Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
Federal Register, Volume 80 Issue 207 (Tuesday, October 27, 2015)
Federal Register Volume 80, Number 207 (Tuesday, October 27, 2015)
Notices
Pages 65807-65822
From the Federal Register Online via the Government Publishing Office www.gpo.gov
FR Doc No: 2015-27042
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NUCLEAR REGULATORY COMMISSION
NRC-2015-0242
Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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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 September 29 to October 9, 2015. The last biweekly notice was published on October 13, 2015.
DATES: Comments must be filed by November 27, 2015. A request for a hearing must be filed by December 28, 2015.
ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0242. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov.
Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
For additional direction on obtaining information and submitting comments, see ``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-2242, email: Paula.Blechman@nrc.gov.
SUPPLEMENTARY INFORMATION:
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Obtaining Information and Submitting Comments
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Obtaining Information
Please refer to Docket ID NRC-2015-0242 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0242.
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 the SUPPLEMENTARY INFORMATION section
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.
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Submitting Comments
Please include Docket ID NRC-2015-0242, facility name, unit number(s), application date, and subject in your comment submission.
The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit
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comment submissions to remove identifying or contact information.
If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.
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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 Sec. 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 should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.
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Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 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/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions, including the opportunity to present evidence and to submit a cross-examination plan for cross-examination of witnesses, consistent with NRC regulations, policies and procedures.
Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii).
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 decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it
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immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of any amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.
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 December 28, 2015. 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 for leave to intervene set forth in this section, except that under Sec. 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. A State, local governmental body, Federally-
recognized Indian tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c).
If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in 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 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. Persons desiring to make a limited appearance are requested to inform the Secretary of the Commission by December 28, 2015.
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Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 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 request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing the E-Submittal server are detailed in the NRC's ``Guidance for Electronic Submission,'' which is available on the agency's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC's Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the documents are submitted through the 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 documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System Help Desk through the ``Contact Us'' link located on the NRC'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 Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North,
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11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-
Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii).
For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: August 28, 2015. A publicly-available version is in ADAMS under Accession No. ML15244B179.
Description of amendment request: The amendment provides a temporary extension to the Completion Time for Technical Specification 3.5.2, ``Emergency Core Cooling Systems (ECCS)--Operating,'' Required Action A.1. The temporary extension will be used to allow the licensee to effect an on-line repair of the Residual Heat Removal (RHR) pump motor air handling unit.
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 ECCS provides a mitigating function, and as such, it does not impact the probability of an accident. The consequences of an accident requiring the ECCS function will continue to be mitigated by the operable 1B RHR system train during the extended period in which the 1A RHR system train is considered inoperable. Each of the two RHR trains are redundant, so the 1B RHR pump is capable of performing the necessary mitigating function.
Additionally, engineering evaluations, as documented in the Engineering Change (EC) process, demonstrate that the 1A RHR pump will continue to be capable of performing its mitigating ECCS function using a defense-in-depth measure that establishes alternate forced cooling to the room.
As such, the proposed amendment does not result in an increase in consequences of an accident.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
No new accident causal mechanisms are created as a result of this proposed license amendment request (LAR). No changes are being made to any SSC structure, system, or component that will introduce any new accident causal mechanisms. The defense-in-depth measure to install alternate forced cooling to the 1A RHR pump motor during the repair evolution has been analyzed and evaluated using the Duke Energy EC process.
The EC concludes that the installation of alternate forced cooling equipment would not adversely impact other components such that a new or different accident scenario is created.
3. Does the proposed amendment involve a significant reduction in the margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of the fuel cladding, reactor coolant and containment systems will not be impacted by the proposed LAR.
The proposed activity only impacts the amount of time that the 1A RHR system can be considered inoperable. The amount of inoperable time still remains small relative to the total operating time, and the 1A RHR train would still be considered available (i.e., capable of performing its ECCS function) during the period of extended inoperability. However, even if the train were considered unavailable, the total hours of unavailability would remain bounded by the limits established by the Maintenance Rule program.
Therefore, it is concluded that the proposed changes do 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: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 28202.
NRC Branch Chief: Robert J. Pascarelli.
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: May 19, 2015, as supplemented by letter dated August 20, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML15146A056 and ML15239B290, respectively.
Description of amendment request: The proposed amendments add a Reactor Protective System Nuclear Overpower--High Setpoint trip for three (3) reactor coolant pump operation to Technical Specification Table 3.3.1-1, ``Reactor Protective System Instrumentation.'' The existing overpower protection for three (3) reactor coolant pump operation is the Nuclear Overpower Flux/Flow/Imbalance trip function. The new setpoint provides an absolute setpoint that can be actuated regardless of the transient or Reactor Coolant System flow conditions and provides a
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significant margin gain for the small steam line break accident.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment adds a high flux trip for three (3) Reactor Coolant Pump (RCP) Operation by modifying the existing Nuclear Overpower-High Setpoint function in Technical Specification (TS) Table 3.3.1-1 to delineate between a setpoint valid for four (4) RCP operation and three (3) RCP operation. TS 3.4.4 is modified to require the Nuclear Overpower--High Setpoint to be reset to less than or equal to the Allowable Value of Table 3.3.1-1 for three (3) RCPs operating. The proposed change provides automatic overpower protection when the plant is operating with three (3) RCPs. The existing overpower protection for three (3) RCP operation is the Nuclear Overpower Flux/Flow/Imbalance trip function. Providing a Nuclear Overpower flux setpoint provides an absolute setpoint that can be actuated regardless of the transient or RCS flow conditions. The proposed TS change does not modify the reactor coolant system pressure boundary, nor make any physical changes to the facility design, material, or construction standards. The probability of any design basis accident (DBA) is not affected by this change, nor are the consequences of any DBA significantly affected by this change. The proposed change does not involve changes to any structures, systems, or components (SSCs) that can alter the probability for initiating a LOCA loss-of-coolant accident event. This amendment request includes the adoption of Option A of Technical Specification Task Force (TSTF) TSTF-493-A, Revision 4, ``Clarify Application of Setpoint Methodology for LSSS Limiting Safety System Setting Functions,'' for the Nuclear Overpower--High Setpoint trip function of TS Table 3.3.1-1. The TS changes associated with the implementation of TSTF-493-A will provide additional assurance that the instrumentation setpoints for the Nuclear Overpower--High Setpoint trip function are maintained consistent with the setpoint methodology to ensure the required automatic trips and safety feature actuations occur such that the safety limits are not exceeded.
Therefore, the proposed TS changes do 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 amendment adds a high flux trip for three (3) Reactor Coolant Pump Operation by modifying the existing Nuclear Overpower-High Setpoint function in TS Table 3.3.1-1 to delineate between a setpoint valid for four (4) RCP operation and three (3) RCP operation. This proposed change and the implementation of TSTF-
493-A do not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. No new failure modes are identified, nor are any SSCs required to be operated outside the design bases.
Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment adds a high flux trip for three (3) Reactor Coolant Pump Operation by modifying the existing Nuclear Overpower-High Setpoint function in TS Table 3.3.1-1 to delineate between a setpoint valid for four (4) RCP operation and three (3) RCP operation. This proposed TS change and the implementation of TSTF-493-A do not involve: (1) A physical alteration of the Oconee Units; (2) the installation of new or different equipment; or (3) any impact on the fission product barriers or safety limits. The proposed change adds a new setpoint, which is more conservative than the existing high flux setpoint that initiates a protective action to provide protection for power excursion events initiated from three (3) RCP operation equivalent to that provided for four (4) RCP operation.
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: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon Street--DEC45A, Charlotte, NC 28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington
Date of amendment request: September 2, 2015. A publicly-available version is in ADAMS under Accession No. ML15245A777.
Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) by adding Limiting Condition for Operation (LCO) 3.0.9 to address the impact of unavailable barriers, not explicitly addressed in the TSs but required for operability of supported systems in TSs. The LCO 3.0.9 establishes conditions under which TS systems would remain operable when required physical barriers are not capable of providing their safety-related function. Also, the proposed amendment would replace the term ``train'' with the term ``division'' in LCO 3.0.8 to be consistent with the terminology proposed in LCO 3.0.9, which is editorial in nature.
The proposed changes to the TS are consistent with the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification change traveler TSTF-427, ``Allowance for Non-Technical Specification Barrier Degradation on Supported System OPERABILITY,'' Revision 2 (ADAMS Accession No. ML061240055). The availability of the TS improvement and the model application was published in the Federal Register on October 3, 2006 (71 FR 58444), as part of the Consolidated Line Item Improvement Process (CLIIP).
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee affirmed the applicability of the model no significant hazards consideration determination, which is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated
The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed LCO 3.0.9 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.9. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported
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system TS when inoperability is due solely to an unavailable barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.
Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in NRC Regulatory Guide 1.177, ``An Approach for Plant-Specific Risk-Informed Decisionmaking: Technical Specifications,'' August 1998 (ADAMS Accession No. ML003740176). A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.9 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP incremental conditional core damage probability and ICLERP incremental large early release probability) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation published in the Federal Register on October 3, 2006 (71 FR 58444).
Therefore, this change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the above 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: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station (PNPS), Plymouth County, Massachusetts
Date of amendment request: July 15, 2015. A publicly available version is in ADAMS under Accession No. ML15205A287.
Description of amendment request: The amendment would revise the PNPS Cyber Security Plan (CSP) Implementation Schedule Milestone 8 full implementation date. The amendment would also revise the PNPS Facility Operating License No. DPR-35.
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 CSP Implementation Schedule is administrative in nature. This change 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. The proposed change 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 have no impact on the probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of 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 to the CSP Implementation Schedule is administrative in nature. This proposed change 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. The proposed change 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 do not create the possibility of a new or different kind of accident from any accident previously evaluated.
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.
Plant safety margins are established through limiting conditions for operation, limiting safety system settings, and safety limits specified in the technical specifications. The proposed change to the CSP Implementation Schedule is administrative in nature. In addition, the milestone date delay for full implementation of the CSP has no substantive impact because other measures have been taken which provide adequate protection during this period of time. Because there is no change to established safety margins as a result of this change, the proposed change does not involve a significant reduction in a margin of safety.
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 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. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.
NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: August 18, 2015. A publically-available version is in ADAMS under Accession No. ML15231A097.
Description of amendment request: The proposed change would revise the reactor steam dome pressure specified in the technical specification (TS) safety limits.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the reactor steam dome pressure in the CPS Clinton Power Station, Unit 1, DNPS Dresdent Nuclear Power Station, Units 2 and 3, and QCNPS Quad Cities Nuclear Power Station, Units 1 and 2 Reactor Core Safety Limits TS 2.1.1.1 and 2.1.1.2 does not alter the use of the analytical methods used to determine the safety limits that have been previously reviewed and approved by the NRC. The proposed change is in accordance with an NRC approved critical power correlation
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methodology, and as such, maintains required safety margins. The proposed change does not adversely affect accident initiators or precursors, nor does it alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained.
The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not require any physical change to any plant SSCs nor does it require any change in systems or plant operations. The proposed change is consistent with the safety analysis assumptions and resultant consequences.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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 reduction in the reactor dome pressure safety limit from 785 psig pounds per square inch gauge to 685 psig is a change based upon previously approved documents and does not involve changes to the plant hardware or its operating characteristics. As a result, no new failure modes are being introduced. There are no hardware changes nor are there any changes in the method by which any plant systems perform a safety function. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change.
The proposed change does not introduce any new accident precursors, nor does it involve any physical plant alterations or changes in the methods governing normal plant operation. Also, the change does not impose any new or different requirements or eliminate any existing requirements. The change does not alter assumptions made in the safety analysis.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The margin of safety is established through the design of the plant structures, systems, and components, and through the parameters for safe operation and setpoints for the actuation of equipment relied upon to respond to transients and design basis accidents. Evaluation of the 10 CFR part 21 condition by General Electric determined that since the Minimum Critical Power Ratio improves during the PRFO pressure regulator failure maximum demand (open) transient, there is no decrease in the safety margin and therefore there is not a threat to fuel cladding integrity. The proposed change in reactor dome pressure supports the current safety margin, which protects the fuel cladding integrity during a depressurization transient, but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The change does not alter the behavior of plant equipment, which remains unchanged.
The proposed change to Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2 is consistent with and within the capabilities of the applicable NRC approved critical power correlation for the fuel designs in use at CPS, DNPS, and QCNPS. No setpoints at which protective actions are initiated are altered by the proposed change. The proposed change does not alter the manner in which the safety limits are determined. This change is consistent with plant design and does not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.
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 requested amendments involve no significant hazards consideration.
Attorney for licensee: Bradley Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: July 2, 2015. A publicly-available version is in ADAMS under Accession No. ML15198A153.
Description of amendment request: The amendments would revise the technical specifications (TSs) related to communications and manipulator crane requirements. The licensee requested that these requirements be relocated to the Updated Final Safety Analysis Report (UFSAR) and related procedures and be controlled in accordance with 10 CFR 50.59, ``Changes, tests, and experiments.''
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 changes remove from the TSs the current necessity of establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits and relocate the requirements to the UFSAR and related procedures, which will have no impact on any safety related structures, systems or components. Once relocated to the UFSAR and related procedures, changes to establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits will be controlled in accordance with 10 CFR 50.59.
The probability of occurrence of a previously evaluated accident is not increased because these changes do not introduce any new potential accident initiating conditions. The consequences of accidents previously evaluated in the UFSAR are not affected because the ability of the components to perform their required functions is not affected.
Therefore, the proposed changes do 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 changes remove from the TSs the current necessity of establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits and relocate the requirements to the UFSAR and related procedures, which will have no impact on any safety related structures, systems or components. Once relocated to the UFSAR and related procedures, changes to establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits will be controlled in accordance with 10 CFR 50.59.
The proposed changes do not introduce new modes of plant operation and do not involve physical modifications to the plant (no new or different type of equipment will be installed). There are no changes in the method by which any safety related plant structure, system, or component (SSC) performs its specified safety function. As such, the plant conditions for which the design basis accident analyses were performed remain valid.
No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of the proposed change. There will be no adverse effect or challenges imposed on any SSC as a result of the proposed changes.
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 change involve a significant reduction in a margin of safety?
Response: No.
Margin of safety is related to confidence in the ability of the fission product barriers to
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perform their accident mitigation functions. The proposed changes remove from the TSs the current necessity of establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits and relocate the requirements to the UFSAR and related procedures, which will have no impact on any safety related structures, systems or components. Once relocated to the UFSAR and related procedures, changes to establishing and maintaining communications between the control room and the refueling station and the minimum load capacities and load limit controls required for the manipulator crane limits will be controlled in accordance with 10 CFR 50.59. The proposed changes do not physically alter any SSC. There will be no effect on those SSCs necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, loss of cooling accident peak cladding temperature (LOCA PCT), or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met.
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 S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, Juno Beach, FL 33408-0420.
NRC Branch Chief: Shana R. Helton.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center (DAEC), Linn County, Iowa
Date of amendment request: August 18, 2015. A publicly-available version is in ADAMS under Accession No. ML15246A445.
Description of amendment request: The proposed amendment would revise the technical specifications (TSs) Section 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' by replacing the reference to the NRC Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test Program,'' with a reference to the Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, appendix J,'' and conditions and limitations specified in NEI 94-01, Revision 2-A, as the implementation document used by DAEC to implement the performance-
based containment leakage rate testing program.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' for development of the DAEC performance-based containment testing program. NEI 94-01 allows, based on risk and performance, an extension of Type A and Type C containment leak test intervals. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses.
The findings of the DAEC risk assessment confirm the general findings of previous studies that the risk impact with extending the containment leak rate is small. Per the guidance provided in Regulatory Guide 1.174, an extension of the leak test interval in accordance with NEI 94-01, Revision 3-A results in an estimated change within the very small change region.
Since the change is implementing a performance-based containment testing program, the proposed amendment does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The requirement for containment leakage rate acceptance will not be changed by this amendment. Therefore, the containment will continue to perform its design function as a barrier to fission product releases.
Therefore, the proposed changes do 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 to implement a performance-based containment testing program, associated with integrated leakage rate test frequency, does not change the design or operation of structures, systems, or components of the plant.
The proposed changes would continue to ensure containment integrity and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by these changes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test frequency, does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation.
The specific requirements and conditions of the TS Primary Containment Leakage 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. This 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 implementation of a performance-based containment testing program.
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: Mr. James Petro, P.O. Box 14000 Juno Beach, FL 33408-0420.
NRC Branch Chief: David L. Pelton.
South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: May 4, 2015. A publicly-available version is in ADAMS under Accession No. ML15124A911.
Description of amendment request: The proposed amendment and exemption identify portions of the licensing basis that would more appropriately be classified as Tier 2, specifically the Tier 2* information on Fire Area Figures 9A-1, 9A-2, 9A-3, 9A-4, 9A-5, and 9A-
201 in the Virgil C. Summer Nuclear Station Units 2 and
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3 Updated Final Safety Analysis Report. With the reclassification, prior NRC approval would continue to be required for any safety significant changes to the Fire Area Figures because any revisions to that information would follow the Tier 2 change process provided in 10 CFR part 52, appendix D, Section VIII.B.5.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would reclassify Fire Area Figures Tier 2* information. The proposed amendment does not modify the design, construction, or operation of any plant structures, systems, or components (SSCs), nor does it change any procedures or method of control for any SSCs. Because the proposed amendment does not change the design, construction, or operation of any SSCs, it does not adversely affect any design function as described in the Updated Final Safety Analysis Report.
Therefore, the proposed amendment does not affect the probability of an accident previously evaluated. Similarly, because the proposed amendment does not alter the design or operation of the nuclear plant or any plant SSCs, the proposed amendment does not represent a change to the radiological effects of an accident, and therefore, does not involve an increase in the consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment would reclassify Fire Area Figures Tier 2* information. The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The only impact of this activity is the reclassification of information in the Updated Final Safety Analysis Report.
Because the proposed amendment only reclassifies information and does not change the design, construction, or operation of the nuclear plant or any plant operations, the 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 amendment would reclassify Fire Area Figures Tier 2* information. The proposed amendment is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The only impact of this activity is the reclassification of information in the Updated Final Safety Analysis Report.
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: Ms. Kathryn M. Sutton, Morgan, Lewis & Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: August 31, 2015. A publicly-available version is in ADAMS under Accession No. ML15261A673.
Description of amendment request: The proposed change would eliminate the current requirement to perform the Residual Heat Removal (RHR) autoclosure interlock Surveillance Requirement (SR) 3.4.14.2 and revise Action Condition C to eliminate the RHR autoclosure interlock from the Action Condition.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The two motor-operated gate valves located in each RHR System suction line are normally-closed to maintain the low pressure RHR System (design pressure of 600 psig) isolated from the high pressure reactor coolant system RCS (normal operating pressure of 2235 psig). An autoclosure interlock ACI was provided to isolate the low pressure RHR System from the RCS when the pressure increases above the ACI setpoint. However, spurious ACI actuation has resulted in RHR System isolation and subsequent loss of decay heat removal capability. The removal of the ACI feature will preclude this inadvertent isolation, thus increasing the likelihood that RHR will be available to remove decay heat. The addition of a control room alarm to alert the operator that a suction/isolation valve(s) is not fully closed when the RCS pressure is above the alarm setpoint and administrative procedures will ensure that the RHR System will be isolated from the RCS, if the RCS pressure increases above the alarm setpoint, which will decrease the likelihood of an interfacing system Loss-of-Coolant Accident LOCA. Therefore, the performance of the RHR System would not be adversely affected by the ACI deletion and the RHR suction isolation valve alarm installation.
The RHR ACI provides automatic closure to the RHR System suction isolation valves on high RCS pressure; however, rapid overpressure protection of the RHR System is provided by the RHR relief valves and not by the slow acting suction isolation valves. This RHR System overpressure protection is not affected by the removal of the ACI, this feature also serves to decrease the likelihood of an interfacing system LOCA. Thus, the RHR System integrity will not be affected by the removal of the ACI feature. In addition, the removal of the ACI feature does not adversely affect any fission barrier, alter any assumptions made in the radiological consequences evaluations, or affect the mitigation of radiological consequences.
The impact of ACI removal on RHR shutdown cooling, low temperature overpressure protection, and interfacing system LOCA initiating event frequency was assessed. For each of these areas that were assessed, it was concluded that the removal of ACI and the accompanying plant changes provides a benefit to plant safety.
With the deletion of the ACI, there is no longer any potential for spurious automatic closure of a RHR System suction isolation valve resulting in inadvertent RHR System isolation and loss of shutdown cooling.
Therefore, it is concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The removal of the RHR System ACI, and corresponding TS requirements, does not result in the initiation of any accident nor create any new credible limiting single failures.
The removal of the ACI eliminates the potential for spurious circuitry actuation causing isolation of the RHR system. Furthermore, the addition of an alarm to alert the operator that a suction valve is not fully closed when RCS pressure is above the alarm setpoint reduces the likelihood that the RHR system will be exposed to high pressure conditions. These modifications and the resulting elimination of the ACI TS Surveillance Requirement will not result in the RHR system being operated in any unanalyzed modes, either during normal or accident conditions. Also, the AHA system will continue to be maintained and surveilled as it is currently.
No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The proposed change does not
Page 65816
challenge the performance or integrity of any safety-related system.
Therefore, it is concluded that 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.
Removal of the ACI interlock, and its corresponding TS Surveillance Requirement, does not alter or prevent any plant response such that the margin of safety to any applicable acceptance criteria is significantly decreased. In fact, the addition of a control room alarm that identifies that the suction valve is not fully open, together with the existing overpressure alarm, ensures that the margin of safety to an AHA overpressure condition is not significantly reduced.
Furthermore, the actuation of safety-related components and the response of plant systems to accident scenarios are not affected, and thus will remain as assumed in the safety analysis.
Therefore, the proposed change will not adversely affect the operation or safety function of equipment assumed in the safety analysis.
For the reasons noted above, it is concluded that the proposed change does not involves 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: Leigh D. Perry, SVP & General Counsel of Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J. Pascarelli.
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: September 1, 2015. A publicly-available version is in ADAMS under Accession No. ML15244A602).
Description of amendment request: The proposed change would amend Combined License (COL) Nos. NPF-91 and NPF-92 for the VEGP Units 3 and 4. The requested amendment proposes to revise the VEGP Units 3 and 4 plant-specific emergency planning inspections, tests, analyses, and acceptance criteria in Appendix C of the VEGP Units 3 and 4 COLs, to remove the copy of Updated Final Safety Analysis Report (UFSAR) Table 7.5-1, ``Post-Accident Monitoring System,'' from Appendix C of the VEGP Units 3 and 4 COLs.
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 VEGP Units 3 and 4 emergency planning inspections, tests, analyses, and acceptance criteria (ITAAC) provide assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's rules and regulations. The proposed change to remove the copy of UFSAR Table 7.5-1 from Appendix C of the VEGP Units 3 and 4 COLs does not affect the design of a system, structure, or component (SSC) used to meet the design bases of the nuclear plant. Nor do the changes affect the construction or operation of the nuclear plant itself, so there is no change to the probability or consequences of an accident previously evaluated. Removing the copy of UFSAR Table 7.5-1 from Appendix C of the COLs does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related SSC or function is adversely affected. The changes do not involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. Because the changes do not involve any safety-related SSC or function used to mitigate an accident, 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 VEGP Units 3 and 4 emergency planning ITAAC provide assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's rules and regulations. The changes do not affect the design of an SSC used to meet the design bases of the nuclear plant, nor do the changes affect the construction or operation of the nuclear plant. Consequently, there is no new or different kind of accident from any accident previously evaluated. The changes do not affect safety-related equipment, nor do they affect equipment which, if it failed, could initiate an accident or a failure of a fission product barrier. In addition, the changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment.
No analysis is adversely affected. No system or design function or equipment qualification is adversely affected by the changes. This activity will not allow for a new fission product release path, result in a new fission product barrier failure mode, nor create a new sequence of events that would 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 VEGP Units 3 and 4 emergency planning ITAAC provide assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's rules and regulations. The changes do not affect the assessments or the plant itself. The changes do not adversely interface with safety-related equipment or fission product barriers. No safety analysis, design basis limit or acceptance criterion are challenged or exceeded by the proposed change.
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: Mr. M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424, 50-425, 52-025, 52-026, Vogtle Electric Generating Plant, Units 1, 2, 3, and 4, Burke County, Georgia and Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, City of Dalton, GA
Date of amendment request: August 31, 2015. A publicly-available version is in ADAMS under Accession Package No. ML15246A045.
Description of amendment request: The amendments request NRC approval of a standard emergency plan for all Southern Nuclear Operating Company, Inc., sites and site-specific annexes.
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:
Page 65817
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes have no effect on normal plant operation or on any accident initiator or precursors, and do not impact the function of plant structures, systems, or components (SSCs). The proposed changes do not alter or prevent the ability of the emergency response organization to perform its intended functions to mitigate the consequences of an accident or event. The ability of the emergency response organization to respond adequately to radiological emergencies has been demonstrated as acceptable through a staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will not change the design function or operation of SSCs. The changes do not impact the accident analysis. The changes do not involve a physical alteration of the plant, a change in the method of plant operation, or new operator actions. The proposed changes do not introduce failure modes that could result in a new accident, and the changes do not alter assumptions made in the safety analysis. As demonstrated by the SNC staffing analysis performed in accordance with 10 CFR 50 Appendix E.IV.A.9, the proposed changes do not alter or prevent the ability of the emergency response organization to perform its intended functions to mitigate the consequences of an accident or event.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the proposed changes involve a significant reduction in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the Emergency Plan and do not impact operation of the plant or its response to transients or accidents. The changes do not affect the Technical Specifications. The changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected. The Standard Emergency Plan will continue to provide the necessary response staff for emergencies as demonstrated by staffing and functional analyses including the necessary timeliness of performing major tasks for the functional areas of the Emergency Plan. The proposed changes do not adversely affect SNC's ability to meet the requirements of 10 CFR 50 Appendix E and the emergency planning standards of 10 CFR 50.47.
Therefore, the proposed changes do 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: Leigh D. Perry, SVP & General Counsel of Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J. Pascarelli.
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Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3 (WF3) St. Charles Parish, Louisiana
Date of amendment request: July 2, 2015, as supplemented by letter dated August 14, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML15197A106 and ML15226A346, respectively.
Brief description of amendment: This notice is being reissued in its entirety to remove information that was inadvertently included in the notice published in the Federal Register on September 29, 2015 (80 FR 58520), for WF3. The proposed amendment will modify the Technical Specification (TS) 3.1.3.4, ``Control Element Assembly CEA Drop Time'' and Final Safety Analysis Report, Chapter 15, ``Accident Analyses.'' The proposed amendment would change TS 3.1.3.4 to revise the arithmetic average of all CEA drop times to be less than or equal to 3.5 seconds.
Date of publication of individual notice in the Federal Register: September 8, 2015 (80 FR 53892).
Expiration date of individual notice: October 8, 2015 (public comments); and November 9, 2015 (hearing requests).
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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 Commissions related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of amendment request: October 21, 2014, as supplemented by letters dated June 18, and July 28, 2015.
Description of amendment: The amendment revised the emergency action level scheme for Fermi 2 based
Page 65818
on the Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors,'' dated November 2012.
Date of issuance: September 29, 2015.
Effective date: As of the date of issuance and shall be implemented within 120 days of issuance.
Amendment No.: 202. A publicly-available version is in ADAMS under Accession No. ML15233A084; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-43: Amendment revised the Facility Operating License.
Date of initial notice in Federal Register: December 23, 2014 (79 FR 77045). The supplemental letters dated June 18, and July 28, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2015.
No significant hazards consideration comments received: None.
Entergy Gulf States Louisiana, LLC and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: June 10, 2014, as supplemented by letters dated October 9, and December 31, 2014, and January 30, 2015.
Brief description of amendment: By order dated August 14, 2015, as published in the Federal Register on August 24, 2015 (80 FR 51329), the NRC approved a direct license transfer for Facility Operating License No. NPF-47 for the River Bend Station, Unit 1. This amendment reflects the direct transfer of the license to Entergy Louisiana, LLC.
Date of issuance: October 1, 2015.
Effective date: As of the date of issuance and shall be implemented 30 days from the date of issuance.
Amendment No.: 189. A publicly-available version of the amendment and the order are in ADAMS under Accession Nos. ML15265A116 and ML15146A410, respectively; documents related to this amendment are listed in the safety evaluation (SE) enclosed with the order dated August 14, 2015. Subsequent to the issuance of the order, the licensee submitted a letter dated September 23, 2015 (ADAMS Accession No. ML15268A338). This letter provided additional notifications of regulatory approvals and the closing transaction date, as was required by the order.
Facility Operating License No. NPF-47: The amendment revised the Facility Operating License.
Date of initial notice in Federal Register: August 24, 2015 (80 FR 51329). The supplements dated October 9, and December 31, 2014, January 30, and September 23, 2015, contained clarifying information, did not expand the application beyond the scope of the notice as originally published in the Federal Register, and did not affect the applicability of the generic no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in an SE dated August 14, 2015.
Comments received: Yes. The comments received on the license transfer request are addressed in the SE dated August 14, 2015.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: March 28, 2014, as supplemented by letters dated April 24, June 9, June 11, and August 13, 2014; and May 4, 2015.
Brief description of amendment: The amendment revised the renewed facility operating license and the associated technical specifications to be consistent with the permanent cessation of reactor operations and permanent defueling of the reactor.
Date of issuance: October 7, 2015.
Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.
Amendment No.: 263. A publicly-available version is in ADAMS under Accession No. ML15117A551; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-28: Amendment revised the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 17, 2015 (80 FR 8358). The supplemental letters dated April 24, June 9, June 11, and August 13, 2014; and May 4, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated October 7, 2015.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: July 23, 2015, as supplemented by letters dated July 28, 2015, and August 25, 2015.
Brief description of amendment: The amendment modified the technical specifications (TSs) to allow for the temporary operation of the borated water storage tank (BWST) under administrative and design controls while connected to seismic Class II piping. This change would support necessary cleanup and surveillance activities associated with the TMI Fall 2015 Refueling Outage and Fuel Cycle 21 operation.
Date of issuance: October 1, 2015.
Effective date: As of the date of issuance and shall be implemented within 7 days.
Amendment No.: 289. A publicly-available version is in ADAMS under Accession No. ML15225A158; documents related to this amendment are listed in the safety evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-50: Amendment revised the Renewed Facility Operating License and TSs.
Date of Initial Notice in Federal Register: August 7, 2015 (80 FR 47529). The supplemental letter dated August 25, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 1, 2015.
No significant hazards consideration comments received: No.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: December 5, 2014.
Brief description of amendments: The amendments revised Technical Specifications (TSs) to adopt Technical Specification Task Force Traveler 439, Revision 2, ``Eliminate Second
Page 65819
Completion Times Limiting Time from Discovery of Failure to Meet an LCO Limiting Condition for Operation.'' The second completion times associated with TS 3.6.2.1, ``Containment Spray and Cooling Systems,'' were deleted.
Date of Issuance: October 5, 2015.
Effective Date: As of the date of issuance and shall be implemented within 90 days of issuance.
Amendment Nos. 228 and 178. A publicly-available version is in ADAMS under Accession No. ML15251A094; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the TSs.
Date of initial notice in Federal Register: February 3, 2015 (80 FR 5801).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 5, 2015.
No significant hazards consideration comments received: No.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: October 7, 2014.
Brief description of amendments: The amendments revised the scheduled completion date for Milestone 8 of the Cyber Security Plan implementation schedule and License Condition 3.E in Renewed Facility Operating License Nos. DPR-31 and DPR-41.
Date of issuance: September 28, 2015.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment Nos.: 266 and 261. The amendments are in ADAMS under Accession No. ML15233A379; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: January 6, 2015 (80 FR 535).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 28, 2015.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No.50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: July 13, 2015.
Brief description of amendment: The amendment revised the Technical Specifications (TSs). The amendment added a note to TS Surveillance Requirement 4.4.1.3.4, which requires verification that residual heat removal loop operations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.
Date of issuance: October 6, 2015.
Effective date: As of its date of issuance and shall be implemented within 30 days of issuance.
Amendment No.: 150. A publicly-available version is in ADAMS under Accession No. ML15231A144; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-86: Amendment revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 4, 2015 (80 FR 46350).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 6, 2015.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant (DCPP), Unit Nos. 1 and 2, San Luis Obispo County, California
Date of amendment request: October 17, 2014, as supplemented by letter dated February 19, 2015.
Brief description of amendments: The amendments revised the DCPP Cyber Security Plan (CSP) Milestone h full implementation schedule as set forth in the CSP implementation schedule.
Date of issuance: September 30, 2015.
Effective date: As of its date of issuance and shall be implemented within 60 days from the date of issuance. All subsequent changes to the NRC-approved CSP implementation schedule as approved by the NRC staff with this license amendment will require prior NRC approval pursuant to 10 CFR 50.90.
Amendment Nos.: Unit 1--220; Unit 2--222. A publicly-available version is in ADAMS under Accession No. ML15245A542; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses.
Date of initial notice in Federal Register: April 7, 2015 (80 FR 18659). The supplemental letter dated February 19, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 30, 2015.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: May 26, 2015, as supplemented by letter dated May 28, 2015 and as revised by letters dated June 9, and June 29, 2015.
Description of amendment: The amendment authorized changes to the VCSNS Units 2 and 3 Updated Final Safety Analysis Report on the applicability of the American Institute of Steel Construction (AISC) N690-1994, ``Specification for the Design, Fabrication and Erection of Steel Safety-Related Structures for Nuclear Facilities,'' to allow use of the American Welding Society (AWS) D1.1-2000, ``Structural Welding Code-Steel,'' in lieu of the AWS D1.1-1992 edition identified in AISC N690-1994.
Date of issuance: September 1, 2015.
Effective date: As of the date of issuance and shall be implemented within 90 days of issuance.
Amendment No.: 30. A publicly-available version is in ADAMS under Accession No. ML15224A750; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised the Facility Combined Licenses.
Date of initial notice in Federal Register: June 8, 2015 (80 FR 32413). However, the June 29, 2015, letter revised the application including the No Significant Hazard Determination. Therefore, the staff issued a revised notice on July 9, 2015, (80 FR 39450).
The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated September 1, 2015.
No significant hazards consideration comments received: Yes. The comments were addressed in the Safety Evaluation.
Page 65820
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California
Date of amendment request: November 12, 2014, as supplemented by letter dated August 27, 2015.
Brief description of amendments: The amendments revised the completion date for Milestone 8, full implementation, of the Cyber Security Plan from December 31, 2015, to December 31, 2017.
Date of issuance: October 1, 2015.
Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.
Amendment Nos.: Unit 2-231; Unit 3-224. A publicly-available version is in ADAMS under Accession No. ML15209A935; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Facility Operating Licenses.
Date of initial notice in Federal Register: April 7, 2015 (80 FR 18659). The supplemental letter dated August 27, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 1, 2015.
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: May 26, 2015, as supplemented by letter dated May 28, 2015, and as revised by letters dated June 9, and June 29, 2015.
Brief description of amendment: The license amendment revised the Combined Licenses (COLs) by revising the VEGP Units 3 and 4 Updated Final Safety Analysis Report on the applicability of the American Institute of Steel Construction (AISC) N690-1994, ``Specification for the Design, Fabrication and Erection of Steel Safety-Related Structures for Nuclear Facilities,'' to allow use of a newer version of the American Welding Society (AWS) D1.1-200, ``Structural Welding Code-
Steel,'' in lieu of the AWS D1.1-1992 edition identified in AISC N690-
1994. The use of AWS D1.1-2000 applies to future and installed structural welding.
Date of issuance: August 31, 2015.
Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.
Amendment No.: 37. A publicly-available version is in ADAMS under Accession No. ML15215A288; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment revised the Facility Combined Licenses.
Date of initial notice in Federal Register: June 9, 2015 (80 FR 32624). A revised notice was issued on July 9, 2015 (80 FR 39454) as the June 29, 2015, letter revised the scope of the amendment request and the licensee revised the original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 31, 2015.
No significant hazards consideration comments received: Yes. The comments were addressed in the Safety Evaluation.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia
Date of amendment request: May 19, 2015.
Brief description of amendments: The amendments revised the minimum indicated nitrogen cover pressure required per the Vogtle Electric Generating Plant Technical Specifications (TS) Surveillance Requirement 3.5.1.3 from the current requirement of 626 pounds per square inch gauge (psig) back to the previous requirement of 617 psig based on installation of updated instrumentation.
Date of issuance: October 5, 2015.
Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.
Amendment Nos.: 177 and 158. A publicly-available version is in ADAMS under Accession No. ML15222A753; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Renewed Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: July 21, 2015 (80 FR 43129).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 5, 2015.
No significant hazards consideration comments received: No.
Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES-1 and 2), Luzerne County, Pennsylvania
Date of amendment request: August 11, 2014, as supplemented by letters dated April 6, 2015, and July 16, 2015.
Brief description of amendments: The amendments changed SSES-1 and 2, Technical Specification (TS) 3.4.10, ``RCS Reactor Coolant System Pressure and Temperature (P/T) Limits,'' specifically revising the P/T Limits curves. The revision provides P/T Limits curves that extend into the vacuum region (e.g., below 0 pounds per square inch gauge) to mitigate the risk of a level transient during startup, account for updated surveillance material and fluence data for the reactor vessel beltline materials, and replace the current 35.7 and 30.2 effective full power year (EFPY) P/T Limits curves for SSES-1 and 2, respectively, with new curves that are valid for 40 EFPY. This license amendment request was submitted by PPL Susquehanna, LLC; however, on June 1, 2015, the NRC staff issued an amendment changing the name on the SSES license from PPL Susquehanna, LLC to Susquehanna Nuclear, LLC (ADAMS Accession No. ML15054A066). These amendments were issued subsequent to an order issued on April 10, 2015, to SSES, approving an indirect license transfer of the SSES license to Talen Energy Corporation (ADAMS Accession No. ML15058A073).
Date of issuance: September 30, 2015.
Effective date: As of the date of issuance and shall be implemented within 30 days.
Amendment Nos.: 263 (Unit 1) and 244 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML15243A140; documents related to these amendments are listed in the safety evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-14 and NPF-22: Amendments revised the Facility Operating License and TSs.
Date of initial notice in Federal Register: November 25, 2014 (79 FR 70217). The supplemental letters dated April 6, 2015, and July 16, 2015, provided additional information that clarified the application, expanded the scope of the application as originally noticed, and changed the staff's original proposed no significant hazards consideration determination as published in the Federal Register. As such, the staff published a subsequent
Page 65821
notice in the Federal Register on July 30, 2015 (80 FR 45559).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 30, 2015.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama
Date of amendment request: March 9, 2015, as supplemented by letter dated August 19, 2015.
Brief description of amendment: The amendment revised Technical Specification (TS) 3.1.4, ``Control Rod Scram Times,'' based on Technical Specification Task Force Change Traveler-460, Revision 0, ``Control Rod Scram Time Testing Frequency,'' revising the frequency of Surveillance Requirement 3.1.4.2 regarding control rod scram time testing from ``120 days cumulative operation in MODE 1'' to ``200 days cumulative operation in MODE 1.'' Implementation of this amendment will also include incorporation of the revised acceptance criterion value of 7.5 percent for ``slow'' control rods into the TS Bases.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment No.: 289. A publicly-available version is in ADAMS under Accession No. ML15251A540; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-33: Amendment revised the Facility Operating License and TSs.
Date of initial notice in Federal Register: June 9, 2015 (80 FR 32629). The supplemental letter dated August 19, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2015.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: November 22, 2013, as supplemented by letters dated December 16, 2014; June 19, 2015; July 24, 2015; August 5, 2015; and August 31, 2015.
Brief description of amendments: The amendments converted the current technical specifications to the improved technical specifications (ITSs) and relocate certain requirements to other licensee-controlled documents. The ITSs are based on NUREG-1431, Rev. 3.0, ``Standard Technical Specifications, Westinghouse Plants,'' Rev. 3.0; ``NRC Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors,'' dated July 22, 1993 (58 FR 39132); and 10 CFR 50.36, ``Technical Specifications.'' Technical Specification Task Force changes were also incorporated. The purpose of the conversion is to provide clearer and more readily understandable requirements in the technical specifications for SQN to ensure safe operation. In addition, the amendments include a number of issues that were considered beyond the scope of NUREG-1431.
Date of issuance: September 30, 2015.
Effective date: As of its date of issuance and shall be implemented within 30 days of issuance.
Amendment Nos.: 334--Unit 1 and 327--Unit 2. A publicly-available version is in ADAMS under Accession No. ML15238B499; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. DPR-77 and DPR-79. The amendments revised the TSs.
Date of initial notice in Federal Register: June 24, 2014 (79 FR 35807). The supplemental letters dated December 16, 2014, June 19, July 24, August 5, and August 31, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 30, 2015.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: August 1, 2013, as supplemented by letters dated April 21, 2014, January 29, 2015, and June 12, 2015.
Brief description of amendment: The amendment revised the Limiting Condition for Operation for the Alternating Current Sources--Operating in Technical Specification 3.8.1 to provide additional time to restore an inoperable offsite circuit, modify Surveillance Requirements, and modify the current licensing basis, as described in the Updated Final Safety Analysis Report for the available maintenance feeder for the Common Station Service Transformers A and B.
Date of issuance: September 29, 2015.
Effective date: As of the date of issuance and shall be implemented after the issuance of the Facility Operating License for Unit 2.
Amendment No.: 103. A publicly-available version is in ADAMS under Accession No. ML15225A094; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-90: Amendment revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 29, 2013 (78 FR 64547). The supplemental letters dated April 21, 2014, January 29 and June 12, 2015, provided additional information that expanded the scope of the application as originally noticed. A notice published in the Federal Register on August 28, 2015, supersedes the original notice in its entirety to update the expanded scope of the amendment description and include the staff's proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 29, 2015.
No significant hazards consideration determination comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri
Date of application for amendment: October 2, 2014, as supplemented by letters dated July 6, July 16, and August 31, 2015.
Brief description of amendment: The amendment adopted the NRC-
endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Methodology for the Development of Emergency Action Levels for Non-
Passive Reactors.''
Date of issuance: October 7, 2015.
Effective date: As of its date of issuance and shall be implemented within 90 days from the date of issuance.
Amendment No.: 212. A publicly-available version is in ADAMS under Accession No. ML15251A493; documents related to this amendment
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are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-30: The amendment revised the Emergency Action Level Technical Bases Document.
Date of initial notice in Federal Register: February 3, 2015 (80 FR 5813). The supplemental letters dated July 6, July 16, and August 31, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 7, 2015.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 19th day of October 2015.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
FR Doc. 2015-27042 Filed 10-26-15; 8:45 am
BILLING CODE 7590-01-P
