Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations
Federal Register, Volume 83 Issue 118 (Tuesday, June 19, 2018)
Federal Register Volume 83, Number 118 (Tuesday, June 19, 2018)
Notices
Pages 28456-28467
From the Federal Register Online via the Government Publishing Office www.gpo.gov
FR Doc No: 2018-12506
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NUCLEAR REGULATORY COMMISSION
NRC-2018-0114
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 May 22, 2018, to June 4, 2018. The last biweekly notice was published on June 5, 2018.
DATES: Comments must be filed by July 19, 2018. A request for a hearing must be filed by August 20, 2018.
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 website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0114. Address questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; 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: TWFN-7-A60M, 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: Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1384, 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-2018-0114, 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-2018-0114.
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-2018-0114, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.
The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.
If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS.
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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 section 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 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 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
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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.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North Carolina
Date of amendment request: January 23, 2018. A publicly-available version is in ADAMS under Accession No. ML18023A896.
Description of amendment request: The proposed amendments would revise Technical Specification 3.6.4.1, ``Secondary Containment,'' Surveillance Requirement (SR) 3.6.4.1.2, for Brunswick Steam Electric Plant, Units 1 and 2. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3, ``Revise Secondary Containment Surveillance Requirements'' (ADAMS Accession No. ML16277A226).
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 addresses conditions during which Secondary Containment SR 3.6.4.1.2 is not met. The Secondary Containment is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed change is no different than the consequences of an accident while utilizing the existing eight hour Completion Time for an inoperable Secondary Containment. As a result, the consequences of an accident previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment 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 protection system design, create new failure
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modes, or change any modes of operation. The proposed change does not involve a physical alteration of the plant; and no new or different kind of equipment will be installed. Consequently, there are no new initiators that could result in a new or different kind of accident.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed change addresses conditions during which Secondary Containment SR 3.6.4.1.2 is not met. The allowance for both an inner and outer Secondary Containment door to be open simultaneously for entry and exit does not affect the safety function of the Secondary Containment as the doors are promptly closed after entry or exit, thereby restoring the Secondary Containment boundary.
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: Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon Street, Mail Code DEC45A, Charlotte, NC 28202.
NRC Acting Branch Chief: Brian W. Tindell.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North Carolina
Date of amendment request: January 23, 2018. A publicly-available version is in ADAMS under Accession No. ML18023A899.
Description of amendment request: The amendments would revise the Technical Specifications to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-208, Revision 0, ``Extension of Time to Reach Mode 2 in LCO Limiting Condition for Operation 3.0.3.''
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 time frame to take response action in accordance with LCO 3.0.3 is not an initiating condition for any accident previously evaluated. The proposed change does not authorize the addition of any new plant equipment or systems, nor does it alter the assumptions of any accident analyses. The small increase in the time allowed to reach Mode 2 would not place the plant in any significantly increased probability of an accident occurring. The unit would already be preparing for a plant shutdown condition because of the 1 hour requirement to initiate shutdown actions. There is no change in the time period to reach Mode 3. The Mode 3 Condition is the point at which the plant reactor core is no longer critical (i.e., Hot Shutdown).
Therefore, since there is no change to the time period to reach the Hot Shutdown condition, the small change in the time to reach Mode 2 status 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 allowed time to reach Mode 2 in LCO 3.0.3 does not require any modification to the plant or change equipment operation. The proposed change will not introduce failure modes that could result in a new accident, and the change does not alter assumptions made in the safety analysis. The proposed change will not alter the design configuration, or method of operation of plant equipment beyond its normal functional capabilities. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from those that have been previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change to the allowed time to reach Mode 2 in LCO 3.0.3 does not alter or exceed a design basis or safety limit. There is no change being made to safety analysis assumptions or the safety limits that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by the proposed change and the applicable requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50, Appendix A will continue to be met.
Therefore, the proposed change does not involve any reduction in a margin of safety.
The NRC staff has reviewed the 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: Kathryn B. Nolan, Deputy General Counsel, 550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
NRC Acting Branch Chief: Brian W. Tindell.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit No. 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: March 8, 2018. A publicly-available version is in ADAMS under Accession No. ML18068A705.
Description of amendment request: The amendment would update Section 15.4.3.1 of the Updated Final Safety Analysis Report for Waterford 3, which describes the dose consequence of the worst undetectable single fuel assembly misload. The updated analysis would reflect the use of Next Generation Fuel and integrated fuel burnable absorbers.
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 the fuel assembly misload event analysis. The analysis of the fuel assembly misload event showed that the total number of failed fuel rods is less than other Waterford 3 Condition 3 events that have already been demonstrated to meet the 10 CFR 50.67 acceptance criteria. For Waterford 3, the Excess Load with Loss of Alternating Current (LOAC) has this same release and fuel failure that has been shown to meet the offsite dose requirements. Since the worst undetectable misload has a fuel failure less than the excess load with LOAC event, the fuel assembly misload event is consistent with the Standard Review Plan 15.4.7 and meets the 10 CFR 50.67 requirements.
This change is only analyzing the consequences of the fuel assembly misload event and no changes are being made that would impact the probability of the event occurring.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the fuel assembly misload event analysis. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing plant operations. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose
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consequences exceed the consequences of accidents previously analyzed.
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.
The proposed change revises the fuel assembly misload event analysis. The worst undetectable misloads have fuel failure less than the excess load with the Excess Load with Loss of Alternating Current (LOAC) event; the fuel assembly misload event meets the 10 CFR 50.67 criteria and is consistent with the Standard Review Plan Section 15.4.7 guidance. The new analysis shows more adverse consequences than were shown in previous fuel assembly misload event analyses, but remains within the regulatory acceptance limits. Since the event remains within the 10 CFR 50.67 requirements and is bounded by the excess load with LOAC event, this is not a significant reduction in margin.
Therefore, this change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: 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, 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, Unit Nos. 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Unit Nos. 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Unit Nos. 1 and 2, Rock Island County, Illinois
Date of amendment request: April 25, 2018. A publicly-available version is in ADAMS under Accession No. ML18116A133.
Description of amendment request: The amendments would revise the technical specification (TS) requirements for inoperable snubbers for each facility. The amendments would also make other administrative changes to the TSs.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration for each site, 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 allows a delay time before declaring supported Technical Specification (TS) systems inoperable when the associated snubber(s) cannot perform its required safety function. Entrance into Actions or delaying entrance into Actions is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on the delay time allowed before declaring a TS supported system inoperable and taking its Conditions and Required Actions are no different than the consequences of an accident under the same plant conditions while relying on the existing TS supported system Conditions and Required Actions. Therefore, the consequences of an accident previously evaluated are not significantly increased by this change. 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 allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. 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.
The proposed change allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change restores an allowance in the pre-Improved Standard Technical Specifications (ISTS) conversion TS that was unintentionally eliminated by the conversion. The pre-ISTS TS were considered to provide an adequate margin of safety for plant operation, as does the post-ISTS conversion TS. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis for each site 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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: David J. Wrona.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: March 7, 2018. A publicly-available version is in ADAMS under Accession No. ML18066A648.
Description of amendment request: The proposed amendment would revise Technical Specification 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' to follow guidance developed by the Nuclear Energy Institute (NEI) in topical report NEI 94-01, ``Industry Guideline for Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' Revision 3-A, dated July 2012, with the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008. The proposed license amendment would also revise Technical Specification 5.5.12 by deleting two of the four listed exceptions to program guidelines.
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 test interval extensions do not involve either a physical change to the plant or a change in the way the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. Therefore, the proposed extensions do not involve a significant increase in the probability of an accident previously evaluated.
The effect resulting from changing the Type A test frequency to 1 per 15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is
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0.0318 person-rem/year. EPRI Electric Power Research Institute Report No. 1009325, Revision 2-A, states that a very small population dose is defined as an increase of less than or equal to 1.0 person-rem per year or less than or equal to 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended integrated leak rate test intervals. The results of the risk assessment calculation for the Type A test extension meet these criteria. The risk impact for the integrated leak rate test extension when compared to other severe accident risks is negligible.
The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and (2) time based. Activity based failure mechanisms are defined as degradation due to system and component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, and Technical Specification requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated.
The proposed amendment also deletes two previously granted exceptions to Primary Containment Leakage Rate Testing Program guidelines. The exception regarding the performance of a Type A test no later than a specified date would be deleted as this Type A test has already been performed. Additionally, the exception to use the corrections to NEI 94-01, Revision 0, would be deleted as those corrections would no longer be in use. These changes to the exceptions in Technical Specification 5.5.12 are administrative in nature and do not affect the probability or consequences of an accident previously evaluated.
Therefore, the proposed changes do not result in a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Containment Type A and Type C testing requirements periodically demonstrate the integrity of the containment and exist to ensure the plant's ability to mitigate the consequences of an accident. These tests do not involve any accident precursors or initiators.
The proposed change does not involve a physical modification to the plant (that is, no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.
The proposed amendment also deletes two previously granted exceptions. The exception regarding the performance of a Type A test no later than a specified date would be deleted as this Type A test has already been performed. Additionally, the exception to use the corrections to NEI 94-01, Revision 0, would be deleted as those corrections would no longer be in use. These changes to the exceptions in Technical Specification 5.5.12 are administrative in nature and do not create the possibility of a new or different kind of accident from any previously evaluated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed license amendment does not alter the way safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the Technical Specification 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 Technical Specifications 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 amendment, since they are not affected by implementation of a performance-based containment testing program. This ensures that the margin of safety in the plant safety analysis is maintained.
The proposed amendment also deletes two previously granted exceptions. The exception regarding the performance of a Type A test no later than a specified date would be deleted as this Type A test has already been performed. Additionally, the exception to use the corrections to NEI 94-01, Revision 0, would be deleted as those corrections would no longer be in use. These changes to the exceptions in Technical Specification 5.5.12 are administrative in nature and do not involve a significant reduction in a margin of safety.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: David J. Wrona.
NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant (PBNP), Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin
Date of amendment request: March 30, 2018. A publicly-available version is in ADAMS under Accession No. ML18092A239.
Description of amendment request: The amendments would revise Technical Specification (TS) 5.5.15, ``Containment Leakage Rate Testing Program,'' to require a program in accordance with 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.'' This proposed change will allow extension of the Type A test interval up to one test in 15 years and extension of the Type C test interval up to 75 months, based on acceptable performance history as defined in NEI 94-01, Revision 3-A.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed change involve a significant increase in the probability 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 PBNP 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 PBNP 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.
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Therefore, the containment will continue to perform its design function as a barrier to fission product releases.
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 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 change 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 this change. Therefore, the proposed change will 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.
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 and local leak rate testing 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 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 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, Managing Attorney--Nuclear, Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, Juno Beach, FL 33408-0420.
NRC Branch Chief: David J. Wrona.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station (HCGS), Salem County, New Jersey
Date of amendment request: April 13, 2018. A publicly-available version is in ADAMS under Accession No. ML18103A218.
Description of amendment request: The amendment would revise Technical Specification (TS) 3.8.3.1, ``Distribution--Operating,'' to increase the alternating current (AC) inverters allowed outage time (AOT) from 24 hours to 7 days. The proposed change is based on application of the HCGS probabilistic risk assessment (PRA) in support of a risk-informed extension, and on additional considerations and compensatory actions.
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 TS amendment does not affect the design of the AC inverters, the operational characteristics or function of the inverters, the interfaces between the inverters and other plant systems, or the reliability of the inverters. An inoperable AC inverter is not considered an initiator of an analyzed event. In addition, TS Actions and the associated Allowed Outage Times are not initiators of previously evaluated accidents. Extending the Allowed Outage Time for an inoperable AC inverter would not have a significant impact on the frequency of occurrence of an accident previously evaluated. The proposed amendment will not result in modifications to plant activities associated with inverter maintenance, but rather, provides operational flexibility by allowing additional time to perform inverter troubleshooting, corrective maintenance, and post-maintenance testing on-line.
The proposed extension of the Completion Time for an inoperable AC inverter will not significantly affect the capability of the inverters to perform their safety function, which is to ensure an uninterruptible supply of 120-volt AC electrical power to the associated power distribution subsystems. An evaluation, using PRA methods, confirmed that the increase in plant risk associated with implementation of the proposed Allowed Outage Time extension is consistent with the NRC's Safety Goal Policy Statement, as further described in RG Regulatory Guide 1.174 and RG 1.177. In addition, a deterministic evaluation concluded that plant defense-in-depth philosophy will be maintained with the proposed Allowed Outage Time extension.
There will be no impact on the source term or pathways assumed in accidents previously evaluated. No analysis assumptions will be changed and there will be no adverse effects on onsite or offsite doses as the result of an accident.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment does not involve physical alteration of the HCGS. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters with in which the HCGS is operated. There are no setpoints at which protective or mitigating actions are initiated that are affected by this proposed action. The use of the alternate Class 1E power source for the AC distribution panel is consistent with the HCGS plant design. The change does not alter assumptions made in the safety analysis. This proposed action will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alteration is proposed to the procedures that ensure the HCGS remains with in analyzed limits, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change, which would increase the AOT from 24 hours to 7 days for one inoperable inverter, does not exceed or alter a setpoint, design basis or safety limit.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
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Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: James G. Danna.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County, Georgia
Date of amendment request: April 26, 2018. A publicly-available version is in ADAMS under Accession No. ML18116A138.
Description of amendment request: The requested amendment proposes changes to combined license (COL) Appendix C, with corresponding changes to the associated plant-specific Tier 1 information, and involves associated Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document (DCD) Tier 2 information). Pursuant to the provisions of 10 CFR 52.63(b)(1), also requested is an exemption from elements of the design as certified in the 10 CFR part 52, appendix D, design certification rule for the plant-specific DCD departures.
The requested amendment proposes changes to COL Appendix C (and plant-specific Tier 1) to reflect a new design of containment sump level sensors that affects the acceptance criterion for the detected containment sump level change test and the associated minimum detectable unidentified leakage rate in plant-specific DCD Tier 2 information.
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 is to the containment sump water level instrumentation and its expected reactor coolant system (RCS) leakage detection capability. The affected equipment is not safety-
related, but the containment sump water level sensors are seismically qualified. The change in containment sump level monitoring instruments has no adverse effect on the ability to detect a 0.5 gallons per minute (gpm) leak in containment, and therefore, has no adverse effect on design criteria for leak-before-
break. The change does not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) accident initiator or initiating sequence of events.
Because the containment sump water level monitoring channels are still capable of detecting a 0.5 gpm leak in containment, the change to the SSC has no effect on plant operations. There is no change to plant systems or the response of systems to postulated accident conditions. There is no change to the predicted radioactive releases due to normal operation or postulated accident conditions.
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 change does not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The proposed change to the containment sump water level instrumentation and its expected RCS leakage detection capability has no adverse effect on the ability to detect a 0.5 gpm leak in containment. The containment sump level instrumentation functions are unchanged and leak-before-break design criteria are not adversely affected.
Loss of coolant accidents for a spectrum of pipe sizes and locations are already postulated in UFSAR Chapter 15, Section 15.6. Breaks in the main steam lines inside containment are also analyzed in UFSAR Chapter 15, Section 15.1. Unidentified leakage detection and operator action in response to unidentified leakage are not postulated for any of the design basis accident analyses described in UFSAR Chapter 15.
Therefore, the requested amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The described change to the containment sump water level instrumentation and its expected RCS leakage detection capability is proposed to verify that the ability to detect a 0.5 gpm leak in containment is maintained. The proposed change does not alter any safety-related equipment, applicable design codes, code compliance, design function, or safety analysis. By ensuring that the chosen equipment can detect a 0.5 gpm leak in containment with the described accuracy, guidance in Regulatory Guide 1.45, Revision 0, as committed to in the UFSAR, and requirements in the Technical Specifications are met which ensures that leak-before-break design criteria are not adversely affected. Consequently, no safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed change, thus the margin of safety is not reduced.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and based on this review it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazard consideration
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue, North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer L. Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County, Georgia
Date of amendment request: April 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18117A464.
Description of amendment request: The requested amendment proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document Tier 2 information) and involves related changes to plant-
specific Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C information. Specifically, the amendment, if approved, would revise the Tier 2 information in the UFSAR and related changes to Tier 1 and the associated COL Appendix C to remove the fire protection system non-safety related containment cable spray and install passive fire stops and radiant energy shields. The changes to Tier 1 require an exemption, which is included in the license amendment request.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not affect the operation or reliability of any system, structure or component (SSC) required to maintain a normal power operating condition or to mitigate anticipated transients without safety-related systems. Testing has demonstrated that the passive fire stops prevent propagation of fires along the length of cable trays and prevent the propagation of cable tray fires to adjacent fire zones. The proposed changes do not affect the operation of equipment whose failure could initiate an accident previously analyzed. The existence or failure of passive fire stops in fire zone 1100 AF 11300B does not affect normal equipment operation.
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The proposed changes do not adversely affect the reliability or function of an SSC relied upon to mitigate an accident previously analyzed. The existence or failure of passive fire stops in fire zone 1100 AF 11300B will not adversely affect passive core cooling system (PXS) performance during containment recirculation because the passive fire stops are located outside of the zone of influence (ZOI) of postulated high energy line breaks, and the passive fire stops' material-of-construction complies with in-containment refueling water storage tank (IRWST) and containment recirculation screens design criteria for debris generation and transport.
The existing active open nozzle cable tray suppression system is not fully automatic, is nonsafety-related, and is not credited in the probabilistic risk assessment (PRA). Therefore, replacing the active open nozzle cable tray suppression system with passive fire stops does not have an impact on PRA calculations and results.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not affect the operation of systems or equipment that could initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created. The use of passive fire stops is recognized by Regulatory Guide 1.189. The passive fire stops in nonsafety-related open cable trays are more reliable than active systems such as the current open nozzle cable tray suppression system because they require no mechanical or human action to perform their protective function. When protection is required, there is no delay for operator or mechanical response. Testing has demonstrated that the passive fire stops prevent propagation of fires along the length of cable trays and prevent the propagation of cable tray fires to adjacent fire zones.
The existence or failure of passive fire stops in fire zone 1100 AF 11300B will not adversely affect passive core cooling system (PXS) performance during containment recirculation because the passive fire stops are located outside of the zone of influence (ZOI) of postulated high energy line breaks, and their material-of-
construction complies with in-containment refueling water storage tank (IRWST) and containment recirculation screens design criteria for debris generation and transport.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed changes do not affect existing safety margins. The current open nozzle cable tray suppression system is nonsafety-
related. The use of passive fire stops is recognized by Regulatory Guide 1.189. The passive fire stops in nonsafety-related open cable trays are more reliable than active systems such as the current open nozzle cable tray suppression system because they require no mechanical or human action to perform their protective function. When protection is required, there is no delay for operator or mechanical response. Testing has demonstrated that the passive fire stops prevent propagation of fires along the length of cable trays and prevent the propagation of cable tray fires to adjacent fire zones.
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 hazard consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue, North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer L. Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, Vogtle Electric Generating Plant, Unit Nos. 3 and 4, Burke County, Georgia
Date of amendment request: April 27, 2018. A publicly-available version is in ADAMS under Accession No. ML18117A464.
Description of amendment request: The requested amendment proposes to depart from Tier 2 information in the Updated Final Safety Analysis Report (UFSAR) (which includes the plant-specific Design Control Document Tier 2 information) and involves related changes to plant-
specific Tier 1 information, with corresponding changes to the associated combined license (COL) Appendix C information. Specifically, the amendment, if approved, would revise the Tier 2 information in the UFSAR and related changes to Tier 1 and the associated COL Appendix C to remove the fire protection system non-safety related containment cable spray and install passive fire stops and radiant energy shields. The changes to Tier 1 require an exemption, which is included in the license amendment request.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not affect the operation of any systems or equipment that initiate an analyzed accident or alter any structures, systems, and components (SSC) accident initiator or initiating sequence of events.
The proposed changes do not affect the physical design and operation of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) or In-containment Refueling Water Storage Tank (IRWST) as described in the Updated Final Safety Analysis Report (UFSAR). The proposed changes do not affect the probability of inadvertent operation or failure. Therefore, the probabilities of the accidents previously evaluated in the UFSAR are not affected.
The proposed changes do not affect the ability of the PRHR HX and IRWST to perform their design functions. The designs of the PRHR HX and IRWST continue to meet the same regulatory acceptance criteria, codes, and standards as required by the UFSAR. In addition, the proposed changes maintain the capabilities of the PRHR HX and IRWST to mitigate the consequences of an accident and to meet the applicable regulatory acceptance criteria.
The proposed changes do not affect the prevention and mitigation of other abnormal events (e.g. anticipated operational occurrences, earthquakes, floods and turbine missiles), or their safety or design analyses. Therefore, the consequences of the accidents evaluated in the UFSAR are not affected.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not affect the operation of any systems or equipment that may initiate a new or different kind of accident, or alter any SSC such that a new accident initiator or initiating sequence of events is created.
The proposed changes do not affect any other SSC design functions or methods of operation in a manner that results in a new failure mode, malfunction, or sequence of events that affect safety-
related or nonsafety related equipment. Therefore, this activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that result in significant fuel cladding failures.
Therefore, the requested amendment does not create the possibility of a new or different type of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed changes maintain existing safety margins. The proposed changes verify and maintain the capabilities of the PRHR HX and IRWST to perform their design functions. Therefore, the proposed changes
Page 28465
satisfy the same design functions in accordance with the same codes and standards as stated in the UFSAR. These changes do not affect any design code, function, design analysis, safety analysis input or result, or design/safety margin.
No safety analysis or design basis acceptance limit/criterion is challenged or exceeded by the proposed changes, and no margin of safety is reduced.
Therefore, the requested amendment does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer L. Dixon-Herrity.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2 (Surry), Surry County, Virginia
Date of amendment request: March 2, 2018. A publicly-available version is in ADAMS under Accession No. ML18075A021.
Description of amendment request: The amendments would revise the Technical Specifications (TSs) consistent with Revision 0 to the Technical Specifications Task Force (TSTF) Standard Technical Specification Change Document TSTF-490, ``Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.'' The proposed amendments would adopt TSTF-490 and make the following associated changes: (1) Adoption of a TS change to replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity, and (2) an update of the Alternative Source Term (AST) analyses for Surry.
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:
Criterion 1. The Proposed Changes Do Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated
Reactor coolant specific activity is not an initiator for any accident previously evaluated, and the allowed time period when primary coolant gross activity is not within limits is not an initiator for any accident previously evaluated. In addition, the current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. Updating the Alternative Source Term analyses does not require any changes to any plant structures, systems, or components (SSCs) and therefore does not affect any accident initiators. As a result, the proposed changes do not significantly increase the probability of an accident. The proposed TS change will limit primary coolant noble gases to concentrations consistent with the accident analyses, and the proposed completion time when the limit may be exceeded has no impact on the consequences of any design basis accident since the consequences of an accident during this time period is the same as the consequences of an accident during the existing time periods. The revised assessments of the radiological consequences due to design basis accidents listed in the Surry Updated Final Safety Analysis Report, using the updated AST methodology and proposed assumptions and inputs, conclude that the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room doses are within the limits of 10 CFR 50.67 and within the limits of Regulatory Guide (RG) 1.183. As a result, the consequences of any accident previously evaluated are not significantly increased.
Criterion 2. The Proposed Changes Do Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated
The proposed TS change in specific activity limits and the updated AST dose consequences analyses do not alter any physical part of the plant, (i.e., no new or different type of equipment will be installed,) nor do they affect any plant operating parameter or create new accident precursors. Therefore, the proposed changes do not create the potential for a new or different kind of accident from any previously calculated.
Criterion 3. The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety.
The proposed TS change in specific activity limits is consistent with the assumptions in -the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses. The proposed changes for radiological events related to the computer code used to calculate dose, revised X/Qs for control room and offsite receptors (including the computer code and method used to determine control room X/Qs for SG releases), the computer code used to determine core inventory, the change in FHA Fuel Handling Accident gap fraction methodology, and removing the LRA Locked Rotor Accident from the radiological design basis have been analyzed and result in acceptable consequences, meeting the criteria as specified in 10 CFR 50.67 and RG 1.183. The proposed changes will not result in plant operation in a configuration outside the analyses or design basis and do not adversely affect systems that are required to respond for safe shutdown of the plant and to maintain the plant in a safe operating condition. Therefore, the 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: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Michael T. Markley.
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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.
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DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of amendment request: August 14, 2017.
Brief description of amendment: The amendment modified Fermi 2 Technical Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' by adopting the format and language of NUREG-1433, ``Standard Technical Specifications for General Electric BWR/4 Plants,'' Revision 4.
Date of issuance: May 24, 2018.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment No: 208. A publicly-available version is in ADAMS under Accession No. ML18108A022; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-43: The amendment revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 26, 2017 (82 FR 44851).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 24, 2018.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Brunswick County, North Carolina
Date of amendment request: June 29, 2017, as supplemented by letters dated January 4, 2018, and January 23, 2018.
Brief description of amendments: The amendments adopted Technical Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control,'' for Brunswick Steam Electric Plant, Units 1 and 2. The amendments replaced existing technical specification (TS) requirements associated with ``operations with the potential for draining the reactor vessel,'' with revised TSs providing alternative requirements for reactor pressure vessel water inventory control. These alternative requirements protect Safety Limit 2.1.1.3, which states, ``Reactor vessel water level shall be greater than the top of active irradiated fuel.''
Date of issuance: April 13, 2018.
Effective date: As of the date of issuance and shall be implemented prior to the 2019 Unit 2 refueling outage. This Notice of Issuance corrects the effective date of License Amendment No. 283, originally noticed in the Federal Register on May 8, 2018 (83 FR 20865).
Amendment Nos.: 283 (Unit 1) and 311 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML18039A444; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendments. Amendment Nos. 283 and 311 were corrected by letter dated May 23, 2018 (ADAMS Accession No. ML18137A143).
Renewed Facility Operating License No. DPR-49: The amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: September 12, 2017 (82 FR 42846). The supplemental letters dated January 4, 2018, and January 23, 2018, 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 April 13, 2018.
No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: April 3, 2017, as supplemented by letters dated April 3, 2017; May 2, 2017; September 28, 2017; and January 8, 2018.
Brief description of amendment: The amendment revised the Technical Specifications (TSs) to extend the required frequency of certain 18-
month Surveillance Requirements to 24 months to accommodate a 24-month refueling cycle. In addition, the amendment revised certain programs in TS Section 5.5, ``Programs and Manuals,'' to change 18-month frequencies to 24 months.
Date of issuance: May 25, 2018.
Effective date: As of the date of issuance and shall be implemented within 120 days from the end of the next refueling outage.
Amendment No.: 258. A publicly-available version is in ADAMS under Accession No. ML18115A150; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-23: The amendment revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: July 5, 2017 (82 FR 31092). The supplemental letters dated September 28, 2017, and January 8, 2018, 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 amendment is contained in a Safety Evaluation dated May 25, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1 (Clinton), DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station (LaSalle), Unit Nos. 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station (Limerick), Unit Nos. 1 and 2, Montgomery County, Pennsylvania
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit No. 2 (Nine Mile), Oswego County, New York
Date of amendment request: November 8, 2017.
Brief description of amendments: The amendments revised the technical specification requirements for secondary containment.
Date of issuance: May 29, 2018.
Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.
Amendment Nos.: Clinton--218; LaSalle, Units 1 and 2--228 and 214; Limerick, Units 1 and 2--229 and 192; and Nine Mile--169. A publicly-
available version is in ADAMS under Accession No. ML18113A045. Documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-62, NPF-11, NPF-18, NPF-39, NPF-85, and NPF-69: The amendments revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: December 19, 2017 (82 FR 60227).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 29, 2018.
No significant hazards consideration comments received: No.
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Southern Nuclear Operating Company, Inc.; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; and City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: April 20, 2017, as supplemented by letters dated September 14, 2017; February 19, 2018; and May 1, 2018.
Brief description of amendments: The amendments revised the Technical Specifications by replacing the existing 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, which requires reactor vessel water level to be greater than the top of active irradiated fuel.
Date of issuance: May 31, 2018.
Effective date: As of the date of issuance and shall be implemented prior to the commencement of the Unit No. 2 refueling outage (U2R25) in February 2019.
Amendment Nos.: Unit 1--290, Unit 2--235. A publicly-available version is in ADAMS under Accession No. ML18123A368; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-57 and NPF-5: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: August 29, 2017 (82 FR 41071). The supplemental letters dated September 14, 2017; February 19, 2018; and May 1, 2018, 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 May 31, 2018.
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, Unit Nos. 1, 2, and 3, Limestone County, Alabama
TVA Docket Nos. 50-327, 50-328, and 72-034, Sequoyah Nuclear Plant, Unit Nos. 1 and 2, Hamilton County, Tennessee
TVA Docket Nos. 50-390, 50-391, and 72-1048, Watts Bar Nuclear Plant, Unit Nos. 1 and 2, Rhea County, Tennessee
Date of amendment request: January 4, 2017, as supplemented by letters dated July 7, 2017, and July 27, 2017. (Note: This Notice of Issuance corrects the amendments by adding the supplement dated July 27, 2017, which was inadvertently omitted from the original Federal Register notice (January 16, 2018; 83 FR 2234).
Brief description of amendments: The amendments revised TVA Emergency Plans for the above nuclear plants. Specifically, the amendments 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.: Browns Ferry Nuclear Plant--303 (Unit 1), 327 (Unit 2), and 287 (Unit 3); Sequoyah Nuclear Plant--339 (Unit 1) and 332 (Unit 2); and Watts Bar Nuclear Plant--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 enclosed with the amendments. These amendments were corrected by letter dated May 29, 2018 (ADAMS Accession No. ML18138A452).
Renewed Facility Operating License Nos. DPR-33, DPR-52, DPR-68, DPR-77, and DPR-79, and Facility Operating License Nos, NPF-90 and NPF-
96: The amendments revised the licenses.
Date of initial notice in Federal Register: June 19, 2017 (82 FR 27891). The supplemental letters dated July 7, 2017, and July 27, 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 amendment is contained in a Safety Evaluation dated December 22, 2017.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit No. 1, Callaway County, Missouri
Date of amendment request: April 6, 2017, as supplemented by letter dated February 5, 2018.
Brief description of amendment: The amendment revised the Final Safety Analysis Report to clearly describe conformance with NRC Regulatory Guide 1.106, Revision 1, ``Thermal Overload Protection for Electric Motors on Motor-Operated Valves.''
Date of issuance: May 30, 2018.
Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.
Amendment No.: 218. A publicly-available version is in ADAMS under Accession No. ML18124A026; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-30: The amendment revised the Final Safety Analysis Report.
Date of initial notice in Federal Register: July 18, 2017 (82 FR 32885). The supplemental letter dated February 5, 2018, 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 amendment is contained in a Safety Evaluation dated May 30, 2018.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 6th day of June 2018.
For the Nuclear Regulatory Commission.
Tara Inverso,
Acting Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
FR Doc. 2018-12506 Filed 6-18-18; 8:45 am
BILLING CODE 7590-01-P
