Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
Federal Register, Volume 83 Issue 10 (Tuesday, January 16, 2018)
Federal Register Volume 83, Number 10 (Tuesday, January 16, 2018)
Notices
Pages 2224-2234
From the Federal Register Online via the Government Publishing Office www.gpo.gov
FR Doc No: 2018-00386
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NUCLEAR REGULATORY COMMISSION
NRC-2018-0005
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 December 19, 2017 to December 29, 2017. The last biweekly notice was published on January 2, 2018.
DATES: Comments must be filed by February 15, 2018. A request for a hearing must be filed by March 19, 2018.
ADDRESSES: You may submit comments by any of the following methods.
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0005. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: email protected. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
Mail comments to: May Ma, Office of Administration, Mail Stop: OWFN-2-A13, 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: Kay Goldstein, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-1506, email: email protected.
SUPPLEMENTARY INFORMATION:
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Obtaining Information and Submitting Comments
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Obtaining Information
Please refer to Docket ID NRC-2018-0005, facility name, unit number(s), plant docket number, application date, and subject when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0005.
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 email protected. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.
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Submitting Comments
Please include Docket ID NRC-2018-0005, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.
The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.
If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that 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
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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 if circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.
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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 persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of the regulations is available at the NRC's Public Document Room, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. If a petition is filed, the Commission or a presiding officer will rule on the petition and, if appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements for standing: (1) The name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set forth the specific contentions which the petitioner seeks to have litigated in the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner must provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to the specific sources and documents on which the petitioner intends to rely to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant or licensee on a material issue of law or fact. Contentions must be limited to matters within the scope of the proceeding. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene. Parties have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that party's admitted contentions, including the opportunity to present evidence, consistent with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of publication of this notice. Petitions and motions for leave to file new or amended contentions that are filed after the deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i) through (iii). The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to establish when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of the amendment unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2.309(h)(1). The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission no later than 60 days from the date of publication of this notice. The petition must be filed in accordance with the filing instructions in the ``Electronic Submissions (E-Filing)'' section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or federally recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. Alternatively, a State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the proceeding and is not affiliated with or represented by a party may, at the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the
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provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of his or her position on the issues but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled.
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Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Detailed guidance on making electronic submissions may be found in the Guidance for Electronic Submissions to the NRC and on the NRC's website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at email protected, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit adjudicatory documents. Submissions must be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the document is submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the document on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before adjudicatory documents are filed so that they can obtain access to the documents via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic Filing Help Desk through the ``Contact Us'' link located on the NRC's public website at http://www.nrc.gov/site-help/e-submittals.html, by email to email protected, or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the ``Obtaining Information and Submitting Comments'' section of this document.
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Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2 (CNS), York County, South Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3 (ONC), Oconee County, South Carolina
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Wake County, North Carolina
Duke Energy Progress, LLC, Docket No. 50-261, H.B. Robinson Steam Electric Plant, Unit No. 2 (RNP), Darlington County, South Carolina
Date of amendment request: November 7, 2017. A publicly-available version is in ADAMS under Accession No. ML17312A362.
Description of amendment request: The amendments would revise the technical specifications (TSs) based on Technical Specification Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing IST Program Removal & Clarify SR Surveillance Requirement Usage Rule Application to Section 5.5 Testing'' (ADAMS Accession No. ML15294A555), with some variations. For each plant, the changes include deleting the current TS for the IST Program, adding a new defined term, ``INSERVICE TESTING PROGAM,'' to the TSs, and revising other TSs to reference this new defined term instead of the deleted TS.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 5 (TS Chapter 6 for HNP), ``Administrative Controls,'' Section 5.5 (Section 6.8.4 for HNP), ``Programs and Manuals,'' by replacing the current contents of the ``Inservice Testing Program'' specification with a note referring to the TS Definition of ``INSERVICE TESTING PROGRAM.'' Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code American Society of Mechanical Engineers Code for Operations and Maintenance of Nuclear Power Plants, as clarified by Code Case OMN-20, ``Inservice Test Frequency.'' The remaining requirements in the Section 5.5 (Section 6.8.4 for HNP) IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, ``INSERVICE TESTING PROGRAM,'' is added to the TS, which references the requirements of 10 CFR 50.55a(f).
Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than or equal to 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.
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 does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged. However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than or equal to 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS SR 3.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the Technical Specifications provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or 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 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: Kathryn B. Nolan, Deputy General Counsel, Duke Energy Corporation, 550 South Tryon Street, Mail Code DEC45A, Charlotte NC 28202.
NRC Branch Chief: Undine Shoop.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington
Date of amendment request: October 23, 2017, as supplemented by letter dated November 15, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17296B380, and ML17320A314, respectively.
Description of amendment request: The proposed amendment would adopt Technical Specification Task Force (TSTF) traveler TSTF-542, Revision 2, ADAMS Accession No. ML16343B008 ``Reactor Pressure Vessel Water Inventory Control.'' The proposed amendment would replace existing technical specification (TS) requirements related to operations with a potential for draining the reactor vessel (OPDRVs) with new requirements on Reactor Pressure Vessel (RPV) Water Inventory Control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires the reactor vessel water level to be greater than the top of active irradiated fuel.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the
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licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces existing TS requirements related to OPDRVs with new requirements RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.
The proposed change reduces the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.
The proposed change reduces the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.
The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?
Response: No.
The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed change will not alter the design function of the equipment involved. Under the proposed change, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.
The event of concern under the current requirements and the proposed change is an unexpected draining event. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street NW, Washington, DC 20006-3817.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: November 28, 2017, as supplemented by letter dated December 7, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17332A898, and ML17341B295, respectively.
Description of amendment request: The proposed amendment would revise Section 4.3.3 of the Waterford 3 Updated Final Safety Analysis Report to indicate that the RAPTOR-M3G code is used for reactor vessel fluence calculations. The use of the RAPTOR-M3G code would meet the criteria present in Regulatory Guide (RG) 1.190, ``Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,'' dated March 2001.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The probability of occurrence of an accident previously evaluated for Waterford 3 is not altered by the proposed license amendment. The accidents currently analyzed in the Waterford 3 Final Safety Analysis Report (FSAR) remain the same. The proposed change does not impact the integrity of the reactor coolant pressure boundary (RCPB) (i.e., there is no change to the operating pressure, materials, loadings, etc.). The proposed change does not affect the probability nor consequences of any design basis accident (DBA). The proposed neutron fluence calculational methodology meets the criteria in RG 1.190 and will be used to ensure that the P/T pressure-temperature limit curves, maximum heatup and cooldown rates, and LTOP low-temperature overpressure protection enable temperature remain acceptable to maintain reactor pressure vessel integrity.
Fracture toughness test data are obtained from material specimens contained in capsules that are periodically withdrawn from the reactor vessel. These data, combined
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with the neutron fluence calculations, permit determination of the conditions under which the vessel can be operated with adequate safety margins against brittle fracture throughout its service life. For each analyzed transient and steady state condition, the allowable pressure is determined as a function of reactor coolant temperature considering postulated flaws in the reactor vessel beltline, inlet nozzle, outlet nozzle, and closure head.
The predicted radiation induced DeltaRTNDT delta reference temperature nil ductility transition is calculated using the respective reactor vessel beltline materials' copper and nickel contents and the neutron fluence determination. The RTNDT and, in turn, the operating limits for Waterford 3 are adjusted, if necessary, to account for the effects of irradiation on the fracture toughness of the reactor vessel materials and maintain reactor vessel integrity within design assumptions.
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 neutron fluence calculational method will not create a new accident scenario. The requirements to have P/
T limits and LTOP protection are part of the licensing basis for Waterford 3. The neutron fluence calculation method will validate, and when necessary, provide input to the development of new operating limits. The data analysis for the vessel surveillance specimens are used to confirm that the vessel materials are responding as predicted based on previous neutron fluence projections.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change to the neutron fluence calculational method conforms to the criteria presented in RG 1.190 and will ensure that Waterford 3 continues to operate within the operating margins allowed by 10 CFR 50.60 and the ASME American Society of Mechanical Engineers Code.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC (Exelon), Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of amendment request: November 16, 2017. A publicly-available version is available in ADAMS under Accession No. ML17320A411.
Description of amendment request: The amendment would revise the OCNGS renewed facility operating license (RFOL) and the associated Technical Specifications (TSs) to Permanently Defueled Technical Specifications (PDTS) consistent with the permanent cessation of reactor operation and permanent defueling of the reactor. By letter dated January 7, 2011 (ADAMS Accession No. ML110070507), Exelon provided formal notification to the NRC of Exelon's contingent determination to permanently cease operations at OCNGS no later than December 31, 2019. The amendment would eliminate those TSs applicable in operating modes or modes where fuel is placed in the reactor vessel. The amendment would change other TS limiting conditions for operation (LCOs), definitions, surveillance requirements (SRs), administrative controls, as well as several license conditions.
Basis for proposed no significant hazards consideration (NSHC) determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of NSHC, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes would not take effect until OCNGS has permanently ceased operation, entered a permanently defueled condition, and at least 60 days of irradiated fuel decay time after reactor shutdown. The proposed changes would revise the OCNGS RFOL and TS by deleting or modifying certain portions of the TS that are no longer applicable to a permanently shutdown and defueled facility. This change is consistent with the criteria set forth in 10 CFR 50.36 for the contents of TS.
Chapter 15 of the OCNGS Updated Final Safety Analysis Report (UFSAR) described the design basis accident (DBA) and transient scenarios applicable to OCNGS during power operations. The analyzed accidents that remains applicable to OCNGS in the permanently shut down and defueled condition is a Fuel Handling Accident (FHA) in the spent fuel pool (SFP) (a dropped fuel assembly onto the top of the core will no longer be applicable) and the Postulated Radioactive Tank Failure and Release of Radioactive Liquid Waste while radioactive liquids are still present. The FHA is the remaining accident with radiological consequences and has been revised for the permanently shutdown and defueled condition. The liquid tank accidents analysis remains bounding and unchanged; therefore, is not discussed further in this NSHC evaluation.
Once the reactor is in a permanently defueled condition, the spent fuel pool (SFP) and its cooling systems will be dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents will be much smaller than for an operational plant. Once the certifications are docketed by OCNGS pursuant to 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactor or to place or retain fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2), the majority of the accident scenarios previously postulated in the UFSAR will no longer be possible and will be removed from the UFSAR under the provisions of 10 CFR 50.59.
The deletion of TS definitions and rules of usage and application, that will not be applicable in a defueled condition, has no impact on facility structures, systems, and components (SSCs) or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shutdown and defueled status of OCNGS has no impact on the remaining applicable DBA. The removal of LCOs or SRs that are related to only the operation of the nuclear reactor or to only the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated since these DBAs are no longer applicable in the defueled mode. The safety functions involving core reactivity control, reactor heat removal, reactor coolant system inventory control, and containment integrity are no longer applicable at OCNGS as a permanently defueled plant. The analyzed accidents involving damage to the reactor coolant system, main steam lines, reactor core, and the subsequent release of radioactive material will no longer be possible at OCNGS.
After OCNGS permanently ceases operation, the future generation of fission products will cease and the remaining source term will decay. The radioactive decay of the irradiated fuel following shutdown of the reactor will have reduced the consequences of the FHA in the SFP below those previously analyzed. The relevant parameter (water level) associated with the fuel pool provides an initial condition for the FHA analysis and is included in the PDTS.
The SFP water level and spent fuel storage TSs are retained to preserve the current requirements for safe storage of irradiated fuel. SFP cooling and makeup related equipment and support equipment (e.g., electrical power systems) are not required to be continuously available since there will be sufficient time to effect repairs, establish alternate sources of makeup flow, or establish alternate sources of cooling in the event of a loss of cooling and makeup flow to the SFP.
The deletion and modification of provisions of the administrative controls do not directly affect the design of SSCs necessary for safe storage of irradiated fuel or the methods used for handling and storage of
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such fuel in the fuel pool. The changes to the administrative controls are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shutdown and defueled condition of the reactor.
The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation will no longer be credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to delete and/or modify certain TS have no impact on facility SSCs affecting the safe storage of spent irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of spent irradiated fuel itself. The removal of TS that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor related transients or accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shutdown and defueled and OCNGS will no longer be authorized to operate the reactor.
The proposed deletion of requirements of the OCNGS RFOL and TS do not affect systems credited in the accident analysis for the FHA in the SFP at OCNGS. The proposed RFOL and PDTS will continue to require proper control and monitoring of safety significant parameters and activities.
The TS regarding SFP water level and spent fuel storage is retained to preserve the current requirements for safe storage of irradiated fuel. The restriction on the SFP water level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA.
The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding and spent fuel cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed changes involve deleting and/or modifying certain TS once the OCNGS facility has been permanently shutdown, defueled, and at least 60 days of irradiated fuel decay time after reactor shutdown. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for OCNGS will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel following submittal of the certifications required by 10 CFR 50.82(a)(1). As a result, the occurrence of certain design basis postulated accidents associated with reactor operation is no longer considered credible. The only remaining credible accidents are a FHA and the Postulated Radioactive Releases Due to Liquid Radwaste Tank Failures. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact either accident.
The proposed changes are limited to those portions of the RFOL and TS that are not related to the safe storage of irradiated fuel. The requirements that are proposed to be revised or deleted from the OCNGS RFOL and TS are not credited in the existing accident analysis for the remaining applicable postulated accidents; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated design basis accidents involving the reactor will no longer be possible because the reactor will be permanently shutdown and defueled and OCNGS will no longer be authorized to operate the reactor.
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 NSHC.
Attorney for licensee: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Douglas A. Broaddus.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa
Date of amendment request: June 9, 2017, as supplemented by letter dated November 1, 2017. Publicly-available versions are in ADAMS under Accession Nos. ML17164A076 and ML17305A910, respectively.
Description of amendment request: The proposed amendment would replace the existing technical specification (TS) requirements related to ``operations with a potential for draining the reactor vessel'' (OPDRVs) with requirements for reactor pressure vessel (RPV) water inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires RPV water level to be greater than the top of active irradiated fuel. The proposed amendment is based on Technical Specification Task Force (TSTF) traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control,'' which was approved by the NRC by letter dated December 20, 2016.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an accident previously evaluated and, therefore, replacing the existing TS controls to prevent or mitigate such an event with a new set of controls has no effect on any accident previously evaluated. RPV water inventory control in Mode 4 or Mode 5 is not an initiator of any accident previously evaluated. The existing OPDRV controls or the proposed RPV WIC controls are not mitigating actions assumed in any accident previously evaluated.
The proposed change reduces the probability of an unexpected draining event (which is not a previously evaluated accident) by imposing new requirements on the limiting time in which an unexpected draining event could result in the reactor vessel water level dropping to the top of the active fuel (TAF). These controls require cognizance of the plant configuration and control of configurations with unacceptably short drain times. These requirements reduce the probability of an unexpected draining event. The current TS requirements are only mitigating actions and impose no requirements that reduce the probability of an unexpected draining event.
The proposed change reduces the consequences of an unexpected draining event (which is not a previously evaluated accident) by requiring an Emergency Core Cooling System (ECCS) subsystem to be operable at all times in Modes 4 and 5. The current TS requirements do not require any water injection systems, ECCS or otherwise, to be Operable in certain conditions in Mode 5. The change in requirement from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does not significantly affect the consequences of an unexpected draining event because the proposed Actions ensure equipment is available within the limiting drain time that is as capable of mitigating the event as the current requirements. The proposed controls provide escalating compensatory measures to be established as calculated drain times decrease, such as verification of a second method of water injection and additional confirmations that containment and/or filtration would be available if needed.
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The proposed change reduces or eliminates some requirements that were determined to be unnecessary to manage the consequences of an unexpected draining event, such as automatic initiation of an ECCS subsystem and control room ventilation. These changes do not affect the consequences of any accident previously evaluated since a draining event in Modes 4 and 5 is not a previously evaluated accident and the requirements are not needed to adequately respond to a draining event.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?
Response: No.
The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed change will not alter the design function of the equipment involved. Under the proposed change, some systems that are currently required to be operable during OPDRVs would be required to be available within the limiting drain time or to be in service depending on the limiting drain time. Should those systems be unable to be placed into service, the consequences are no different than if those systems were unable to perform their function under the current TS requirements.
The event of concern under the current requirements and the proposed change is an unexpected draining event. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators that would cause a draining event or a new or different kind of accident not previously evaluated or included in the design and licensing bases.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed change replaces existing TS requirements related to OPDRVs with new requirements on RPV WIC. The current requirements do not have a stated safety basis and no margin of safety is established in the licensing basis. The safety basis for the new requirements is to protect Safety Limit 2.1.1.3. New requirements are added to determine the limiting time in which the RPV water inventory could drain to the top of the fuel in the reactor vessel should an unexpected draining event occur. Plant configurations that could result in lowering the RPV water level to the TAF within one hour are now prohibited. New escalating compensatory measures based on the limiting drain time replace the current controls. The proposed TS establish a safety margin by providing defense-in-depth to ensure that the Safety Limit is protected and to protect the public health and safety. While some less restrictive requirements are proposed for plant configurations with long calculated drain times, the overall effect of the change is to improve plant safety and to add safety margin.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, P. O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: David J. Wrona.
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50-
296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone County, Alabama
TVA, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
TVA, Docket Nos. 50-390 and 50-391, Watts Bar Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee
Date of amendment request: November 17, 2017. A publicly-available version is in ADAMS under Accession Nos. ML17324A349.
Description of amendment request: The amendments would add a new level of protection, ``Unbalanced Voltage,'' to the Technical Specifications for the loss of power instrumentation. The NRC issued Bulletin 2012-01, ``Design Vulnerability in Electric Power System,'' which requested addressees to submit specific information regarding plant design and operating configurations relative to the regulatory requirements of General Design Criterion (GDC) 17, ``Electric power systems.'' The Nuclear Energy Institute notified the NRC that the nuclear industry's chief nuclear officers approved a formal initiative to address the open phase condition (OPC). It further stated that the initiative represented a formal commitment among nuclear power plant licensees to address the OPC design vulnerability for operating reactors.
The licensee stated, in its November 17, 2017, submittal, that the primary reason for the proposed change is to provide equipment protection from the effects of an unbalanced voltage in a similar fashion to the existing degraded and loss of voltage protection schemes. The identification of the vulnerability was based on industry operating experience and subsequent commitment to meet the voluntary Nuclear Strategic Issues Advisory Committee Open Phase Industry Initiative, also known as the ``Voluntary Industry Initiative'' (VII) for GDC 17 Compliance.
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 consequence of an accident previously evaluated?
Response: No.
The proposed change to add a new unbalanced voltage relay (UVR) function at BFN, SQN, and WBN provides another level of undervoltage protection for the Class 1E electrical equipment. The new relay setpoints ensure that the normally operating Class 1E motors and equipment, which are powered from the Class 1E buses, are appropriately isolated from the normal offsite power source and would not be damaged in the event of sustained unbalanced voltage. The addition of the UVR function continues to allow the existing undervoltage protection circuitry to function as originally designed (i.e., degraded and loss of voltage protection remain in place and are unaffected by this change). The addition of the new UVR function has no impact on accident initiators or precursors; does not alter the accident analysis assumptions or the manner in which the plant is operated or maintained; and does not affect the probability of operator error.
Based on the above, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to add a new UVR function at BFN, SQN, and WBN provides another level of undervoltage protection for the Class 1E electrical equipment. This change ensures that the assumption in the previously evaluated accidents, which may involve a degraded voltage condition, continue to be valid. The proposed change does not result in the creation of any new accident precursors; does not result in changes to any existing accident scenarios; and does not introduce any operational changes or mechanisms that would create the possibility of a new or different kind of accident. The UVR function would not affect the existing loss of voltage and degraded voltage protection schemes, would not affect the number of occurrences of degraded voltage conditions that would cause the actuation of the existing Loss of Voltage Relays, Degraded Voltage Relays or the new UVRs; would not affect the failure rate of the existing protection relays; and would not impact the assumptions in any existing accident scenario.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The current undervoltage protection circuitry is designed to isolate the normally operating Class 1E motors/equipment, which are powered from the Class 1E buses, from the offsite power source such that the subject equipment would not be damaged in the event of sustained degraded bus voltage. After the Class 1E buses are isolated from the offsite power supply, the Class 1E motors would be sequenced back on the Class 1E bus powered by the diesel generators (DGs) and continue to perform their design basis function to mitigate the consequences of an accident, with a specified margin of safety. With the addition of the new level of undervoltage protection, the capability of the Class 1E equipment is assured. Thus the equipment would continue to perform its design basis function to mitigate the consequences of the previously analyzed accidents and maintain the existing margin to safety currently assumed in the accident analyses. A DG start due to a safety injection signal (i.e., loss of coolant accident) and the subsequent sequencing of Class 1E loads back onto the Class 1E buses, powered by the DG, are not adversely affected by this change. If an actual loss of voltage condition were to occur on the Class 1E buses, the loss of voltage time delays would continue to isolate the Class 1E distribution system from the offsite power source prior to the DG assuming the Class 1E loads. The Class 1E loads would sequence back on the bus in a specified order and timer interval, again ensuring that the existing accident analysis assumptions remain valid and the existing margin to safety is unaffected.
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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine Shoop.
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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 Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the ``Obtaining Information and Submitting Comments'' section of this document.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3 (Indian Point 2 and 3), Westchester County, New York
Date of amendment request: December 14, 2016, as supplemented by letters dated April 19, 2017; August 16, 2017; and October 2, 2017.
Brief description of amendments: The amendments revised the Appendix A Technical Specifications Limiting Condition for Operation (LCO) 3.7.13, ``Spent Fuel Pit Storage,'' for Indian Point 2 and Appendix C Technical Specifications LCO 3.1.2, ``Shielded Transfer Canister (STC) Loading,'' for Indian Point 2 and 3. These LCOs ensure that the fuel to be loaded into the STC meets the design basis for the STC and has an acceptable rack location in the Indian Point 2 spent fuel pool before the STC is loaded with fuel. The proposed changes increase the population of Indian Point 3 fuel eligible for transfer via the STC to the Indian Point 2 spent fuel pool.
Date of issuance: December 22, 2017.
Effective date: As of the date of issuance, and shall be implemented within 60 days.
Amendment Nos.: 287 (Unit No. 2) and 264 (Unit No. 3). A publicly-
available version is in ADAMS under Accession No. ML17320A354; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. DPR-26 and DPR-64: The amendments revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: June 19, 2017 (82 FR 27885). The supplemental letters dated August 16, 2017, and October 2, 2017, 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 the amendments is contained in a Safety Evaluation dated December 22, 2017.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: May 1, 2017, as supplemented by letter dated June 13, 2017.
Brief description of amendment: The amendment revised the completion date for Milestone 8, full implementation of the Cyber Security Plan, from December 15, 2017, to July 31, 2019.
Date of issuance: December 15, 2017.
Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.
Amendment No.: 266. A publicly-available version is in ADAMS under Accession No. ML17339A097; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License No. DPR-28: The amendment revised the Facility Operating License.
Date of initial notice in Federal Register: August 15, 2017 (82 FR 38717). The supplemental letter dated June 13, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 15, 2017.
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No significant hazards consideration comments received: No.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania
Date of amendment request: January 30, 2017, as supplemented by letters dated August 11, 2017, September 8, 2017, and December 20, 2017.
Brief description of amendments: The amendments replaced existing Technical Specification requirements related to ``operations with a potential for draining the reactor vessel'' with new requirements on reactor pressure vessel water inventory control to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires reactor pressure vessel water level to be greater than the top of active irradiated fuel. The changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-
542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control.''
Date of issuance: December 27, 2017.
Effective date: As of the date of issuance and shall be implemented prior to the Unit 2 fall 2018 refueling outage (P2R22).
Amendments Nos.: 317 (Unit 2) and 320 (Unit 3). A publicly-
available version is in ADAMS under Accession No. ML17325B708; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 28, 2017 (82 FR 15382). The supplemental letters dated August 11, 2017, September 8, 2017, and December 20, 2017, 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 the amendments is contained in a Safety Evaluation dated December 27, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: April 5, 2017.
Brief description of amendment: The amendment revised the Nine Mile Point Nuclear Station, Unit 2, Technical Specifications to allow greater flexibility in performing surveillance testing in Modes 1, 2, or 3 of emergency diesel generators. The changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-283A, Revision 3, ``Modify Section 3.8 Mode Restrictions Notes.''
Date of issuance: December 21, 2017.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment No.: 165. A publicly-available version of the amendment is in ADAMS under Accession No. ML17324B178; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-69: Amendment revised the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 19, 2017 (82 FR 27887).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 21, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of amendment request: April 10, 2017, as supplemented by letters dated October 4 and December 15, 2017.
Brief description of amendment: The amendment revised the OCNGS renewed facility operating license for the Cyber Security Plan (CSP) Milestone 8 full implementation completion date, as set forth in the CSP implementation schedule, and revised the physical protection license condition. The amendment revised the CSP Milestone 8 completion date from December 31, 2017, to August 31, 2021.
Date of issuance: December 22, 2017.
Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.
Amendment No.: 292. A publicly-available version is in ADAMS under Accession No. ML17289A222; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-16: The amendment revised the renewed facility operating license.
Date of initial notice in Federal Register: May 23, 2017 (82 FR 23626). The supplemental letters dated October 4 and December 15, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's 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 December 22, 2017.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment request: March 24, 2017.
Brief description of amendments: The amendments revised the Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Technical Specification (TS) 3.7.2, ``Steam Generator Stop Valves (SGSVs),'' to incorporate the SGSV actuator trains into the Limiting Condition for Operation statement and to provide associated Conditions, Required Actions, and Completion Times to the ACTIONS table. In addition, Surveillance Requirement (SR) 3.7.2.2 was revised to clearly identify that the SGSV actuator trains are required to be tested in accordance with the SR.
Date of issuance: December 19, 2017.
Effective date: As of the date of issuance and shall be implemented within 90 days of issuance.
Amendment Nos.: Unit No. 1-338; Unit No. 2-320. A publicly-
available version is in ADAMS under Accession No. ML17312B030; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: May 23, 2017 (82 FR 23626).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 19, 2017.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston County, Alabama
Date of amendment request: November 22, 2016, as supplemented by letters dated May 23, 2017; June 8, 2017;
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September 7, 2017; November 21, 2017; and December 18, 2017.
Brief description of amendments: The amendments revised the licensing basis of FNP to support a full scope application of an Alternative Source Term methodology and modified Technical Specifications (TSs) 3.7.10, 3.9.3, and TS 5.5.18, consistent with Technical Specifications Task Force (TSTF) Travelers TSTF-448-A, ``Control Room Habitability,'' Revision 3, and TSTF-312, ``Administratively Control Containment Penetrations.''
Date of issuance: December 20, 2017.
Effective date: As of the date of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 216 (Unit 1) and 213 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17271A265; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: The amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: January 3, 2017 (82 FR 160). The supplemental letters dated May 23, 2017; June 8, 2017; September 7, 2017; November 21, 2017; and December 18, 2017, 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 the amendments is contained in a Safety Evaluation dated December 20, 2017.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3 (BFN), Limestone County, Alabama
Date of amendment request: June 7, 2017. As supplemented by letters dated September 18 and October 23, 2017.
Brief description of amendment: The amendments revised fire protection license condition 2.C.(13) for Unit 1, license condition 2.C.(14) for Unit 2, and license condition 2.C.(7) for Unit 3.
Date of issuance: December 19, 2017.
Effective date: As of the date of issuance and shall be implemented as indicated in Items 2 and 3 under ``Transition License Conditions'' of the Operating Licenses, as shown in the attachment to the license amendments.
Amendment Nos.: 302 (Unit 1), 326 (Unit 2), and 286 (Unit 3). A publicly-available version is in ADAMS under Accession No. ML17317A422; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: Amendments revised the licenses.
Date of initial notice in Federal Register: September 5, 2017 (82 FR 41997). The supplemental letters dated September 18 and October 23, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluations of the amendments are contained in Safety Evaluations dated December 19, 2017.
No significant hazards consideration comments received: No.
Tennessee Valley Authority (TVA) Docket Nos. 50-259, 50-260, 50-296, and 72-052, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone County, Alabama
TVA Docket Nos. 50-327, 50-328, and 72-034, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
TVA Docket Nos. 50-390, 50-391, and 72-1048, Watts Bar Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee
Date of amendment request: January 4, 2017, as supplemented by letter dated July 7, 2017.
Brief description of amendments: The amendments revised TVA Emergency Plans for the above nuclear plants. Specifically, they adopted the NRC-endorsed Radiological Emergency Plan Emergency Action Level schemes developed by the Nuclear Energy Institute (NEI 99-01, Revision 6, ``Development of Emergency Action Levels for Non-Passive Reactors'').
Date of issuance: December 22, 2017.
Effective date: As of the date of issuance and shall be implemented within 180 days from the date of its issuance or July 3, 2018, whichever comes later.
Amendment Nos.: BFN, 303 (Unit 1), 327 (Unit 2), and 287 (Unit 3); SQN, 339 (Unit 1) and 332 (Unit 2); and WBN, 118 (Unit 1) and 18 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML17289A032; documents related to these amendments are listed in the Safety Evaluations (SEs) enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-33, DPR-52, DPR-68, DPR-77, DPR-79 and Facility Operating License Nos, NPF-90 and NPF-96: Amendments revised the licenses.
Date of initial notice in Federal Register: June 19, 2017 (82 FR 27891). The supplemental letter dated July 7, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in SEs dated December 22, 2017.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, on January 8, 2018.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Acting Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
FR Doc. 2018-00386 Filed 1-12-18; 8:45 am
BILLING CODE 7590-01-P
