ML15106A916

Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication in April 28, 2015, Federal Register

ML15106A916
Person / Time
Site:	Beaver Valley, Millstone, Indian Point, Catawba, Grand Gulf, Braidwood, Summer, Columbia, Brunswick, Turkey Point, Diablo Canyon, Vogtle, Farley, Cook
Issue date:	04/20/2015
From:	Louise Lund Division of Operating Reactor Licensing
To:
Figueroa S
References
NRC-2015-0104
Download: ML15106A916 (45)

Text

[7590-01-P]

NUCLEAR REGULATORY COMMISSION

[NRC-2015-0104]

Biweekly Notice Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

SUMMARY

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice.

The Act requires the Commission to 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 or combined license, as applicable, 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 April 2, 2015, to April 14, 2015. The last biweekly notice was published on April 14, 2015.

DATES: Comments must be filed by May 28, 2015. A request for a hearing must be filed by June 29, 2015.

2 ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0104. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; e-mail: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.

• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

For additional direction on obtaining information and submitting comments, see Obtaining Information and Submitting Comments in the SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Sandra Figueroa, Office of U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1262, e-mail:

Sandra.Figueroa@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments.

A. Obtaining Information.

Please refer to Docket ID NRC-2015-0104 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

3

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0104.

• NRCs Agencywide Documents Access and Management System (ADAMS):

You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ADAMS Public Documents and then select Begin Web-based ADAMS Search. For problems with ADAMS, please contact the NRCs Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section.

• NRCs PDR: You may examine and purchase copies of public documents at the NRCs PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments.

Please include Docket ID NRC-2015-0104, facility name, unit number(s), application date, and subject in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that

4 they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination.

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commissions regulations in

§ 50.92 of Title 10 of the Code of Federal Regulations (10 CFR), 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

5 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.

A. Opportunity to Request a Hearing and Petition for Leave to Intervene.

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 or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commissions Agency Rules of Practice and Procedure in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRCs PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRCs regulations are accessible electronically from the NRC Library on the NRCs Web site at http://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.

6 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 requestors/petitioners right under the Act to be made a party to the proceeding; 3) the nature and extent of the requestors/petitioners 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 requestors/petitioners 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.

7 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, then any hearing held would take place before the issuance of any amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

B. Electronic Submissions (E-Filing).

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 NRCs 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

8 hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1) a digital identification (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 the NRCs public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing the E-Submittal server are detailed in the NRCs Guidance for Electronic Submission, which is available on the agencys public Web site at http://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 NRCs 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 NRCs online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRCs Web site.

Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRCs public Web site at http://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.

9 Submissions should be in Portable Document Format (PDF) in accordance with the NRCs guidance available on the NRCs public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRCs 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 document to the NRCs 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 NRCs adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the Contact Us link located on the NRCs public Web site at http://www.nrc.gov/site-help/e-submittals.html, by e-mail to 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,

10 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 the NRCs electronic hearing docket which is available to the public at http://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. However, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. 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. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be

11 entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii).

For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRCs PDR. For additional direction on accessing information related to this document, see the Obtaining Information and Submitting Comments section of this document.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit 2, New London County, Connecticut Date of amendment request: October 22, 2014. A publicly-available version is in ADAMS under Accession No. ML14301A112.

Description of amendment request: The amendment would revise the Millstone Power Station, Unit 2 (MPS2) technical specification (TS) by relocating surveillance frequencies to a licensee-controlled program.

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 any accident previously evaluated?

Response: No.

The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the TSs for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As

12 a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

No new or different [kinds of] accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, DNC will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the Surveillance Frequency Control Program. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

13 The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc.,

120 Tredegar Street, RS-2, Richmond, VA 23219.

Acting NRC Branch Chief: Michael I. Dudek.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power Station, Unit 3, New London County, Connecticut Date of amendment request: October 14, 2014. A publicly-available version is in ADAMS under Accession No. ML14294A454.

Description of amendment request: The amendment would revise the Millstone Power Station, Unit 3 (MPS3) Surveillance Requirement (SR) 4.4.4.2 to remove the requirement to perform the surveillance for a pressurizer power-operated relief valve (PORV) block valve that is being maintained closed in accordance with technical specification (TS) 3.4.4 Action a.

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, with NRC staff revisions provided in [brackets], which is presented below:

Criterion 1 Will operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The block valve for the pressurizer PORV is not a potential accident initiator.

Therefore, not requiring a surveillance of the block valve while it is being used to isolate its associated PORV will not increase the probability of an accident

14 previously evaluated. Not requiring the surveillance of the block valve may slightly reduce the probability of a loss of coolant accident from a stuck open PORV since it will eliminate the challenge to the PORV from the pressure transient that results from cycling the block valve.

The PORVs are credited in the MPS3 Final Safety Analysis Report (FSAR),

Chapter 15, Accident Analysis, for event mitigation (Section 15.5.1, Inadvertent Operation of the Emergency Core Cooling System during Power, and Section 15.5.2, CVCS [chemical and volume control system] Malfunction that Increases Reactor Coolant Inventory). Not performing the surveillance on the block valve does not significantly reduce the assurance that the block valve is capable of opening to allow operation of the PORV. The block valves have been demonstrated by operating experience to be reliable and are also subject to the motor-operated valve testing program. Consequently, the proposed amendment does not significantly reduce the confidence that the block valve can be opened to permit automatic or manual actuation of the PORV to depressurize the RCS.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2 Will operation of the facility in accordance with this proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment only affects the performance of the surveillance test for the block valve and does not involve any physical alteration of plant equipment or introduce any operating configurations not previously evaluated.

The pressurizer PORV block valves provide isolation for a postulated stuck-open or leaking PORV. Isolation is satisfied with the block valve closed in accordance with TS 3.4.4 Action a.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3 Will operation of the facility in accordance with this proposed amendment involve a significant reduction in the margin of safety?

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident.

These barriers include the fuel cladding, the reactor coolant system, and the containment system. These barriers are not significantly affected by the changes proposed herein. The margin of safety is established through the design of the

15 plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of setpoints for the actuation of equipment relied upon to respond to an event, and thereby protect the fission product barriers. The proposed amendment to the surveillance requirement for the pressurizer PORV block valve does not affect the assumptions in any accident analysis.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resource Services, Inc.,

120 Tredegar Street, RS-2, Richmond, VA 23219.

Acting NRC Branch Chief: Michael I. Dudek.

Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North Carolina Date of amendment request: January 30, 2015. A publicly-available version is in ADAMS under Accession No. ML15044A198.

Description of amendment request: The amendments would revise the emergency action levels (EALs) from a scheme based on Revision 5 of Nuclear Energy Institute (NEI) 99-01 Methodology for Development of Emergency Action Levels, to a scheme based on NRC-endorsed Revision 6 of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors.

16 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, with NRC staff revisions provided in [brackets], 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 to the BSEP emergency action levels does not impact the physical function of plant structures, systems, or components (SSC) or the manner in which SCCs perform their design function. The proposed change does not authorize the addition of any new plant equipment or systems, nor does it alter the assumptions of any accident analyses. The proposed change does not adversely affect accident initiators or precursors, nor does it alter the design assumptions, conditions, and configuration or the manner in which the plant is operated and maintained.

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 change to BSEPs EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6, does not authorize any physical changes to the plant systems or equipment. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. The proposed change will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The BSEP ERO [Emergency Response Organization]

functions will continue to be performed as required. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from those that have been previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

17 Response: No.

The proposed change to BSEPs EAL scheme does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.

The proposed change does not affect the Technical Specifications or the operating license. There are no changes to setpoints or environmental conditions of any SSC or the manner in which any SSC is operated.

Margins of safety are unaffected by the proposed change to adopt the NEI 99-01, Revision 6, EAL scheme guidance. The applicable requirements or 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.

Therefore, the proposed change does not involve any reduction in a margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: David T. Conley, Associate General Counsel II - Legal Department, Progress Energy Service Company, LLC, P.O. Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Shana R. Helton.

Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of amendment request: February 19, 2015. A publicly-available version is in ADAMS under Accession No. ML15075A021.

Description of amendment request: The amendments would (1) revise Technical Specifications (TSs) by replacing AREVA Topical Report ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation, Revision 0, March 2010, with Revision 1, March 2014, of the same topical report;

18 and (2) revise Appendix B, Additional Conditions, by removing the license condition issued by Amendment Nos. 262 and 290 for Units 1 and Unit 2, respectively.

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 probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The proposed license amendments only involve an update to a currently-approved methodology for determining core operating limits. As such, the proposed license amendments do not involve any plant modifications or operational changes that could affect system reliability or performance, or that could affect the probability of operator error. As such, the proposed changes do not affect any postulated accident precursors. Since no individual precursors of an accident are affected, the proposed license amendments do not involve a significant increase in the probability of a previously analyzed event.

The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences.

AREVA Topical Report ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014, is being adopted to resolve a previously identified concern with the calculation of the K-factor, which is a modelling parameter that characterizes the effect on critical power ratio of radial fuel rod peaking distribution within a fuel bundle. Adoption of AREVA Topical Report ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, also eliminates the need to perform a confirmatory evaluation as described in the Appendix B license condition issued as part of License Amendments 262 and 290 for Units 1 and 2.

Therefore, the license condition is being eliminated.

The adoption of AREVA Topical Report ANP-10298P-A, ACE/ATRIUM 10XM Critical Power Correlation, Revision 1, March 2014, continues to ensure that the SLMCPR [safety limit minimum critical power ratio],

setpoint, and core operating limit values determined using NRC-approved methods continue to satisfy the acceptance criteria that at least 99.9 percent of all fuel rods in the core do not experience boiling transition. Based on these considerations, the proposed change does not

19 involve a significant increase in the consequences of a previously analyzed accident.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation. The proposed amendments do neither.

Core operating limit values are calculated using NRC-approved methodology identified in the TS. AREVA Topical Report ANP-10298PA, Revision 0, is an NRC-approved methodology listed in TS 5.6.5.b for determining core operating limits. Replacing the analytical methodology described in Topical Report ANP-10298PA, Revision 0, with the methodology contained in ANP-10298P-A, Revision 1, will ensure that (1) core operating limits are no longer affected by the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1, and (2) the current level of fuel protection is maintained by continuing to ensure that the fuel design safety criterion is met (i.e., that at least 99.9 percent of all fuel rods in the core do not experience boiling transition if the MCPR [minimum critical power ratio] Safety Limit is not exceeded).

The update of AREVA analytical methodology does not involve any new modes of plant operation or any plant modifications and does not directly or indirectly affect the failure modes of any plant systems or components.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The SLMCPR ensures that at least 99.9 percent of the fuel rods do not experience boiling transition during normal operation and anticipated operational occurrences, if the SLMCPR is not exceeded. Topical Report ANP-10298PA is listed as an NRC-approved analytical method in Technical Specification 5.6.5.b. Replacing the analytical methodology described in Topical Report ANP-10298PA, Revision 0, with the methodology contained in ANP-10298P-A, Revision 1, will ensure that (1) core operating limits are no longer affected by the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274,

20 Revision 1, and (2) the current level of fuel protection is maintained by continuing to ensure that the fuel design safety criterion is met (i.e., that no more than 0.1 percent of the rods are expected to be in boiling transition if the MCPR Safety Limit is not exceeded).

Meeting the fuel design criterion that at least 99.9 percent of all fuel rods in the core do not experience boiling transition and establishing core operating limits ensures the margin of safety required by the fuel design criterion is maintained. Therefore, the proposed amendments do not result in a significant reduction in the margin of safety.

The NRC staff has reviewed the licensees 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 T. Conley, Associate General Counsel II - Legal Department, Progress Energy Service Company, LLC, P.O. Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Shana R. Helton.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of amendment request: December 15, 2014. A publicly-available version is in ADAMS under Accession No. ML14351A069.

Description of amendment request: The proposed amendment would revise the requirements of Technical Specification (TS) 3.6.4.3, Standby Gas Treatment (SGT) System, and TS 3.7.3, Control Room Fresh Air (CRFA) System, to operate the ventilation systems with charcoal filters from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> each month to 15 minutes each month, consistent with Technical Specification Task Force (TSTF) traveler TSTF-522, Revision 0, Revise Ventilation System Surveillance Requirements to Operate for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per Month. The Notice of Availability and

21 model safety evaluation of TSTF-522, Revision 0, were published in the Federal Register on September 20, 2012 (77 FR 58421).

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, with NRC staff revisions provided in [brackets], 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 replaces an existing Surveillance Requirement to operate the BWR [boiling water reactor]/6 SGT System and CRFA Systems equipped with electric heaters for a continuous 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period every 31 days with a requirement to operate the systems for 15 continuous minutes with heaters operating, if needed.

These systems are not accident initiators and therefore, these changes do not involve a significant increase in the probability of an accident. The proposed system and filter testing changes are consistent with current regulatory guidance for these systems and will continue to assure that these systems perform their design function which may include mitigating accidents. Thus the change does not involve a significant increase in the consequences of an accident.

Therefore, it is concluded that this 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.

The proposed change replaces an existing Surveillance Requirement to operate the BWR/6 SGT System and CRFA Systems equipped with electric heaters for a continuous 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period every 31 days with a requirement to operate the systems for 15 continuous minutes with heaters operating, if needed.

The change proposed for these ventilation systems does not change any system operations or maintenance activities. Testing requirements will be revised and will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are capable of performing

22 their intended safety functions. The change does not create new failure modes or mechanisms and no new accident precursors are generated.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change replaces an existing Surveillance Requirement to operate the BWR/6 SGT System and CRFA Systems equipped with electric heaters for a continuous 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period every 31 days with a requirement to operate the systems for 15 continuous minutes with heaters operating, if needed.

The design basis for the ventilation systems heaters is to heat the incoming air which reduces the relative humidity. The heater testing change proposed will continue to demonstrate that the heaters are capable of heating the air and will perform their design function. The proposed change is consistent with [the NRCs] regulatory guidance.

Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Joseph A. Aluise, Associate General Counsel - Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.

NRC Branch Chief: Meena K. Khanna.

23 Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne County, Mississippi Date of amendment request: January 6, 2015, as supplemented by letter dated March 27, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML15006A238 and ML15089A126, respectively.

Description of amendment request: The amendment would revise Technical Specification (TS) 5.6.5.b, Core Operating Limits Report [COLR] by adding the following reference, NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution -

Confirmation Density [DSS-CD].

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. The NRC staff has reviewed the licensees analysis against the standards of 10 CFR 50.92(c). The NRC staffs review 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 NRC staff completed its review of NEDC-33075P-A, Revision 6, General Electric Boiling Water Reactor Detect and Suppress Solution -

Confirmation Density, a licensing topical report (LTR) and issued its safety evaluation on January 25, 2008 (ADAMS Accession No. ML080310388). The NRC staff had concluded that this LTR is acceptable for referencing in licensing applications for nuclear power plants to the extent specified and under the limitations delineated in the accepted versions of the LTR. In addition, by letter dated November 19, 2013, LTR NEDE-33075P, Revision 8, has been approved for use in future licensing actions. The licensee proposes to add NEDC-33075P-A, Revision 8, to TS 5.6.5.b as Reference 27. The licensee demonstrated the applicability of this LTR for the GGNS in its submittal dated September 25, 2013 (ADAMS Accession No. ML13269A140). Adding this approved LTR to the TS 5.6.5.b will allow the licensee to use the approved DSS-CD methodology for preparing the COLR for the Maximum

24 Extended Load Line Limit Analysis Plus (MELLLA+) reloads following the approval of the MELLLA+ license amendment request. As such, adding this reference to TS 5.6.5.b, is administrative in nature.

Therefore, it is concluded that this 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.

The licensee proposes to add LTR NEDC-33075P-A, Revision 8, to TS 5.6.5.b as Reference 27. The licensee demonstrated the applicability of this LTR for the GGNS in its submittal dated September 25, 2013.

Adding this approved LTR to TS 5.6.5.b will allow the licensee to use the approved DSS-CD methodology for preparing the COLR for the MELLLA+ reloads following the approval of the MELLLA+ license amendment request. As such, adding this reference to TS 5.6.5.b, is administrative in nature.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The licensee proposes to add LTR NEDC-33075P-A, Revision 8, to TS 5.6.5.b as Reference 27. The licensee demonstrated the applicability of this LTR for the GGNS in its submittal dated September 25, 2013.

Adding this approved LTR to TS 5.6.5.b will allow the licensee to use the approved DSS-CD methodology for preparing the COLR for the MELLLA+ reloads following the approval of the MELLLA+ license amendment request. As such, adding this reference to TS 5.6.5.b, is administrative in nature.

Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

25 proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Joseph A. Aluise, Associate General Counsel - Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.

NRC Branch Chief: Meena K. Khanna.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Units 3 and 4, Miami-Dade County, Florida Date of amendment request: December 23, 2014. A publicly-available version is in ADAMS under Accession No. ML15029A297.

Description of amendment request: The amendments would modify the Technical Specifications (TSs) related to Completion Times (CTs) for Required Actions (RAs) to provide the option to calculate a longer, risk-informed CT (RICT). A new program, the Risk-Informed Completion Time Program, would be added to TS Section 6.0, Administrative Controls. The methodology for using the RICT Program is described in Nuclear Energy Institute (NEI) 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0 - A (ADAMS Accession No. ML12286A322). Adherence to NEI 06-09 would be required by the RICT Program. The licensee stated that the proposed amendments would be consistent with Technical Specification Task Force (TSTF) Traveler 505 TSTF-505, Revision 1, Provide Risk-Informed Extended Completion Times - RITSTF [Risk Informed TSTF] Initiative 4b (ADAMS Accession No. ML111650552). The licensee requested that not all the modified RAs in TSTF-505 be included in the amendments. The licensee also requested that some plant-specific RAs be included in the amendments that were not included

26 in TSTF-505. The Federal Register notice published on March 15, 2012 (77 FR 15399),

announced the availability of TSTF-505, Revision 1.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee 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 permits the extension of Completion Times provided the associated risk is assessed and managed in accordance with the NRC approved Risk-Informed Completion Time Program. The proposed change does not involve a significant increase in the probability of an accident previously evaluated because the change involves no change to the plant or its modes of operation. The proposed change does not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Completion Time are no different from those during the existing Completion Time.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not change the design, configuration, or method of operation of the plant. The proposed change does not involve a physical alteration of the plant (no new or different kind of equipment will be installed).

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 change involve a significant reduction in the margin of safety?

Response: No.

The proposed change permits the extension of Completion Times provided risk is assessed and managed in accordance with the NRC approved Risk-Informed Completion Time Program. The proposed change implements a risk-informed configuration management program

27 to assure that adequate margins of safety are maintained. Application of these new specifications and the configuration management program considers cumulative effects of multiple systems or components being out of service and does so more effectively than the current TS.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: William S. Blair, Managing Attorney - Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Shana R. Helton.

Pacific Gas and Electric Company (PGandE), Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, California Date of amendment request: February 25, 2015. A publicly-available version is in ADAMS under Accession No. ML15056A773.

Description of amendment request: The amendments propose to incorporate into the licensing basis an analysis of pressurizer reaching a water-solid (filled) condition associated with the main feedwater pipe rupture accident summarized in Updated Final Safety Analysis Report (UFSAR)

Section 15.4.2.2. Further, the proposed amendments involve the addition of time critical operator actions and modifications of the PG&E Design Class I backup nitrogen accumulators, which are credited in the new pressurizer filling analysis.

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:

28

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment provides an analysis of the FLB [feedwater line break] accident assuming the worst-case conditions that could result in pressurizer filling wherein water relief through the PSVs [pressurizer safety valves] may challenge the integrity of the reactor coolant boundary.

The purpose of the pressurizer filling analysis is to determine the operator actions that preclude water relief through the PSVs if a FLB accident has occurred. The pressurizer filling analysis assumes an accident occurs and evaluates the plant response to the accident; therefore, the proposed amendment results in no change in the probability of an accident previously evaluated.

The proposed amendment does not change any design functions of existing structures, systems and components (SSCs) and does not increase the likelihood of the malfunction of an SSC. The operator actions added by the amendment are designed to ensure the capability of SSCs to perform their design function by ensuring a PORV [power operated relief valve] is available to provide reactor coolant pressure relief and by terminating the pressurizer filling event before water is relieved from the PSVs.

Therefore, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

The proposed amendment does not change any design functions of existing SSCs and does not affect the SSCs' operation or ability to perform their design function. The new FLB pressurizer filling analysis identifies operator actions that will prevent water relief through the PSVs.

Simulator runs for the FLB pressurizer filling scenario have demonstrated that operator actions credited in the analysis are consistently completed in time to prevent water relief through the PSVs.

Therefore the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

29 The UFSAR (Section 15.4.2.2.3) currently credits the SSI [spurious safety injection] pressurizer filling analysis (in UFSAR Section 15.2.15.3) for the FLB pressurizer filling condition. The results of the new FLB pressurizer filling analysis indicate the response time for the operator action to ensure a PORV available during a FLB is not bounded by the existing analysis for the SSI pressurizer filling event. In addition, the analysis determined the PORVs need to cycle longer than accommodated by the current nitrogen supply to prevent water relief through the PSVs.

The new analysis identifies operator actions to mitigate the pressurizer filling condition specific to a FLB accident. Simulator runs for a FLB scenario have demonstrated that operator actions credited in the analysis are consistently completed in time to prevent water relief through the PSVs.

The new FLB analysis credits an increased number of PORV water-relief cycles, which will be provided by modifications to increase the nitrogen supply to the PORV[s]. The PORVs have been qualified to perform the increased number of water-relief cycles and are environmentally qualified to withstand the harsh environment that could result from a FLB.

Increasing the required number of PORV water-relief cycles does not alter the overall thermal hydraulic response of the RCS [reactor coolant system] and, therefore, has no effect on overall atmospheric steam releases. The PORV relief is not a source of radiological release since the RCS fluid remains inside containment and therefore is a negligible source of radiological release to the environment.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC proposes to determine that the amendment requests involve no significant hazards consideration.

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: Michael T. Markley.

30 South Carolina Electric and Gas Company Docket Nos.52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, South Carolina Date of amendment request: December 18, 2014. A publicly-available version is in ADAMS under Accession No. ML14353A107.

Description of amendment request: The proposed change would amend Combined License Nos. NPF-93 and NPF-94 for the VCSNS Units 2 and 3 by revising line number information in Tier 1 and promote consistency with the Updated Final Safety Analysis Report (UFSAR) Tier 2 information. The line number information includes the Automatic Depressurization System, the Passive Containment Cooling System, the Passive Core Cooling System, the Normal Residual Heat Removal System, the Containment Air Filtration System, Spent Fuel Pool Cooling System, and the Sanitary Discharge System piping line numbers to reflect the as-designed configuration resulting from changes in piping layout or rerouting.

Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Advanced Passive 1000 design control document (DCD), the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with the provisions of part 52, appendix D, section III.B of Title 10 of the Code of Federal Regulations (10 CFR), Design Certification Rule for the AP1000 Design, Scope and Contents.

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 COL Appendix C Tables and corresponding plant-specific Tier 1 Tables proposed changes involve updating piping line name/number or functional capability requirements. These changes do not affect any

31 system design function. Adding or updating information for existing ASME Section III piping does not involve (i.e., cannot affect) any accident initiating event or component failure, thus, the probabilities of the accidents previously evaluated are not affected. The maximum allowable leakage rate specified in the Technical Specifications is unchanged, and radiological material release source terms are not affected, thus, the radiological releases in the accident analyses are not affected.

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 COL Appendix C Tables and corresponding plant-specific Tier 1 Tables proposed changes to update piping line name/number or functional capability requirements do not adversely affect the design or quality of any structure, system, or component. Adding or updating ASME Section III piping line information for existing process piping lines to a licensing table does not create a new fault or sequence of events that could result in a radioactive material release.

Therefore, the proposed amendment 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.

The COL Appendix C Tables and corresponding plant-specific Tier 1 Tables proposed changes involve updating piping line name/number or functional capability requirements information for new/existing process piping lines. Adding or updating the ASME Section III piping line name/number or functional capability requirements in the tables would not affect any radioactive material barrier. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus, no margin of safety is reduced.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

32 The NRC staff has reviewed the licensees analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue, NW, Washington, DC 20004-2514.

NRC Branch Chief: Lawrence Burkhart.

III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined 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 Commissions rules and regulations. The Commission has made appropriate findings as required by the Act and the Commissions rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, 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

33 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 Commissions related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the Obtaining Information and Submitting Comments section of this document.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of application for amendments: May 31, 2012, as supplemented by letters dated March 13, 2013, and November 25, 2014.

Brief description of amendments: The amendments approve a conditional exception to the end of cycle moderator temperature coefficient surveillance requirement if certain conditions are met.

Date of issuance: April 14, 2015.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

Amendment Nos.: 275, 271, 278, and 258. A publicly-available version of the application is in ADAMS under Accession No. ML12153A328; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, and NPF-17: Amendments revised the Renewed Facility Operating licenses and Technical Specifications.

34 Date of initial notice in Federal Register: February 5, 2013 (78 FR 8198). The licensees March 13, 2013, and November 25, 2014, supplements provided clarifying information that did not change the scope of the proposed amendment as described in the original notice of proposed action published in the Federal Register, and did not change the initial proposed no significant hazards consideration determination.

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated April 14, 2015.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington Date of application for amendment: August 22, 2014, as supplemented by letter dated December 23, 2014.

Brief description of amendment: The amendment changed Technical Specification 3.7.1, Standby Service Water (SW) System and Ultimate Heal Sink (UHS), TS Surveillance Requirement 3.7.1.1 related to verifying that the average water level in the UHS spray ponds is the average of the level in both ponds.

Date of issuance: April 15, 2015.

Effective date: As of its date of issuance and shall be implemented within 10 days from the date of issuance.

Amendment No.: 233. A publicly-available version is in ADAMS under Accession No. ML15076A122; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-21: The amendment revised the Facility Operating License and Technical Specifications.

35 Date of initial notice in Federal Register: September 5, 2014 (79 FR 53085). The supplemental letter dated December 23, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated April 15, 2015.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit 2, Westchester County, New York Date of amendment request: February 12, 2015, as supplemented by letters dated March 10, and April 1, 2015.

Brief description of amendment: The amendment revised the acceptance criteria for Surveillance Requirement 3.1.4.2 for Control Rod G-3. The change defers subsequent testing of Control Rod G-3 until repaired during the next forced outage of sufficient duration prior to the refuel outage of 2016 or during the refuel outage of 2016.

Date of issuance: April 2, 2015.

Effective date: As of the date of issuance, and shall be implemented within 30 days.

Amendment No.: 280. A publicly-available version is in ADAMS under Accession No. ML15083A490; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License Nos. DPR-26: The amendment revised the Facility Operating License and the Technical Specifications.

36 Date of initial notice in Federal Register: March 2, 2015 (80 FR 11236). The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment and final NSHC determination are contained in a Safety Evaluation dated April 2, 2015.

No significant hazards consideration comments received: Yes. The comments are addressed in the Safety Evaluation referenced above.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, Ogle County, Illinois Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-10, 50-237 and 50-249, Dresden Nuclear Power Station, Units 1, 2 and 3, Grundy County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of application for amendments: August 11, 2014 (ADAMS Accession No. ML14224A245).

Brief description of amendments: The amendments revise the description of the Emergency Response Organization requalification training frequency for Exelon personnel defined in

37 Exelons governing Emergency Plans for the named stations from annually to once per calendar year not to exceed 18 months between training sessions.

Date of issuance: April 8, 2015.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment Nos.: 182, 182, 188, 188, 203, 44, 243, 236, 213, 199, 256, and 251. A publicly-available version is in ADAMS under Accession No. ML14323A522. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66, NPF- 62, DPR-2, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, DPR-30: The amendments revised the Licenses.

Date of initial notice in Federal Register: September 30, 2014 (79 FR 58815).

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated April 8, 2015.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey Date amendment request: April 30, 2014, as supplemented by letter dated October 16, 2014.

Brief description of amendment: The amendment revised Oyster Creek Nuclear Generating Station (OCNGS) Technical Specifications (TS) 4.5M, Shock Suppressors (Snubbers), to conform the TS to the revised OCNGS Snubber Inspection Program.

Date of issuance: April 3, 2015.

Effective date: As of the date of issuance and shall be implemented within 60 days.

38 Amendment No.: 286. A publicly-available version is in ADAMS under Accession No. ML15040A721; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License No. DPR-16: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: July 8, 2014 (79 FR 38590). The supplemental letter dated October 16, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated April 3, 2015.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Units 1 and 2, Beaver County, Pennsylvania Date of amendment request: April 16, 2014, as supplemented by letters dated November 4, 2014, and March 23, 2015.

Description of amendment request: The amendment changed the Beaver Valley Power Station Units 1 and 2 (BVPS-1 and BVPS-2) technical specifications (TS). Specifically, the amendment revised TS 5.5.12, Containment Leakage Rate Testing Program, Item a, by deleting reference to the BVPS-1 exemption transmittal letter dated December 5, 1984 (ADAMS Accession No. ML003766713), and requiring compliance with Nuclear Energy Institute (NEI) topical report NEI

39 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, (ADAMS Accession No. ML12221A202) instead of Regulatory Guide 1.163, Performance-Based Containment Leak Test Program, (ADAMS Accession No. ML003740058) including listed exceptions. In summary, the amendment allows extension of the Type A Reactor Containment Integrated Leak test, required by 10 CFR Part 50, Appendix J, interval to one test in 15 years and an extension of the Type C test interval to 75 months, with a permissible extension period of 9 months (total of 84 months) for non-routine emergent conditions, based on acceptable performance history of the containment test as defined in NEI 94-01, Revision 3-A.

Date of Issuance: April 8, 2015.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

Amendment Nos.: 293 and 180. A publicly-available version is in ADAMS under Accession No. ML14322A461.

Facility Operating License Nos DPR-66 and NPF-73: Amendments revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: August 5, 2014 (79 FR 45477). The supplemental letters dated November 4, 2014, and March 23, 2015, 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 as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated April 8, 2015.

No significant hazards consideration comments received: No.

40 Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan Date of amendment requests: February 6, 2015.

Brief description of amendments: The amendments modified the technical specifications requirements for unavailable barriers by adding limiting condition for operation 3.0.8. The changes are consistent with the NRC approved Technical Specification Task Force (TSTF)

Standard Technical Specification change TSTF-427, Allowance for Non-Technical Specification Barrier Degradation on Supported System OPERABILITY, Revision 2.

Date of issuance: April 6, 2015.

Effective date: As of the date of issuance and shall be implemented within 90 days of issuance.

Amendment Nos.: 327 - Unit 1; 310 - Unit 2. A publicly-available version is in ADAMS under Accession No. ML15076A226; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-58 and DPR-74: Amendments revise the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: March 3, 2015 (80 FR 11478).

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated April 6, 2015.

No significant hazards consideration comments received: No.

South Carolina Electric and Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, Fairfield County, South Carolina

41 Date of amendment request: April 7, 2014, as supplemented by letters dated October 3, 2014, and March 18, 2015.

Date of issuance: April 13, 2015.

Brief description of amendment: The amendment approves a revision to the emergency action levels from a scheme based on NEI 99-01, Revision 5, Methodology for Development of Emergency Action Levels to a scheme based on NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors.

Effective date: As of the date of its issuance and shall be implemented within 180 days of issuance.

Amendment No.: 200. A publicly-available version is in ADAMS under Accession No. ML15063A355; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-12: Amendment revised the Renewed Facility Operating License.

Date of initial notice in Federal Register: June 6, 2014 (79 FR 32771). The supplemental letters dated October 3, 2014, and March 18, 2015, 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 as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated April 13, 2015.

No significant hazards consideration comments received: No.

42 Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama Date of application for amendment: June 3, 2014.

Brief description of amendments: The amendments revise the Technical Specification Limiting Condition for Operation 3.3.1 and Surveillance Requirement 3.2.4.2 related to the reactor trip system instrumentation.

Date of issuance: April 8, 2015.

Effective date: As of its date of issuance and shall be implemented within 120 days from the date of issuance.

Amendment Nos.: Unit 1 - 197, Unit 2 - 193. A publicly-available version is in ADAMS under Accession No. ML15028A165, documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: July 22, 2014 (79 FR 42551).

The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated April 8, 2015.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia Date of amendment request: July 14, 2014, and supplemented by the letter dated December 12, 2014.

43 Brief description of amendment: The license amendment revised the Combined Licenses (COLs) to modify the fire area fire barriers of the turbine building switchgear rooms of the turbine building to accommodate the revised layout of the low and medium voltage switchgear and associated equipment.

Date of issuance: April 1, 2015.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

Amendment No.: 32. A publicly-available version is in ADAMS under Accession No. ML15037A045; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: September 30, 2014 (79 FR 58812). The supplemental letter dated December 12, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated April 1, 2015.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia

44 Date of amendment request: July 29, 2014, and supplemented by the letter dated November 5, 2014.

Brief description of amendment: The license amendment revised the Combined Licenses (COLs) with regard to Tier 1 material and promoted consistency with the Updated Final Safety Analysis Report Tier 2.

Date of issuance: February 13, 2015.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

Amendment No.: 30. A publicly-available version is in ADAMS under Accession No. ML14350B012; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

45 Date of initial notice in Federal Register: September 30, 2014 (79 FR 58812). The supplemental letter dated November 5, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated February 13, 2015.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 20th day of April 2015.

For the Nuclear Regulatory Commission.

/RA/

Louise Lund, Acting Director Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.