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Notice

Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

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Start Preamble Start Printed Page 25832

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

Pursuant to 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 7, 2019, to May 20, 2019. The last biweekly notice was published on May 21, 2019.

DATES:

Comments must be filed by July 5, 2019. A request for a hearing must be filed by August 5, 2019.

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-2019-0129. Address questions about NRC docket IDs in Regulations.gov to Jennifer Borges; telephone: 301-287-9127; email: Jennifer.Borges@nrc.gov. For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • Mail Comments To: Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Program Management, Announcements and Editing Staff.

For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

Start Further Info

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: Janet.Burkhardt@nrc.gov.

End Further Info End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2019-0129, 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-2019-0129.
  • 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 “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in this document.
  • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2019-0129, facility name, unit number(s), plant docket number, application date, and subject in your comment submission.

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

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS.

II. Background

Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the 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.

III. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses and Proposed No Significant Hazards Consideration Determination

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in § 50.92 of title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

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

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period if circumstances change during the 30-day comment Start Printed Page 25833period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. If the Commission takes action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. If the Commission makes a final no significant hazards consideration determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

Within 60 days after the date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and petition for leave to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309. The NRC's regulations are accessible electronically from the NRC Library on the NRC's 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.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic Start Printed Page 25834storage 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 hearing.docket@nrc.gov, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public 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 MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 6 p.m., Eastern Time, Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing adjudicatory documents in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at https://adams.nrc.gov/​ehd, unless excluded pursuant to an order of the Commission or the presiding officer. If you do not have an NRC-issued digital ID certificate as described above, click cancel when the link requests certificates and you will be automatically directed to the NRC's electronic hearing dockets where you will be able to access any publicly available documents in a particular hearing docket. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or personal phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. For example, in some instances, individuals provide home addresses in order to demonstrate proximity to a facility or site. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

For further details with respect to these license amendment applications, see the application for amendment which is available for public inspection in ADAMS and at the NRC's PDR. For additional direction on accessing information related to this document, see the “Obtaining Information and Submitting Comments” section of this document.

Dominion Energy Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, Millstone Power Station, Unit Nos. 2 and 3 (Millstone or MPS), New London County, Connecticut

Date of amendment request: April 11, 2019. A publicly-available version is in ADAMS under Accession No. ML19109A100.

Description of amendment request: The amendments would adopt Technical Specifications Task Force (TSTF) Traveler TSTF-522, “Revise Ventilation System Surveillance Requirements to Operate for 10 Hours per Month,” and decrease ventilation system flow test requirements from 10 hours at the frequency specified in the Millstone, Unit Nos. 2 and 3, Surveillance Frequency Control Program (SFCP) to 15 continuous minutes at the frequency specified in the SFCP. Additionally, Millstone, Unit No. 2, Technical Specification (TS) Surveillance Requirement (SR) 4.6.5.1.a would be revised to remove the requirement to run the flow test with the duct heaters energized since the charcoal adsorption test is performed at 95 percent relative humidity.

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

1. Does the proposed amendment involve a significant increase in the probability or Start Printed Page 25835consequences of an accident previously evaluated?

Response: No.

The proposed change modifies existing SRs to operate the EBFS [Enclosure Building Filtration System] system for MPS2 and ABFS [Auxiliary Building Filter System], CREVS [Control Room Emergency Ventilation System], and SLCRS [Supplementary Leak Collection and Release System] systems for MPS3 that are equipped with electric heaters for a 10 hour period at the frequency specified in the SFCP with a requirement to operate the systems for 15 continuous minutes. Additionally, the SR for EBFS will be revised to remove the requirement [to] conduct the flow test with the duct heaters energized since the charcoal adsorption test is performed at 95% relative humidity.

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

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

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

Response: No.

The proposed change modifies existing SRs to operate the EBFS, ABFS, CREVS, and SLCRS systems equipped with electric heaters for a 10 hour period at the frequency specified in the SFCP with a requirement to operate the systems for 15 continuous minutes. Additionally, the SR for EBFS will be revised to remove the requirement [to] conduct the flow test with the duct heaters energized since the charcoal adsorption test is performed at 95% relative humidity.

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

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

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

Response: No.

The proposed change modifies existing SRs to operate the EBFS, ABFS, CREVS, and SLCRS systems equipped with electric heaters for a 10 hour period at the frequency specified in the SFCP with a requirement to operate the systems for 15 continuous minutes. Additionally, TSTF-522 identifies a regulatory position which indicates that plants which test ventilation system absorption at a relative humidity of 95% do not require heaters for the ventilation system to perform its specified safety function systems and that reference to the heaters can be removed from the TS. Based on justification provided in TSTF-522, the existing SR for EBFS will be revised to remove the requirement to complete the ventilation system test with the duct heaters energized since the adsorption test is performed at 95% relative humidity. EBFS will continue to have the heaters, but they will not be credited in the TS.

The design basis for the ventilation systems' heaters is to heat the incoming air which reduces the relative humidity. Per TSTF-522, the monthly 10 hour system operation utilizing the heaters was intended to remove moisture from the charcoal adsorber banks. Because the ASTM D3803-1989 Standard no longer requires this 10 hour operation utilizing the heaters, the duration is replaced with a continuous 15 minute operation requirement. The proposed change is consistent with guidance provided in Regulatory Position 4.9 of Regulatory Guide 1.52, Revision 3.

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

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

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.

NRC Branch Chief: James G. Danna.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: February 27, 2019. A publicly-available version is in ADAMS under Accession No. ML19058A251.

Description of amendment request: The amendment would place a Note prior to the surveillance requirements (SRs) section of Technical Specification (TS) 3.3.5.3 that allows delayed entry into the associated conditions and required actions, when a channel is placed in an inoperable status solely for testing, provided the associated Function maintains emergency core cooling system (ECCS) initiation capability.

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 change to TS 3.3.5.3 adds a note that previously applied when the Surveillance Requirements for Modes 4 and 5 were included in TS 3.3.5.1. There are no new requirements or actions added that have not been previously approved. Applying the note cannot increase probability of an accident because it does not change plant equipment or SR method or surveillance frequency.

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 duplicates existing TS Surveillance Requirements that will continue to protect Safety Limit 2.1.1.3. The note requires ECCS initiation function to be maintained in order to allow the delayed entry into the Condition. The proposed change will not alter the design function of the equipment involved. The event of concern 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 changes have no adverse effect on plant operation. The plant response to the design basis accidents do not change. The proposed changes do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analyses. There is no change being made to safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. The analysis in NEDC-30936-P-A demonstrates that the testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary. The note can only be used when initiation capability is maintained.

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: Jon P. Christinidis, DTE Energy, Expert Start Printed Page 25836Attorney—Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.

NRC Branch Chief: David J. Wrona.

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

Date of amendment request: February 25, 2019. A publicly-available version is in ADAMS under Accession No. ML19057A549.

Description of amendment request: The proposed amendment would allow use of the control room chilled water (CCH) system or the emergency service water (SW) system as acceptable cooling sources in support of the main control room (MCR) air conditioning (AC) system.

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 CCH system is not an initiator of an accident and does not have the function of preventing any accidents. Therefore, the proposed change does not involve an increase in the probability of an event.

The CCH system utilizes active components to perform its design function in support of MCR cooling, however, the CCH system utilizes safety-related equipment which meet the design requirements stated in the Columbia FSAR [Final Safety Analysis Report]. System performance and reliability will be monitored by the Maintenance Rule, the IST [Inservice Testing] Program and TS [technical specification] surveillance. Procedures are available for CCH system use and the CCH system components are accessible post-accident. Analyses have been performed and conclude there is adequate time to initiate MCR cooling following a design basis event. The proposed change does not impact radiological consequences of any accident described in the FSAR. Therefore, the proposed change does not involve a significant increase in the consequences of an event.

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

Response: No.

The proposed change allows the use of either CCH or SW, when capable of the required heat removal, as cooling support to the [Main] Control Room AC system for the purpose of meeting both the equipment qualification temperature limit and the bounding control room habitability steady state temperature. The proposed change will align CCH to both the Division 1 and Division 2 emergency cooling coils for emergency standby service. If normal MCR cooling is lost, emergency MCR cooling will be manually initiated post-accident and is supported by analyses that conclude the manual actions are feasible and adequate time is available to perform the actions. The [Main] Control Room AC system cooling function is not an accident initiator and is not postulated to create a new or different kind of accident than previously analyzed.

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

Response: No.

The proposed LAR [license amendment request] provides additional flexibility to utilize either the CCH or SW system to meet the MCR required equipment qualification temperature limit and the long term steady state temperature for 30 days continuous control room occupancy. The SW system will be evaluated to ensure it is capable of the required heat removal prior to crediting it as the available cooling source. Operator training will be provided to reflect use of CCH as the preferred cooling source to support the Control Room AC system in both Division 1 and Division 2 following approval of this LAR. Analyses have been performed and conclude that there is adequate time to initiate MCR cooling following a design basis event. Surveillances will be performed on both the CCH and SW systems in support of MCR cooling and the systems will be maintained as safety-related. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Robert J. Pascarelli.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 1, Pope County, Arkansas

Date of amendment request: March 25, 2019. A publicly-available version is in ADAMS under Accession No. ML19084A217.

Description of amendment request: The amendment would modify the Arkansas Nuclear One, Unit 1 Technical Specifications (TSs) to remove second completion times consistent with NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-439, Revision 2, “Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO [Limiting Condition for Operation” (ADAMS Accession No. ML051860296).

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

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

Response: No.

The proposed change eliminates certain Completion Times from the TSs. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident with respect to the proposed change are no different than the consequences of the same accident when applying the existing Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. The proposed change does not alter or prevent the ability of [a] structure, system, or component (SSC) from performing the credited function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, reactor building isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed change is consistent with the safety analysis assumptions and resultant consequences.

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

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

Response: No.

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

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

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

Response: No.

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

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

The NRC staff has reviewed the licensee's analysis and, based on this Start Printed Page 25837review, 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 Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: April 9, 2019. A publicly-available version is in ADAMS under Accession No. ML19099A367.

Description of amendment request: The amendments would modify Technical Specification requirement 6.8.4.g, “Primary Containment Leakage Rate Testing Program,” to allow for a permanent extension of Types A and C integrated leakage rate test frequencies from 10 years to 1 year. In addition, the proposed request seeks approval for drywell-to-suppression chamber bypass leak rate test frequency from 120 months (10 years) to 180 months (15 years) to align this test with the proposed Type A test frequency (Surveillance Requirement 4.6.2.1.e).

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 activity involves the revision of the Limerick Generating Station (LGS), Units 1 and 2 Technical Specification (TS) 6.8.4.g, “Primary Containment Leakage Rate Testing Program,” to allow the extension of the Type A integrated leakage rate test (ILRT) containment test interval to 15 years and the extension of the Type C local leakage rate test (LLRT) interval to 75 months. The proposed activity also involves the extension of the drywell-to-suppression chamber bypass leak test (DWBT) from 120 months to 180 months to align the test with the proposed Type A test frequency. Per the guidance provided in Nuclear Energy Institute (NEI) 94-01, “Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J,” Revision 3-A, the current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions.

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

The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen years, measured as an increase to the total integrated dose risk for all internal events accident sequences for LGS, is 6.60E-02 person-roentgen equivalent man(rem)/yr (0.36 percent) using the Electric Power Research Institute (EPRI) guidance with the base case corrosion included. The change in dose risk drops to 1.16E-02 person-rem/yr (0.06 percent) when using the EPRI Expert Elicitation methodology. The values calculated per the EPRI guidance are all lower than the acceptance criteria of ≤1.0 person-rem/yr or <1.0% person-rem/yr. The change in dose risk for changing the DWBT frequency from once-per-ten years to once-per-fifteen years, measured as an increase to the total integrated dose risk for all internal events accident sequences for LGS, is 1.5E-02 person-rem/yr. The results of the risk assessment for this amendment meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

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

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) Activity based, and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Containment Maintenance Rule Structures Monitoring Program, Containment Coatings Program and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test (ILRT). Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes Units 1 and 2 TS 6.8.4.g exceptions previously granted via TS Amendments No. 190 (Unit 1) and No. 151 (Unit 2) to allow one-time extensions of the ILRT test frequency for LGS. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.

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 amendment to the LGS, Units 1 and 2 TS 6.8.4.g involves the extension of the LGS, Units 1 and 2 Type A (ILRT) containment test interval to 15 years and the extension of the Type C (LLRT) test interval to 75 months. The proposed activity also involves the extension of the DWBT from 120 months to 180 months to align the test with the proposed Type A test frequency. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.

The proposed amendment also deletes Units 1 and 2 TS 6.8.4.g(a) exceptions previously granted to allow one-time extensions of the ILRT test frequency for LGS. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.

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 amendment to Units 1 and 2 TS 6.8.4.g involves the extension of the LGS Type A containment test interval to 15 years and the extension of the Type C test Start Printed Page 25838interval to 75 months for selected components. The proposed activity also involves the extension of the DWBT from 120 months to 180 months to align the test with the proposed Type A test frequency. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type A containment leak rate tests and Type C tests for LGS. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Types B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

The current frequency associated with a DWBT leakage test is 120 months. If any DWBT test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the NRC. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed. The proposed change will modify this leakage test frequency from 120 months to 180 months. The proposed change is acceptable as the results from previous tests show that the measured drywell-to-suppression chamber bypass leakage at the current TS frequency has been a small percentage of the allowable leakage. Acceptability is further demonstrated by the design requirements applied to the primary containment components and other periodically performed primary containment inspections.

LGS, Units 1 and 2 TS SR 4.6.2.1.e DWBT monitors the combined leakage of three types of pathways: (1) The drywell floor and downcomers, (2) piping externally connected to both the drywell and suppression chamber air space and (3) the suppression chamber to drywell vacuum breakers. This amendment would extend the surveillance interval on the passive components of the test (the first two types of pathways), while retaining the current surveillance interval on the active components (suppression chamber to drywell vacuum breakers).

The proposed amendment also deletes Units 1 and 2 TS 6.8.4.g(a) exceptions previously granted to allow one-time extensions of the ILRT test frequency for LGS. These exceptions were for activities that would have already taken place by the time this amendment is approved; therefore, the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.

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: Tamra Domeyer, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: James G. Danna.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: February 28, 2019. A publicly-available version is in ADAMS under Accession No. ML19071A111.

Description of amendment request: The proposed amendment would revise the Cooper Nuclear Station Technical Specifications (TSs) to define a new time limit for restoring inoperable reactor coolant system (RCS) leakage detection instrumentation to operable status; establish alternate methods of monitoring RCS leakage when one or more required monitors are inoperable; and make TS Bases changes that reflect the proposed changes and more accurately reflect the contents of the facility design basis related to operability of the RCS leakage detection instrumentation. The proposed changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-514, Revision 3, “Revise BWR [Boiling Water Reactor] Operability Requirements and Actions for RCS Leakage Instrumentation.” The availability of this TS improvement was announced in the Federal Register on December 17, 2010 (75 FR 79048), as part of the consolidated line item improvement process.

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 clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. The monitoring of RCS leakage is not a precursor to any accident previously evaluated. The monitoring of RCS leakage is not used to mitigate the consequences of any accident previously evaluated. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. 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, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. Reducing the amount of time the plant is allowed to operate with only the drywell atmospheric gaseous radiation monitor operable increases the margin of safety by increasing the likelihood that an increase in RCS leakage will be detected before it potentially results in gross failure. Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power Start Printed Page 25839District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Branch Chief: Robert J. Pascarelli.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 18, 2019. A publicly-available version is in ADAMS under Accession No. ML19108A143.

Description of amendment request: The amendment would revise Hope Creek Generating Station Technical Specification (TS) 3.6.5.1, “Secondary Containment Integrity,” Surveillance Requirements (SRs) 4.6.5.1.a and 4.6.5.1.b.2.a. SR 4.6.5.1.a would be revised to address conditions during which the secondary containment pressure may not meet the SR pressure requirements. SR 4.6.5.1.b.2.a would be modified to acknowledge that secondary containment access openings may be open for entry and exit. Additionally, TS Definitions 1.39.d and 1.39.g would be revised to conform to the proposed changes to these two SRs.

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 addresses conditions during which the secondary containment SRs are not met. The secondary containment is not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not increased. The consequences of an accident previously evaluated while utilizing the proposed changes are no different than the consequences of an accident while utilizing the existing four hour Completion Time for an inoperable secondary containment. In addition, the proposed Note for SR 4.6.5.1.a provides an alternative means to ensure the secondary containment safety function is met. 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 change create the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

The proposed change does not alter the protection system design, create new failure 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 change involve a significant reduction in a margin of safety?

Response: No.

The proposed change addresses conditions during which the secondary containment SRs are not met. Conditions in which the secondary containment is not at a negative pressure are acceptable provided the conditions do not affect the ability of the FRVS [filtration recirculation and ventilation system] to establish the required secondary containment vacuum under post-accident conditions within the time assumed in the accident analysis. This condition is incorporated in the proposed change by requiring an analysis of actual environmental and secondary containment pressure conditions to confirm the capability of the FRVS is maintained within the assumptions of the accident analysis. Therefore, the safety function of the secondary containment is not affected. 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: Steven Fleischer, PSEG Services Corporation, 80 Park Plaza, T-5, Newark, NJ 07102.

NRC Branch Chief: James G. Danna.

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

Date of amendment request: March 29, 2019. A publicly-available version is in ADAMS under Accession No. ML19088A126.

Description of amendment request: The amendment proposes a change in Tier 1 (and associated Combined License Appendix C) Figure 2.2.4-1 (Sheet 3) to relocate the auxiliary steam header isolation valve from the same header as the turbine bypass valves to a new header.

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 does 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. There is no change to the auxiliary steam header isolation valve safety class or nonsafety-related functions. With the proposed change, the auxiliary steam header isolation valve will continue to perform its nonsafety-related design function of providing isolation at the system interface between the main steam system and auxiliary steam supply system. The operation of the auxiliary steam header isolation valve is not changed, and it remains downstream of the main steam isolation valve (MSIV). The auxiliary steam header isolation valve is not, nor was it, credited in limiting blowdown of a second steam generator in the event of a steam line break upstream of an MSIV concurrent with the failure of the other MSIV. Therefore, there is no impact to the MSS [main steam system] design function of limiting blowdown of a second steam generator in the event of a steam line break upstream of an MSIV concurrent with the failure of the other MSIV, and there is no impact to Chapter 15 evaluations.

Therefore, the proposed amendment does not involve a significant 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 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. There is no change to the auxiliary steam header isolation valve safety class or nonsafety-related functions. With the proposed change, the auxiliary steam header isolation valve will continue to perform its nonsafety-related design function of providing isolation at the system interface between the main steam system and auxiliary steam supply system. The operation of the auxiliary steam header isolation valve is not changed, and it remains downstream of the main steam isolation valve (MSIV). The auxiliary steam header isolation valve is not, nor was it, credited in limiting blowdown of a second steam generator in the event of a steam line break upstream of an MSIV concurrent with the failure of the other MSIV. Therefore, there is no impact to the MSS design function of limiting blowdown of a second steam generator in the event of a steam line break upstream of an MSIV concurrent with the failure of the other MSIV, and there is no impact to Chapter 15 evaluations.Start Printed Page 25840

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 change does not affect existing safety margins. There is no change to the auxiliary steam header isolation valve safety class or nonsafety-related functions. With the proposed change, the auxiliary steam header isolation valve will continue to perform its nonsafety-related design function of providing isolation at the system interface between the main steam system and auxiliary steam supply system. The operation of the auxiliary steam header isolation valve is not changed, and it remains downstream of the main steam isolation valve (MSIV). The auxiliary steam header isolation valve is not, nor was it, credited in limiting blowdown of a second steam generator in the event of a steam line break upstream of an MSIV concurrent with the failure of the other MSIV. Therefore, there is no impact to the MSS design function of limiting blowdown of a second steam generator in the event of a steam line break upstream of an MSIV concurrent with the failure of the other MSIV, and there is no impact to Chapter 15 evaluations.

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

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

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

NRC Branch Chief: Jennifer L. Dixon-Herrity.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendment request: July 27, 2018, as supplemented by letters dated May 3, 2019, and May 17, 2019. A publicly-available version is in ADAMS under Accession Nos. ML18208A619, ML19123A253, and ML19137A343, respectively.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) requirements to permit use of Risk-Informed Completion Times in accordance with Nuclear Energy Institute (NEI) topical report NEI 06-09, Revision 0-A, “Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines.” Notice of this action was previously published in the Federal Register on September 25, 2018 (83 FR 48466). The re-noticing of this action is provided to include two supplements dated May 3, 2019, and May 17, 2019, to the licensee's original application dated July 27, 2018. This re-notice supersedes the Federal Register notice of September 25, 2018, in its entirety. The supplements added a new Condition B in Technical Specification 3.7.8, “Service Water System (SWS)”.

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 [change] involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change permits the extension of completion times provided risk is assessed and managed within the Risk Informed Completion Time Program. The proposed change does not involve a significant increase in the probability of an accident previously evaluated because the changes involve no change to the plant or its mode of operation. The proposed change does not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended completion time are no different from those during the existing COMPLETION TIME.

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 TS revision does not change the design, configuration, or method of plant operation. The proposed change does not involve a physical alteration of the plant in that no new or different kind of equipment will be installed.

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

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

Response: No.

The proposed change permits the extension of completion times provided risk is assessed and managed within the Risk Informed Completion Time Program. The proposed change implements a risk-informed configuration management program to assure that adequate safety margins are maintained. Application of these new specifications and the configuration management program considers cumulative effects of multiple systems or components being out of service and does so more effectively than the current 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 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: Jennifer M. Buettner, Associate General Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center Parkway, Birmingham, AL 35242.

NRC Branch Chief: Michael T. Markley.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: April 24, 2019. A publicly-available version is in ADAMS under Accession No. ML19114A535.

Description of amendment request: The amendments would revise the South Texas Project, Units 1 and 2, Technical Specification Tables 2.2-1, 3.3-1, and 4.3-1, to change the description of the P-13 permissive interlock for the Reactor Trip System instrumentation. Specifically, the phrases “Turbine Impulse Chamber Pressure” and “Turbine Impulse Pressure” would be replaced with “Turbine Inlet Pressure.”

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

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

Response: No.

The proposed change to replace the words “Turbine Impulse Chamber Pressure” or “Turbine Impulse Pressure”, as appropriate, with “Turbine Inlet Pressure” in the descriptive text associated with the P-13 function of the Reactor Trip System does not involve any physical or design change to the P-13 function. The proposed change is intended to eliminate potential confusion by making the description generically applicable for other turbine types.

Therefore, there is no impact to the probability or consequences of an accident previously evaluated due to the proposed change.

2. Does the proposed amendment create the possibility of a new or different kind of Start Printed Page 25841accident from any accident previously evaluated?

Response: No.

Replacing the words “Turbine Impulse Chamber Pressure” with “Turbine Inlet Pressure” in the descriptive text associated with the P-13 function will not create the possibility of a new or different kind of accident from any accident previously evaluated. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. No new operator actions are created as a result of the proposed change.

Changing the descriptive text associated with the P-13 permissive has no impact on the accidents analyzed in the STPNOC [STP Nuclear Operating Company] Updated Final Safety Analysis Report (UFSAR) and is not an accident initiator. Since this change does not impact any conditions that would initiate an accident, there is no possibility of a new or different kind of accident resulting from this change.

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.

Changing the descriptive text associated with the P-13 permissive will not affect the margin of safety. The margin of safety presently provided by the Technical Specifications remains unchanged.

The proposed amendment does not affect the design of the facility or system operating parameters, does not physically alter safety-related systems and does not affect the method in which safety-related systems perform their functions.

Therefore, the proposed change does not impact margin of safety.

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

Attorney for licensee: Kym Harshaw, Vice President and General Counsel, STP Nuclear Operating Company, P.O. Box 289, Wadsworth, TX 77483.

NRC Branch Chief: Robert J. Pascarelli.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1 (Callaway), Callaway County, Missouri

Date of amendment request: March 12, 2019. A publicly-available version is in ADAMS under Accession No. ML19071A281.

Description of amendment request: The amendment would revise the Callaway technical specifications (TSs) to remove slave relay K620 from the scope of TS Surveillance Requirement (SR) 3.3.2.14 testing during shutdown conditions at 18-month intervals and incorporate it into the scope of SR 3.3.2.6 for surveillance testing during power operations, at a frequency in accordance with the Surveillance Frequency Control Program.

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

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

Response: No.

Testing slave relay K620 more frequently than currently required will not increase the probability or the consequences of a previously evaluated accident.

The new turbine controls being installed under a plant modification include new EHC [Electrohydraulic Control] trip bus coils with an impedance sized to allow a small test current to be applied to the trip logic without activating the trip coils. This permits the K620 slave relay to be tested on-line at the frequency used for testing other, similar slave relays in the plant and without any significant increase in the probability of an inadvertent turbine trip. Consequently, the new test scheme for this relay does not increase the probability of a previously evaluated transient (i.e., turbine trip) for Callaway.

Slave relay K620 provides trip signals to the Main Turbine and the Main Feedwater trip logic. Performing this test at the increased frequency will not adversely affect the relay's performance since the new frequency is typical for slave relays that can be tested during plant operation. It is thus reasonable to conclude that the likelihood of relay failure is not increased.

In regard to accident consequences, the change in test frequency for the K620 relay does not affect its required operability. Since the relay's function is not affected, there is no change to how the function is credited or assumed in the plant's accident analysis. The analyzed consequences are thus unaffected.

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.

Testing slave relay K620 more frequently than currently required does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Currently, slave relay K620 is tested with the turbine offline since under the current (unmodified) design, the testing of slave relay K620 produces a test current sufficient to trip the main turbine. The new proposed turbine controls include new EHC trip bus coils with an impedance sized to allow a small test current to be applied to the trip logic without activating the trip coils, thus allowing the slave relay test to be performed online. There is no change to the design or function of the relay itself or its associated logic. Thus, no new failure modes are introduced by the replacement of these trip coils.

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 TS change only affects the testability of the K620 relay (and thus the frequency at which the relay is tested). The design and function of the K620 slave relay itself are unchanged. No changes to the accident analyses, including any associated assumptions such as instrument setpoints or credited trip functions, are required or being made for this proposed change.

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

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

Attorney for licensee: John O'Neill, Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW, Washington, DC 20037.

NRC Branch Chief: Robert J. Pascarelli.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1 (Callaway), Callaway County, Missouri

Date of amendment request: March 22, 2019. A publicly-available version is in ADAMS under Accession No. ML19081A173.

Description of amendment request: The amendment would revise the Callaway technical specifications (TSs) to eliminate TS Section 5.5.8, “lnservice Testing Program.” The proposed change eliminates the Callaway TS Section 5.5.8, to remove requirements duplicated in the American Society of Mechanical Engineers Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) Code Case OMN-20, “Inservice Test Frequency,” which is approved for use in the Callaway Plant inservice testing program (IST). A new defined term, “INSERVICE TESTING PROGRAM,” will be added to TS Section 1.1, “Definitions.” The proposed change to the TSs is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, “TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing.”Start Printed Page 25842

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, “Administrative Controls,” Section 5.5, “Programs and Manuals,” by eliminating the “lnservice Testing Program” specification (i.e., TS 5.5.8). Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, “lnservice Test Frequency.” The remaining requirements in the Section 5.5 IST Program description 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 Section 1.1 of 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. lnservice test frequencies under Code Case OMN-20 are equivalent to the current testing periods allowed by the TS with the exception that test intervals greater than 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 test intervals greater than 2 years to be extended by 6 months (consistent with code case OMN-20) 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 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 amendment request involves no significant hazards consideration.

Attorney for licensee: John O'Neill, Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street NW, Washington, DC 20037.

NRC Branch Chief: Robert J. Pascarelli.

IV. 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.

Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida

Date of application for amendment: January 16, 2019.

Brief description of amendment: The amendment approved revision 1 to the Independent Spent Fuel Storage Installation-Only Emergency Plan for the CR-3 Site.

Date of issuance: May 3, 2019.

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

Amendment No.: 257. A publicly-available version is in ADAMS under Accession No. ML19080A186; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. DPR-72: This amendment revised the License.

Date of initial notice in Federal Register : February 12, 2019 (84 FR 3507).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 3, 2019.

No significant hazards consideration comments received: No.Start Printed Page 25843

Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of amendment request: April 25, 2018, as supplemented by letter dated March 26, 2019.

Brief description of amendments: The amendments revised the expiration date of an existing Note for Technical Specification 3.8.3, “Diesel Fuel Oil,” to allow, on a one-time basis, the main fuel oil storage tank to be inoperable for up to 14 days for the purpose of performing required inspection, cleaning, and any necessary repair activities.

Date of issuance: May 6, 2019.

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

Amendment Nos.: 290 and 318. A publicly-available version is in ADAMS under Accession No. ML19018A206; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-71 and DPR-62: The amendments revised the Renewed Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : July 3, 2018 (83 FR 31183).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 6, 2019.

No significant hazards consideration comments received: No.

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1 (River Bend), West Feliciana Parish, Louisiana

Date of amendment request: April 30, 2018, as supplemented by letter dated October 18, 2018.

Brief description of amendment: The amendment revised the River Bend Emergency Plan to adopt an Emergency Action Level scheme based on Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6, “Development of Emergency Action Levels for Non-Passive Reactors,” dated November 2012, which was endorsed by the NRC by letter dated March 28, 2013.

Date of issuance: May 14, 2019.

Effective date: As of the date of issuance and shall be implemented within 365 days from the date of issuance.

Amendment No.: 197. A publicly-available version is in ADAMS under Accession No. ML19070A062; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-47: The amendment revised the River Bend Emergency Plan.

Date of initial notice in Federal Register : July 31, 2018 (83 FR 36975). The supplemental letter dated October 18, 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 14, 2019.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear Plant, Van Buren County, Michigan

Date of amendment request: May 30, 2018, as supplemented by letters dated February 7 and April 17, 2019.

Brief description of amendment: The amendment revised Technical Specification 3.3.5, “Diesel Generator (DG)—Undervoltage Start (UV Start),” Surveillance Requirement 3.3.5.2a by adding a channel calibration requirement for the combined time delay setpoints for the degraded voltage sensing relay and the degraded voltage time delay relay.

Date of issuance: May 13, 2019.

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

Amendment No.: 268. A publicly-available version is in ADAMS under Accession No. ML19107A053; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-20: Amendment revised the Renewed Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : August 14, 2018 (83 FR 40347). The supplemental letters dated February 7 and April 17, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 13, 2019.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 28, 2018, as supplemented by letter dated February 4, 2019.

Brief description of amendment: The amendments authorized changes to Appendix E of the VEGP Units 3 and 4 Physical Security Plan to describe the Transitional Security Measures that will be implemented in the event that Unit 3 is ready to load fuel and begin operation with a contiguous Protected Area boundary and vehicle barrier system, and where a secure boundary is needed between VEGP Units 3 and 4. In addition, the amendment revised the plant-specific emergency planning inspections, tests, analyses, and acceptance criteria in Appendix C of the VEGP Unit 4 Combined License, associated with the presence of a security barrier between the Technical Support Center and the Unit 4 control room.

Date of issuance: April 30, 2019.

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

Amendment Nos.: 160 (Unit 3) and 158 (Unit 4). A publicly-available version is in ADAMS under Accession No. ML19092A449. The documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

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

Date of initial notice in Federal Register : January 8, 2019 (84 FR 88). The February 4, 2019, supplemental letter provided additional information that did not change the scope or the conclusions of 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 April 30, 2019.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: March 9, 2018, as supplemented by letters dated April 11, 2018, and January 30, 2019.

Brief description of amendments: The amendments authorized changes to the Essential Raw Cooling Water Motor Control Center Breakers and authorized revision of the Updated Final Safety Start Printed Page 25844Analysis Report (UFSAR) to describe the normal and alternate power sources for the ERCW system.

Date of issuance: May 7, 2019.

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

Amendment Nos.: 344—Unit 1 and 337—Unit 2. A publicly-available version is in ADAMS under Accession No. ML19058A029; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the UFSAR.

Date of initial notice in Federal Register : June 5, 2018 (83 FR 26107). The supplemental letter dated January 30, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 7, 2019.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 29th day of May 2019.

For the Nuclear Regulatory Commission.

Gregory F. Suber,

Deputy Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

End Signature End Supplemental Information

[FR Doc. 2019-11453 Filed 6-3-19; 8:45 am]

BILLING CODE 7590-01-P