ML15022A653

Biweekly Sholly FRN - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations - Publication Date: February 3, 2015

ML15022A653
Person / Time
Site:	Monticello, Harris, Saint Lucie, Arkansas Nuclear, Summer, Columbia, Limerick, Callaway, Vogtle, Farley, Quad Cities, Washington State University, Fort Calhoun
Issue date:	01/26/2015
From:	Marilyn Evans Division of Operating Reactor Licensing
To:
Clayton B
References
NRC-2015-0015
Download: ML15022A653 (71)

Text

[7590-01-P]

NUCLEAR REGULATORY COMMISSION

[NRC-2015-0015]

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 January 8, 2015, to January 21, 2015. The last biweekly notice was published on January 20, 2015.

2 DATES: Comments must be filed by [INSERT DATE 30 DAYS FROM DATE OF PUBLICATION IN THE FEDERAL REGISTER]. A request for a hearing must be filed by

[INSERT DATE 60 DAYS FROM DATE OF PUBLICATION IN THE FEDERAL REGISTER].

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-2015-0015. Address questions about NRC dockets to Carol Gallagher; telephone: 301-287-3422; e-mail: Carol.Gallagher@nrc.gov.

• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: 3WFN A44M, 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: Beverly A. Clayton, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone:

301-415-3475, e-mail: Beverly.Clayton@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments.

A. Obtaining Information.

3 Please refer to Docket ID NRC-2015-0015 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:

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0015.

• 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-2015-0015 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

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

4 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 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.

5 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.

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

6 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 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

7 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, 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

8 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 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

9 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.

Submissions should be in Portable Document Format (PDF) in accordance with NRC 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.

10 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 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

11 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 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.

Energy Northwest, Docket No. 50-397, Columbia Generating Station (Columbia), Benton County, Washington Date of amendment request: November 17, 2014. A publicly-available version is in ADAMS under Accession No. ML14336A100.

Description of amendment request: The proposed amendment would modify the Technical Specifications to revise values for the safety limit minimum critical power ratio (SLMCPR) due to core loading fuel management changes for the upcoming Columbia operating cycle.

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.

12 The basis of the Safety Limit Minimum Critical Power Ratio (SLMCPR) is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. The new SLMCPR values preserve the existing margin to transition boiling. The derivation of the revised SLMCPR for Columbia, for incorporation into the Technical Specifications and its use to determine plant and cycle-specific thermal limits, has been performed using NRC approved methods. The revised SLMCPR values do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient.

Based on the above, Energy Northwest has concluded that the proposed change will not result in 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 analyzed?

Response: No.

The proposed changes result only from a specific analysis for the Columbia core reload design. These changes do not involve any new or different methods for operating the facility. No new initiating events or transients result from these changes.

Based on the above, Energy Northwest has concluded that the proposed change will not create the possibility of a new or different kind of accident from those previously evaluated.

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

Response: No.

The new SLMCPR is calculated using NRC approved methods with plant and cycle specific parameters for the current core design. The SLMCPR value remains conservative enough to ensure that at least 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. The operating limit minimum critical power ratio (MCPR) is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs).

Accordingly, the margin of safety is maintained with the revised values.

As a result, Energy Northwest has determined that the proposed change will not result in 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 staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street, N.W.,

Washington, DC 20006-3817.

Acting NRC Branch Chief: Eric R. Oesterle.

Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida Date of amendment request: December 5, 2014. A publicly-available version is in ADAMS under Accession No. ML14351A074.

Description of amendment request: The amendment would revise Technical Specifications (TSs) Section 3.6.2.1, regarding containment spray and cooling systems, by eliminating second completion times limiting time from discovery of failure to meet a limiting condition for operation (LCO). The proposed revision is consistent with NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-439, Revision 2, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO (Adams Accession No. ML051860296).

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 any accident previously evaluated?

Response: No.

14 The proposed change that incorporated TSTF-439, Revision 2, [will eliminate] certain Completion Times from the TS. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the revised Completion Times are no different [from] the consequences of the same accident during the existing Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. The proposed change does not alter or prevent the ability of structures, systems, or components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed change is consistent with the [previous] safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 does 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. The proposed change does not alter any assumptions made in the safety analysis.

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 to delete the second Completion Times does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis

15 acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis.

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

The NRC staff has reviewed the licensees analysis and determined 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: William S. Blair, Managing Attorney - Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, Juno Beach, FL 33408-0420.

NRC Branch Chief: Shana R. Helton.

Omaha Public Power District (OPPD), Docket No. 50-285, Fort Calhoun Station, Unit 1, Washington County, Nebraska Date of amendment request: December 26, 2014. A publicly-available version is in ADAMS under Accession No. ML14365A123.

Description of amendment request: The proposed amendment upgrades the Emergency Action Level (EAL) scheme by adopting NRC-endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, issued January 2011 (ADAMS Accession No. ML110240324).

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.

16 The proposed changes to OPPDs EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors, do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed changes do not reduce the functionality, performance, or capability of OPPDs ERO

[emergency response organization] to respond in mitigating the consequences of any design basis accident.

The probability of a reactor accident requiring implementation of Emergency Plan EALs has no relevance in determining whether the proposed changes to the EALs reduce the effectiveness of the Emergency Plans. As discussed in Section D, Planning Basis, of NUREG-0654, Revision 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants [issued November 1980; ADAMS Accession No. ML040420012]:

The overall objective of emergency response plans is to provide dose savings (and in some cases immediate life saving) for a spectrum of accidents that could produce offsite doses in excess of Protective Action Guides (PAGs). No single specific accident sequence should be isolated as the one for which to plan because each accident could have different consequences, both in nature and degree. Further, the range of possible selection for a planning basis is very large, starting with a zero point of requiring no planning at all because significant offsite radiological accident consequences are unlikely to occur, to planning for the worst possible accident, regardless of its extremely low likelihood Therefore, OPPD did not consider the risk insights regarding any specific accident initiation or progression in evaluating the proposed changes.

The proposed changes do not involve any physical changes to plant equipment or systems, nor do they alter the assumptions of any accident analyses. The proposed changes do not adversely affect accident initiators or precursors nor do they alter the design assumptions, conditions, and configuration or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of Structures, Systems, or Components (SSCs) to perform their intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits.

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

17

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 changes to OPPDs EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6, do not involve any physical changes to plant systems or equipment. The proposed changes do not involve the addition of any new plant equipment. The proposed changes will not alter the design configuration, or method of operation of plant to be performed as required. The proposed changes do not create any new credible failure mechanisms, malfunctions, or accident initiators.

Therefore, the proposed changes do 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?

Response: No.

The proposed changes to OPPDs EAL scheme to adopt the NRC-endorsed guidance in NEI 99-01, Revision 6, do 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.

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 changes to adopt the NEI 99-01, Revision 6, EAL scheme guidance. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.

Therefore, the proposed changes do 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 staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street, N.W.,

Washington, DC 20006-3817.

Acting NRC Branch Chief: Eric R. Oesterle.

18 South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina Date of amendment request: July 17, 2014. A publicly-available version is in ADAMS under Accession No. ML14202A088.

Description of amendment request: The proposed changes would revise the Combined Licenses (COLs) by 1) providing additional detail to describe the mechanical connection between the internal containment structural module steel faceplates and the base concrete,

2) allowing for increases in the thickness of the structural wall module faceplates, 3) identifying changes to the wall thicknesses for portions of some internal containment structural wall modules, and 4) identifying the use of steel plates, structural shapes, reinforcement bars, or tie bars between the faceplates of the structural wall modules, where needed to meet applicable code requirements.

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 10 CFR 52.63(b)(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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The design function of the internal containment structures is to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in those structures. These structures are structurally designed to meet seismic Category I requirements as defined in Regulatory Guide 1.29.

19 The changes to the design details for the structural modules do not have an adverse impact on the response of the nuclear island structures to safe shutdown earthquake ground motions or loads due to anticipated transients or postulated accident conditions, nor do they change the seismic Category I classification. Evaluations have been performed which determined that the proposed changes do not have a significant impact on the calculated loads for the affected structural modules, or critical locations, and no significant impact on the global seismic model.

The changes to the design details for the structural modules do not impact the support, design, or operation of mechanical and fluid systems. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to postulated accident conditions.

The plant response to previously evaluated accidents or external events is not adversely affected, nor does the change described create any new accident precursors.

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 changes are to revise design details for the internal containment structural modules. The changes do not change the design requirements of the nuclear island structures, nor do they change the seismic Category I classification. The changes to the design details for the internal containment structural modules do not change the design function, support, design, or operation of mechanical and fluid systems.

The changes to the design details for the internal containment structural modules do not result in a new failure mechanism for the nuclear island structures or introduce any new accident precursors. As a result, the design function of the nuclear island structures is not adversely affected by the proposed change.

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.

20 The requested amendment proposes changes to the structural details associated with the in-containment structural modules. The purpose of these changes is to ensure that the requirements contained in the applicable construction codes are met. As discussed in UFSAR

[Updated Final Analysis Report], Section 3.8.3.5, Design Procedures and Acceptance Criteria, the in-containment structural modules are designed in accordance with ACI [American Concrete Institute] 349 and AISC [American Institute of Steel Construction] N690. Thus, the identification of additional structural module connection details, the increase in structural module faceplate and wall thicknesses, and the addition of additional reinforcement in specific areas are proposed to ensure that the codes of record, and the associated margins contained therein, continue to be met as specified in the design basis. Structural and seismic analysis of the modified sections in accordance with the methodologies identified in the UFSAR has confirmed that the applicable requirements of ACI 349 and AISC N690 continue to be met for affected in-containment structural modules.

As a result, the proposed changes do not adversely affect any safety related equipment or other design functions, design code compliance, design analysis, safety analysis input or result, or design/safety margin.

No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes.

Therefore, the requested 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 staff 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 J. Burkhart.

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

21 Date of amendment request: December 19, 2014. A publicly-available version is in ADAMS Package Accession No. ML14363A422.

Description of amendment request: The licensee proposes to expand the emergency planning zone (EPZ) boundary, to revise the evacuation time estimates (ETA) analysis, and revise the alert and notification system (ANS) design reports to encompass the expanded EPZ boundary.

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 changes, which include expansion of the EPZ boundary and revision of the ETE analysis and ANS design reports to encompass the expanded EPZ boundary, do not impact the physical function of plant structures, systems, or components (SSC) or the manner in which SSCs perform their design function. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are affected by the proposed changes. 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.

The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed or removed) or a change in the method of plant operation. The proposed changes 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 changes, which include expansion of the EPZ boundary and revision of the ETE analysis and ANS design reports to encompass the expanded EPZ boundary, are not initiators of any accidents. Therefore,

22 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 amendment involve a significant reduction in a margin of safety?

Response: No.

Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public.

The proposed changes, which include expansion of the EPZ boundary and revision of the ETE analysis and ANS design reports to encompass the expanded EPZ boundary, do not impact operation of the plant or its response to transients or accidents. The proposed changes do not alter requirements of the Technical Specifications or the Unit 1 Operating License. The proposed changes do not involve a change in the method of plant operation and no accident analyses will be affected by the proposed changes.

Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these proposed changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

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: J. Hagood Hamilton, Jr., South Carolina Electric & Gas Company, Post Office Box 764, Columbia, SC 29218.

NRC Branch Chief: Robert J. Pascarelli.

23 Southern Nuclear Operating Company, Inc. Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, Georgia Date of amendment request: January 8, 2015. A publicly-available version is in ADAMS under Accession No. ML15008A466.

Description of amendment request: The proposed change would amend Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4 by departing from the plant-specific Design Control Document (DCD) Tier 1 (and corresponding Combined License Appendix C information) and Tier 2 material by making changes to specify the use of latching control relays in lieu of breakers to de-energize the control rod drive mechanism (CRDM) motor generator (MG) set generator field on a diverse actuation system (DAS) signal.

Because this proposed change requires a departure from Tier 1 information in the Westinghouse Advanced Passive 1000 DCD, the licensee also requested an exemption from the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change to use field control relays in lieu of field circuit breakers to de-energize the CRDM MG Set excitation field does not result in a change to the basic MG Set design function, which is to supply reliable electrical power to the CRDMs while providing a trip function on a DAS signal, allowing the control rods to drop. The Probabilistic Risk Assessment (PRA) is not adversely affected. No safety-related structure, system, or component (SSC) or function is adversely affected. The change does not involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. Because the change maintains

24 the CRDM MG set trip function used to mitigate an accident, the consequences of the accidents evaluated in the UFSAR are not affected.

Therefore, there is no 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.

There is no safety-related SSC or function adversely affected by this proposed change to use control relays instead of breakers to de-energize the CRDM MG set generator field on demand. This proposed change does not change any equipment qualification or fission product barrier.

The change does not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment.

This activity will not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that would result in significant fuel cladding failures.

Therefore, this activity 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.

There is no safety-related SSC or function adversely affected by this proposed change to use relays instead of breakers to control the CRDM MG set generator field. The function to trip the MG set generator field on a DAS signal, allowing the control rods to drop, is not adversely affected by the use of relays as the device to de-energize the generator field. The proposed change does not affect any safety-related design code, function, design analysis, safety analysis input or result, or design/safety margin. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the requested change, thus, no margin of safety is reduced.

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 staff

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

Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.

NRC Branch Chief: Lawrence Burkhart.

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 amendment request: November 24, 2014. A publicly-available version is in ADAMS under Accession Package No. ML14335A689.

Description of amendment request: The licensee requested 24 revisions to the Technical Specifications. Twenty two revisions adopt various previously NRC approved Technical Specifications Task Force Travelers and two revisions are not associated with Travelers. A list of the requested revisions is included in Enclosure 1 of the application.

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 for each of the 24 changes requested, which is presented below:

Request No. 1: TSTF-27-A, Revision 3, Revise SR Frequency for Minimum Temperature for Criticality

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 Surveillance Frequency for monitoring RCS temperature to ensure the minimum temperature for criticality is met.

The Frequency is changed from a 30 minute Frequency when certain conditions are met to a periodic Frequency that it is controlled in accordance with the Surveillance Frequency Control Program. The measurement of RCS [reactor coolant system] temperature is not an

26 initiator of any accident previously evaluated. The minimum RCS temperature for criticality is not changed. As a result, the mitigation of any accident previously evaluated is not affected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change revises the Surveillance Frequency for monitoring RCS temperature to ensure the minimum temperature for criticality is met.

The current, condition based Frequency represents a distraction to the control room operator during the critical period of plant startup. RCS temperature is closely monitored by the operator during the approach to criticality and temperature is recorded on charts and computer logs.

Allowing the operator to monitor temperature as needed by the situation and logging RCS temperature at a periodic Frequency that it is controlled in accordance with the Surveillance Frequency Control Program is sufficient to ensure that the LCO [limiting condition for operation] is met while eliminating a diversion of the operators attention.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

27 Request No. 2: TSTF-46-A, Revision 1, Clarify the CIV Surveillance to Apply Only to Automatic Isolation Valves

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 requirements in Technical Specification SR 3.6.3.4, and the associated Bases, to delete the reference to verifying the isolation time of each power operated containment isolation valve (CIV) and only require verification of each automatic power operated containment isolation valve. The closure times for CIVs that do not receive an automatic closure signal are not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The CIVs are used to respond to accidents previously evaluated. Power operated CIVs that do not receive an automatic closure signal are not assumed to close in a specified time. The proposed change does not change how the plant would mitigate an accident previously evaluated.

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 does not result in a change in the manner in which the CIVs provide plant protection or introduce any new or different operational conditions. Periodic verification that the closure times for CIVs that receive an automatic closure signal are within the limits established by the accident analysis will continue to be performed under SR 3.6.3.4. The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. There are also no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type 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 amendment involve a significant reduction in a margin of safety?

Response: No.

28 The proposed change provides clarification that only CIVs that receive an automatic isolation signal are within the scope of the SR 3.6.3.4. The proposed change does not result in a change in the manner in which the CIVs provide plant protection. Periodic verification that closure times for CIVs that receive an automatic isolation signal are within the limits established by the accident analysis will continue to be performed. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any Safety Analysis Limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 3: TSTF-87-A, Revision 2, Revise RTBs Open and CRDM De-energized Actions to Incapable of Rod Withdrawal

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

Response: No.

This change revises the Required Actions for LCO 3.4.5, RCS Loops -

Mode 3, Conditions C.2 and D.1, from De-energize all control rod drive mechanisms, to Place the Rod Control System in a condition incapable of rod withdrawal. It also revises LCO 3.4.9, Pressurizer, Required Action A. 1, from requiring the Reactor Trip Breakers to be open after reaching MODE 3 to Place the Rod Control System in a condition incapable of rod withdrawal, and to require full insertion of all rods.

Inadvertent rod withdrawal can be an initiator for design basis accidents or events during certain plant conditions, and therefore must be prevented under those conditions. The proposed Required Actions for LCO 3.4.5 and LCO 3.4.9 satisfy the same intent as the current Required Actions, which is to prevent inadvertent rod withdrawal when an applicable Condition is not met, and is consistent with the assumptions of the accident analysis. As a result, the proposed change does not

29 increase the probability of any accident previously evaluated. The proposed change does not change how the plant would mitigate an accident previously evaluated as in both the current and proposed requirements, rod withdrawal is prohibited.

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 provides less specific, but equivalent, direction on the manner in which inadvertent control rod withdrawal is to be prevented when the Conditions of LCO 3.4.5 and LCO 3.4.9 are not met. Rod withdrawal will continue to be prevented when the applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis.

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.

The proposed change provides the operational flexibility of allowing alternate, but equivalent, methods of preventing rod withdrawal when LCO 3.4.5 and LCO 3.4.9 are not met. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

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

Request No. 4: TSTF-245-A, Revision 1, AFW Train Operable When in Service

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 requirements in Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System, to clarify the operability of an AFW train when it is aligned for manual steam generator level control.

The AFW System is not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The AFW System is used to respond to accidents previously evaluated. The proposed change does not affect the design of the AFW System, and no physical changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated.

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 does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Manual control of AFW level control valves is not an accident initiator.

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?

31 Response: No.

The proposed change provides the operational flexibility of allowing an AFW train(s) to be considered operable when it is not in the normal standby alignment and is temporarily incapable of automatic initiation, such as during alignment and operation for manual steam generator level control, provided it is capable of being manually realigned to the AFW heat removal mode of operation. The proposed change does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any Safety Analysis Limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 5: TSTF-247-A, Revision 0, Provide Separate Condition Entry for Each PORV and Block Valve

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 requirements in Technical Specification 3.4.11, Pressurizer PORVs [power operated relief valves], to clarify that separate Condition entry is allowed for each block valve. Additionally, the Actions are modified to no longer require that the PORVs be placed in manual operation when both block valves are inoperable and cannot be restored to operable status within the specified Completion Time. This preserves the overpressure protection capabilities of the PORVs. The pressurizer block valves are used to isolate their respective PORV in the

32 event it is experiencing excessive leakage, and are not an initiator of any design basis accident or event. Therefore the proposed change does not increase the probability of any accident previously evaluated. The PORV and block valves are used to respond to accidents previously evaluated.

The proposed change does not affect the design of the PORV and block valves, and no physical changes are made to the plant. The proposed change does not change how the plant would mitigate an accident previously evaluated.

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 does not result in a change in the manner in which the PORV and block valves provide plant protection. The PORVs will continue to provide overpressure protection, and the block valves will continue to provide isolation capability in the event a PORV is experiencing excessive leakage. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Operation of the PORV block valves is not an accident initiator.

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.

The proposed changes provide clarification that separate Condition entry is allowed for each block valve. Additionally, the Actions are modified to no longer require that the PORVs be placed in manual operation when both block valves are inoperable and cannot be restored to operable status within the specified Completion Time. This preserves the overpressure protection capabilities of the PORVs. The proposed change does not result in a change in the manner in which the PORV and block valves provide plant protection. The PORVs will continue to provide overpressure protection, and the block valves will continue to provide isolation capability in the event a PORV is experiencing excessive leakage. The proposed change does not affect the safety analysis

33 acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions.

The proposed change will not result in plant operation in a configuration outside the design basis.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 6: TSTF-248-A, Revision 0, Revise Shutdown Margin Definition for Stuck Rod Exception

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

Response: No.

The proposed change modifies the definition of Shutdown Margin to eliminate the requirement to assume the highest worth control rod is fully withdrawn when calculating Shutdown Margin if it can be verified by two independent means that all control rods are inserted. The method for calculating shutdown margin is not an initiator of any accident previously evaluated. If it can be verified by two independent means that all control rods are inserted, the calculated Shutdown Margin, without the conservatism of assuming the highest worth control rod is withdrawn, is accurate and consistent with the assumptions in the accident analysis.

As a result, the mitigation of any accident previously evaluated is not affected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change

34 to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change modifies the definition of Shutdown Margin to eliminate the requirement to assume the highest worth control rod is fully withdrawn when calculating Shutdown Margin if it can be verified by two independent means that all control rods are inserted. The additional margin of safety provided by the assumption that the highest worth control rod is fully withdrawn is unnecessary if it can be independently verified that all controls rods are inserted.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 7: TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls

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

Response: No.

The proposed change removes the list of Remote Shutdown System instrumentation and controls from the Technical Specifications and places them in the Bases. The Technical Specifications continue to require that the instrumentation and controls be operable. The location of the list of Remote Shutdown System instrumentation and controls is not an initiator to any accident previously evaluated. The proposed change will have no effect on the mitigation of any accident previously evaluated because the instrumentation and controls continue to be required to be operable.

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

35

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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change removes the list of Remote Shutdown System instrumentation and controls from the Technical Specifications and places it in the Bases. The review performed by the NRC when the list of Remote Shutdown System instrumentation and controls is revised will no longer be needed unless the criteria in 10 CFR 50.59 are not met such that prior NRC review is required. The Technical Specification requirement that the Remote Shutdown System be operable, the definition of operability, the requirements of 10 CFR 50.59, and the Technical Specifications Bases Control Program are sufficient to ensure that revision of the list without prior NRC review and approval does not introduce a significant safety risk.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 8: TSTF-272-A, Revision 1, Refueling Boron Concentration Clarification

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

Response: No.

36 The proposed change modifies the Applicability of Specification 3.9.1, Boron Concentration, to clarify that the boron concentration limits are only applicable to the refueling canal and the refueling cavity when those volumes are attached to the Reactor Coolant System (RCS). The boron concentration of water volumes not connected to the RCS are not an initiator of an accident previously evaluated. The ability to mitigate any accident previously evaluated is not affected by the boron concentration of water volumes not connected to the RCS.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change modifies the Applicability of Specification 3.9.1, Boron Concentration, to clarify that the boron concentration limits are only applicable to the refueling canal and the refueling cavity when those volumes are attached to the RCS. Technical Specification SR 3.0.4 requires that Surveillances be met prior to entering the Applicability of a Specification. As a result, the boron concentration of the refueling cavity or the refueling canal must be verified to satisfy the LCO prior to connecting those volumes to the RCS. The margin of safety provided by the refueling boron concentration is not affected by this change as the RCS boron concentration will continue to satisfy the LCO.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

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

Request No. 9: TSTF-273-A, Revision 2, Safety Function Determination Program Clarifications

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

Response: No.

The proposed TS changes add explanatory text to the programmatic description of the Safety Function Determination Program (SFDP) in Specification 5.5.15 to clarify in the requirements that consideration does not have to be made for a loss of power in determining loss of function.

The Bases for LCO 3.0.6 is revised to provide clarification of the appropriate LCO for loss of function, and that consideration does not have to be made for a loss of power in determining loss of function. The changes are editorial and administrative in nature, and therefore do not increase the probability of any accident previously evaluated. No physical or operational changes are made to the plant. The proposed change does not change how the plant would mitigate an accident previously evaluated.

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 changes are editorial and administrative in nature and do not result in a change in the manner in which the plant operates. The loss of function of any specific component will continue to be addressed in its specific TS LCO and plant configuration will be governed by the required actions of those LCOs. The proposed changes are clarifications that do not degrade the availability or capability of safety related equipment, and therefore do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no design changes associated with the proposed changes, and the changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analysis, and are consistent with the safety analysis assumptions and current plant operating practice. Due to the administrative nature of the changes, they cannot be an accident initiator.

38 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.

The proposed changes to TS 5.5.15 are clarifications and are editorial and administrative in nature. No changes are made the LCOs for plant equipment, the time required for the TS Required Actions to be completed, or the out of service time for the components involved. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed changes will not result in plant operation in a configuration outside the design basis.

Therefore, the proposed changes do 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 proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 10: TSTF-283-A, Revision 3, Modify Section 3.8 Mode Restriction Notes

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

Response: No.

The proposed change modifies Mode restriction Notes on four diesel generator (DG) Surveillances to allow performance of the Surveillance in whole or in part to reestablish DG Operability. The emergency diesel generators and their associated emergency loads are accident mitigating features, and are not an initiator of any accident previously evaluated. As a result the probability of any accident previously evaluated is not increased. The proposed change allows Surveillance testing to be performed in whole or in part to reestablish Operability of a DG. The consequences of an accident previously evaluated during the period that

39 the DG is being tested to reestablish Operability are no different from the consequences of an accident previously evaluated while the DG is inoperable. As a result, the consequences of any accident previously evaluated are not increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The purpose of Surveillances is to verify that equipment is capable of performing its assumed safety function. The proposed change will only allow the performance of the Surveillances to reestablish Operability and the proposed changes may not be used to remove a DG from service. In addition, the proposed change will potentially shorten the time that a DG is unavailable because testing to reestablish Operability can be performed without a plant shutdown. The proposed changes also require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the normally prohibited Modes.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

40 Request No. 11: TSTF-284-A, Revision 3, Add Met vs. Perform to Technical Specification 14, Frequency

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

Response: No.

The proposed changes insert a discussion paragraph into Specification 1.4, and several new examples are added to facilitate the use and application of SR Notes that utilize the terms met and perform. The changes also modify SRs in multiple Specifications to appropriately use met and perform exceptions. The changes are administrative in nature because they provide clarification and correction of existing expectations, and therefore the proposed change does not increase the probability of any accident previously evaluated. No physical or operational changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated.

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 changes are administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes provide clarification and correction of existing expectations that do not degrade the availability or capability of safety related equipment, and therefore do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no design changes associated with the proposed changes, and the changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analysis, and are consistent with the safety analysis assumptions and current plant operating practice. Due to the administrative nature of the changes, they cannot be an accident initiator.

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.

41 The proposed changes are administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes provide clarification and correction of existing expectations that do not degrade the availability or capability of safety related equipment, or alter their operation. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed changes will not result in plant operation in a configuration outside the design basis.

Therefore, the proposed changes do 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 proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 12: TSTF-308-A, Revision 1, Determination of Cumulative and Projected Dose Contributions in RECP

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 Specification 5.5.4, Radioactive Effluent Controls Program, paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are not an assumption in any accident previously evaluated and have no effect on the mitigation of any accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to

42 the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change revises Specification 5.5.4, Radioactive Effluent Controls Program, paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are administrative tools to assure compliance with regulatory limits. The proposed change revises the requirement to clarify the intent, thereby improving the administrative control over this process. As a result, any effect on the margin of safety should be minimal.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 13: TSTF-312-A, Revision 1, Administrative Control of Containment Penetrations

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

Response: No.

The proposed change would allow containment penetrations to be unisolated under administrative controls during core alterations or movement of irradiated fuel assemblies within containment. The status of containment penetration flow paths (i.e., open or closed) is not an initiator for any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The proposed change does not affect the design of the primary containment, or alter plant operating practices such that the probability of an accident previously evaluated would be significantly increased. The proposed change does not significantly change how the

43 plant would mitigate an accident previously evaluated, and is bounded by the fuel handling accident (FHA) 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.

Allowing penetration flow paths to be open is not an initiator for any accident. The proposed change to allow open penetration flow paths will not affect plant safety functions or plant operating practices such that a new or different accident could be created. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice.

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.

TS 3.9.3 provides measures to ensure that the dose consequences of a postulated FHA inside containment are minimized. The proposed change to LCO 3.9.3 will allow penetration flow path(s) to be open during refueling operations under administrative control. These administrative controls will provide assurance that prompt closure of open penetrations flow paths can and will be achieved in the event of an FHA inside containment, and will minimize dose consequences. The proposed change is bounded by the existing FHA analysis. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.

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

44 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.

Request No. 14: TSTF-314-A, Revision 0, Require Static and Transient FQ Measurement

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 Required Actions of Specification 3.1.4, Rod Group Alignment Limits, and Specification 3.2.4, Quadrant Power Tilt Ratio, to require measurement of both the steady state and transient portions of the Heat Flux Hot Channel Factor, FQ(Z). This change will ensure that the hot channel factors are within their limits when the rod alignment limits or quadrant power tilt ratio are not within their limits. The verification of hot channel factors is not an initiator of any accident previously evaluated. The verification that both the steady state and transient portion of FQ(Z) are within their limits will ensure this initial assumption of the accident analysis is met should a previously evaluated accident occur.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

45 The proposed change revises the Required Actions in the Specifications for Rod Group Alignment Limits and Quadrant Power Tilt Ratio to require measurement of both the steady state and transient portions of the Heat Flux Hot Channel Factor, FQ(Z). This change is a correction that ensures that the plant conditions are as assumed in the accident analysis.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 15: TSTF-315-A, Revision 0, Reduce Plant Trips Due to Spurious Signals to the NIS During Physics Testing

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 Specification 3.1.8, PHYSICS TESTS Exceptions - MODE 2, to allow the number of channels required by LCO 3.3.1, RTS Instrumentation, to be reduced from "4" to "3" to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition. A reduction in the number of required nuclear instrumentation channels is not an initiator to any accident previously evaluated. With the nuclear instrumentation channel placed in bypass instead of in trip, reactor protection is provided by the intermediate range neutron flux detectors and the nuclear instrumentation system operating in a two-out-of-three channel logic. As a result, the ability to mitigate any accident previously evaluated is not significantly affected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change

46 to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change reduces the probability of a spurious reactor trip during physics testing. The reactor trip system continues to be capable of protecting the reactor utilizing the intermediate range neutron flux reactor trip and the power range neutron flux trips operating in a two-out-of-three trip logic. As a result, the reactor is protected and the probability of a spurious reactor trip is significantly reduced.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 16: TSTF-325, Revision 0, ECCS Conditions and Required Actions with Less Than 100% Equivalent ECCS Flow

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

Response: No.

The proposed change corrects the structure of Technical Specification 3.5.2 to assure its proper application. There is no change in intent or in the way the Technical Specification is applied. The literal (and unintended) interpretation of the existing LCO structure could, under some circumstances, provide longer than intended Completion Times for restoration of operability. The proposed change only clarifies the requirements of the Required Actions. Since the proposed change affects neither the Technical Specification intent, nor its application, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

47

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 corrects the structure of the Technical Specification to assure its correct application. There is no change in intent or in the way the Technical Specification is applied. The proposed changes would not result in any physical alterations to the plant configuration, no new equipment is added, no equipment interfaces are modified, and no changes to any equipments function or the method of operating the equipment are being made. As the proposed changes would not change the design, configuration or operation of the plant, no new or different kinds of accident modes are created.

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.

The proposed change corrects the structure of the Technical Specification to assure its correct application. There is no change in intent or in the way the Technical Specification is applied.

Therefore, the proposed changes do 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 proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 17: TSTF-340-A, Revision 3, Allow 7 Day Completion Time for a Turbine-Driven AFW Pump Inoperable

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 Specification 3.7.5, Auxiliary Feedwater (AFW) System, to allow a 7 day Completion Time to restore an inoperable turbine-driven pump in Mode 3 immediately following a

48 refueling outage, if Mode 2 has not been entered. An inoperable AFW turbine-driven pump is not an initiator of any accident previously evaluated. The ability of the plant to mitigate an accident is no different while in the extended Completion Time than during the existing Completion Time.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change revises Specification 3.7.5, Auxiliary Feedwater (AFW) System, to allow a 7 day Completion Time to restore an inoperable turbine-driven AFW pump in Mode 3 immediately following a refueling outage if Mode 2 has not been entered. In Mode 3 immediately following a refueling outage, core decay heat is low and the need for AFW is also diminished. The two operable motor driven AFW pumps are available and there are alternate means of decay heat removal if needed.

As a result, the risk presented by the extended Completion Time is minimal.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

49 Request No. 18: TSTF-343, Revision 1, Containment Structural Integrity

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

Response: No.

The proposed changes revise the Technical Specifications (TS)

Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed changes affect the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program, and allows those examinations to be performed during power operation in addition to during a refueling outage.

The frequency of visual examinations of the containment and the mode of operation during which those examinations are performed does not affect the initiation of any accident previously evaluated. The use of NRC approved methods and frequencies for performing the inspections will ensure the containment continues to perform the mitigating function assumed for accidents previously evaluated.

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.

The proposed changes revise the TS Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed changes affect the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program, and allows those examinations to be performed during power operation in addition to during a refueling outage.

The proposed changes do not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed changes will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure.

50 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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes revise the Technical Specifications (TS)

Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed changes affect the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program, and allows those examinations to be performed during power operation in addition to during a refueling outage. The safety function of the containment as a fission product barrier will be maintained.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 19: TSTF-349-A, Revision 1, Add Note to LCO 3.9.5 Allowing Shutdown Cooling Loops Removal from Operation

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

Response: No.

The proposed change adds an LCO Note to LCO 3.9.5, RHR and Coolant Circulation - Low Water Level, to allow securing the operating train of Residual Heat Removal (RHR) for up to 15 minutes to support switching operating trains. The allowance is restricted to conditions in which core outlet temperature is maintained at least 10 degrees F below the saturation temperature, when there are no draining operations, and when operations that could reduce the reactor coolant system (RCS) boron concentration are prohibited. Securing an RHR train to facilitate the changing of the operating train is not an initiator to any accident previously evaluated. The restrictions on the use of the allowance ensure that an RHR train will not be needed during the 15 minute period to mitigate any accident previously evaluated.

51 Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change adds an LCO Note to LCO 3.9.5, RHR and Coolant Circulation - Low Water Level, to allow securing the operating train of RHR to support switching operating trains. The allowance is restricted to conditions in which core outlet temperature is maintained at least 10 degrees F below the saturation temperature, when there are no draining operations, and when operations that could reduce the reactor coolant system (RCS) boron concentration are prohibited. With these restrictions, combined with the short time frame allowed to swap operating RHR trains and the ability to start an operating RHR train if needed, the occurrence of an event that would require immediate operation of an RHR train is extremely remote.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 20: TSTF-355-A, Revision 0, Changes to RTS and ESF Tables

52

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

Response: No.

The RTS [Reactor Trip System] and ESFAS [Engineered Safety Feature Actuations System] instrument functions are part of the accident mitigation response and are not themselves an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly affected by the proposed changes. The changes ensure that automatic protective actions will be initiated at or before the condition assumed in the safety analysis, and are in accordance with the intent of the Technical Specifications. The proposed changes will not cause any design or analysis acceptance criteria to be exceeded. Since there will be no adverse effect on the trip setpoints or the instrumentation associated with the trip setpoints, there will be no significant increase in the consequences of any 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 changes include modifications to the format of the nominal trip setpoints that preserve safety analysis assumptions related to accident mitigation. The protection system will continue to initiate the protective actions as assumed in the safety analysis. The proposed changes will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis.

As the proposed changes do not change the design, configuration or operation of the plant, no new or different kinds of accident modes are created.

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.

The proposed changes do not alter any nominal trip setpoints, allowable values, or limiting safety system settings, and will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis. The response of protection systems to accident transients reported in the Final Safety Analysis Report is unaffected by this change, and accident analysis acceptance criteria are consequently not affected.

53 Therefore, the proposed changes do 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 proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 21: TSTF-371-A, Revision 1, NIS Power Range Channel Daily SR TS Change to Address Low Power Decalibration

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 Specification 3.3.1, RTS Instrumentation, Surveillances 3.3.1.2 and 3.3.1.3 to move requirements currently in a Note to the Surveillance itself. The change in presentation is editorial and does not affect the application of the Surveillances. The proposed change does not affect any accident initiators or analyzed events or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility.

Therefore, this proposed change does not represent 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

54 The proposed change revises Specification 3.3.1, RTS Instrumentation, Surveillances 3.3.1.2 and 3.3.1.3 to move requirements currently in a Note to the Surveillance itself. The proposed change represents an editorial preference and does not affect the performance of the Surveillance or plant operation. The safety function tested by the Surveillance is unaffected.

Therefore, this 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 22: TSTF-439-A, Revision 2, Eliminate Second Completion Times Limiting Time From Discovery of Failure To Meet an LCO

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

Response: No.

The proposed change eliminates certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the remaining Completion Time are no different than the consequences of the same accident during the removed Completion Times.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

55 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.

The proposed change to delete the second Completion Time does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside of the design basis.

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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 23: ISTS Adoption #1 - Revise LCO 3.3.2 ESFAS Interlock P-4 Required Action Completion Time

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 Condition to be entered when the ESFAS Interlock P-4 is inoperable. Current Technical Specifications require restoring the channel to Operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in Mode 3 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 5 within the following 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br />. The proposed change provides 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to restore the inoperable channel, or be in Mode 3 in 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> and Mode 4 in 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />. The ESFAS P-4 interlock is not an initiator to any accident previously evaluated. The consequences of any accident previously evaluated during the proposed Completion Time are no different from the consequences during the existing Completion Time. As a result, the proposed change does not result in a significant increase in the consequences of any accident previously evaluated.

Therefore, this proposed change does not represent a significant increase in the probability or consequences of an accident previously evaluated.

56

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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change provides an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore an inoperable ESFAS P-4 Interlock. During the proposed Completion Time, manual actions can perform the functions provided by the inoperable P-4 interlock. Also, the proposed Completion Time is reasonable given the available redundant channel, and the low probability of an event occurring during this interval.

Therefore, this 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Request No. 24: Revise LCO 3.5.5 to 8-hour Completion Time and Note allowance

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

Response: No.

The proposed change modifies the LCO 3.5.5, Seal Injection Flow, Action A, Seal injection flow not within limit, Completion Time from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and the Note to SR 3.5.5.1 to allow 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> instead of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to stabilize reactor coolant system (RCS) pressure prior to verifying the seal injection throttle valves are properly adjusted. The

57 proposed change does not involve the addition or removal of any equipment, or any design changes to the facility. Seal injection flow is not an initiator of any accident previously evaluated. The consequences of any accident previously evaluated during the extended Completion Time or Note allowance are the same as during the existing Completion Time and Note allowance.

Therefore, this proposed change does not represent 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 physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

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.

The proposed change provides additional time to verify seal injection flow is within limit or to restore seal injection flow to within limit if it is discovered that it is not within limit. The additional time is acceptable on the basis that there is little likelihood of an event that would challenge the ECCS occurring during the 8-hour window, and it reduces the pressure on the operations staff should iterations in the adjustment procedure be necessary to balance seal injection flow.

Therefore, this 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

58 Attorney for licensee: Leigh D. Perry, SVP & General Counsel of Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201.

NRC Branch Chief: Robert J. Pascarelli.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri Date of amendment request: October 2, 2014. A publicly-available version is in ADAMS under Accession No. ML14275A441.

Description of amendment request: The proposed amendment upgrades the Emergency Action Level scheme by adopting NRC-endorsed Nuclear Energy Institute 99-01, Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, issued January 2011 (ADAMS Accession No. ML110240324).

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 changes to the Callaway Plant emergency action levels do not impact the physical function of plant structures, systems, or components (SSC) or the manner in which SSCs perform their design function. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within assumed acceptance limits. No operating procedures or administrative controls that function to prevent or mitigate accidents are affected by the proposed changes.

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

59

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 changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed or removed) or a change in the method of plant operation. The proposed changes 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 changes to the Callaway Plant emergency action levels are not initiators of any accidents.

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.

Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public.

The proposed changes do not impact operation of the plant or its response to transients or accidents. The changes do not affect the Technical Specifications or the operating license. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. The emergency plan will continue to activate an emergency response commensurate with the extent of degradation of plant 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.

60 Attorney for licensee: John ONeill, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, N.W., Washington, DC 20037.

Acting NRC Branch Chief: Eric R. Oesterle.

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 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.

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 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.

61 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 can be accessed as described in the Obtaining Information and Submitting Comments section of this document.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina Date of amendment request: April 24, 2014.

Brief description of amendment: The amendment revises Technical Specification (TS) 3/4.4.5, Steam Generator Tube Integrity, TS 6.8.4.I, Steam Generator Program, and TS 6.9.1.7, Steam Generator Tube Inspection Report to address implementation associated with the inspections and reporting requirements as described in Technical Specifications Task Force (TSTF) TSTF-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.

Date of issuance: January 9, 2015.

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

Amendment No.: 145. A publicly-available version is in ADAMS under Accession No. ML14307A800; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-63 The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: July 22, 2014 (79 FR 42543).

62 The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated January 9, 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: October 31, 2013, as supplemented by letters dated May 29, 2014, and September 9, 2014.

Brief description of amendment: The amendment revised Technical Specification Surveillance Requirements 3.5.1.4 and 3.5.2.5 for low pressure core spray and low pressure coolant injection pump flows.

Date of issuance: January 7, 2015.

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

Amendment No.: 229. A publicly-available version is in ADAMS under Accession No. ML14335A189; 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.

Date of initial notice in Federal Register: April 8, 2014 (79 FR 19399). The supplemental letters dated May 29, 2014, and September 9, 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.

63 The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated January 7, 2015.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 2, Pope County, Arkansas Date of application for amendment: January 21, 2014, as supplemented by letters dated March 17 and September 24, 2014.

Brief description of amendment: The amendment revised the Technical Specification 6.5.16 requirements for the local leak test required for the containment building emergency escape air lock doors, in that it would require a seal contact verification in lieu of the current seal pressure test to verify leak tightness.

Date of issuance: January 22, 2015.

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

Amendment No.: 299. A publicly-available version is in ADAMS under Accession No. ML14350B285; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-6: Amendment revised the Technical Specifications/license.

Date of initial notice in Federal Register: April 15, 2014 (79 FR 21296). The supplemental letter dated September 24, 2014, 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

64 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 January 22, 2015.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: November 15, 2013, as supplemented by letters dated April 16, 2014; September 11, 2014; and November 7, 2014.

Brief description of amendments: The amendments revise the Technical Specification (TS) requirements related to the response time for the main steam line flow-high isolation function.

Date of issuance: January 7, 2015.

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

Amendment Nos.: 214 and 175. A publicly-available version is in ADAMS under Accession No. ML14344A681; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-39 and NPF-85: Amendments revised the Renewed Facility Operating License and TSs.

Date of initial notice in Federal Register: February 4, 2014 (79 FR 6642). The supplemental letters dated April 16, 2014; September 11, 2014; and November 7, 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.

65 The Commissions related evaluation of the amendments is contained in a Safety Evaluation dated January 7, 2015.

No significant hazards consideration comments received: No.

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 amendment requests: July 16, 2013, as supplemented by letters dated September 18, 2013, January 22, April 7, August 12, and November 11, 2014.

Brief description of amendments: The amendments revises the Technical Specifications to include the use of neutron absorbing spent fuel pool rack inserts (i.e., NETCO-SNAP-IN rack inserts) for the purpose of criticality control in the spent fuel pools.

Date of issuance: December 31, 2014.

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

Amendment Nos.: 253 - Unit 1; 248 - Unit 2. A publicly-available version is in ADAMS under Accession No. ML14346A306; documents related to these amendments are listed in the safety evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-29 and DPR-30: The amendments revised the Technical Specifications and Facility Operating License.

Date of initial notice in Federal Register: July 8, 2014 (79 FR 38577). The supplemental letters dated September 18, 2013, January 22, April 7, August 12, and November 11, 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.

66 The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated December 31, 2014.

No significant hazards consideration comments received: No.

Northern States Power Company - Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota Date of amendment request: November 14, 2013.

Brief description of amendment: The amendment revises Technical Specification (TS) 5.5.11, Primary Containment Leakage Rate Testing Program, by removing TS 5.5.11.d.2.b, the reduced pressure testing option for drywell airlock door leakage testing. This testing methodology is not required and does not reflect the current testing practice at MNGP. As such, the drywell airlock door seals will be tested by performing an overall airlock leakage test as specified in current TS 5.5.11.d.2.a.

Date of issuance: January 8, 2015.

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

Amendment No.: 187. A publicly-available version is in ADAMS under Accession No. ML14323A033; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-22: This amendment revises the Renewed Facility Operating License and the Technical Specifications.

Date of initial notice in Federal Register: August 5, 2014 (79 FR 45478).

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

No significant hazards consideration comments received: No.

67 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: April 3, 2014, as supplemented by letter dated May 19, 2014.

Brief description of amendment: The amendment revises Tier 2* information, incorporated into the VCSNS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR). Specifically, the amendment revises the details regarding the structural floor of the Auxiliary Building and its constructability. Notes are added to drawings in Subsection 3H.5 of the UFSAR in order to clarify variations in detail design such as size and spacing or reinforcement and spans of the noncritical sections of floors.

Date of issuance: July 18, 2014.

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

Amendment No.: 14. A publicly-available version is in ADAMS under Accession No. ML14188B185; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: April 29, 2014 (79 FR 24024).

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated July 18, 2014.

No significant hazards consideration comments received: No.

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

68 Date of amendment request: March 17, 2014, and revised by letters dated May 8, September 2, and October 2, 2014.

Brief description of amendment: The amendment revises the VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) by clarifying how human diversity was applied during the design process for the Component Interface Module and Diverse Actuation System. The changes to the VEGP Units 3 and 4 UFSAR include changes to Table 1.6, Material Referenced, Chapter 7, Sections 7.1.2.14.1, 7.1.7 and 7.2.4 and the addition of Appendix 7A to Chapter 7. The changes to the VEGP Units 3 and 4 UFSAR modify information related to human diversity, as presented in a Tier 2* document, WCAP-17179-P and WCAP-17179-NP, AP1000 Component Interface Module Technical Report, Revision 2, and two Tier 2 documents, WCAP-15775, AP1000 Instrumentation and Control Defense-in-Depth and Diversity Report, Revision 4 and WCAP-17184-P, AP1000 Diverse Actuation System Planning and Functional Design Summary Technical Report, that are incorporated by reference in the VEGP Units 3 and 4 UFSAR.

Date of issuance: December 24, 2014.

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

Amendment No.: 28. A publicly-available version is in ADAMS under Accession No. ML14329A298; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

Date of initial notice in Federal Register: April 29, 2014 (79 FR 24021).

The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated December 24, 2014.

69 No significant hazards consideration comments received: No.

Southern Nuclear Operating Company Docket Nos.52-025 and 52-026, Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, Georgia Date of amendment request: August 22, 2014, and revised by letter dated September 23, 2014, and supplemented by letters dated October 30 and November 6, 2014.

Brief description of amendment: The amendment revises the VEGP Units 3 and 4 Updated Final Safety Analysis Report to reflect changes related to:

(a) Installation of an additional non-safety-related battery; (b) Revision to the annex building internal configuration by converting a shift turnover room to a battery room, adding an additional battery equipment room, and moving a fire area wall; (c) Increase in the height of a room in the annex building; and (d) Increase in thicknesses of certain annex building floor slabs.

In addition, the proposed changes also include reconfiguring existing rooms and related rooms, wall, and access path changes and making changes to the corresponding Tier 1 information in Appendix C to the Combined Licenses.

Date of issuance: December 23, 2014.

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

Amendment No.: 27. A publicly-available version is in ADAMS under Accession No. ML14323A609; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.

70 Date of initial notice in Federal Register: October 14, 2014 (79 FR 61662).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 23, 2014.

No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri Date of application for amendment: January 23, 2014.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.4.12, Cold Overpressure Mitigation System (COMS), to reflect the mass input transient analysis that assumes an Emergency Core Cooling System centrifugal charging pump and the normal charging pump capable of injecting into the reactor coolant system when TS 3.4.12 is applicable. The amendment also revised TS Table 3.3.1-1, Reactor Trip System Instrumentation, to remove unnecessary page number references.

Date of issuance: January 20, 2015.

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

Amendment No.: 210. A publicly-available version is in ADAMS under Accession No. ML14350B239; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: April 1, 2014 (79 FR 18348).

71 The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 20, 2015.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 26th day of January 2015.

For the Nuclear Regulatory Commission.

/RA/

Michele G. Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.