2014-17257. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Biweekly notice.

SUMMARY:

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from June 26, 2014, to July 9, 2014.

DATES:

Comments must be filed by August 21, 2014. A request for a hearing must be filed by September 22, 2014.

ADDRESSES:

You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

• Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0169. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail].
• Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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

FOR FURTHER INFORMATION CONTACT:

Mable Henderson, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: [phone], email: [e-mail].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2014-0169 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

• Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0169.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section
• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2014-0169 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

The NRC cautions you not to include identifying or contact information in comment submissions 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, and the NRC does not routinely edit comment submissions to remove identifying or contact information.

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

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

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

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

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

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

Within 60 days after the date of publication of this notice, any person(s) whose interest may be affected by this action may file a request for a hearing and a petition to intervene with respect to issuance of the amendment to the subject facility operating license or combined license. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Agency Rules of Practice and Procedure” in 10 CFR Part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the requestor/petitioner seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, then any hearing held would take place before the issuance of any amendment.

B. Electronic Submissions (E-Filing)

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least ten 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at [e-mail], or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​getting-started.html. System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through the Electronic Information Exchange System, users will be required to install a Web browser plug-in from the NRC's Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an email notice confirming receipt of the document. The E-Filing system also distributes an email notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by email to [e-mail], or by a toll-free call at 1-866-672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​,, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, a request to intervene will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 60-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1)(i)-(iii).

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

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

Date of amendment request: April 23, 2014, as supplemented by letter dated June 19, 2014. Publicly available versions are in ADAMS under Accession Nos. ML14113A445 and ML14170B201, respectively.

Description of amendment request: The proposed amendment would revise the technical specification (TS) surveillance requirements (SRs) associated with TS 3.8.4, “DC Sources—Operating” and TS 3.8.6, “Battery Cell Parameters.”

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. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Performing the proposed changes in battery parameter surveillance testing and verification is not a precursor of any accident previously evaluated. Furthermore, these changes will help to ensure that the voltage and capacity of the batteries is such that they will provide the power assumed in calculations of design basis accident mitigation.

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

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

Response: No.

The proposed changes do not involve any modification of the plant or how the plant is operated; they only involve surveillance testing and verification activities.

Therefore, DTE concludes that these proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of the fuel cladding, reactor coolant, and containment systems will not be impacted by the proposed changes.

The proposed Fermi 2 revisions of the SRs ensure the continued availability and operability of the batteries. As such, sufficient [direct current] capacity to support operation of mitigation equipment remains within the design basis.

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

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

Attorney for licensee: Bruce R. Maters, DTE Energy, General Counsel—Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.

NRC Branch Chief: Robert D. Carlson.

Duke Energy Progress Inc., Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: February 10, 2014, as supplemented by letter dated April 4, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14052A065 and ML14107A339, respectively.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.3.1 for the Reactor Protection System Instrumentation Turbine Trip function on Low Auto Stop Oil (ASO) Pressure to a Turbine Trip function on Low Electro-Hydraulic (EH) Fluid Oil Pressure. The amendment would revise the Allowable Value and Nominal Trip Setpoint and revise the TS by applying additional testing requirements listed in Technical Specifications Task Force Traveler 493-A Revision 4, “Clarify Application of Setpoint Methodology for Limiting Safety System Setting Functions,” for Low EH Fluid Oil Pressure trip only.

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

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

Response: No.

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

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

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

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

Response: No.

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

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

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

Response: No.

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

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

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

Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202.

NRC Acting Branch Chief: Lisa M. Regner.

Duke Energy Progress Inc., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit No. 1, New Hill, North Carolina

Date of amendment request: April 24, 2014. A publicly-available version is in ADAMS under Accession No. ML14114A743.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3/4.4.5, “Steam Generator Tube Integrity,” TS 6.8.4.I, “Steam Generator Program,” and TS 6.9.1.7, “Steam Generator Tube Inspection Report” to address implementation associated with the inspections and reporting requirements as described in Technical Specifications Task Force (TSTF) TSTF-510-A, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.”

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

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

Response: No.

The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions.

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 changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

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 SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes. Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

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

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

Attorney for licensee: Lara S. Nichols, Deputy General Counsel, Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, Charlotte, NC 28202.

NRC Acting Branch Chief: Lisa M. Regner.

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

Date of amendment request: March 18, 2014. A publicly-available version is in ADAMS under Accession No. ML14086A389.

Description of amendment request: The amendment would adopt Technical Specification (TS) Task Force (TSTF) change traveler TSTF-535, Revision 0, “Revise Shutdown Margin [SDM] Definition to Address Advanced Fuel Designs,” at Columbia Generating Station. The notice of availability of TSTF-535, Revision 0, was announced in the Federal Register on February 26, 2013 (78 FR 13100).

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

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

Response: No.

The proposed change revises the definition of SDM. SDM is not an initiator to any accident previously evaluated. Accordingly, the proposed change to the definition of SDM has no effect on the probability of any accident previously evaluated. SDM is an assumption in the analysis of some previously evaluated accidents and inadequate SDM could lead to an increase in consequences for those accidents. However, the proposed change revises the SDM definition to ensure that the correct SDM is determined for all fuel types at all times during the fuel cycle. As a result, the proposed change does not adversely affect the consequences of any accident previously evaluated.

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

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

Response: No.

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

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

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

Response: No.

The proposed change revises the definition of SDM. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change ensures that the SDM assumed in determining safety limits, limiting safety system settings or limiting conditions for operation is correct for all BWR fuel types at all times during the fuel cycle.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: March 24, 2014, as supplemented by letter dated May 8, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML14098A400 and ML14141A538, respectively.

Description of amendment request: The amendment would revise Columbia Generating Station Technical Specification (TS) Table 3.3.1.1-1 to update Scram Discharge Volume (SDV) instrumentation nomenclature, add a Surveillance Requirement (SR) which was previously omitted, and add footnotes to an SR consistent with TS Task Force (TSTF) change traveler TSTF-493, Revision 4, “Clarify Application of Setpoint Methodology for LSSS [Limiting Safety System Settings] Functions,” Option A. The notice of availability of the models for plant-specific adoption of TSTF-493, Revision 4, was announced in the Federal Register on May 11, 2010 (75 FR 26294).

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 Function 7 names are administrative in nature and ensure that the description of SDV Water Level—High instrumentation in TS matches the plant configuration. The addition of a missing channel check SR and TSTF-493 footnotes for the new Function 7.b instruments makes the TS more comprehensive by ensuring the appropriate surveillances and footnotes are applied to this instrumentation.

The replacement instruments for Function 7.b meet the high functional reliability standard of GDC 21 [General Design Criteria 21, “Protection system reliability and testability,” of 10 CFR Part 50, Appendix A] and all pertinent requirements of 10 CFR 50.55a(h)(2). The instrumentation modification was reviewed under 10 CFR 50.59(c)(1) and determined to not meet any of the criteria in 10 CFR 50.59(c)(2).

The addition of a channel check to Function 7.a and addition of TSTF-493 notes (d) and (e) to SR 3.3.1.1.10 for the Function 7.b instrumentation do not change accident frequency or consequences. TS requirements that govern operability or routine testing of plant instruments are not assumed to be initiators of any analyzed event because these instruments are intended to prevent, detect, or mitigate accidents. Additionally, these proposed changes will not increase the consequences of an accident previously evaluated because the proposed changes do not adversely impact structures, systems, or components. The proposed TS changes establish requirements that ensure components are operable when necessary for the prevention or mitigation of accidents or transients. Furthermore, there will be no change in the types or significant increase in the amounts of any effluents released offsite.

In summary, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes to administratively revise instrument descriptions, incorporate a new SR, and add footnotes to an existing SR do not change the parameters within which Columbia is operated.

The proposed changes do not adversely impact the manner in which the SDV Water Level—High RPS [Reactor Protection System] instrumentation will operate under normal and abnormal operating conditions. The instrumentation design changes were reviewed under 10 CFR 50.59(c)(1) and determined to not meet any of the criteria of 10 CFR 50.59(c)(2). The proposed changes will not alter the functional demands on credited equipment. No alteration in the procedures which ensure that Columbia remains within analyzed limits are proposed and no change is being made to procedures relied upon to respond to an off-normal event.

Therefore, these proposed changes provide an equivalent level of safety and will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed changes to the function descriptions in TS Table 3.3.1.1-1 Functions 7.a and 7.b are considered administrative in nature, and do not impact plant safety.

Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the establishment of setpoints to initiate alarms and actions. The proposed changes support a planned upgrade of the SDV instrumentation that preserves the reliability of the RPS system. The proposed changes do not adversely affect the probability of failure or availability of the affected instrumentation. The instrumentation design changes were evaluated under 10 CFR 50.59(c)(1) and determined not to meet any of the criteria of 10 CFR 50.59(c)(2).

The addition of a Channel Check SR to TS Table 3.3.1.1-1 Function 7.a and the addition of TSTF-493 notes (d) and (e) to SR 3.3.1.1.10 for the new scram discharge instrumentation in TS Table 3.3.1.1-1 Function 7.b are conservative changes that align the SRs for proper determination of operability with that of similar instrumentation.

On this basis, is concluded that the proposed changes do not result in a reduction in the margin of safety.

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

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

NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: November 26, 2013. A publicly-available version is in ADAMS under Accession No. ML13346A026.

Description of amendment request: The amendment would revise Technical Specification 4.3.4, “Heavy Loads” limitation imposed on maximum weight that could travel over the irradiated fuel in the spent fuel pool.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (CFR) Section 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 Reactor Building crane is being upgraded to meet the applicable single-failure-proof criteria of NUREG 0554 and NUREG 0612 for the modification of the existing non single-failure-proof crane. While loads in excess of 2,000 lbs [pounds] shall continue to be prohibited from travel over irradiated fuel assemblies in the spent fuel pool by the PNPS [Pilgrim Nuclear Power Station] Technical Specifications, a Multi-Purpose Canister (MPC) lid will be permitted to travel over irradiated fuel assemblies in a transfer cask, using a single-failure-proof handling system as described in NUREG-0800 Section 9.1.5 Paragraph llI.4.C, to enable the conduct of dry cask storage loading and unloading operations. Specifically, this will enable the MPC lid and its associated lifting apparatus to travel over irradiated fuel assemblies in a MPC. The probability of dropping this load onto an irradiated fuel assembly in the canister is reduced as a result of the reliability of the single-failure-proof handling system.

The proposed change does not affect the consequences of any accidents previously evaluated in the PNPS UFSAR [Updated Final Safety Analysis Report]. The change involves the travel of heavy loads over irradiated fuel assemblies in a transfer cask using a single-failure-proof handling system. Under these circumstances, no new load drop accidents are postulated and no changes to the probabilities or consequences of accidents previously evaluated are involved.

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

Response: No.

Section 10.3 of the PNPS UFSAR evaluates fuel storage and handling operations. Section 14 of the PNPS UFSAR discusses the analysis of design basis fuel handling accidents involving drop of an irradiated assembly resulting in multiple fuel rod failures and consequent release of radioactivity. The change involves the travel of heavy loads over irradiated fuel assemblies in a transfer cask using a single-failure-proof handling system. Under these circumstances, no new or different load drop accidents are postulated to occur and there are no changes in any of the load drop accidents previously evaluated.

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

Response: No.

The revised Technical Specification changes do not involve a reduction in any margin of safety. Technical Specification 4.3.4 currently prohibits travel of heavy loads in excess of 2,000 lbs over irradiated fuel assemblies in the spent fuel pool. The proposed change will continue to restrict travel of heavy loads in excess of 2,000 lbs over irradiated fuel assemblies in the spent fuel pool, with the exception of the MPC lid over irradiated fuel assemblies in the canister to enable dry cask storage operations. This exception is only permitted when the heavy load is handled using a single-failure-proof handling system. Due to the reliability of this upgraded handling system that complies with the guidance of NUREG-0800 Section 9.1.5 for a single-failure-proof handling system, a load drop accident is not considered a credible event. Under these circumstances, no new load drop accidents are postulated and no reductions in margins of safety are involved.

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

Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Benjamin G. Beasley

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: March 24, 2014. A publicly available version is in ADAMS under Accession No. ML14085A257.

Description of amendment request: The proposed amendment would revise the site emergency plan for the permanently defueled condition to reflect changes in the on-shift staffing and Emergency Response Organization staffing.

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

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

Response: No.

The proposed changes to the SEP [Site Emergency Plan] do not impact the function of plant structures, systems, or components (SSCs). The proposed changes do not affect accident initiators or precursors, nor does it alter design assumptions. The proposed changes do not prevent the ability of the on-shift staff and ERO [Emergency Response Organization] to perform their intended functions to mitigate the consequences of any accident or event that will be credible in the permanently defueled condition. The proposed changes only remove positions that will no longer be credited in the SEP in the permanently defueled condition.

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

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

Response: No.

The proposed changes reduce the number of on-shift and ERO positions commensurate with the hazards associated with a permanently shutdown and defueled facility. The proposed changes do not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed changes do not result in a change to the way that the equipment or facility is operated so that no new accident initiators are created.

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.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes are associated with the SEP staffing and do not impact operation of the plant or its response to transients or accidents. The change does not affect the Technical Specifications. The proposed changes do not involve a change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Safety analysis acceptance criteria are not affected by the proposed changes. The revised SEP will continue to provide the necessary response staff with the proposed changes.

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

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

Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Douglas A. Broaddus.

Exelon Generation Company (EGC), LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Date of amendment request: April 17, 2014. A publicly-available version is in ADAMS under Accession No. ML14111A257.

Description of amendment request: The proposed amendment would revise required action notes in the Braidwood and Byron TS 3.3.1 and TS 3.3.2 to reflect the specific functions in TS 3.3.1 and TS 3.3.2 that have bypass test capability installed and the specific functions that do not have bypass test capability installed. The current wording is no longer applicable because the installation and implementation of the bypass test instrumentation modifications for certain functions have been completed.

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 is administrative in nature as it revises previously approved specific TS [Technical Specifications] Required Actions Notes that are no longer applicable following plant modification installation and implementation to reflect the applicable RTS [Reactor Trip System] and ESFAS [Engineered Safety Feature Actuation System] Functions with installed bypass test capability.

The proposed change does not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events modeled in the safety analyses. The proposed change does not alter the design assumptions, conditions, or configuration of the facility, nor does it affect the structural and functional integrity of the RTS and ESFAS. The proposed change does not alter or prevent the ability of any structures, systems, and components from performing their intended design function to mitigate the consequences of an initiating event within the applicable acceptance criteria.

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

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

Response: No.

The proposed change to revise previously approved specific TS Required Actions Notes that are no longer applicable to specific RTS and ESFAS Functions with installed bypass test capability is administrative in nature. The proposed change does not result in a change to any design function or the manner in which the RTS and ESFAS operates to provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed change is implemented. In addition, this change does not install or modify any plant equipment. Therefore, no new failure modes are being created nor does the change result in the creation of any changes to the existing accident scenarios or do they create any new or different accident scenarios. The types of accidents defined in the UFSAR [updated final safety analysis report] continue to represent the credible spectrum of events to be analyzed which determine safe plant operation.

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.

No safety analyses are changed or modified as a result of the proposed change to revise previously approved specific TS Required Actions Notes that are no longer applicable to RTS and ESFAS Functions with installed bypass test capability. The proposed change does not alter the manner in which the safety limits, limiting safety system settings, or limiting conditions for operation are determined. Margins associated with the current applicable safety analyses acceptance criteria are unaffected. The current safety analyses remain bounding since their conclusions are not affected by this change and the plant will continue to operate in a manner consistent with the safety analyses. The safety systems credited in the safety analyses will continue to be available to perform their mitigation functions.

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

Based on the above evaluation, EGC concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and, accordingly, a finding of no significant hazards consideration is justified.

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

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Travis L. Tate.

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendment request: October 31, 2013. A publicly-available version is in the ADAMS System under Accession No. ML13308A387.

Description of amendments request: The amendments would modify the Technical Specification requirements regarding steam generator tube inspections and reporting as described in Technical Specification Task Force 510-A, Revision 2, “Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.”

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. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

No.

The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture event (SGTR) is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed these assumptions.

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

2. Create the possibility of a new or different type of accident from any accident previously evaluated; or

No.

The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

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

3. Involve a significant reduction in a margin of safety.

No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

Therefore, the proposed amendment would 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 amendments request involves no significant hazards consideration.

Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Benjamin G. Beasley

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendment request: November 13, 2013. A publicly-available version is in the ADAMS System under Accession No. ML13318A892.

Description of amendments request: The amendments would modify the Technical Specification requirements to adopt the changes described in Technical Specification Task Force 426-A, Revision 5, “Revise or Add Actions to Preclude Entry into LCO 3.0.3—RITSTF Initiatives 6b and 6c.”

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. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

No.

The proposed change provides a short Completion Time to restore an inoperable system for conditions under which the existing Technical Specifications require a plant shutdown to begin within one hour in accordance with Limiting Condition for Operation 3.0.3. Entering into Technical Specification Actions is not an initiator of any accident previously evaluated. As a result, the probability of an accident previously evaluated is not significantly increased. The consequences of any accident previously evaluated that may occur during the proposed Completion Times are no different from the consequences of the same accident during the existing one hour allowance. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

2. Create the possibility of a new or different type of accident from any accident previously evaluated; or

No.

No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis.

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

3. Involve a significant reduction in a margin of safety.

No.

The proposed change increases the time the plant may operate without the ability to perform an assumed safety function. The analyses in WCAP-16125-NP-A, “Justification for Risk-Informed Modifications to Selected Technical Specifications for Conditions Leading to Exigent Plant Shutdown,” Revision 2, August 2010, demonstrated that there is an acceptably small increase in risk due to a limited period of continued operation in these conditions and that this risk is balanced by avoiding the risks associated with a plant shutdown. As a result, the change to the margin of safety provided by requiring a plant shutdown within one hour is not significant.

Therefore, the proposed amendment would 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 amendments request involves no significant hazards consideration.

Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 Exelon Way, Kennett Square, PA 19348

NRC Branch Chief: Benjamin G. Beasley.

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendment request: January 13, 2014. A publicly-available version is in the ADAMS under Accession No. ML14015A138.

Description of amendments request: The amendments would add a Technical Specification (TS) for the atmospheric dump valves (ADVs).

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 addition of a new TS to address the operability of the ADVs does not alter the assumed initiators to any analyzed event. The probability of an accident previously evaluated will not be increased by this proposed change. This proposed change will not affect radiological dose consequence analyses. The radiological dose consequence analyses assume a certain release of radioactive material through the ADVs following a steam generator tube rupture (SGTR), which is not affected by the addition of the ADVs to the TS. The addition of a Surveillance Requirement for the ADVs will continue to ensure that the ADVs can perform their specified function. The consequences of an accident previously evaluated will not be increased by this proposed change.

Therefore, operation of the facility in accordance with the proposed TS for the ADVs will 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 addition of a new TS to address the operability of the ADVs has been evaluated to determine the effect of adding the new TS to the operation of the plant. This change does not involve any alteration in the plant configuration (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Therefore, operation of the facility in accordance with the proposed addition of a new TS to address the operability of the ADVs would not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The margin of safety is related to the ability of the ADV to release enough steam to cool the Reactor Coolant System down and be isolated when required to limit the radioactive release from a SGTR. The inclusion of the ADVs in the TS will provide limited time for continued operation without both ADVs available. This ensures that the margin of safety is maintained by ensuring that the ADV can meet the assumptions for its operation specified in the SGTR analysis. Since the radiological consequences of a SGTR are not affected by the addition of the proposed TS, the margin of safety is not changed significantly.

Therefore, the proposed addition of a new TS to address the operability of the ADVs does not involve a significant reduction in the margin of safety.

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

Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Benjamin G. Beasley.

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendment request: February 13, 2014. A publicly-available version is in the ADAMS under Accession No. ML14050A374.

Description of amendments request: The amendments would modify the as-found lift tolerances in the surveillance requirement for the pressurizer safety valves (PSVs).

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?

No.

The proposed change, modifying the as-found and as-left lift settings in the Surveillance Requirement of the PSVs, does not change the design function or operation of the PSVs and it does not change the way the PSVs are maintained, tested, or inspected. The PSVs are not accident initiators; they operate in response to the pressurization of the Reactor Coolant System (RCS). They limit the pressure of the RCS to less than the allowable American Society of Mechanical Engineers Boiler and Pressure Vessel, Section III Code during an accident or transient. Analyses were performed of peak pressure events, which are evaluated against the RCS limit. Action of the PSVs is required to mitigate the consequences of these events. The change in the setpoint tolerance and a change in one valve's nominal setpoint were explicitly considered in the analysis of these events. The RCS pressure remained below the required limits with these changes considered. Therefore, this change does not impact the ability of the PSVs to perform their safety function during evaluated accidents.

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

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

No.

The proposed change, modifying the as-found and as-left lift settings in the Surveillance Requirement of the PSVs, does not change the PSVs design function to maintain RCS pressure below the RCS pressure Safety Limit of 2750 psia [pounds per square inch absolute] during design basis accidents nor does it affect the PSVs ability to perform this design function. The proposed change does not require any modification to the plant (other than the setpoint change) or change equipment operation or testing. It also does not create any credible new failure mechanisms, malfunctions, or accident initiators that would cause an accident not previously considered.

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

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

No.

The proposed change, modifying the as-found and as-left lift settings in the Surveillance Requirement of the PSVs, does not involve a significant reduction in the margin of safety in maintaining RCS pressure below Safety Limits of 2750 psia during design basis accidents. The analyses conducted in support of this proposed change evaluated the ability of the PSVs to maintain an adequate safety margin assuming the change in setpoint tolerances and a change in one valve's nominal setpoint. The analysis determined that the response of the PSVs would maintain an adequate safety margin to the reactor coolant Safety Limit of 2750 psia.

Therefore the proposed change does not involve a significant reduction in the margin of safety of maintaining RCS pressure the below RCS pressure Safety Limit.

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 amendments request involves no significant hazards consideration.

Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Benjamin G. Beasley.

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendment request: May 1, 2014. A publicly-available version is in the ADAMS under Accession No. ML14125A015.

Description of amendments request: The amendments would modify the Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute 04-10, “Risk Informed Method for Control of Surveillance Frequencies.”

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. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the Technical Specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

2. Create the possibility of a new or different type of accident from any accident previously evaluated; or

Response: No.

No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

3. Involve a significant reduction in a margin of safety.

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures and components specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the updated final safety analysis report and the bases to the TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Calvert Cliffs will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Revision 1 in accordance with the TS Surveillance Frequency Control Program. Nuclear Energy Institute 04-10, Revision 1 methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

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

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

Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Benjamin G. Beasley.

Florida Power and Light Company (FPL), et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: February 20, 2014. Available in ADAMS under Accession No. ML14070A087.

Description of amendment request: The amendment would revise the Technical Specifications (TSs) by relocating specific surveillance frequency requirements to a licensee-controlled program with implementation of Nuclear Energy Institute (NEI) 04-10, “Risk Informed Technical Specification Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies” (ADAMS Accession No. ML071360456). The licensee stated that the NEI 04-10 methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Guide 1.177, “An Approach for Plant-Specific Risk-Informed Decision-Making: Technical Specifications” (ADAMS Accession No. ML003740176). The licensee stated that the changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specifications change TSTF-425, “Relocate Surveillance Frequencies to Licensee Control—RITSTF [Risk Informed Technical Specifications Task Force] Initiative 5b,” Revision 3 (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of TSTF-425, Revision 3.

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

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the Technical Specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be [sic] capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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.

No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, FPL will perform a probabilistic risk evaluation using the guidance contained in NRC-approved NEI 04-10, Revision 1 in accordance with the TS Surveillance Frequency Control Program. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide (RG) 1.177.

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

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

Attorney for licensee: William S. Blair, Managing Attorney—Nuclear, Florida Power & Light Company, 700 Universe Blvd. MS LAW/JB, Juno Beach, Florida 33408-0420.

NRC Acting Branch Chief: Lisa M. Regner.

Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251, Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida

Date of amendment request: April 9, 2014. Available in ADAMS under Accession No. ML14105A042.

Description of amendment request: The amendment would revise the Technical Specifications (TSs) by relocating specific surveillance frequency requirements to a licensee-controlled program with implementation of Nuclear Energy Institute (NEI) 04-10, “Risk Informed Technical Specification Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies” (ADAMS Accession No. ML071360456). The licensee stated that the NEI 04-10 methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Guide 1.177, “An Approach for Plant-Specific Risk-Informed Decision-Making: Technical Specifications” (ADAMS Accession No. ML003740176). The licensee stated that the changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specifications change TSTF-425, “Relocate Surveillance Frequencies to Licensee Control—RITSTF [Risk Informed Technical Specifications Task Force] Initiative 5b,” Revision 3 (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of TSTF-425, Revision 3.

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

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the Technical Specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be [sic] capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

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

Response: No.

The proposed changes relocate the surveillance frequencies for Surveillance Requirements that have a set periodicity from the TS to a licensee controlled Surveillance Frequency Control Program. This change does not alter any existing surveillance frequencies. Within the constraints of the Program, the licensee will be able to change the periodicity of these surveillance requirements. Relocating the surveillance frequencies does not impact the ability of structures, systems or components (SSCs) from performing there [sic] design functions, and thus, does not create the possibility of a new or different kind of accident from any previously evaluated.

No new or different accidents result from utilizing the proposed change. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

The design, operation, testing methods, and acceptance criteria for structures, systems, and components (SSCs) specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, FPL will perform a probabilistic risk evaluation using the guidance contained in NRC-approved NEI 04-10, Revision 1 in accordance with the TS Surveillance Frequency Control Program. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide (RG) 1.177, An Approach for Plant-Specific Risk-Informed Decision-Making: Technical Specifications.

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

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

Attorney for licensee: William S. Blair, Managing Attorney—Nuclear, Florida Power & Light Company, 700 Universe Blvd. MS LAW/JB, Juno Beach, Florida 33408-0420.

NRC Acting Branch Chief: Lisa M. Regner.

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: June 3, 2014. A publicly available version is in ADAMS under Accession No. ML14154A136.

Description of amendment request: The amendments would revise the Technical Specification Limiting Condition for Operation 3.3.1 and Surveillance Requirement 3.2.4.2 regarding the reactor trip system instrumentation.

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

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

Response: No.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do 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 changes are consistent with safety analysis assumptions and resultant consequences.

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

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

Response: No.

The proposed changes do not result in a change in the manner in which the Reactor Trip System (RTS) and engineered safety features actuation system (ESFAS) provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed changes are implemented. There are no design changes associated with the license amendment.

The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

The proposed changes do 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 impacted by these changes. Redundant RTS and ESFAS trains are maintained, and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also maintained. All signals credited as primary or secondary, and all operator actions credited in the accident analyses will remain the same. The proposed changes will not result in plant operation in a configuration outside the design basis.

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

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

Attorney for licensee: Leigh D. Perry, SVP & General Counsel of Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201.

NRC Branch Chief: Robert J. Pascarelli.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant (HNP), Unit Nos. 1 and 2, Appling County, Georgia

Date of amendment request: March 17, 2014. A publicly-available version is in ADAMS under Accession No. ML14076A141.

Description of amendment request: The proposed amendments would modify Technical Specification (TS) definition of Shutdown Margin (SDM) to require calculation of the SDM at a reactor moderator temperature of 68 °F or a higher temperature that represents the most reactive state throughout the operating cycle. This change is needed to address new Boiling Water Reactor (BWR) fuel designs which may be more reactive at shutdown temperatures above 68 °F.

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

SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

The proposed change revises the definition of SDM. SDM is not an initiator to any accident previously evaluated. Accordingly, the proposed change to the definition of SDM has no effect on the probability of any accident previously evaluated. SDM is an assumption in the analysis of some previously evaluated accidents and inadequate SDM could lead to an increase in consequences for those accidents. However, the proposed change revises the SDM definition to ensure that the correct SDM is determined for all fuel types at all times during the fuel cycle. As a result, the proposed change does not adversely affect the consequences of any accident previously evaluated.

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

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

Response: No.

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

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

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

Response: No.

The proposed change revises the definition of SDM. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed change ensures that the SDM assumed in determining safety limits, limiting safety system settings or limiting conditions for operation is correct for all BWR fuel types at all times during the fuel cycle.

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

Based on the above, SNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92

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: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

NRC Branch Chief: Robert Pascarelli.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 And 50-281 Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of amendment request: April 11, 2014. A publicly-available version is in ADAMS under Accession No. ML14112A073.

Description of amendment request: The proposed license amendment requests the changes to the Technical Specification (TS) TS 4.2, “Augmented Inspections,” and TS 4.15, “Augmented Inservice Inspection Program for High Energy Lines Outside of Containment,” by relocating to the Surry Technical Requirements Manual (TRM). In addition, TS 6.4.U, “Augmented Inspections and Examinations,” will be added to the Administrative Controls Section 6.4, “Unit Operating Procedures and Programs.” The proposed relocation of the TS 4.2 and TS 4.15 requirements to the TRM is appropriate since these requirements do not satisfy the categories and criteria of 10 CFR 50.36(c) for inclusion in the TS. Along with the relocation of the TS 4.2 and TS 4.15 requirements to the TRM, the Bases for TS 4.2 and TS 4.15 are also being relocated to the TRM.

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 relocates Technical Specification (TS) 4.2, “Augmented Inspections,” TS 4.15, “Augmented Inservice Inspection Program for High Energy Lines Outside of Containment,” and the associated TS Bases to the Surry Technical Requirements Manual (TRM). In addition, TS 6.4.U, “Augmented Inspections and Examinations,” will be added to the Surry TS. The proposed relocation of the TS 4.2 and TS 4.15 requirements to the TRM is appropriate since these requirements do not satisfy the categories and criteria of 10CFR50.36(c), which specifies what items qualify for inclusion in the TS.

Specifically, the TS 4.2 augmented inspections of the low head safety injection piping located in the valve pit, the reactor coolant pump flywheel, the low pressure turbine rotor blades, sensitized stainless steel, and TS 4.15 augmented inspections of the welds in the main steam and main feedwater lines in the main steam valve house of each unit will be relocated to the TRM. The augmented inspections, which are performed in addition to required ASME Code Section Xl inspections/examinations, will continue to be performed as required by the TRM.

The plant systems and components to which the augmented inspections apply will not be operated in a different manner. The proposed relocation of the augmented inspections does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled.

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 involve any physical alteration of plant equipment. As such, no new or different types of equipment will be installed, and the basic operation of installed plant systems and components, to which the augmented inspections apply, is unchanged.

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 will not reduce a margin of safety because the relocation of the augmented inspections to the TRM has no impact on any safety analysis assumptions, as indicated by the fact that the requirements do not meet the 10CFR50.36(c) criteria for inclusion in the TS. In addition, the augmented inspections will be moved to the TRM without change and will continue to be performed as required by the TRM.

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 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: Robert Pascarelli.

ZionSolutions LLC (ZS), Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station (ZNPS), Units 1 and 2, Lake County, Illinois

Date of amendment request: May 27, 2014. A publicly-available version is in ADAMS under Accession No. ML14148A295.

Description of amendment request: The license amendment request proposes changes to ZNPS Defueled Station Emergency Plan (DSEP) in accordance with 10 CFR 50.54(q). ZS proposes removal of the various emergency actions related to the former spent fuel pool, the transfer of responsibility for implementing the Emergency Plan to the Independent Spent Fuel Storage Installation (ISFSI) Shift Supervisor, a revised emergency plan organization, abandonment of the Control Room consistent with the current state of decommissioning, transition to NEI 99-01 Revision 6 and reformatting consistent with current industry practice.

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

No. ZS has, in effect, a U.S. Nuclear Regulatory Commission-approved (NRC) emergency plan. The remaining ZNPS accident (Radioactive Waste Handling Accident) and the credible accidents involving the ISFSI and the Modular, Advanced Generation, Nuclear All-purpose Storage (MAGNASTOR) system have been analyzed and determined that none result in doses to the public beyond the owner controlled area boundary that would exceed the U.S. Environmental Protection Agency's (EPA) Protective Action Guides (PAGs). These analyses have not changed. With spent fuel relocated to the ISFSI, the Spent Fuel Pool previously analyzed events (Loss of Spent Fuel Pool Cooling, Loss of Spent Fuel Pool Inventory, and Fuel Handling Accident in the Fuel Building) are no longer credible.

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

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

No. ZS has, in effect, an NRC-approved emergency plan. The remaining ZNPS accident (Radioactive Waste Handling Accident) and the credible accidents involving the ISFSI and MAGNASTOR system have been analyzed and determined that none result in doses to the public beyond the owner controlled area boundary that would exceed the EPA's PAGs. These analyses have not changed. With spent fuel relocated to the ISFSI, the Spent Fuel Pool previously analyzed events (Loss of Spent Fuel Pool Cooling, Loss of Spent Fuel Pool Inventory, and Fuel Handling Accident in the Fuel Building) are no longer credible. Accidents associated with the ISFSI are addressed in the MAGNASTOR Final Safety Analysis Report.

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

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

No. Margin of safety is related to the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. ZS has, in effect, an NRC-approved emergency plan. The remaining ZNPS accident (Radioactive Waste Handling Accident) and the credible accidents involving the ISFSI and MAGNASTOR system have been analyzed and determined that none result in doses to the public beyond the owner controlled area boundary that would exceed the EPA's PAGs These analyses have not changed. With spent fuel relocated to the ISFSI, the Spent Fuel Pool previously analyzed events (Loss of Spent Fuel Pool Cooling, Loss of Spent Fuel Pool Inventory, and Fuel Handling Accident in the Fuel Building) are no longer credible.

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: Russ Workman, Deputy General Counsel, EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 84101.

NRC Branch Chief: Bruce Watson.

III. Notice of Issuance of Amendments to Facility Operating Licenses and Combined Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

A notice of consideration of issuance of amendment to facility operating license or combined license, as applicable, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions, was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) The applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items can be accessed as described in the “Obtaining Information and Submitting Comments” section of this document.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

Date of amendment request: May 23, 2013, as supplemented by letter dated October 11, 2013.

Brief description of amendment(s): The amendments revise the Technical Specifications to risk-inform requirements regarding selected Required Action End States. Specifically, the changes permit an end state of Mode 4 rather than an end state of Mode 5 consistent with Technical Specification Task Force (TSTF) Traveler TSTF 432-A, Revision 1, “Change in Technical Specifications End States WCAP-16294.”

Date of issuance: July 7, 2014.

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

Amendment No.: Unit 2-275; Unit 3-252. A publicly-available version is in ADAMS under Accession No. ML14122A303; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. DPR-26 and DPR-64: The amendment revised the Facility Operating License and the Technical Specifications.

Date of initial notice in Federal Register: July 23, 2013 (78 FR 44170). The supplemental letter 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 July 7, 2014.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 11, 2013, as supplemented by letter dated December 11, 2013.

Brief description of amendment: The amendment revised Technical Specification 2.1.1.1, to add a provision for the determination of the maximum local fuel pin centerline temperature using the NRC reviewed and approved COPERNIC fuel performance computer code.

Date of issuance: July 9, 2014.

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

Amendment No.: 249. A publicly-available version is in ADAMS under Accession No. ML14169A475; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications and the renewed facility operating license.

Date of initial notice in Federal Register: April 1, 2014 (79 FR 18331). The supplemental letter dated December 11, 2013, 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 July 9, 2014.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of application for amendment: November 8, 2013.

Brief description of amendment: The amendment revised the Technical Specification (TS) definition of “Shutdown Margin” (SDM) to require calculation of the SDM at a reactor moderator temperature of 68 degrees Fahrenheit (°F) or a higher temperature that represents the most reactive state throughout the operating cycle. This change is needed to address new Boiling Water Reactor (BWR) fuel designs which may be more reactive at shutdown temperatures above 68 °F.

This TS change is part of the Consolidated Line Item Improvement Process (CLIIP) TS Task Force (TSTF) Traveler TSTF-535, Revision 0, “Revise Shutdown Margin Definition to Address Advanced Fuel Designs.” The licensee stated there are no variations or deviations from the NRC staff's model safety evaluation.

Date of issuance: June 30, 2014.

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

Amendment No: 198. A publicly-available version is in ADAMS under Accession No. ML14106A133; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-29: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : March 4, 2014 (79 FR 12244).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 30, 2014.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1, Oswego County, New York

Date of application for amendment: June 11, 2012, as supplemented by letters dated February 27, March 27, April 30, and December 9, 2013; and January 22, March 14, April 15, May 9, and May 23, 2014.

Brief description of amendment: The amendment authorizes the transition of the Nine Mile Point Nuclear Station, Unit 1, fire protection program to a risk-informed, performance-based program based on National Fire Protection Association (NFPA) 805, in accordance with 10 CFR 50.48(c). NFPA 805 allows the use of performance-based methods such as fire modeling and risk-informed methods such as fire probabilistic risk assessment to demonstrate compliance with the nuclear safety performance criteria.

Date of issuance: June 30, 2014.

Effective date: As of its date of issuance and shall be implemented by 180 days from the date of issuance.

Amendment No.: 215. A publicly available version is in ADAMS under Accession No. ML14126A003; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-63: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register : September 11, 2012 (77 FR 55874).

The supplements dated February 27, March 27, April 30, and December 9, 2013; and January 22, March 14, April 15, May 9, and May 23, 2014, 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 June 30, 2014.

No significant hazards consideration comments received: No.

Amendment No.: 215. A publicly available version is in ADAMS under Accession No. ML14126A003; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. DPR-63: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register : September 11, 2012 (77 FR 55874).

The supplements dated February 27, March 27, April 30, and December 9, 2013; and January 22, March 14, April 15, May 9, and May 23, 2014, 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 June 30, 2014.

No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: March 28, 2013, as supplemented by letters dated July 16, October 22, November 26, and December 17, 2013, and January 16, April 17, and May 1, 2014.

Brief description of amendment: The amendments revised Technical Specification (TS) 3.7.16, “Fuel Storage Pool Boron Concentration,” TS 3.7.17, “Spent Fuel Assembly Storage,” TS 4.3, “Fuel Storage,” and TS 5.5, “Programs and Manuals,” for storage of uprated fuel in Region II of the spent fuel pool. Changes to TS 3.7.16 reflect a change in the required fuel storage pool soluble boron concentration based on the results of a new criticality analysis. Changes to TS 3.7.17 include new spent fuel pool loading restrictions in terms of allowable storage patterns, and minimum burnup requirements as a function of enrichment, fuel type, and fuel reactivity category. The revised TS 4.3 section includes updates to the minimum soluble boron concentration, Region I fuel assembly spacing, specific new or partially spent fuel assembly storage restrictions in Region II consistent with TS 3.7.17, and general Region II storage restrictions consistent with TS 3.7.17. The change to TS 5.5 adds TS program 5.5.22, “Neutron Absorber Monitoring Program.”

Date of issuance: July 1, 2014.

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

Amendment Nos.: Unit 1-162; Unit 2-162. A publicly-available version is in ADAMS under Accession No. ML14160A035; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : November 5, 2013 (78 FR 66391). The NRC staff's original proposed no significant hazards consideration determination was based on letters dated March 28, and July 16, 2013. The supplements dated October 22, November 26, and December 17, 2013, and January 16, April 17, and May 1, 2014, 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 July 1, 2014.

No significant hazards consideration comments received: No.

NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of amendment request: January 15, 2013, as supplemented on March 1, April 18, and September 12, 2013, and March 11, 2014.

Description of amendment: The license amendment revised Technical Specifications 5.6.5, “Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),” to allow the use of two new methodologies for determining RCS pressure and temperature limits at the Point Beach Nuclear Plant, Units 1 and 2.

Date of issuance: June 30, 2014.

Effective date: As of the date of issuance and shall be implemented with 180 days.

Amendment Nos.: 250 (Unit 1) and 254 (Unit 2). A publicly-available version is in ADAMS under Accession No. ML14126A378; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: The amendment revised the Renewed Facility Operating License and the Technical Specifications.

Date of initial notice in Federal Register : June 11, 2013 (78 FR 35062). The supplemental letters dated March 1, April 18, and September 12, 2013, and March 11, 2014, 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 June 30, 2014.

No significant hazards consideration comments received: No.

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

Date of amendment requests: August 12, 2013, as supplemented by letters dated January 24, March 13, and March 25, 2014.

Brief description of amendments: The licensee requested to revise the Technical Specifications to, in effect, extend the Type A primary containment Integrated Leak Rate Test intervals to fifteen years and the Type C local leak rate test intervals to 75 months, and incorporate the regulatory positions stated in RG 1.163.

Date of issuance: July 3, 2014.

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

Amendment Nos.: Unit 1, 282; Unit 2, 282. A publicly-available version is in ADAMS under Accession No. ML14148A235; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.

Renewed Facility Operating License Nos. DPR-32 and DPR-37: The amendments revise the Renewed Facility Operating Licenses and the Technical Specifications.

Date of initial notice in Federal Register : October 29, 2013 (78 FR 64548). The supplemental letters dated January 24, March 13, and March 25, 2014, 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 July 3, 2014.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 26, 2013.

Brief description of amendment: The amendment revised Technical Specification Surveillance Requirement (SR) 3.7.10.1 and SR 3.7.13.1 to reduce the required run time for periodic operation of the control room pressurization system filter trains and emergency exhaust system filter trains, with heaters on, from 10 hours to 15 minutes. The amendment is consistent with plant-specific options provided in the NRC's model safety evaluation in Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, “Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month,” as part of the consolidated line item improvement process.

Date of issuance: July 1, 2014.

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

Amendment No.: 209. A publicly-available version is in ADAMS under Accession No. ML14175A390; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register : January 21, 2014 (79 FR 3418).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 1, 2014.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: December 17, 2013.

Brief description of amendment: The amendment revised Technical Specification Surveillance Requirement (SR) 3.7.10.1 and SR 3.7.13.1 to reduce the required run time for periodic operation of the control room pressurization system filter trains and emergency exhaust system filter trains, with heaters on, from 10 hours to 15 minutes. The amendment is consistent with plant-specific options provided in the NRC's model safety evaluation in Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, “Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month,” as part of the consolidated line item improvement process.

Date of issuance: July 1, 2014.

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

Amendment No.: 208. A publicly-available version is in ADAMS under Accession No. ML14157A082; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.

Renewed Facility Operating License No. NPF-42. The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register : March 18, 2014 (79 FR 15151).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 1, 2014.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 14th day of July 2014.

For the Nuclear Regulatory Commission.

Louise Lund,

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