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 May 9, 2017, to May 22, 2017. The last biweekly notice was published on May 23, 2017.

DATES:

Comments must be filed by July 6, 2017. A request for a hearing must be filed by August 7, 2017.

ADDRESSES:

You may submit comments by any of the following methods:

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0131. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: TWFN-8-D36M, 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:

Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-5411; email: shirley.rohrer@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2017-0131, facility name, unit number(s), plant docket number, application date, and subject 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-2017-0131.
• NRC's 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 NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email 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 this document.
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2017-0131, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

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

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that 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 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 Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if 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. If the Commission takes 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. If the Commission makes 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 persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued.

As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) The name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest.

In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in 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 petitioner must 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 petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) 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. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures.

Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the 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) through (iii). The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish 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 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 the 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.

A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by August 7, 2017. The petition must be filed in accordance with the filing instructions in the “Electronic Submissions (E-Filing)” section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c).

If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), 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 that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html. 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 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (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 NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's 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 document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system.

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC's Electronic Filing Help Desk through the “Contact Us” link located on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html,, by email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and 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, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents 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 NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/ehd,, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly-available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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 NRC's 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: March 27, 2017. A publicly-available version is in ADAMS under Accession No. ML17086A586.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) for Columbia and proposes changes to the containment leakage rate testing programs of Type A, B and C. These tests are required by TS 5.5.12, “Primary Containment Leakage Rate Testing Program,” and these changes would adopt the more conservative allowable test internal extension of Nuclear Energy Institute (NEI) 94-01, Revision 3-A and also adopt American National Standards Institute/American Nuclear Society 56.8-2002, “Containment System Leakage Testing Requirements.”

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 activities involve the revision of Columbia Generating Station (Columbia) Technical Specification (TS) 5.5.12 to allow the extension of the Type A containment test interval to 15 years, and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions.

The proposed extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

The change in Type A test frequency to once-per-fifteen-years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 2.77E-4 person-rem [roentgen equivalent man]/yr (a 0.00761% increase). EPRI [Electric Power Research Institute] Report No. 1009, Revision 2-A states that a very small population dose is defined as an increase of less than 1.0 person-rem per year or less than 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT [integrated leakage rate test] intervals. Moreover, the risk impact when compared to other severe accident risks is negligible. Therefore, the proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

In addition, as documented in NUREG-1493, “Performance-Based Containment Leak-Test Program,” dated January 1995, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The Columbia Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME [American Society of Mechanical Engineers] Section XI, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes two exceptions previously granted. The first exception allowed a one-time extension of the ILRT test frequency for Columbia. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how Columbia is operated. The second exemption to compensate for flow metering inaccuracies in excess of those specified in the American National Standards Institute (ANSI)/American Nuclear Society (ANS) ANSI/ANS 56.8-1994 will be deleted as new test equipment has been acquired with accuracies within the tolerances specified in ANSI/ANS 56.8-1994 and 2002.

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

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

Response: No.

The proposed amendment to the TS 5.5.12, “Primary Containment Leakage Rate Testing Program,” involves the extension of the Columbia Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident.

The proposed change does not involve a physical modification to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.

The proposed amendment also deletes two exceptions previously granted. The first exception granted under TS Amendment No. 191 allowed a one-time extension of the ILRT test frequency for Columbia. This exception was for an activity that has already occurred; therefore, this deletion is solely an administrative action that does not result in any change in how Columbia is operated. The second exemption which was originally granted via Amendment No. 144 to compensate for flow meter inaccuracies in excess of those specified in ANSI/ANS 56.8-1994, will be deleted as new test equipment has been acquired with accuracies within the tolerances specified in ANSI/ANS 56.8-1994 and 2002. These changes to the exceptions in TS 5.5.12 are administrative in nature and do not create the possibility of a new or different kind of accident from any previously evaluated.

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

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

Response: No.

The proposed amendment to TS 5.5.12 involves the extension of the Columbia Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves the extension of the interval between Type A containment leak rate tests and Type C tests for Columbia. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals. The proposed amendment also deletes exceptions previously granted to allow one time extension of the ILRT test frequency for Columbia. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how Columbia is operated and maintained. Thus, there is no reduction in any margin of safety.

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are 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 NW., Washington, DC 20006-3817.

NRC Branch Chief: Robert J. Pascarelli.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of amendment request: March 27, 2017. A publicly-available version is in ADAMS under Accession No. ML17086A587.

Description of amendment request: The proposed amendment would revise or add surveillance requirements (SRs) to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances, which permit performance of the verification. The changes are being made to address the concerns discussed in Generic Letter 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems.” The proposed amendment is consistent with Technical Specification Task Force (TSTF) TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change revises or adds SRs that require verification that the Emergency Core Cooling System (ECCS), Reactor Core Isolation Cooling (RCIC) System, Residual Heat Removal (RHR) Shutdown Cooling System, RHR Drywell Spray System, and RHR Suppression Pool Cooling System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable to perform their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased.

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

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

Response: No.

The proposed change revises or adds SRs that require verification that the ECCS, RCIC System, RHR Shutdown Cooling System, RHR Drywell Spray System, and RHR Suppression Pool Cooling System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. 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. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change does not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions.

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 revises or adds SRs that require verification that the ECCS, RCIC System, RHR Shutdown Cooling System, RHR Drywell Spray System, and RHR Suppression Pool Cooling System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.

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

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

Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817.

NRC Branch Chief: Robert J. Pascarelli.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa

Date of amendment request: March 31, 2017. A publicly-available version is in ADAMS under Package Accession No. ML17102B194.

Description of amendment request: The proposed amendment by NextEra Energy Duane Arnold, LLC (NextEra Duane Arnold) would modify the DAEC Emergency Plan (E Plan) that revises the Emergency Planning Zone (EPZ) boundary for an area beyond the 10 mile required EPZ, specifically, subarea 24 of the EPZ by designating U.S. Highway 30 as its southern boundary. Currently, there is a tract within the DAEC EPZ subarea 24 that is to the south of US Highway 30. This tract in subarea 24 is unique—otherwise, the entire DAEC EPZ is to the north of US Highway 30, which is a four lane, divided highway. Subarea 24 is within Linn County, Iowa. The EPZ boundary change requires that a new Evacuation Time Estimates (ETE) study be performed for the DAEC host counties of Linn and Benton, Iowa, and this revision is also included in the proposal. The proposed change to the southern boundary of the EPZ is considered a reduction in effectiveness as defined in 10 CFR 50, Paragraph 50.54(q)(1)(iv) due to the reduction in EPZ area beyond the 10 mile boundary, and as such, it requires prior NRC approval in accordance with the requirements of 10 CFR 50.54(q)(4). The proposed change to the subarea 24 boundary will enhance law enforcement's ability to evacuate subareas in the Cedar Rapids area as well as improve their ability to control the access back into evacuated metro areas. Further, the proposed change to subarea 24 will make the overall DAEC EPZ boundary more consistent.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

This amendment request would alter portions of the southern, outer EPZ boundary defined in the DAEC E Plan to align with the EPZ boundaries requested by the Linn County Emergency Management Commission. The proposed amendment does not involve any modifications or physical changes to plant systems, structures, or components. The proposed amendment does not change plant operations or maintenance of plant systems, structures, or components, nor does the proposed amendment alter any DAEC E Plan facility or equipment. Changing the EPZ boundaries cannot increase the probability of an accident since emergency plan functions would be implemented after a postulated accident occurs. The proposed amendment does not alter or prevent the ability of the DAEC emergency response organization to perform intended emergency plan functions to mitigate the consequences of, and to respond adequately to, radiological emergencies.

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

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

Response: No.

This amendment request alters the EPZ boundary described in the DAEC E Plan. The proposed amendment does not involve any design modifications or physical changes to the plant, does not change plant operation or maintenance of equipment, and does not alter DAEC E Plan facilities or equipment. The proposed amendment to the DAEC E Plan does not alter any DAEC emergency actions that would be implemented in response to postulated accident events.

The proposed amendment does not create any credible new failure mechanisms, malfunctions, or accident initiators not previously considered.

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 a margin of safety?

Response: No.

This amendment request would alter one subarea in the EPZ boundary defined in the DAEC E Plan. The proposed amendment does not involve any design or licensing bases functions of the plant, no physical changes to the plant are to be made, it does not impact plant operation or maintenance of equipment, and it does not alter DAEC E Plan facilities or equipment. This change does not alter any DAEC emergency actions that would be implemented in response to postulated accident events. The DAEC E Plan continues to meet 10 CFR 50.47 and 10 CFR 50, Appendix E requirements for emergency response.

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

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

Attorney for licensee: William Blair, P. O. Box 14000 Juno Beach, FL 33408-0420.

NRC Branch Chief: David J. Wrona.

Northern States Power Company—Minnesota (NSPM), Docket Nos. 50-263, 50-282 and 50-306, Monticello Nuclear Generating Plant (MNGP), Wright County, and Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP), Goodhue County, Minnesota

Date of amendment request: March 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17090A201.

Description of amendment request: The proposed amendment would revise the PINGP technical specification (TS) 5.3, “Plant Staff Qualifications” and MNGP TS 5.3, “Unit Staff Qualifications,” subsections 5.3.1 to add an exception for licensed operators from the education and experience eligibility requirements of American National Standards Institute (ANSI) N18.1-1971, “Selection and Training of Nuclear Power Plant Personnel,” by requiring that licensed operators comply only with the requirements of 10 CFR part 55, “Operators' Licenses.” Additionally, the proposed change would revise the PINGP and MNGP TS 5.0, “Administrative Controls,” sub-sections 5.1-5.3 by making changes to standardize and align formatting to the extent possible between the TSs.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change revises TS 5.3.1 to take exception to ANSI N18.1-1971 requirements for the education and experience qualifications requirements for licensed operators and requires compliance with 10 CFR 55 and standardizes language between the TS without modifying meaning. An allowance for utilization of a Commission-approved training program that is based upon a SAT [site access training] is contained within 10 CFR 55. The NRC has also stated that the NANT [National Academy for Nuclear Training] guidelines, as endorsed, for initial licensed operator training and qualification are an acceptable way to meet the requirements of 10 CFR 55.

The proposed changes are administrative and do not affect any system that is a contributor to initiating events for previously evaluated accidents. Nor do the changes affect any system that is used to mitigate any previously evaluated accidents.

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 previously evaluated?

Response: No.

The proposed change revises TS 5.3.1 to take exception to ANSI N18.1-1971 requirements for the education and experience qualifications requirements for licensed operators and requires compliance with 10 CFR 55 and standardizes language between the TS without modifying the meaning. An allowance for utilization of a Commission-approved training program that is based upon a SAT is contained within 10 CFR 55. The NRC has also stated that the NANT guidelines, as endorsed, for initial licensed operator training and qualification are an acceptable way to meet the requirements of 10 CFR 55. The proposed change is administrative and does not alter the design, function, or operation of any plant component, nor do they involve installation of any new or different equipment.

Therefore, the proposed change does not create the possibility of a new or difference [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 revises TS 5.3.1 to take exception to ANSI N18.1-1971 requirements for the education and experience qualifications requirements for licensed operators and requires compliance with 10 CFR 55 and standardizes language between the TS without modifying the meaning. An allowance for utilization of a Commission-approved training program that is based upon a SAT is contained within 10 CFR 55. The NRC has also stated that the NANT guidelines, as endorsed, for initial licensed operator training and qualification are an acceptable way to meet the requirements of 10 CFR 55. The proposed change is administrative and does not alter the design, function, or operation of any plant component, nor do they involve installation of any new or different equipment.

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

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

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David J. Wrona.

Northern States Power Company—Minnesota (NSPM), Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

Date of amendment request: March 31, 2017. A publicly-available version is in ADAMS under Accession Package No. ML17095A107.

Description of amendment request: The proposed amendment would revise the current emergency action levels (EAL) scheme used at MNGP to the EAL scheme contained in NEI 99-01, Revision 6, “Development of Emergency Action Levels.”

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change to the MNGP EAL scheme does not impact the physical function of plant structures, systems or components (SSC) or the manner in which the SSCs perform their design function. The proposed change neither adversely affects accident initiators or precursors, nor alters design assumptions. Therefore, the proposed change does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an event. The Emergency Plan, including the associated EALs, is implemented when an event occurs and cannot increase the probability of an accident. Further, the proposed change does not reduce the effectiveness of the Emergency Plan to meet the emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E.

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

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

Response: No.

The proposed change does not involve any physical alteration to the plant, that is, no new or different type of equipment will be installed. The proposed change also does not change the method of plant operation and does not alter assumptions made in the safety analysis. Therefore, the proposed change will not create new failure modes or mechanisms that could result in a new or different kind of accident. The Emergency Plan, including the associated EAL scheme, is implemented when an event occurs and is not an accident initiator.

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

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

Response: No.

Margin of safety is provided by the ability of accident mitigation SSCs to perform at their analyzed capability. The change proposed in this license amendment request does not modify any plant equipment and there is no impact to the capability of the equipment to perform its intended accident mitigation function. The proposed change does not impact operation of the plant or its response to transients or accidents. Additionally, the proposed changes will not change any criteria used to establish safety limits or any safety system settings. The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue to be met.

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

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

Attorney for licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: David J. Wrona.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1 (FCS), Washington County, Nebraska

Date of amendment request: March 24, 2017. A publicly-available version is available in ADAMS under Accession No. ML17094A810.

Description of amendment request: The amendment would revise the renewed facility operating license Paragraph 3.C, “Security and Safeguards Contingency Plans.” The amendment would revise the FCS Cyber Security Plan (CSP) implementation schedule for Milestone 8 (MS8) full implementation date from December 31, 2017, to December 28, 2018.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review 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 amendment request proposes a change to the FCS CSP MS8 completion date as set forth in the CSP implementation schedule and associated regulatory commitments. The NRC staff has concluded that the proposed change: (1) Does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected; (2) does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents; and (3) has no impact on the probability or consequences of an accident previously evaluated. In addition, the NRC staff has concluded that the proposed change to the CSP implementation schedule is administrative in nature.

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 NRC staff has concluded the proposed change: (1) Does not alter accident analysis assumptions, add any initiators, or affect the function of plant systems or the manner in which systems are operated, maintained, modified, tested, or inspected; and (2) does not require any plant modifications which affect the performance capability of the structures, systems, and components relied upon to mitigate the consequences of postulated accidents and does not create the possibility of a new or different kind of accident from any accident previously evaluated. In addition, the NRC staff has concluded that the proposed change to the FCS CSP MS8 implementation schedule is administrative in nature.

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.

Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The delay of the full implementation date for the FCS CSP MS8 has no substantive impact because other measures have been taken which provide adequate protection for the plant during this period of time. Therefore, the NRC staff has concluded that there is no significant reduction in a margin of safety. In addition, the NRC staff has concluded that the proposed change to the FCS CSP MS8 implementation schedule is administrative in nature.

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

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

Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817.

NRC Branch Chief: Douglas A. Broaddus.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1 (FCS), Washington County, Nebraska

Date of amendment request: March 31, 2017. A publicly-available version is available in ADAMS under Accession No. ML17093A309.

Description of amendment request: The amendment would revise the FCS license conditions, definitions, and Technical Specifications (TS) sections to align with those required for the Permanently Defueled Technical Specifications (PDTS) that will reflect decommissioning requirements.

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.

Because the 10 CFR part 50 license for FCS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel with the certifications required by 10 CFR part 50.82(a)(1) submitted, as specified in 10 CFR part 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The only remaining credible accident is a [fuel handling accident (FHA)]. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact the FHA.

The only remaining [Update Safety Analysis Report (USAR)] Chapter 14 postulated accident scenario that could potentially occur at a permanently defueled facility would be a[n] FHA. Remaining Chapter 14 events include an accidental release of waste liquid and heavy load drop. Since the waste gas decay tanks have been purged of their content, and the volume control tanks, liquid holdup tanks, reactor coolant drain tank, and associated systems, contain waste that does not exceed any of the 10 CFR 50.67 limits if an event were to occur. The analyzed accident that remains applicable to FCS in the permanently shutdown and defueled condition is a[n] FHA in the auxiliary building where the SFP is located. The FHA analyses for FCS shows that, following 100 days of decay time after reactor shutdown and provided the [spent fuel pool (SFP)] water level requirements of TS 2.8.3(2) are met, the dose consequences are acceptable without relying on [structures, systems, and components (SSCs)] remaining functional for accident mitigation during and following the event. The one exception to this is the continued function of the passive SFP structure.

The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition and safe storage and handling of fuel will be the only operations performed, and therefore bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation will no longer be credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

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 have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The removal of TS that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor is permanently shutdown and defueled and FCS is no longer authorized to operate the reactor.

The proposed modification or deletion of requirements in the FCS 10 CFR part 50 License and TS do not affect systems credited in the accident analysis for the FHA at FCS.

The proposed license and TS will continue to require proper control and monitoring of systems associated with significant parameters and activities. The TSs continue to preserve the requirements for safe storage and movement of irradiated fuel.

The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining credited barriers for defueled plants (fuel cladding, spent fuel racks, SFP integrity, and SFP water level). Since extended operation in a defueled condition and safe fuel handling will be the only operations performed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

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

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

Response: No.

Because the 10 CFR part 50 license for FCS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel with the certifications required by 10 CFR part 50.82(a)(1) submitted, as specified in 10 CFR part 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. The only remaining credible postulated accident is a[n] FHA. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact the FHA.

The proposed changes are limited to those portions of the license and TS that are not related to the safe storage or movement of irradiated fuel. The requirements that are proposed to be revised or deleted from the FCS license and TS are not credited in the existing accident analysis for the remaining applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated [design-basis accidents (DBAs)] involving the reactor will no longer be possible because the reactor will be permanently shutdown and defueled and FCS will no longer be authorized to operate the reactor.

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

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

Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 K Street NW., Washington, DC 20006-3817.

NRC Branch Chief: Douglas A. Broaddus.

PSEG Nuclear, LLC, and Exelon Generation Company, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: March 6, 2017, as supplemented by letter dated May 4, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17065A241 and ML17125A051, respectively.

Description of amendment request: The amendments would revise Technical Specification (TS) 3.6.2.3, “Containment Cooling System,” to extend the containment fan coil unit allowed outage time (AOT) from 7 days to 14 days for one or two inoperable containment fan coil units.

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 containment fan cooling units (CFCUs) are safety related components which provide the minimum containment cooling as assumed by the containment response analysis for a design-basis loss of coolant accident (LOCA) or main steam line break (MSLB) event. The CFCUs are not accident initiators; the CFCUs are designed to mitigate the consequences of previously evaluated accidents including a design basis LOCA or MSLB event. Extending the AOT for one or two inoperable CFCUs would not affect the previously evaluated accidents since the remaining three CFCUs supplying cooling to containment would continue to be available to perform the accident mitigation functions. Thus allowing one or two CFCUs to be inoperable for an additional 7 days for performance of maintenance or testing does not increase the probability of a previously evaluated accident.

Deterministic and probabilistic risk assessments evaluated the effect of the proposed Technical Specification change on the acceptability of operating with one or two CFCUs inoperable for up to 14 days. These assessments concluded that the proposed Technical Specification change does not involve a significant increase in the risk from CFCU unavailability.

The calculated impact on risk associated with continued operation for an additional 7 days with one or two CFCUs inoperable is very small and is consistent with the acceptance guidelines contained in Regulatory Guides 1.174 and 1.177. This risk is judged to be reasonably consistent with the risk associated with operations for 7 days with one or two CFCUs inoperable as allowed by the current Technical Specifications. The remaining 3 operable CFCUs, in conjunction with the Containment Spray System, are adequate to supply cooling to remove sufficient heat from the reactor containment, following the initial LOCA/MSLB containment pressure transient, to keep the containment pressure from exceeding the design pressure.

The consequences of previously evaluated accidents will remain the same during the proposed 14 day AOT as during the current 7 day AOT. The ability of the remaining 3 TS required CFCUs to maintain containment pressure and temperature within limits following a postulated design basis LOCA or MSLB event will not be affected.

There will be no impact on the source term or pathways assumed in accidents previously evaluated. No analysis assumptions will be changed and there will be no adverse effects on onsite or offsite doses as the result of an accident.

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

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

Response: No.

The proposed Technical Specification change does not involve a change in the plant design, system operation, or procedures involved with the CFCUs. The proposed changes allow one or two CFCUs to be inoperable for additional time. There are no new failure modes or mechanisms created due to plant operation for an extended period to perform CFCU maintenance or testing. Extended operation with one or two inoperable CFCUs does not involve any modification in the operational limits or physical design of plant systems. There are no new accident precursors generated due to the extended AOT.

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.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change, which would increase the AOT from 7 days to 14 days for one or two inoperable CFCUs, does not exceed or alter a setpoint, design basis or safety limit.

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

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

Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: James G. Danna.

South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield, South Carolina

Date of amendment request: May 2, 2017. A publicly-available version is in ADAMS under Accession No. ML17122A353.

Description of amendment request: The amendment request proposes changes to the Protection and Safety Monitoring System (PMS) including the reactor trip system (RTS) and the engineered safety feature actuation system (ESFAS), the passive core cooling system (PXS), the steam generator blowdown system (BDS), and the spent fuel pool cooling system (SFS). In addition, revisions are proposed to COL Appendix A, Technical Specifications. Because, this proposed change requires a departure from Tier 1 information in the Westinghouse Electric Company's AP1000 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 proposed change to add IRWST lower narrow range level instruments addresses the accuracy required to initiate IRWST containment recirculation following a design basis accident in order to mitigate the consequences of the accident. The proposed change to add the new defense-in-depth refueling cavity and SFS isolation on Low IRWST wide range level addresses a seismic or other event resulting in a pipe rupture in the nonsafety-related, nonseismic SFS when connected to the IRWST that could potentially result in a loss of IRWST inventory. Isolation of the SFS from the IRWST to mitigate the consequences of a design basis accident continues to be implemented by the existing containment isolation function, and does not rely on the new defense-in-depth refueling cavity and SFS isolation on Low IRWST wide range level. The addition of RTS and ESFAS P-9 interlocks and blocks does not affect the availability of the actuated equipment to perform their design functions to mitigate the consequences of an accident. The proposed changes do not involve any accident initiating component/system failure or event, thus the probabilities of the accidents previously evaluated are not affected.

The affected equipment does not adversely affect or interact with safety-related equipment or a radioactive material barrier, and this activity does not involve the containment of radioactive material. Thus, the proposed changes would not adversely affect any safety-related accident mitigating function. The radioactive material source terms and release paths used in the safety analyses are unchanged, thus the radiological release in the UFSAR accident analyses are not affected.

These proposed changes to the PMS design do not have an adverse effect on any of the design functions of the affected actuated systems. The proposed changes do not affect the support, design, or operation of mechanical and fluid Systems required to mitigate the consequences of an accident. 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 do the proposed changes create any new accident precursors.

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

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

Response: No.

The proposed change to add IRWST lower narrow range level instruments include requirements similar in function and qualification to many safety-related instruments already performing the affected safety functions as described in the current licensing basis to enable the RTS and ESFAS to perform required design functions, and are consistent with other Updated Final Safety Analysis Report (UFSAR) information. The proposed change to add the new defense-in-depth refueling cavity and SFS isolation on Low IRWST wide range level addresses a seismic or other event resulting in a postulated pipe rupture in the nonsafety-related, nonseismic SFS when connected to the IRWST that could potentially result in a loss of IRWST inventory. Isolation of the SFS from the IRWST to mitigate the consequences of a design basis accident continues to be implemented by the existing containment isolation function, and does not rely on the new defense-in-depth refueling cavity and SFS isolation on Low IRWST wide range level. The addition of RTS and ESFAS P-9 interlocks and blocks does not affect the availability of the actuated equipment to perform their design functions to mitigate the consequences of an accident. This activity does not allow for a new radioactive material release path, result in a new radioactive material barrier failure mode, or create a new sequence of events that would result in significant fuel cladding failures.

The proposed changes revise the PMS design. The proposed changes do not adversely affect the design requirements for the PMS, or the design requirements of associated actuated systems. The proposed changes do not adversely affect the design function, support, design, or operation of mechanical and fluid systems. The proposed changes to the PMS do not result in a new failure mechanism or introduce any new accident precursors. No design function described in the UFSAR is adversely affected by the proposed changes.

Therefore, the requested 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.

No safety analysis or design basis acceptance limit or acceptance criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced. The proposed change to add the new defense-in-depth refueling cavity and SFS isolation of Low IRWST wide range level addresses a seismic or other event resulting in a postulated pipe rupture in the nonsafety-related, nonseismic SFS when connected to the IRWST, maintaining the required IRWST inventory and preserving the original margin of safety assumed for the PXS and SFS.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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: May 10, 2017. A publicly-available version is in ADAMS under Accession No. ML17130A999.

Description of amendment request: The VEGP amendment request proposes changes which involve departures from incorporated plant-specific Tier 2 and Tier 2* Updated Final Safety Analysis Report (UFSAR) information in order to make changes to the design of certain components of the auxiliary building roof reinforcement and roof girders, and other related changes.

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 functions of the auxiliary building roof are to provide support, protection, and separation for the seismic Category I mechanical and electrical equipment located in the auxiliary building. The auxiliary building is a seismic Category I structure and is designed for dead, live, thermal, pressure, safe shutdown earthquake loads, and loads due to postulated pipe breaks. The auxiliary building roof is designed for snow, wind, and tornado loads and postulated external missiles. The proposed changes to UFSAR descriptions and figures are intended to address changes in the detail design of the auxiliary building roof. The thickness and strength of the auxiliary building roof are not reduced. As a result, the design function of the auxiliary building structure is not adversely affected by the proposed changes. 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 do the changes 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 to UFSAR descriptions and figures are proposed to address changes in the detail design of the auxiliary building roof. The thickness, geometry, and strength of the structures are not adversely altered. The concrete and reinforcement materials are not altered. The properties of the concrete are not altered. The changes to the design details of the auxiliary building structure do not create any new accident precursors. As a result, the design function of the auxiliary building structure is not adversely affected by the proposed changes.

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

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

Response: No.

The criteria and requirements of American Concrete Institute (ACI) 349 and American Institute of Steel Construction (AISC) N690 provide a margin of safety to structural failure. The design of the auxiliary building structure conforms to applicable criteria and requirements in ACI 349 and AISC N690 and therefore maintains the margin of safety. The proposed changes to the UFSAR address changes in the detail design of the auxiliary building roof. There is no change to design requirements of the auxiliary building structure. There is no change to the method of evaluation from that used in the design basis calculations. There is not a significant change to the in structure response spectra. No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, thus no margin of safety is reduced.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: March 31, 2017. A publicly-available version is in ADAMS under Accession No. ML17090A209.

Description of amendment request: The requested amendment proposes changes to combined operating license (COL) Appendix C (and plant-specific Tier 1) and Updated Final Safety Analysis Report (UFSAR) Tier 2 that describe; (1) the inspection and analysis of, and specifies the maximum calculated flow resistance acceptance criteria for, the fourth-stage (automatic depressurization system (ADS) loops; (2) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that describes the testing of, and specifies the allowable flow resistance acceptance criteria for, the in-containment refueling water storage tank (IRWST) injection line; (3) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that describes the testing of, and specifies the maximum flow resistance acceptance criteria for, the containment recirculation line; (4) revises licensing basis text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that specifies acceptance criteria for the maximum flow resistance between the IRWST drain line and the containment; and (5) removes licensing basis text from UFSAR Tier 2 that discusses the operation of swing check valves in current operating plants. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule is also requested for the plant-specific Design Control Document Tier 1 material departures.

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 do not adversely affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSCs) accident initiator or initiating sequence of events. The proposed changes do not adversely affect the physical design and operation of the in-containment refueling water storage tank (IRWST) injection, drain, containment recirculation, or fourth-stage automatic depressurization system (ADS) valves, including as-installed inspections and maintenance requirements as described in the Updated Final Safety Analysis Report (UFSAR). Inadvertent operation or failure of the fourth-stage ADS valves are considered as an accident initiator or part of an initiating sequence of events for an accident previously evaluated. However, the proposed change to the test methodology and calculated flow resistance for the fourth-stage ADS lines does not adversely affect the probability of inadvertent operation or failure. Therefore, the probabilities of the accidents previously evaluated in the UFSAR are not affected.

The proposed changes do not adversely affect the ability of IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to perform their design functions. The designs of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves continue to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the proposed changes maintain the capabilities of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to mitigate the consequences of an accident and to meet the applicable regulatory acceptance criteria. The proposed changes do not adversely affect the prevention and mitigation of other abnormal events, e.g., anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR are not affected.

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

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

Response: No.

The proposed changes do not affect the operation of any systems or equipment that might initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed changes do not adversely affect the physical design and operation of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves, including as-installed inspections, and maintenance requirements, as described in the UFSAR. Therefore, the operation of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves is not adversely affected. These proposed changes do not adversely affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-related or nonsafety-related equipment. Therefore, this activity does 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 result in significant fuel cladding failures.

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

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

Response: No.

The proposed changes maintain existing safety margins. The proposed changes verify and maintain the capabilities of the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves to perform their design functions. The proposed changes maintain existing safety margin through continued application of the existing requirements of the UFSAR, while updating the acceptance criteria for verifying the design features necessary to ensure the IRWST injection, drain, containment recirculation, and fourth-stage ADS valves perform the design functions required to meet the existing safety margins in the safety analyses. Therefore, the proposed changes satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not adversely affect any 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 proposed changes, and no margin of safety is reduced.

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

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

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

NRC Branch Chief: Jennifer Dixon-Herrity.

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

Date of amendment request: January 25, 2017, as supplemented by letter dated March 21, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17044A149 and ML17080A405.

Description of amendment request: The amendments would revise certain Surveillance Requirements (SRs) in Technical Specification (TS) 3.8.1, “AC [Alternating Current] Sources—Operating.” The request is for changes in the use of steady state voltage and frequency acceptance criteria for onsite standby power source of the diesel generators (DGs), allowing for the use of new and more conservative design analysis.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below, with NRC edits in square brackets:

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

Response: No.

The proposed amendment would provide more restrictive acceptance criteria for certain DG technical specification surveillance tests. The proposed acceptance criteria changes would help to ensure the DGs are capable of carrying the electrical loading assumed in the safety analyses that take credit for the operation of the DGs. [The proposed changes] would not affect the capability of other structures, systems, and components to perform their design function, and would not increase the likelihood of a malfunction.

Therefore, the proposed amendment does not significantly increase 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 would provide more restrictive acceptance criteria to be applied to existing technical specification surveillance tests that demonstrate the capability of the facility DGs to perform their design function. The proposed acceptance criteria changes would not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

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

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

Response: No.

The proposed DG surveillance requirement changes to voltage and frequency test acceptance criteria are conservative because the minimum steady state voltage increase and the narrowing of the acceptable steady-state frequency range validates use of existing design basis analysis for these test acceptance criteria. Both changes support the use of conservative administrative controls that remain in place, allowing [the] use of the new test acceptance criteria in test procedures until technical specifications reflect these new requirements. The conduct of surveillance tests on safety related plant equipment is a means of assuring that the equipment is capable of maintaining the margin of safety established in the safety analyses for the facility. The proposed amendment does not affect DG performance as described in the design basis analyses, including the capability for the DG to attain and maintain required voltage and frequency for accepting and supporting plant safety loads, should a DG start signal occur. The proposed amendment does not introduce changes to limits established in accident analysis.

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

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

Attorney for licensee: Damon D. Obie, Associate General Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 18101.

NRC Branch Chief: James G. Danna.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3 (BFN), Limestone County, Alabama

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar Nuclear Plant, Units 1 and 2 (WBN), Rhea County, Tennessee

Date of amendment request: April 5, 2017. A publicly-available version is in ADAMS under Accession No. ML17096A620.

Description of amendment request: The amendments would modify technical specification surveillance requirements (SRs) that currently operate ventilation systems with charcoal filters for 10 hours each month in accordance with Technical Specification Task Force (TSTF) Traveler TSTF-522, Revision 0, “Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month.” Specifically, BFN SRs 3.6.4.3.1 and 3.7.3.1, and WBN SRs 3.6.9.1 and 3.7.12.1 are being revised to require operation of the systems for 15 continuous minutes.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change replaces existing Surveillance Requirements to operate the SGT [Standby Gas Treatment] and CREV [Control Room Emergency Ventilation] systems for BFN and the EGT [Emergency Gas Treatment] and ABGT [Auxiliary Building Gas Treatment] systems for WBN, equipped with electric heaters for a continuous 10 hour period every 31 days with a requirement to operate the systems for 15 continuous minutes with heaters operating.

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

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change replaces existing Surveillance Requirements to operate the SGT and CREV systems for BFN and the EGT and ABGT systems for WBN, equipped with electric heaters for a continuous 10 hour period every 31 days with a requirement to operate the systems for 15 continuous minutes with heaters operating.

The change proposed for these ventilation systems does not change any system operations or maintenance activities. Testing requirements will be revised and will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are capable of performing their intended safety functions. The change does not create new failure modes or mechanisms and no new accident precursors are generated.

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

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

Response: No.

The proposed change replaces existing Surveillance Requirements to operate the SGT and CREV systems for BFN and the EGT and ABGT systems for WBN, equipped with electric heaters for a continuous 10 hour period every 31 days with a requirement to operate the systems for 15 continuous minutes with heaters operating.

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

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

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.

NRC Branch Chief: Benjamin G. Beasley.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: March 16, 2017. A publicly-available version is in ADAMS under Accession No. ML17075A229.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.3.1, “Reactor Trip System (RTS) Instrumentation,” Table 3.3.1-1, to increase the values for the nominal trip setpoint and the allowable value for Function 14.a. “Turbine Trip—Low Fluid Oil Pressure.” The proposed amendment also requests changes in accordance with Technical Specifications Task Force (TSTF) Traveler TSTF-493, Revision 4, “Clarify Application of Setpoint Methodology for LSSS [Limiting Safety System Settings] Functions,” Option A, for the affected turbine trip on low fluid oil pressure function setpoints only.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change reflects a design change to the turbine control system that results in the use of an increased control oil [system pressure], necessitating a change to the value at which a low fluid oil pressure initiates a reactor trip on turbine trip. The low fluid oil pressure is an input to the reactor trip instrumentation in response to a turbine trip event. The value at which the low fluid oil initiates a reactor trip is not an accident initiator. A change in the nominal control oil pressure does not introduce any mechanisms that would increase the probability of an accident previously analyzed. The reactor trip on turbine trip function is initiated by the same protective signal as used for the existing auto stop low fluid oil system trip signal. There is no change in form or function of this signal and the probability or consequences of previously analyzed accidents are not impacted.

The proposed change also adds test requirements to the low fluid oil pressure TS instrument function related to those variables to ensure that instruments will function as required to initiate protective systems or actuate mitigating systems at the point assumed in the applicable setpoint calculation. Surveillance tests are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the low fluid oil pressure TS instrument function for which surveillance tests are added are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function.

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

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

Response: No.

The EHC [electrohydraulic control] fluid oil pressure rapidly decreases in response to a turbine trip signal. The value at which the low fluid oil pressure switches initiates a reactor trip is not an accident initiator. The proposed TS change reflects the higher pressure that will be sensed after the pressure switches are relocated from the auto stop low fluid oil system to the EHC high pressure header. Failure of the new switches would not result in a different outcome than is considered in the current design basis. Further, the change does not alter assumptions made in the safety analysis but ensures that the instruments perform as assumed in the accident analysis.

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

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

Response: No.

The change involves a parameter that initiates an anticipatory reactor trip following a turbine trip. The safety analyses do not credit this anticipatory trip for reactor core protection. The original pressure switch configuration and the new pressure switch configuration both generate the same reactor trip signal. The difference is that the initiation of the trip will now be adjusted to a different system of higher pressure. This system function of sensing and transmitting a reactor trip signal on turbine trip remains the same. Also, the proposed change adds test requirements that will assure that technical specifications instrumentation allowable values: (1) Will be limiting settings for assessing instrument channel operability and; (2) will be conservatively determined so that evaluation of instrument performance history and the as left tolerance requirements of the calibration procedures will not have an adverse effect on equipment operability. The testing methods and acceptance criteria for systems, structures, and components, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis including the updated Final Safety Analysis Report. There is no impact to safety analysis acceptance criteria as described in the plant licensing basis because no change is made to the accident analysis assumptions.

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

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

Attorney for licensee: Sherry A. Quirk, General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.

NRC Branch Chief: Benjamin G. Beasley.

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.

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (PVNGS), Maricopa County, Arizona

Date of amendment request: June 29, 2016.

Description of amendment request: The amendments revised the Technical Specifications (TSs) for PVNGS, by modifying the TS requirements to address Generic Letter 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” as described in TS Task Force [TSTF]-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”

Date of issuance: May 16, 2017.

Effective date: As of the date of issuance and shall be implemented within 1 year from the date of issuance.

Amendment Nos.: Unit 1—202, Unit 2—202, and Unit 3—202. A publicly available version is in ADAMS under Accession No. ML17123A435; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revised the Operating Licenses and TSs.

Date of initial notice in Federal Register: August 16, 2016 (81 FR 54613).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 16, 2017.

No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: July 21, 2016. A publicly-available version is in ADAMS under Accession No. ML16209A223.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) for the Oconee Nuclear Station, Units 1, 2, and 3 (ONS); specifically, TS 2.1.1.1, “Reactor Core SLs [Safety Limits],” and TS 5.6.5, “Core Operating Limits Report (COLR),” to allow the use of the COPERNIC fuel performance code.

Date of issuance: May 11, 2017.

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

Amendment Nos.: 403, 405, and 404. A publicly-available version is in ADAMS under Accession No. ML17103A509; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: February 14, 2017 (82 FR 10593).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 11, 2017.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of application for amendment: June 28, 2016, as supplemented by letter dated, August 11, 2016, August 18, 2016, November 14, 2016, December 8, 2016, December 12, 2016, January 9, 2017, January 12, 2017, February 16, 2017, February 21, 2017, March 7, 2017.

Brief description of amendment: The amendment would revise the operating license and technical specifications to implement an increase in rated thermal power from the current licensed thermal power of 3486 megawatts (MWt) to a measurement uncertainty recapture thermal power of 3544 MWt.

Date of issuance: May 11, 2017.

Effective date: As of its date of issuance and shall be implemented within 120 days from the date of issuance, or during the 2017 Refueling Outage if issued on May 13, 2017, or earlier.

Amendment No.: 241. A publicly-available version is in ADAMS under Accession No. ML17095A117; 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: October 4, 2016 (81 FR 68470). The supplemental letter(s) dated August 11, 2016, August 18, 2016, November 14, 2016, December 8, 2016, December 12, 2016, January 9, 2017, January 12, 2017, February 16, 2017, February 21, 2017, and March 7, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 11, 2017.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 2 (ANO-2), Pope County, Arkansas

Date of application for amendment: October 27, 2016, as supplemented by letters dated December 2, 2016, and February 21, 2017.

Brief description of amendment: The amendment authorized a new risk-informed, performance-based fire protection licensing basis for ANO-2, with revised modifications, recovery actions, ignition frequencies, and the application of an NRC-approved fire modeling method. The amendment also revised Attachments M, “License Condition Changes”; Attachment S, “Plant Modifications and Items to be Completed during Implementation”; and Attachment W, “Fire PRA [Probabilistic Risk Assessment] Insights,” of the previously approved National Fire Protection Association (NFPA) 805 amendment.

Date of issuance: May 12, 2017.

Effective date: As of the date of issuance and shall be implemented as described in the transition license conditions.

Amendment No.: 306. A publicly-available version is in ADAMS under Accession No. ML17096A235; 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 renewed facility operating license.

Date of initial notice in Federal Register: January 31, 2017 (82 FR 8869). The supplemental letter dated February 21, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 12, 2017.

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: July 26, 2016.

Brief description of amendments: The amendments revised the Technical Specification (TS) requirements relating to the inservice inspection program required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Code and the inservice testing program required by the ASME Code for Operation and Maintenance of Nuclear Power Plants. The changes are based in part on Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing.”

Date of issuance: May 16, 2017.

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

Amendment Nos.: 225 (Unit 1) and 188 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17103A081; 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 Licenses and Technical Specifications.

Date of initial notice in Federal Register: October 25, 2016 (81 FR 73435).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 16, 2017.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, Pennsylvania

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

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

Date of application for amendments: May 24, 2016, as supplemented by letter dated October 25, 2016.

Brief description of amendments: The amendments eliminated the technical specifications (TS), Section 5.5, “Inservice Testing Program,” to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, “Inservice Test Frequency.” A new defined term, “INSERVICE TESTING PROGRAM,” was added to TS Section 1.1, “Definitions.” This change to the TS is consistent with TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveilance Requirement] Usage Rule Application to Section 5.5 Testing,” with deviations as described in the license amendment request dated May 24, 2016 (ADAMS Accession No. ML16148A047).

Date of issuance: May 11, 2017.

Effective date: As of the date of issuance and shall be implemented within 150 days from the date of issuance.

Amendment Nos.: 298 for DPR-66, 186 for NPF-73, 295 for NPF-3, and 175 for NPF-58. A publicly-available version is in ADAMS under Accession No. ML17081A509; the documents related to these amendments are listed in the Safety Evaluation enclosed with the amendment(s).

Facility Operating License Nos. DPR-66, NPF-73, NPF-3, and NPF-58: The amendments revised the Technical Specifications and the Licenses.

Date of initial notice in Federal Register: August 2, 2016 (81 FR 50732). The supplement dated October 25, 2016, contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 11, 2017.

No significant hazards consideration comments received: No.

PSEG Nuclear, LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek Generating Station, and Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: June 30, 2016.

Brief description of amendments: The amendments revised the Cyber Security Plan (CSP) Milestone 8 implementation schedule for Hope Creek Generating Station (Hope Creek) and Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. Specifically, this change extended the PSEG Nuclear LLC (PSEG) CSP Milestone 8 full implementation date as set forth in the PSEG CSP implementation schedule and revised the Renewed Facility Operating Licenses.

Date of issuance: May 16, 2017.

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

Amendment Nos.: 204 (Hope Creek), 318 (Salem, Unit No. 1), and 299 (Salem, Unit No. 2). A publicly-available version is in ADAMS under Accession No. ML17093A870; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. NPF-57, DPR-70, and DPR-75: The amendments revised the Renewed Facility Operating Licenses.

Date of initial notice in Federal Register: October 4, 2016 (81 FR 68471).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 16, 2017.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company and South Carolina Public Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield, South Carolina

Date of amendment request: July 19, 2016.

Brief description of amendments: The amendments change Combined License (COL) Nos. NPF-93 and NPF-94 for the VCSNS, Units 2 and 3. The amendments change the station's Updated Final Safety Analysis Reports (UFSAR) by departing from the incorporated AP1000 Design Control Document Tier 2 information and involve related changes to the combined operating license (COL) Appendix A Technical Specifications (TS). Specifically, the changes revise the COLs and plant-specific UFSAR Tier 2 information and TS to update the Protection and Safety Monitoring System (PMS) to align with the standards of the Institute of Electrical and Electronics Engineers (IEEE) 603-1991, “IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations.”

Date of issuance: April 10, 2017.

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

Amendment No.: 69. Publicly-available versions are in ADAMS under Accession Nos. ML17041A020 and ML17041A022; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-93 and NPF-94: Amendments revised the COL UFSAR in the form of departures from the incorporated plant-specific DCD Tier 2 information and COL Appendix A TS.

Date of initial notice in Federal Register: August 30, 2016 (81 FR 59659).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 10, 2017.

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: December 16, 2016, and supplemented by letters dated January 12 and February 22, 2017.

Description of amendment: The amendment consists of changes to the VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant specific Design Control Document Tier 2 information. Specifically, the amendment consists of changes to the UFSAR to provide clarification of the interface criteria for nonsafety-related instrumentation that monitors safety-related fluid systems.

Date of issuance: May 1, 2017.

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

Amendment Nos.: 76 and 75. A publicly-available version is in ADAMS under Accession Package No. ML17094A845; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: February 28, 2017 (82 FR 12130). The supplemental letters dated January 12, and February 22, 2017, provided additional information that clarified the application, did not expand the scope of the application request as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated May 1, 2017.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 4, 2016, as supplemented on January 31, 2017.

Description of amendment: This amendment revises License Condition (LC) 2.D(12)(d) related to initial Emergency Action Levels (EALs). The LC will require SNC to submit a fully-developed set of EALs before initial fuel load in accordance with the criteria defined in this license amendment.

Date of issuance: May 18, 2017.

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

Amendment Nos.: 77 (Unit 3) and 76 (Unit 4). A publicly-available version is in ADAMS under Accession Package No. ML17045A537; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

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

Date of initial notice in Federal Register: August 2, 2016 (81 FR 50736). The supplemental letter dated January 31, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated May 18, 2017

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, Unit 2, Rhea County, Tennessee

Date of amendment request: February 16, 2017.

Brief description of amendment: The amendment revised the Technical Specification Containment Leakage Rate Testing Program to allow a one-time extension for the Type C local leak rate test for certain containment isolation valves.

Date of issuance: May 18, 2017.

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

Amendment No.: 11. A publicly-available version is in ADAMS under Accession No. ML17123A228; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-96: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: March 14, 2017 (82 FR 13671).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 18, 2017.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: May 10, 2016, as supplemented by letter dated October 18, 2016.

Brief description of amendments: The amendments would expand primary grade water lockout requirements in Technical Specification (TS) 3.2.E from being applicable in refueling shutdown (RSD) and cold shutdown (CSD) modes to being applicable in RSD, CSD, intermediate shutdown, and hot shutdown modes.

Date of issuance: May 10, 2017.

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

Amendment Nos.: 288 (Unit 1) and 288 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17039A513; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: October 11, 2016 (81 FR 70187). The supplemental letter dated October 18, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 10, 2017.

No significant hazards consideration comments received: No.