. The proposed change provides the same assurance of RPV integrity as previously provided. The change will require that the reactor pressure vessel and interfacing coolant system continue to be operated within their design, operational or testing limits. Also, the change will not alter any assumptions previously made in evaluating the radiological consequences of accidents. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change does not involve a modification of the design of plant structures, systems, or components. The change will not impact the manner in which the plant is operated and will not degrade the reliability of structures, systems, or components important to safety as equipment protection features will not be deleted or modified, equipment redundancy or independence will not be reduced, supporting system performance will not be affected and no severe testing of equipment will be imposed. No new failure modes or mechanisms will be introduced as a result of this proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Appendix G to 10 CFR 50 describes the conditions that require pressure-temperature (P–T) limits and provides the general bases for these limits. Operating limits based on the criteria of Appendix G, as defined by applicable regulations, codes and standards, provide reasonable assurance that nonductile or rapidly propagating failure will not occur. The P–T limits are prescribed for all plant modes to avoid encountering pressure, temperature and temperature rate of change conditions that might cause undetected flaws VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 to propagate and cause non-ductile failure of the reactor coolant pressure boundary. Calculation of P–T limits in accordance with the criteria of Appendix G to 10 CFR 50 and applicable regulatory requirements ensures that adequate margins of safety are maintained and there is no significant reduction in a margin of safety. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There is no change or impact on any safety analysis assumption or in any other parameter affecting the course of an accident analysis supporting the basis of any Technical Specification. The proposed change does not involve any increase in calculated off-site dose consequences. Therefore, operation of VY in accordance with the proposed amendment will not involve a significant reduction in a margin to safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: Nancy Salgado. FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50–440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio Date of amendment request: December 15, 2010. Description of amendment request: The proposed amendment would modify the requirements for testing control rod scram times following fuel movement within the reactor pressure vessel by incorporating Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) change traveler TSTF–222–A, Revision 1. Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (CFR) 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The control rod drive system is not an initiator to any accident sequence analyzed in the PNPP Updated Final Safety Analysis Report (USAR), including Appendix 15C, ‘‘Anticipated Transients Without Scram (ATWS).’’ The proposed TS changes improve PO 00000 Frm 00085 Fmt 4703 Sfmt 4703 existing surveillance requirements by eliminating unnecessary control rob scram time testing, while continuing to provide adequate assurance of control rod performance for those control rods in core cells in which fuel is moved or replaced, or control rod maintenance was performed. Historically, testing results indicate the control rod drive system is highly reliable. Since the fall 1996 implementation of Improved Technical Specifications, during 6036 control rod tests covering all 177 control rods, only 7 control rod tests (0.12 percent) yielded results slower than the required insertion time limit, and no control rods were inoperable as a result of scream time testing. All seven slow insertion time test results have been attributed to control rod scream solenoid pilot valves (SSPVs). These seven slow tests occurred prior to May 1999, and prior to a control rod SSPV upgrade program during which all 177 SSPV’s were replaced. As such, this type of change does not affect initiators of analyzed events and does not affect the mitigation of any accidents or transients. Therefore, the proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed TS changes do not involve a physical alteration of the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There are no setpoints affected by the changes at which protective or mitigative actions are initiated. The changes will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the USAR. This change does not alter assumptions made in the safety analysis and licensing basis. As such, no new failures modes are being introduced. Accordingly, the proposed changes do not create any new credible failure mechanisms, malfunction, or accident initiators not previously considered in PNPP design and licensing basis. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Margin of safety is related to the ability of the fission product barriers to perform their design functions during and following accident conditions. These barriers include the fuel cladding, the reactor coolant system, and the containment. This request does not involve a change to the fuel cladding, the reactor coolant system, or the containment. The proposed TS changes associated with TSTF–222–1 modify current frequency requirements for scram time testing control rods following refueling outages and for control rod requiring testing due to work E:\FR\FM\22FEN1.SGM 22FEN1 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices activities. Scram times for control rods not affected by fuel movement or control rod maintenance remain unaffected. The proposed TS changes have no affect on any safety analysis assumptions or methods of performing safety analyses. The changes do not adversely affect system design or operational requirements, and the equipment continues to be tested in a manner and at a frequency necessary to provide confidence that the equipment can perform its intended safety functions. Therefore, the proposed TS changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop. A–GO–15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Robert D. Carlson. mstockstill on DSKH9S0YB1PROD with NOTICES FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50–440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio Date of amendment request: December 15, 2010 Description of amendment request: The proposed amendment would revise the required testing frequency of Surveillance Requirement (SR) 3.1.4.2 from ‘‘120 days cumulative operation in MODE 1’’ to ‘‘200 days cumulative operation in MODE 1’’ by incorporating Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) change traveler TSTF–460, Revision 0. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The frequency of surveillance testing is not an initiator of any accident previously evaluated. The frequency of surveillance testing does not affect the ability to mitigate any accident previously evaluated, as the tested component is still required to be operable. Therefore, the proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change does not result in any new or different modes of plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change continues to test the control rod scram time to ensure the assumptions in the safety analysis are protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A–GO–15, 76 South Main Street, Akron, OH 44308. NRC Branch Chief: Robert D. Carlson. FPL Energy Duane Arnold, LLC, Docket No. 50–331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: October 15, 2010. Description of amendment request: The proposed amendment would revise Operating License No. DPR–49 by modifying the License to delete the parent guarantee License Condition. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment is an administrative change deleting the parent guarantee License Condition, as well as other minor editorial changes in format. Deletion of this License Condition does not involve any modifications to the safety-related structures, PO 00000 Frm 00086 Fmt 4703 Sfmt 4703 9825 systems or components (SSCs). Deletion of this License Condition will not alter previously evaluated Final Safety Analysis Report (FSAR) design basis accident analysis assumptions, add any accident initiators, or affect the function of the plant safety-related SSCs as to how they are operated, maintained, modified, tested, or inspected. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment only deletes the parent guarantee License Condition and makes other minor editorial changes. Deletion of this License Condition does not result in the need for any new or different FSAR design basis accident analysis. It does not introduce new equipment that could create a new or different kind of accident, and no new equipment failure modes are created. As a result, no new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed amendment. Therefore, the proposed amendment does not create a possibility for an accident of a new or different type than those previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The margin of safety is associated with the confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant pressure boundary, and containment structure) to limit the level of radiation to the public. The proposed amendment would not alter the way any safety-related SSC functions and would not alter the way the plant is operated. The amendment only involves deletion of the parent guarantee License Condition and minor editorial changes. The proposed amendment would not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. The proposed amendment would have no impact on the structural integrity of the fuel cladding, reactor coolant pressure boundary, or containment structure. Based on the above considerations, the proposed amendment would not degrade the confidence in the ability of the fission product barriers to limit the level of radiation to the public. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Ms. Marjan Mashhadi, Florida Power & Light Company, 801 Pennsylvania Avenue, NW., Suite 220, Washington, DC 20004. E:\FR\FM\22FEN1.SGM 22FEN1 9826 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices NRC Branch Chief: Robert J. Pascarelli. Indiana Michigan Power Company (the licensee), Docket No. 50–315, Donald C. Cook Nuclear Plant, Unit 1 (DCCNP–1), Berrien County, Michigan mstockstill on DSKH9S0YB1PROD with NOTICES Date of amendment request: December 16, 2010. Description of amendment request: The proposed amendment would revise Technical Specification (TS) 4.2.1, adding Optimized ZIRLO TM fuel rods to the fuel matrix in addition to Zircaloy or ZIRLO fuel rods that are currently in use. The proposed amendment would also add a Westinghouse topical report regarding Optimized ZIRLO TM as reference 8 in TS 5.6.5.b, which lists the analytical methods used to determine the core operating limits. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change would allow the use of Optimized ZIRLO TM clad nuclear fuel in the reactors. The NRC approved topical report WCAP–12610–P–A and CENPD–404– P–A, Addendum 1–A ‘‘Optimized ZIRLO TM,’’ prepared by Westinghouse Electric Company LLC (Westinghouse), addresses Optimized ZIRLO TM and demonstrates that Optimized ZIRLO TM has essentially the same properties as currently licensed ZIRLO TM. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO TM fuel cladding has been shown to meet all 10 CFR 50.46 acceptance criteria and, therefore, will not increase the consequences of an accident. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. Use of Optimized ZIRLO TM clad fuel will not result in changes in the operation or configuration of the facility. Topical Report WCAP–12610–P–A and CENPD–404–P–A demonstrated that the material properties of Optimized ZIRLO TM are similar to those of standard ZIRLO TM. Therefore, Optimized ZIRLO TM fuel rod cladding will perform similarly to those fabricated from standard ZIRLO TM, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident. Therefore, the proposed change does not create the possibility of a new or different VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the Optimized ZIRLO TM are not significantly different from those of standard ZIRLO TM. Optimized ZIRLO TM is expected to perform similarly to standard ZIRLO TM for all normal operating and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference in Optimized ZIRLO TM material properties relative to standard ZIRLO TM could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO TM properties will be performed prior to the use of fuel assemblies with fuel rods containing Optimized ZIRLO TM. These LOCA analyses will demonstrate that the acceptance criteria of 10 CFR 50.46 will be satisfied when Optimized ZIRLO TM fuel rod cladding is implemented. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: James M. Petro, Jr., Senior Nuclear Counsel, Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106. NRC Branch Chief: Robert J. Pascarelli. Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50–220, Nine Mile Point Nuclear Station Unit 1 (NMP1), Oswego County, New York Date of amendment request: September 29, 2010. Description of amendment request: The proposed amendment would revise the NMP1 Technical Specifications (TSs) Section 3/4.1.5, ‘‘SolenoidActuated Pressure Relief Valves (Automatic Depressurization System),’’ and 3/4.2.9, ‘‘Pressure Relief Systems— Solenoid-Actuated Pressure Relief Valves (Overpressurization),’’ to provide for an alternative means of testing the main steam electromatic relief valves (ERVs). Specifically, the proposed amendment would revise TS Surveillance Requirements (SRs) 4.1.5.a and 4.2.9.b to verify each ERV actuator strokes when manually actuated at least once each operating cycle. The functional testing requirements for the ERVs would be described in the PO 00000 Frm 00087 Fmt 4703 Sfmt 4703 Inservice Testing (IST) Program and controlled pursuant to TS Administrative Controls Section 6.5.4, ‘‘Inservice Testing Program.’’ The proposed change would allow demonstration of the capability of the valves to perform their safety function without requiring the ERVs to be cycled with reactor steam pressure while installed in the plant. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed amendment revises the TS Surveillance Requirements (SRs) to provide for an alternative means of testing the main steam ERVs. The ERVs perform automatic depressurization system (ADS) and overpressurization relief mode safety functions to mitigate the consequences of a small break loss of coolant accident (SBLOCA) and other accidents and transients. The ERVs are not considered an initiator for any accident previously evaluated except for the stuck-open ERV event, which is evaluated in Section XV– B.3.11 of the NMP1 Updated Final Safety Analysis Report (UFSAR). The proposed amendment would allow demonstration of the capability of the valves to perform their safety function through a series of tests, inspections, and maintenance activities without requiring the ERVs to be cycled with reactor steam pressure while installed in the plant, thereby eliminating the possibility of a stuck-open ERV event due to testing. Thus, the proposed amendment does not increase the probability of a stuck-open ERV event. The testing methodology, comprehensive inspections and preventive maintenance, and associated programmatic controls will provide an equivalent level of assurance that the ERVs are capable of performing their intended accident mitigation safety functions and, as such, will have no effect on the types or amounts of radiation released or the predicted offsite doses in the event of an accident. Accordingly, the proposed amendment does not alter the initial conditions, assumptions, or conclusions of any accident analysis. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed amendment does not affect the assumed accident performance of the ERVs, or of any plant structure, system, or component previously evaluated. The proposed amendment does not involve the E:\FR\FM\22FEN1.SGM 22FEN1 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices mstockstill on DSKH9S0YB1PROD with NOTICES installation of new equipment, and installed equipment is not being operated in a new or different manner. The proposed amendment provides for an alternative means of testing the ERVs that does not involve opening the valves with reactor steam while installed in the plant. The alternative testing and associated programmatic controls will provide an equivalent level of assurance that the ERVs are capable of performing their accident mitigation safety functions. No setpoints are being changed that would alter the dynamic response of plant equipment. As such, the proposed amendment will not introduce any new failure modes. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed amendment provides for an alternative means of testing the ERVs, in that the testing requirements will be satisfied by a combination of required testing in accordance with the Inservice Testing Program (controlled in accordance with TS administrative controls) and the revised TS SRs. The proposed changes will provide a complete verification of the functional capability of the ERVs by performing a series of tests, inspections, and maintenance activities without opening the valves with reactor steam while installed in the plant. The alternative testing and associated programmatic controls will provide an equivalent level of assurance that the ERVs are capable of performing their intended accident mitigation safety functions. The proposed amendment does not affect the valve setpoints or adversely affect any other operational criteria assumed for accident mitigation. No changes are proposed that alter the setpoints at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation. Moreover, it is expected that the alternative testing methodology will increase the margin of safety by reducing the potential for ERV leakage resulting from testing the ERVs with reactor steam pressure while installed in the plant. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Carey W. Fleming, Senior Counsel, Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, Baltimore, MD 21202. NRC Branch Chief: Nancy L. Salgado. VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 Northern States Power Company— Minnesota, Docket Nos. 50–282 and 50– 306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota Date of amendment request: February 3, 2011. Description of amendment request: The proposed amendments would revise the Technical Specification (TS) 3.8.1, ‘‘AC Sources—Operating’’, Surveillance Requirement 3.8.1.10 footnote, which concerns battery charger modifications to be installed during or prior to the Unit 1 2011 refueling outage. The proposed change will allow use of different battery charger modifications to those considered when the footnote was added to the TS. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. This license amendment request proposes to revise the footnote to the emergency diesel generator Technical Specification surveillance requirement for loss of offsite power with safety injection actuation. The proposed footnote revision removes some specific requirements for battery charger modifications but will continue to assure that the applicable emergency diesel generator and its associated battery charger perform their required safety functions. The emergency diesel generators and their associated battery chargers are not accident initiators and therefore, these changes do not involve a significant increase [in] the probability of an accident. The proposed changes to the Technical Specification footnote will assure that the emergency diesel generator and the associated battery charger continue to perform their required safety function. Since the emergency diesel generator and the associated battery charger will provide required electrical power as assumed in the accident analyses, the results of the previous accident analyses are not changed and the changes proposed in this license amendment request do not involve a significant increase in the consequences of an accident. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. This license amendment request proposes to revise the footnote to the emergency diesel PO 00000 Frm 00088 Fmt 4703 Sfmt 4703 9827 generator Technical Specification surveillance requirement for loss of offsite power with safety injection actuation. The proposed footnote revision removes some specific requirements for battery charger modifications but will continue to assure that the applicable emergency diesel generator and its associated battery charger perform their required safety functions. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. The proposed change does involve modification of plant battery chargers, however, failures of battery chargers has been previously considered and bounded by assuming one safety related train of equipment fails. The modified battery chargers do not create new failure modes or mechanisms and no new accident precursors are generated. Surveillance testing requirements for the emergency diesel generator and battery charger will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are functional. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. This license amendment request proposes to revise the footnote to the emergency diesel generator Technical Specification surveillance requirement for loss of offsite power with safety injection actuation. The proposed footnote revision removes some specific requirements for battery charger modifications but will continue to assure that the applicable emergency diesel generator and its associated battery charger perform their required safety functions. The proposed Technical Specification footnote change does not affect the availability, operability, or performance of safety-related systems and components: The affected emergency diesel generator and its associated battery will continue to perform their safety functions. The ability of operable structures, systems, and components to perform their designated safety function is unaffected by this proposed change. The proposed change does not involve a significant reduction in a margin of safety because the proposed footnote changes do not reduce the margin of safety that exists in the present Technical Specifications or Updated Safety Analysis Report. The operability requirements of the Technical Specifications are consistent with the initial condition assumptions of the safety analyses and the surveillance testing requirements will continue to demonstrate the operability of the emergency diesel generator. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are E:\FR\FM\22FEN1.SGM 22FEN1 9828 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration. Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401 NRC Branch Chief: Robert J. Pascarell. mstockstill on DSKH9S0YB1PROD with NOTICES PPL Susquehanna, LLC, Docket Nos. 50– 387 and 50–388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania Date of amendment request: November 10, 2010. Description of amendment request: The change to the PPL Susquehanna, LLC (PPL) Unit 1 and Unit 2 Technical Specification (TS) Surveillance Requirement (SR) 3.4.3.1 ‘‘Safety/Relief Valves (S/RVs)’’ proposes a new safety function lift setpoint lower tolerance for the S/RVs. The proposed change will revise the lower tolerances from ¥3% to ¥5%. This change would be limited to the lower tolerances and does not affect the upper tolerances. This change only applies to the lower as-found tolerance and not to the as-left tolerance, which will remain unchanged at ±1% of the safety lift setpoint. The asfound tolerances are used for determining past operability and to increase sample sizes for S/RV testing should the upper tolerance be exceeded. There will be no revision to the actual setpoints of the valves installed in the plant due to this change. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. This change has no influence on the probability or consequences of any accident previously evaluated. The lower setpoint tolerance change does not affect the operation of the valves and it does not change the as-left setpoint tolerance. The change only affects the lower tolerance for opening the valve and does not change the upper tolerance, which is the limit that protects from overpressurization. The proposed action does not involve physical changes to the valves, nor does it change the safety function of the valves. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions and no changes to existing structures, systems, or components. The proposed action does not change any other behavior or operation of any S/RVs, VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 and, therefore, has no significant impact on reactor operation. It also has no significant impact on response to any perturbation of reactor operation including transients and accidents previously analyzed in the Final Safety Analysis Report (FSAR). Therefore, the proposed amendment does not result in a significant increase in the probability or consequences of any previously evaluated accident. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed lower setpoint tolerance change only affects the criteria to determine when an as-found S/RV test is considered to be acceptable. This change does not affect the criteria for the upper setpoint tolerance. The proposed lower setpoint tolerance change does not adversely affect the operation of any safety-related components or equipment. Since the proposed action does not involve hardware changes, significant changes to the operation of any systems or components, nor change to existing structures, systems, or components, there is no possibility that a new or different kind of accident is created. The proposed change does not involve physical changes to the S/RVs, nor does it change the safety function of the S/RVs. The proposed change does not require any physical change or alteration of any existing plant equipment. No new or different equipment is being installed, and installed equipment is not being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated. This change does not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the FSAR. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. The proposed lower setpoint tolerance change only affects the criteria to determine when an as-found S/RV test is considered to be acceptable. This change does not affect the criteria for the upper setpoint tolerance. The TS setpoints for the S/RVs are not changed. The as-left setpoint tolerances are not changed by this proposed change. The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not significantly impact the condition or PO 00000 Frm 00089 Fmt 4703 Sfmt 4703 performance of structures, systems, and components relied upon for accident mitigation. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101–1179. NRC Branch Chief : Nancy L. Salgado. Southern Nuclear Operating Company, Inc., Docket Nos. 50–424 and 50–425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia, and Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50– 321 and 50–366, Edwin I. Hatch Nuclear Plant, Unit 1 and 2, Appling County, Georgia Date of amendment request: December 16, 2010. Description of amendment request: The proposed amendments would revise Technical Specification (TS) Section 2.0 ‘‘Safety Limits.’’ Specifically, the removal of the requirement to report a Safety Limit Violation, that is redundant to existing regulations, Title 10 of the Code of Federal Regulations (10 CFR), Part 50.36(c)(8) ‘‘Written Reports.’’ The proposed change is described in Technical Specification Task Force Traveler TSTF–5–A, Revision 1, ‘‘Delete Safety Limit Violation Notification Requirements,’’ (Agencywide Documents Access and Management System (ADAMS) Accession No. ML052010227), and was described in the Notice of Availability published in the Federal Register (FR) on November 4, 2005 (70 FR 67202). The proposed changes are consistent with the NRC-approved TSTF–5–A, Revision 1. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. E:\FR\FM\22FEN1.SGM 22FEN1 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices The proposed change to remove the duplicative safety limit reporting, notification, and restart constraint requirements from the TS does not affect the plant or operation of the plant. The change simply removes duplicative information from the TS that is covered in the NRC regulations. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated? Response: No. The proposed change to remove the duplicative safety limit reporting, notification, and restart constraint requirements from the TS does not introduce any new accident scenarios, failure mechanisms, or limiting single failures. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safetyrelated system or component and does not challenge the performance or integrity of any safety related system. This change is considered an administrative action to remove duplicative reporting, notification, and restart constraint requirements. Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes are administrative and do not involve any reduction in a margin of safety. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safety-related system or component and does not [involve a significant reduction in a margin of safety.] mstockstill on DSKH9S0YB1PROD with NOTICES The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308–2216. NRC Branch Chief: Gloria Kulesa. Notice of Issuance of Amendments to Facility Operating Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission’s rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing in connection with these actions was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission’s Public Document Room (PDR), located at One White Flint North, Room O1–F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible from the Agencywide Documents Access and Management System (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/ reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1–800–397–4209, 301– 415–4737 or by e-mail to pdr.resource@nrc.gov. Dairyland Power Cooperative, Docket No. 50–409, La Crosse Boiling Water Reactor, Vernon County, Wisconsin Date of application for amendment: July 28, 2009, and supplemented August 7, 2009, May 19, 2010, and August 12, 2010. Brief description of amendment: The amendment revises the La Crosse Boiling Water Rector (LACBWR) Technical Specifications, in support of the dry cask storage project at LACBWR. Date of issuance: January 25, 2011. PO 00000 Frm 00090 Fmt 4703 Sfmt 4703 9829 Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 71. Facility Operating License No. DPR–7: This amendment revises the Technical Specifications. Date of initial notice in Federal Register: October 6, 2009 (74 FR 51326). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated January 25, 2011. No significant hazards consideration comments received: No. Dominion Electric Kewaunee, Inc. Docket No. 50–305, Kewaunee Power Station (KPS), Kewaunee County, Wisconsin Date of application for amendment: August 24, 2009 (Agencywide Documents and Management System (ADAMS) Accession No. ML092440398), as supplemented by letters dated October 22, 2009 (ADAMS Accession No. ML093070092), April 13, 2010 (ADAMS Accession Nos. ML101060517 and ML101040090), May 12, 2010 (ADAMS Accession No. ML101380399), July 1, 2010 (ADAMS Accession No. ML101890404), July 16, 2010 (ADAMS Accession No. ML102370370), August 18, 2010 (ADAMS Accession No. ML102371064), September 7, 2010 (ADAMS Accession No. ML102730383), September 8, 2010 (ADAMS Accession No. ML102580700), October 15, 2010 (ADAMS Accession No. ML102920037), and December 2, 2010 (ADAMS Accession No. ML103400328). Brief description of amendment: This amendment converts the current technical specifications (CTSs) to the improved TSs (ITSs) and relocates certain requirements to other licenseecontrolled documents. The ITSs are based on NUREG–1431, Rev. 3.0, ‘‘Standard Technical Specifications, Westinghouse Plants,’’ Revision 3.0; ‘‘NRC Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors,’’ dated July 22, 1993 (58 FR 39132); and 10 CFR 50.36, ‘‘Technical Specifications.’’ Technical Specification Task Force changes were also incorporated. The purpose of the conversion is to provide clearer and more readily understandable requirements in the TSs for KPS to ensure safe operation. In addition, the amendment includes a number of issues that were considered beyond the scope of NUREG–1431. Date of issuance: February 2, 2011. Effective date: As of the date of issuance and shall be implemented on or before February 23, 2011. E:\FR\FM\22FEN1.SGM 22FEN1 9830 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices Amendment No.: 207. Facility Operating License No. DPR– 43: Amendment revised the Technical Specifications and License. Date of initial notice in Federal Register: December 15, 2009 (74 FR 66384). The supplements provided, contained clarifying information and did not expand the scope of the application as originally noticed. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated February 2, 2011. No significant hazards consideration comments received: No. mstockstill on DSKH9S0YB1PROD with NOTICES Entergy Nuclear Operations, Inc., Docket No. 50–255, Palisades Nuclear Plant, Van Buren County, Michigan Date of application for amendment: January 27, 2010. Brief description of amendment: The amendment revises Section 2.E. of the Palisades Nuclear Plant (PNP) Renewed Facility Operating License to remove the name of the former operator of the plant in the title of the PNP physical security plan and replace it with Entergy Nuclear. The change also removes the security plan revision number and the date the plan was submitted to the Nuclear Regulatory Commission. Date of issuance: January 25, 2011. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 241. Facility Operating License No. DPR– 20: Amendment revised the Technical Specifications. Public comments requested as to proposed no significant hazards considerations (NSHC): The notice provided an opportunity to submit comments on the Commission’s proposed NSHC determination. No comments have been received. Date of initial notice in Federal Register: November 18, 2010 (75 FR 70708), followed by the repeat biweekly notice in the Federal Register on January 25, 2011 (76 FR 4389). The Commission’s related evaluation of the amendment, state consultation, and final NSHC determination are contained in a Safety Evaluation dated January 25, 2011. Attorney for licensee: Mr. William Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601. NRC Branch Chief: Robert J. Pascarelli. VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 Entergy Nuclear Operations, Inc., Docket No. 50–293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts Date of application for amendment: January 24, 2010, as supplemented by letters dated September 7 and November 4, 2010. Brief description of amendment: This amendment request would revise the Technical Specifications (TSs) Section 1.0, Definitions, TS Section 3.6, Primary System Boundary Specifications 3.6.A, and TS Programs and Manuals Section 5.5, to include reference to the Pressure and Temperature Limits Report (PTLR). The proposed PTLR would include revised 43 effective full-power years pressure-temperature curves, neutron fluence, and adjusted reference temperature values. Date of issuance: January 26, 2011. Effective date: As of the date of issuance, and shall be implemented within 60 days. Amendment No.: 234. Facility Operating License No. DPR– 35: The amendment revised the License and Technical Specifications. Date of initial notice in Federal Register: April 6, 2010 (75 FR 17443). The supplemental letters dated September 7 and November 4, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of this amendment is contained in a Safety Evaluation dated January 26, 2011. No significant hazards consideration comments received: No. Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50–271, Vermont Yankee Nuclear Power Station, Vernon, Vermont Date of amendment request: April 13, 2010 as supplemented by letter dated. February 2, 2011. Description of amendment request: The amendment would revise Technical Specification (TS) to update the Table of Contents and the Applicability and Objective portions of TS 4.12 as a result of changes made by License Amendment Nos. 230 and 239 and to revise wording in TS 3.7.A.8. The changes are considered administrative in nature and do not materially change any technical requirement. Date of Issuance: February 9, 2011. Effective date: As of the date of issuance, and shall be implemented within 60 days. PO 00000 Frm 00091 Fmt 4703 Sfmt 4703 Amendment No.: 245. Facility Operating License No. DPR– 28: Amendment revised the License and Technical Specifications. Date of initial notice in Federal Register: June 29, 2010 (75 FR 37474). The supplement letter dated February 2, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of this amendment is contained in a Safety Evaluation dated February 9, 2011. No significant hazards consideration comments received: No. Entergy Operations, Inc., Docket No. 50– 382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: July 20, 2010. Brief description of amendment: The amendment revised Technical Specification (TS) 3.7.1.2, ‘‘Emergency Feedwater System,’’ Limiting Condition for Operation (LCO) 3/4.7.1.2, ‘‘Emergency Feedwater,’’ to clarify the acceptability of transitioning from Mode 4, Hot Shutdown, to Mode 3, Hot Standby, with the turbine-driven emergency feedwater (EFW) pump inoperable but available. The amendment granted an exception to TS LCO 3.0.4 and Surveillance Requirement 4.0.4 allowing entry into operational Mode 3 with TS LCO equipment, the turbine-driven EFW pump, associated with a shutdown action inoperable. Date of issuance: January 31, 2011. Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance. Amendment No.: 230. Facility Operating License No. NPF– 38: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: September 21, 2010 (75 FR 57523). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2011. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50–237 and 50–249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Date of amendment request: February 4, 2010 as supplemented by letters E:\FR\FM\22FEN1.SGM 22FEN1 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices mstockstill on DSKH9S0YB1PROD with NOTICES dated September 15, 2010, October 6, 2010, and December 13, 2010. Description of amendment request: The proposed amendments would revise Technical Specification (TS) 3.3.6.1, ‘‘Primary Containment Isolation Instrumentation,’’ ‘‘Table 3.3.6.1–1, ‘‘Primary Containment Isolation Instrumentation,’’ Function 6.a ‘‘Shutdown Cooling System Isolation, Recirculation Line Water Temperature— High,’’ to enable implementation with reactor pressure-based isolation instrumentation, for the Dresden Nuclear Power Station, Units 2 and 3. Date of issuance: February 7, 2011. Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment Nos.: 236/229. Facility Operating License Nos. DPR– 19 and DPR–25: The amendment revised the Technical Specifications and License. Date of initial notice in Federal Register: April 20, 2010 (75 FR 20635). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated February 7, 2011. No significant hazards consideration comments received: No. Exelon Generation Company, LLC, Docket Nos. 50–373 and 50–374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Date of application for amendments: Dated October 5, 2009 as supplemented by letters dated June 10, November 23, December 14, and December 22, 2010, and January 11, 24, and 28, 2011. Brief description of amendments: The proposed amendment would revise Technical Specification (TS) 4.3.1, ‘‘Criticality,’’ to address a nonconservative TS. The proposed change addresses the Boraflex degradation issue in the LSCS Unit 2 spent fuel storage racks by revising TS Section 4.3.1 to allow the use of NETCO–SNAP–IN® rack inserts in LSCS Unit 2 spent fuel storage rack cells as a replacement for the neutron absorbing properties of the existing Boraflex panels. Date of issuance: January 28, 2011. Effective date: As of the date of issuance and shall be implemented within 120 days after the end of Unit 2 refueling outage 13. Amendment Nos.: 199 and 186. Facility Operating License Nos. NPF– 11 and NPF–18: The amendments revised the Technical Specifications and License. Date of initial notice in Federal Register: January 5, 2010 (75 FR 463). The June 10, November 23, December 14, and December 22, 2010, and January VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 11, 24, and 28, 2011, submittals contained clarifying information and did not change the NRC staff’s initial proposed finding of no significant hazards consideration. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated January 28, 2011. No significant hazards consideration comments received: No. FirstEnergy Nuclear Operating Company, et al., Docket No. 50–346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio Date of amendment request: April 15, 2009, as supplemented by letters dated December 18, 2009, October 8, 2010 and January 10, 2011. Brief description of amendment request: The amendment request and proposed exemption request were to incorporate a new methodology for the development of Reactor Coolant System (RCS) pressure-temperature limits into Technical Specification (TS) 5.6.4, ‘‘Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).’’ The amendment also requested a revision to the period of validity of the analysis for the low temperature overpressure protection (LTOP) system contained in Operating License Condition 2.C(3)(d). An associated revision to the Technical Specification Basis 3.4.12 ‘‘Low Temperature Overpressure Protection (LTOP)’’ supports the change to the operating license condition. Date of issuance: January 28, 2011. Effective date: As of the date of issuance and shall be implemented within 90 days. Amendment No.: 282. Facility Operating License No. NPF–3: The amendment revised the TS and license. Date of initial notice in Federal Register: June 16, 2009 (72 FR 28577). The supplemental letters contained clarifying information, did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 2011. No significant hazards consideration comments received: No. Nine Mile Point Nuclear Station, LLC, Docket No. 50–410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York Date of application for amendment: December 9, 2009. PO 00000 Frm 00092 Fmt 4703 Sfmt 4703 9831 Brief description of amendment: The amendment changes the NMP2 Technical Specification (TS) 3.8.4, ‘‘DC Sources—Operating,’’ to remove the Mode restrictions for performance of TS Surveillance Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division 3 direct current (DC) electrical power subsystem battery. The Division 3 DC electrical power subsystem feeds emergency DC loads associated with the high-pressure core spray (HPCS) system. These surveillances verify that the battery capacity is adequate for the battery to perform its required functions. The amendment removes these Mode restrictions for the Division 3 battery, thereby allowing performance of the SRs during Mode 1, 2, or 3 in conjunction with scheduled HPCS system outages. Date of issuance: January 31, 2011. Effective date: As of the date of issuance to be implemented within 90 days. Amendment No.: 136. Renewed Facility Operating License No. NPF–069: The amendment revises the License and TSs. Date of initial notice in Federal Register: April 6, 2010 (75 FR 17444). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2011. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Docket Nos. 50–424 and 50–425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of application for amendments: February 2, 2010. Brief description of amendments: The amendments revised the Technical Specifications (TSs) Table 3.3.1–1 ‘‘Reactor Trip System Instrumentation [RTS],’’ Function 3, ‘‘Power Range Neutron Flux High Positive Rate.’’ Specifically, the revision added surveillance requirement 3.3.1.15 to verify the RTS response time is within limits. Date of issuance: February 7, 2011. Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. Amendment Nos.: 159 and 141. Facility Operating License Nos. NPF– 68 and NPF–81: Amendments revised the licenses and the TSs. Date of initial notice in Federal Register: May 4, 2010 (75 FR 23817). The supplement dated October 29, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC E:\FR\FM\22FEN1.SGM 22FEN1 9832 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated February 7, 2011. No significant hazards consideration comments received: No. NUCLEAR REGULATORY COMMISSION Virginia Electric and Power Company, Docket Nos. 50–338 and 50–339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia I Date of application for amendment: March 30, 2010. Brief description of amendment: The amendments revised the North Anna Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04–10, ‘‘Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.’’ Date of issuance: January 31, 2011. Effective date: As of the date of issuance and shall be implemented within 180 days from the date of issuance. Amendment Nos.: 262 and 243. Renewed Facility Operating License Nos. NPF–4 and NPF–7: Amendments changed the licenses and the technical specifications. Date of initial notice in Federal Register: May 18, 2010 (75 FR 27833). The supplements dated August 30, 2010, and January 18, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated January 31, 2011. No significant hazards consideration comments received: No. mstockstill on DSKH9S0YB1PROD with NOTICES Dated at Rockville, Maryland, this 10th day of February 2011. For the Nuclear Regulatory Commission. Joseph G. Giitter, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2011–3721 Filed 2–18–11; 8:45 am] BILLING CODE 7590–01–P VerDate Mar<15>2010 16:51 Feb 18, 2011 Jkt 223001 [Docket No. 03034542; License No: 37– 30412–01; EA–10–077; NRC–2011–0041] In the Matter of Superior Well Services, Ltd. Indiana, PA; Confirmatory Order Modifying License (Effective Immediately) Superior Well Services, Ltd. (SWS or Licensee) is the holder of radioactive material License No. 37–30412–01 issued by the U.S. Nuclear Regulatory Commission (NRC or agency) pursuant to 10 CFR Part 30. The license authorizes the possession, storage, and use of radioactive sources for oil and gas well logging at the Licensee’s facilities in Buckhannon, West Virginia, Sophia, West Virginia, and Gaylord, Michigan, and at temporary jobsites within the NRC’s jurisdiction, in accordance with conditions specified therein. II On October 21, 2010, the NRC issued a Notice of Violation (Notice) and Proposed Imposition of Civil Penalty (CP) in the amount of $34,000 for five violations that were categorized into two severity level (SL) III problems. The violations were identified during an NRC inspection as well as an investigation conducted by the NRC Office of Investigations (OI). (Reference: NRC Inspection Report No. 03034542/ 2009001 and OI Investigation Report No. 1–2009–035). The violations were also discussed at a predecisional enforcement conference (PEC) on September 2, 2010. The first SL III problem described in the Notice related to an event that occurred on September 20, 2008, when two well-logging sealed sources fell off of a company truck during transport. One violation involved the failure to secure the packages containing the licensed material from shifting during transport. On September 20, 2008, when the truck transporting these sources reportedly hit a large pothole, the weld securing the source plate to the truck broke, and the sources fell off of the truck and remained unattended by the side of a public highway. The second violation involved the failure to control and maintain constant surveillance of the sources while they were on the highway (an unrestricted area). Since SWS did not recognize that the sources had fallen out of the truck until the truck reached its destination at the SWS facility in Buckhannon, WV, the sources remained unattended for approximately ninety minutes until SWS personnel PO 00000 Frm 00093 Fmt 4703 Sfmt 4703 located and retrieved the sources. The third violation involved the failure to immediately report this occurrence by telephone to the NRC Operations Center. The involved SWS employees, including the site Radiation Safety Officer (RSO) for the associated SWS facility, did not recognize the need to report this event to the NRC. As a result, SWS did not provide the required immediate telephone notification of this event to the NRC Operations Center until July 23, 2009, after an NRC inspector informed SWS of the reportability requirement while conducting a routine inspection. The second SL III problem described in the Notice involved SWS’s failure to conduct required radiological surveys of vehicles before transporting licensed material and the deliberate falsification of survey records for these vehicles. Specifically, former SWS employees informed the NRC inspector and investigator that on numerous occasions, they did not perform the surveys and they instead completed the survey forms by copying data from previously completed forms. The employees’ failure to perform the required radiological surveys of vehicles prevented SWS from assuring that the dose rates inside and outside the trucks did not exceed limits set by the NRC and the U.S. Department of Transportation. The employees who admitted to the NRC that they had falsified survey records indicated that they did so because they did not know how to use the survey instruments. III In response to the October 21, 2010, NRC letter, SWS requested the use of the NRC’s Alternative Dispute Resolution (ADR) process to resolve differences it had with the NRC regarding the Notice. ADR is a process in which a neutral professional mediator with no decision-making authority assists the parties in reaching an agreement to resolve any differences regarding the enforcement action. On January 4, 2011, the NRC and SWS met in an ADR mediation session, arranged through Cornell University’s Scheinman Institute on Conflict Resolution. During that ADR mediation session, an agreement in principle was reached. This Confirmatory Order is the result of that agreement, the elements of which consisted of the following: 1. SWS did not take issue with the NRC conclusion set forth in the October 21, 2010, letter and enclosed Notice that the subject violations regarding the temporary loss of two well-logging sources occurred as identified. Further, SWS did not take issue with the NRC E:\FR\FM\22FEN1.SGM 22FEN1

Agencies

[Federal Register Volume 76, Number 35 (Tuesday, February 22, 2011)]
[Notices]
[Pages 9821-9832]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2011-3721]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0040]


Biweekly Notice Applications and Amendments to Facility Operating 
Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC) is publishing this regular biweekly notice. The Act 
requires the Commission publish notice of any amendments issued, or 
proposed to be issued and grants the Commission the authority to issue 
and make immediately effective any amendment to an operating license 
upon a determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 27, 2011, to February 10, 2011. The 
last biweekly notice was published on February 8, 2011 (76 FR 6830).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR), Section 50.92, this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's

[[Page 9822]]

``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR part 
2. Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or a presiding officer designated by the Commission or by 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, any hearing held 
would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
hearing.docket@nrc.gov, or by telephone at (301) 415-1677, to request 
(1) a digital ID certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
https://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through EIE, users will be required to install a Web 
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser 
plug-in, is available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
E-Filing system also distributes an e-mail notice that provides access 
to the

[[Page 9823]]

document to the NRC Office of the General Counsel and any others who 
have advised the Office of the Secretary that they wish to participate 
in the proceeding, so that the filer need not serve the documents on 
those participants separately. Therefore, applicants and other 
participants (or their counsel or representative) must apply for and 
receive a digital ID certificate before a hearing request/petition to 
intervene is filed so that they can obtain access to the document via 
the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/EHD/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. With 
respect to copyrighted works, except for limited excerpts that serve 
the purpose of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the ADAMS 
Public Electronic Reading Room on the Internet at the NRC Web site, 
https://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: January 13, 2011.
    Description of amendment request: The proposed amendment would 
modify the Facility Operating License (FOL) by deleting references to 
specific Safety Evaluation Reports (SER), Technical Specification (TS) 
Amendments, and Exemptions from License Condition 2.C(3), Fire 
Protection, and replacing them with the words ``as supplemented.'' This 
is an administrative amendment to the FOL.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed FOL change is administrative and does not involve a 
plant or design function change. It has no effect on reactor 
operation or accident analyses, and thus, the proposed FOL change 
does not increase the probability or consequence of an accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed FOL change is administrative and does not involve a 
plant or design function change. Because the proposed amendment 
would not change the design, configuration, or method of operation 
of the plant, it would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed FOL change is administrative and does not involve a 
plant or design function change. No design or safety margin is 
involved. Therefore, the proposed change does not involve a 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy L. Salgado.

Entergy Nuclear Vermont Yankee (VY), LLC and Entergy Nuclear 
Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power 
Station, Vernon, Vermont

    Date of amendment request: December 21, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Section 3.6.A ``Pressure and 
Temperature Limitation'' to reflect the pressure and temperature (P-T) 
limits for the reactor coolant system through, approximately the end of 
the prospective 20-year

[[Page 9824]]

renewed license period, depending on the plant capacity factor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the period of applicability of the 
P-T limits. The technical bases for the new period of applicability 
have been previously reviewed and approved by the NRC as discussed 
in the submittal. Because the applicable regulatory requirements 
continue to be met, the change does not significantly increase the 
probability of any accident previously evaluated. The proposed 
change provides the same assurance of RPV integrity as previously 
provided.
    The change will require that the reactor pressure vessel and 
interfacing coolant system continue to be operated within their 
design, operational or testing limits. Also, the change will not 
alter any assumptions previously made in evaluating the radiological 
consequences of accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a modification of the 
design of plant structures, systems, or components. The change will 
not impact the manner in which the plant is operated and will not 
degrade the reliability of structures, systems, or components 
important to safety as equipment protection features will not be 
deleted or modified, equipment redundancy or independence will not 
be reduced, supporting system performance will not be affected and 
no severe testing of equipment will be imposed. No new failure modes 
or mechanisms will be introduced as a result of this proposed 
change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Appendix G to 10 CFR 50 describes the conditions that require 
pressure-temperature (P-T) limits and provides the general bases for 
these limits. Operating limits based on the criteria of Appendix G, 
as defined by applicable regulations, codes and standards, provide 
reasonable assurance that non-ductile or rapidly propagating failure 
will not occur. The P-T limits are prescribed for all plant modes to 
avoid encountering pressure, temperature and temperature rate of 
change conditions that might cause undetected flaws to propagate and 
cause non-ductile failure of the reactor coolant pressure boundary. 
Calculation of P-T limits in accordance with the criteria of 
Appendix G to 10 CFR 50 and applicable regulatory requirements 
ensures that adequate margins of safety are maintained and there is 
no significant reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. There is no change or impact on any safety 
analysis assumption or in any other parameter affecting the course 
of an accident analysis supporting the basis of any Technical 
Specification. The proposed change does not involve any increase in 
calculated off-site dose consequences.
    Therefore, operation of VY in accordance with the proposed 
amendment will not involve a significant reduction in a margin to 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy Salgado.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio

    Date of amendment request: December 15, 2010.
    Description of amendment request: The proposed amendment would 
modify the requirements for testing control rod scram times following 
fuel movement within the reactor pressure vessel by incorporating 
Nuclear Regulatory Commission (NRC) approved Technical Specification 
Task Force (TSTF) change traveler TSTF-222-A, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (CFR) 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration which is presented 
below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The control rod drive system is not an initiator to any accident 
sequence analyzed in the PNPP Updated Final Safety Analysis Report 
(USAR), including Appendix 15C, ``Anticipated Transients Without 
Scram (ATWS).'' The proposed TS changes improve existing 
surveillance requirements by eliminating unnecessary control rob 
scram time testing, while continuing to provide adequate assurance 
of control rod performance for those control rods in core cells in 
which fuel is moved or replaced, or control rod maintenance was 
performed.
    Historically, testing results indicate the control rod drive 
system is highly reliable. Since the fall 1996 implementation of 
Improved Technical Specifications, during 6036 control rod tests 
covering all 177 control rods, only 7 control rod tests (0.12 
percent) yielded results slower than the required insertion time 
limit, and no control rods were inoperable as a result of scream 
time testing. All seven slow insertion time test results have been 
attributed to control rod scream solenoid pilot valves (SSPVs). 
These seven slow tests occurred prior to May 1999, and prior to a 
control rod SSPV upgrade program during which all 177 SSPV's were 
replaced.
    As such, this type of change does not affect initiators of 
analyzed events and does not affect the mitigation of any accidents 
or transients.
    Therefore, the proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS changes do not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints affected by the changes at which protective or 
mitigative actions are initiated. The changes will not alter the 
manner in which equipment operation is initiated, nor will the 
functional demands on credited equipment be changed. No alterations 
in the procedures that ensure the plant remains within analyzed 
limits are being proposed, and no changes are being made to the 
procedures relied upon to respond to an off-normal event as 
described in the USAR. This change does not alter assumptions made 
in the safety analysis and licensing basis. As such, no new failures 
modes are being introduced. Accordingly, the proposed changes do not 
create any new credible failure mechanisms, malfunction, or accident 
initiators not previously considered in PNPP design and licensing 
basis.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to the ability of the fission 
product barriers to perform their design functions during and 
following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment. This 
request does not involve a change to the fuel cladding, the reactor 
coolant system, or the containment.
    The proposed TS changes associated with TSTF-222-1 modify 
current frequency requirements for scram time testing control rods 
following refueling outages and for control rod requiring testing 
due to work

[[Page 9825]]

activities. Scram times for control rods not affected by fuel 
movement or control rod maintenance remain unaffected.
    The proposed TS changes have no affect on any safety analysis 
assumptions or methods of performing safety analyses. The changes do 
not adversely affect system design or operational requirements, and 
the equipment continues to be tested in a manner and at a frequency 
necessary to provide confidence that the equipment can perform its 
intended safety functions.
    Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop. A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Robert D. Carlson.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio

    Date of amendment request: December 15, 2010
    Description of amendment request: The proposed amendment would 
revise the required testing frequency of Surveillance Requirement (SR) 
3.1.4.2 from ``120 days cumulative operation in MODE 1'' to ``200 days 
cumulative operation in MODE 1'' by incorporating Nuclear Regulatory 
Commission (NRC) approved Technical Specification Task Force (TSTF) 
change traveler TSTF-460, Revision 0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable.
    Therefore, the proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Robert D. Carlson.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: October 15, 2010.
    Description of amendment request: The proposed amendment would 
revise Operating License No. DPR-49 by modifying the License to delete 
the parent guarantee License Condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment is an administrative change deleting the 
parent guarantee License Condition, as well as other minor editorial 
changes in format. Deletion of this License Condition does not 
involve any modifications to the safety-related structures, systems 
or components (SSCs). Deletion of this License Condition will not 
alter previously evaluated Final Safety Analysis Report (FSAR) 
design basis accident analysis assumptions, add any accident 
initiators, or affect the function of the plant safety-related SSCs 
as to how they are operated, maintained, modified, tested, or 
inspected. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment only deletes the parent guarantee License 
Condition and makes other minor editorial changes. Deletion of this 
License Condition does not result in the need for any new or 
different FSAR design basis accident analysis. It does not introduce 
new equipment that could create a new or different kind of accident, 
and no new equipment failure modes are created. As a result, no new 
accident scenarios, failure mechanisms, or limiting single failures 
are introduced as a result of this proposed amendment. Therefore, 
the proposed amendment does not create a possibility for an accident 
of a new or different type than those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety is associated with the confidence in the 
ability of the fission product barriers (i.e., fuel cladding, 
reactor coolant pressure boundary, and containment structure) to 
limit the level of radiation to the public. The proposed amendment 
would not alter the way any safety-related SSC functions and would 
not alter the way the plant is operated. The amendment only involves 
deletion of the parent guarantee License Condition and minor 
editorial changes. The proposed amendment would not introduce any 
new uncertainties or change any existing uncertainties associated 
with any safety limit. The proposed amendment would have no impact 
on the structural integrity of the fuel cladding, reactor coolant 
pressure boundary, or containment structure. Based on the above 
considerations, the proposed amendment would not degrade the 
confidence in the ability of the fission product barriers to limit 
the level of radiation to the public. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Marjan Mashhadi, Florida Power & Light 
Company, 801 Pennsylvania Avenue, NW., Suite 220, Washington, DC 20004.

[[Page 9826]]

    NRC Branch Chief: Robert J. Pascarelli.

Indiana Michigan Power Company (the licensee), Docket No. 50-315, 
Donald C. Cook Nuclear Plant, Unit 1 (DCCNP-1), Berrien County, 
Michigan

    Date of amendment request: December 16, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.2.1, adding Optimized ZIRLO \TM\ 
fuel rods to the fuel matrix in addition to Zircaloy or ZIRLO fuel rods 
that are currently in use. The proposed amendment would also add a 
Westinghouse topical report regarding Optimized ZIRLO \TM\ as reference 
8 in TS 5.6.5.b, which lists the analytical methods used to determine 
the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow the use of Optimized ZIRLO \TM\ 
clad nuclear fuel in the reactors. The NRC approved topical report 
WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized ZIRLO 
\TM\,'' prepared by Westinghouse Electric Company LLC 
(Westinghouse), addresses Optimized ZIRLO \TM\ and demonstrates that 
Optimized ZIRLO \TM\ has essentially the same properties as 
currently licensed ZIRLO \TM\. The fuel cladding itself is not an 
accident initiator and does not affect accident probability. Use of 
Optimized ZIRLO \TM\ fuel cladding has been shown to meet all 10 CFR 
50.46 acceptance criteria and, therefore, will not increase the 
consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Use of Optimized ZIRLO \TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical Report 
WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material 
properties of Optimized ZIRLO \TM\ are similar to those of standard 
ZIRLO \TM\. Therefore, Optimized ZIRLO \TM\ fuel rod cladding will 
perform similarly to those fabricated from standard ZIRLO \TM\, thus 
precluding the possibility of the fuel becoming an accident 
initiator and causing a new or different type of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the Optimized ZIRLO \TM\ are not 
significantly different from those of standard ZIRLO \TM\. Optimized 
ZIRLO \TM\ is expected to perform similarly to standard ZIRLO \TM\ 
for all normal operating and accident scenarios, including both loss 
of coolant accident (LOCA) and non-LOCA scenarios. For LOCA 
scenarios, where the slight difference in Optimized ZIRLO \TM\ 
material properties relative to standard ZIRLO \TM\ could have some 
impact on the overall accident scenario, plant-specific LOCA 
analyses using Optimized ZIRLO \TM\ properties will be performed 
prior to the use of fuel assemblies with fuel rods containing 
Optimized ZIRLO \TM\. These LOCA analyses will demonstrate that the 
acceptance criteria of 10 CFR 50.46 will be satisfied when Optimized 
ZIRLO \TM\ fuel rod cladding is implemented.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Senior Nuclear Counsel, 
Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Robert J. Pascarelli.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit 1 (NMP1), Oswego County, New York

    Date of amendment request: September 29, 2010.
    Description of amendment request: The proposed amendment would 
revise the NMP1 Technical Specifications (TSs) Section 3/4.1.5, 
``Solenoid-Actuated Pressure Relief Valves (Automatic Depressurization 
System),'' and 3/4.2.9, ``Pressure Relief Systems--Solenoid-Actuated 
Pressure Relief Valves (Overpressurization),'' to provide for an 
alternative means of testing the main steam electromatic relief valves 
(ERVs). Specifically, the proposed amendment would revise TS 
Surveillance Requirements (SRs) 4.1.5.a and 4.2.9.b to verify each ERV 
actuator strokes when manually actuated at least once each operating 
cycle. The functional testing requirements for the ERVs would be 
described in the Inservice Testing (IST) Program and controlled 
pursuant to TS Administrative Controls Section 6.5.4, ``Inservice 
Testing Program.'' The proposed change would allow demonstration of the 
capability of the valves to perform their safety function without 
requiring the ERVs to be cycled with reactor steam pressure while 
installed in the plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment revises the TS Surveillance Requirements 
(SRs) to provide for an alternative means of testing the main steam 
ERVs. The ERVs perform automatic depressurization system (ADS) and 
overpressurization relief mode safety functions to mitigate the 
consequences of a small break loss of coolant accident (SBLOCA) and 
other accidents and transients. The ERVs are not considered an 
initiator for any accident previously evaluated except for the 
stuck-open ERV event, which is evaluated in Section XV-B.3.11 of the 
NMP1 Updated Final Safety Analysis Report (UFSAR). The proposed 
amendment would allow demonstration of the capability of the valves 
to perform their safety function through a series of tests, 
inspections, and maintenance activities without requiring the ERVs 
to be cycled with reactor steam pressure while installed in the 
plant, thereby eliminating the possibility of a stuck-open ERV event 
due to testing. Thus, the proposed amendment does not increase the 
probability of a stuck-open ERV event. The testing methodology, 
comprehensive inspections and preventive maintenance, and associated 
programmatic controls will provide an equivalent level of assurance 
that the ERVs are capable of performing their intended accident 
mitigation safety functions and, as such, will have no effect on the 
types or amounts of radiation released or the predicted offsite 
doses in the event of an accident. Accordingly, the proposed 
amendment does not alter the initial conditions, assumptions, or 
conclusions of any accident analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not affect the assumed accident 
performance of the ERVs, or of any plant structure, system, or 
component previously evaluated. The proposed amendment does not 
involve the

[[Page 9827]]

installation of new equipment, and installed equipment is not being 
operated in a new or different manner. The proposed amendment 
provides for an alternative means of testing the ERVs that does not 
involve opening the valves with reactor steam while installed in the 
plant. The alternative testing and associated programmatic controls 
will provide an equivalent level of assurance that the ERVs are 
capable of performing their accident mitigation safety functions. No 
setpoints are being changed that would alter the dynamic response of 
plant equipment. As such, the proposed amendment will not introduce 
any new failure modes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment provides for an alternative means of 
testing the ERVs, in that the testing requirements will be satisfied 
by a combination of required testing in accordance with the 
Inservice Testing Program (controlled in accordance with TS 
administrative controls) and the revised TS SRs. The proposed 
changes will provide a complete verification of the functional 
capability of the ERVs by performing a series of tests, inspections, 
and maintenance activities without opening the valves with reactor 
steam while installed in the plant. The alternative testing and 
associated programmatic controls will provide an equivalent level of 
assurance that the ERVs are capable of performing their intended 
accident mitigation safety functions. The proposed amendment does 
not affect the valve setpoints or adversely affect any other 
operational criteria assumed for accident mitigation. No changes are 
proposed that alter the setpoints at which protective actions are 
initiated, and there is no change to the operability requirements 
for equipment assumed to operate for accident mitigation. Moreover, 
it is expected that the alternative testing methodology will 
increase the margin of safety by reducing the potential for ERV 
leakage resulting from testing the ERVs with reactor steam pressure 
while installed in the plant.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey W. Fleming, Senior Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200C, Baltimore, MD 21202.
    NRC Branch Chief: Nancy L. Salgado.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: February 3, 2011.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) 3.8.1, ``AC Sources--
Operating'', Surveillance Requirement 3.8.1.10 footnote, which concerns 
battery charger modifications to be installed during or prior to the 
Unit 1 2011 refueling outage. The proposed change will allow use of 
different battery charger modifications to those considered when the 
footnote was added to the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes to revise the footnote 
to the emergency diesel generator Technical Specification 
surveillance requirement for loss of offsite power with safety 
injection actuation. The proposed footnote revision removes some 
specific requirements for battery charger modifications but will 
continue to assure that the applicable emergency diesel generator 
and its associated battery charger perform their required safety 
functions.
    The emergency diesel generators and their associated battery 
chargers are not accident initiators and therefore, these changes do 
not involve a significant increase [in] the probability of an 
accident.
    The proposed changes to the Technical Specification footnote 
will assure that the emergency diesel generator and the associated 
battery charger continue to perform their required safety function. 
Since the emergency diesel generator and the associated battery 
charger will provide required electrical power as assumed in the 
accident analyses, the results of the previous accident analyses are 
not changed and the changes proposed in this license amendment 
request do not involve a significant increase in the consequences of 
an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This license amendment request proposes to revise the footnote 
to the emergency diesel generator Technical Specification 
surveillance requirement for loss of offsite power with safety 
injection actuation. The proposed footnote revision removes some 
specific requirements for battery charger modifications but will 
continue to assure that the applicable emergency diesel generator 
and its associated battery charger perform their required safety 
functions.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. The proposed change does involve 
modification of plant battery chargers, however, failures of battery 
chargers has been previously considered and bounded by assuming one 
safety related train of equipment fails. The modified battery 
chargers do not create new failure modes or mechanisms and no new 
accident precursors are generated. Surveillance testing requirements 
for the emergency diesel generator and battery charger will continue 
to demonstrate that the Limiting Conditions for Operation are met 
and the system components are functional.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes to revise the footnote 
to the emergency diesel generator Technical Specification 
surveillance requirement for loss of offsite power with safety 
injection actuation. The proposed footnote revision removes some 
specific requirements for battery charger modifications but will 
continue to assure that the applicable emergency diesel generator 
and its associated battery charger perform their required safety 
functions.
    The proposed Technical Specification footnote change does not 
affect the availability, operability, or performance of safety-
related systems and components: The affected emergency diesel 
generator and its associated battery will continue to perform their 
safety functions. The ability of operable structures, systems, and 
components to perform their designated safety function is unaffected 
by this proposed change. The proposed change does not involve a 
significant reduction in a margin of safety because the proposed 
footnote changes do not reduce the margin of safety that exists in 
the present Technical Specifications or Updated Safety Analysis 
Report. The operability requirements of the Technical Specifications 
are consistent with the initial condition assumptions of the safety 
analyses and the surveillance testing requirements will continue to 
demonstrate the operability of the emergency diesel generator.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 9828]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401
    NRC Branch Chief: Robert J. Pascarell.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: November 10, 2010.
    Description of amendment request: The change to the PPL 
Susquehanna, LLC (PPL) Unit 1 and Unit 2 Technical Specification (TS) 
Surveillance Requirement (SR) 3.4.3.1 ``Safety/Relief Valves (S/RVs)'' 
proposes a new safety function lift setpoint lower tolerance for the S/
RVs. The proposed change will revise the lower tolerances from -3% to -
5%. This change would be limited to the lower tolerances and does not 
affect the upper tolerances. This change only applies to the lower as-
found tolerance and not to the as-left tolerance, which will remain 
unchanged at 1% of the safety lift setpoint. The as-found 
tolerances are used for determining past operability and to increase 
sample sizes for S/RV testing should the upper tolerance be exceeded. 
There will be no revision to the actual setpoints of the valves 
installed in the plant due to this change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This change has no influence on the probability or consequences 
of any accident previously evaluated. The lower setpoint tolerance 
change does not affect the operation of the valves and it does not 
change the as-left setpoint tolerance. The change only affects the 
lower tolerance for opening the valve and does not change the upper 
tolerance, which is the limit that protects from overpressurization.
    The proposed action does not involve physical changes to the 
valves, nor does it change the safety function of the valves. The 
proposed TS revision involves no significant changes to the 
operation of any systems or components in normal or accident 
operating conditions and no changes to existing structures, systems, 
or components.
    The proposed action does not change any other behavior or 
operation of any S/RVs, and, therefore, has no significant impact on 
reactor operation. It also has no significant impact on response to 
any perturbation of reactor operation including transients and 
accidents previously analyzed in the Final Safety Analysis Report 
(FSAR).
    Therefore, the proposed amendment does not result in a 
significant increase in the probability or consequences of any 
previously evaluated accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed lower setpoint tolerance change only affects the 
criteria to determine when an as-found S/RV test is considered to be 
acceptable. This change does not affect the criteria for the upper 
setpoint tolerance.
    The proposed lower setpoint tolerance change does not adversely 
affect the operation of any safety-related components or equipment. 
Since the proposed action does not involve hardware changes, 
significant changes to the operation of any systems or components, 
nor change to existing structures, systems, or components, there is 
no possibility that a new or different kind of accident is created.
    The proposed change does not involve physical changes to the S/
RVs, nor does it change the safety function of the S/RVs. The 
proposed change does not require any physical change or alteration 
of any existing plant equipment. No new or different equipment is 
being installed, and installed equipment is not being operated in a 
new or different manner. There is no alteration to the parameters 
within which the plant is normally operated. This change does not 
alter the manner in which equipment operation is initiated, nor will 
the functional demands on credited equipment be changed. No 
alterations in the procedures that ensure the plant remains within 
analyzed limits are being proposed, and no changes are being made to 
the procedures relied upon to respond to an off-normal event as 
described in the FSAR. As such, no new failure modes are being 
introduced. The change does not alter assumptions made in the safety 
analysis and licensing basis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed lower setpoint tolerance change only affects the 
criteria to determine when an as-found S/RV test is considered to be 
acceptable. This change does not affect the criteria for the upper 
setpoint tolerance. The TS setpoints for the S/RVs are not changed. 
The as-left setpoint tolerances are not changed by this proposed 
change.
    The margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change does not significantly impact the condition or 
performance of structures, systems, and components relied upon for 
accident mitigation.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief : Nancy L. Salgado.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia, and Southern Nuclear Operating Company, Inc., Georgia Power 
Company, Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin 
I. Hatch Nuclear Plant, Unit 1 and 2, Appling County, Georgia

    Date of amendment request: December 16, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Section 2.0 ``Safety Limits.'' 
Specifically, the removal of the requirement to report a Safety Limit 
Violation, that is redundant to existing regulations, Title 10 of the 
Code of Federal Regulations (10 CFR), Part 50.36(c)(8) ``Written 
Reports.'' The proposed change is described in Technical Specification 
Task Force Traveler TSTF-5-A, Revision 1, ``Delete Safety Limit 
Violation Notification Requirements,'' (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML052010227), and was described 
in the Notice of Availability published in the Federal Register (FR) on 
November 4, 2005 (70 FR 67202). The proposed changes are consistent 
with the NRC-approved TSTF-5-A, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.

[[Page 9829]]

    The proposed change to remove the duplicative safety limit 
reporting, notification, and restart constraint requirements from 
the TS does not affect the plant or operation of the plant. The 
change simply removes duplicative information from the TS that is 
covered in the NRC regulations. Therefore, the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed change to remove the duplicative safety limit 
reporting, notification, and restart constraint requirements from 
the TS does not introduce any new accident scenarios, failure 
mechanisms, or limiting single failures. All systems, structures, 
and components previously required for the mitigation of a transient 
remain capable of fulfilling their intended design functions. The 
proposed change has no adverse effect on any safety-related system 
or component and does not challenge the performance or integrity of 
any safety related system. This change is considered an 
administrative action to remove duplicative reporting, notification, 
and restart constraint requirements. Therefore, this proposed change 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are administrative and do not involve any 
reduction in a margin of safety. All systems, structures, and 
components previously required for the mitigation of a transient 
remain capable of fulfilling their intended design functions. The 
proposed change has no adverse effect on any safety-related system 
or component and does not [involve a significant reduction in a 
margin of safety.]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Gloria Kulesa.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluatio
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