Document Headings

                    Document headings vary by document type but may contain                    the following:                  
                -                       the agency or agencies that issued and signed a document                    
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###### Nuclear Regulatory Commission

          - [NRC-2015-0041]

# 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 February 5, 2015 to February 18, 2015. The last biweekly notice was published on February 17, 2015.

# DATES:

Comments must be filed by April 2, 2015. A request for a hearing must be filed by May 4, 2015.

# ADDRESSES:

You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):
      -  Federal Rulemaking Web site:                          Go to                          http://www.regulations.gov                          and search for Docket ID NRC-2015-0041. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email:                          Carol.Gallagher@nrc.gov. 
      -  Mail comments to:                          Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.                    

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

# FOR FURTHER INFORMATION CONTACT:

                        Shirley Rohrer, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-5411, email:                         *Shirley.Rohrer@nrc.gov.*

      ( print page 11473)    

# SUPPLEMENTARY INFORMATION:

## I. Obtaining Information and Submitting Comments

### A. Obtaining Information

Please refer to Docket ID NRC-2015-0041 when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:
    -  Federal rulemaking Web site:                      Go to                      http://www.regulations.gov                      and search for Docket ID NRC-2015-0041.                
    -  NRC's Agencywide Documents Access and Management System (ADAMS):                      You may obtain publicly-available documents online in the ADAMS Public Documents collection at                      http://www.nrc.gov/​reading-rm/​adams.html.                      To begin the search, select “                     ADAMS Public Documents”                      and then select “                     Begin Web-based ADAMS Search.”                      For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to                      pdr.resource@nrc.gov.                      The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the                      SUPPLEMENTARY INFORMATION                      section.                
    -  NRC's PDR:                      You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.                

### B. Submitting Comments

Please include Docket ID NRC-2015-0041, facility name, unit number(s), application date, and subject in your comment submission.

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

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

## II. 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                           ( print page 11474)                        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 unless the Commission finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2.

### 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                     *hearing.docket@nrc.gov*, or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.                

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

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

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

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

                    Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 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                           ( print page 11475)                        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.

#### Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2 (ANO-2), Pope County, Arkansas

*Date of amendment request:*                     February 6, 2015. A publicly-available version is in ADAMS under Accession No. ML15041A068.                

*Description of amendment request:*                     The amendment would revise a Note to Technical Specification (TS) Surveillance Requirement (SR) 4.1.3.1.2 to exclude Control Element Assembly (CEA) 18 from being exercised per the SR for the remainder of Cycle 24 due to a degrading upper gripper coil.                

*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.

One function of the CEAs is to provide a means of rapid negative reactivity addition into the core. This occurs upon receipt of a signal from the Reactor Protection System. This function will continue to be accomplished with the approval of the proposed change. Typically, once per 92 days each CEA is moved at least five inches to ensure the CEA is free to move. CEA 18 remains trippable (free to move) as illustrated by the last performance of SR 4.1.3.1.2 in January 2015. However, due to abnormally high coil voltage and current measured on the CEA 18 Upper Gripper Coil (UGC), future exercising of the CEA could result in the CEA inadvertently inserting into the core, if the UGC were to fail during the exercise test. The mis-operation of a CEA, which includes a CEA drop event, is an abnormal occurrence and has been previously evaluated as part of the ANO-2 accident analysis. Inadvertent CEA insertion will result in a reactivity transient and power reduction, and could lead to a reactor shutdown if the CEA is deemed to be unrecoverable. The proposed change would minimize the potential for inadvertent insertion of CEA 18 into the core by maintaining the CEA in place using the Lower Gripper Coil (LGC), which is operating normally. The proposed change will not affect the CEAs ability to insert fully into the core upon receipt of a reactor trip signal.

No modifications are proposed to the Reactor Protection System or associated Control Element Drive Mechanism Control System logic with regard to the ability of CEA 18 to remain available for immediate insertion. The accident mitigation features of the plant are not affected by the proposed amendment. Because CEA 18 remains trippable, no additional reactivity considerations need to be taken into consideration. Nevertheless, Entergy has evaluated the reactivity consequences associated with failure of CEA 18 to insert upon a reactor trip in accordance with TS requirements for Shutdown Margin (SDM) and has determined that SDM requirements would be met should such an event occur at any time during the remainder of Cycle 24 operation.

Therefore, 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.

CEA 18 remains trippable. The proposed change will not introduce any new design changes or systems that can prevent the CEA from [performing] its specified safety function. As discussed previously, CEA mis-operation has been previously evaluated in the ANO-2 accident analysis. Furthermore, SDM has been shown to remain within limits should an event occur at any time during the remainder of operating Cycle 24 such that CEA 18 fails to insert into the core upon receipt of a reactor trip signal.

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

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

Response: No.

                        SR 4.1.3.1.2 is intended to verify CEAs are free to move (                        *i.e.*, not mechanically bound). The physical and electrical design of the CEAs, and past operating experience, provides high confidence that CEAs remain trippable whether or not exercised during each SR interval. Eliminating further exercising of CEA 18 for the remainder of Cycle 24 operation does not directly relate to the potential for CEA binding to occur. No mechanical binding has been previously experienced at ANO-2. CEA 18 is contained within a Shutdown CEA Group and is not used for reactivity control during power maneuvers (the CEA must remain fully withdrawn at all times when the reactor is critical). In addition, Entergy has concluded that required SDM will be maintained should CEA 18 fail to insert following a reactor trip at any point during the remainder of Cycle 24 operation.                    

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

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

*Attorney for licensee:*                     Joseph A. Aluise, Associate General Counsel—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.                

*NRC Acting Branch Chief:*                     Eric R. Oesterle.                

#### Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

*Date of amendment request:*                     October 1, 2014, as supplemented by letter dated February 2, 2015. Publicly-available versions are in ADAMS under Accession Nos. ML14275A374 and ML15033A482.                

*Description of amendment request:*                     The amendment would relocate Technical Specifications 3.9.6, “Refuel Machine,” and 3.9.7, “Crane Travel,” to the Technical Requirements Manual.                

*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                           ( print page 11476)                        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.

This proposed change relocates Technical Specifications (TS) 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane Travel) to the Waterford 3 Technical Requirements Manual (TRM). This is consistent with the requirements of [10 CFR 50.36(c)(2)(ii)] and aligns with NUREG-1432 (Combustion Engineering Standard Technical Specifications).

The applicable TS 3.9.6 and TS 3.9.7 design basis accident is the Fuel Handling Accident (FHA) described in the Updated Final Safety Analysis Report (UFSAR) Section 15.7.3.4. The limiting FHA results in all the fuel pins in the dropped and impacted fuel assemblies failing (472 pins or 236 per assembly). The analysis assumes that a fuel assembly is dropped as an initial condition and no equipment or intervention can prevent the initiating condition. The proposed change was evaluated against [10 CFR 50.36(c)(2)(ii)] criteria and shows no impact to the lowest functional capability or performance levels of equipment required for safe operation of the facility because the TS 3.9.6 and TS 3.9.7 requirements do not prevent the accident conditions from occurring and do not limit the severity of the accident. Since, the dropped fuel assembly and the impacted fuel assembly are both already failed in the design basis accident scenario, this change could not result in a significant increase in the accident consequences. The TS 3.9.6 and TS 3.9.7 equipment are not required to respond, mitigate, or terminate any design basis accident, thus this change will not adversely impact the likelihood or probability of a design basis accident.

The TS 3.9.6 and TS 3.9.7 requirements do not prevent the accident conditions from occurring and do not limit the severity of the accident.

Therefore the TS 3.9.6 and TS 3.9.7 relocation to the TRM would not cause a significant increase in the accident probability or accident consequences.

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

Response: No.

This proposed change relocates TS 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane Travel) to the Waterford 3 TRM. In general, Technical Specifications are based upon the accident analyses. The accident analyses assumptions and initial conditions must be protected by the Technical Specifications. This is a requirement as outlined in [10 CFR 50.36].

[10 CFR 50.36(b)] states the technical specifications will be derived from the analyses and evaluation included in the safety analysis report.

[10 CFR 50.36(c)(2)(i)] states that [“]the limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility[. . . .”] [10 CFR 50.36(c)(2)(ii)] provides the four criteria in which any one met requires a limiting condition for operation. The proposed change demonstrated that the [10 CFR 50.36(c)(2)(ii)] criteria were not met and the relocation to the TRM is allowable. By not meeting the [10 CFR 50.36(c)(2)(ii)] criteria for inclusion into the TS means that TS 3.9.6 and TS 3.9.7 do not impact the accident analyses previously evaluated and would not create the possibility of a new or different kind of accident.

Specifically, TS 3.9.6 and TS 3.9.7 equipment are not instrumentation used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary (Criterion 1). TS 3.9.6 and TS 3.9.7 do not contain a process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier (Criterion 2). TS 3.9.6 and TS 3.9.7 does not contain a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier (Criterion 3). Lastly, TS 3.9.6 and TS 3.9.7 do not contain a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety (Criterion 4).

TS 3.9.6 and 3.9.7 are not required to meet the lowest functional capability or performance levels of equipment required for safe operation of the facility.

Therefore, the accident analyses are not impacted and the possibility of a new or different kind of accident from any accident previously evaluated has not changed.

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

Response: No.

The proposed TS 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane Travel) relocation to the Waterford 3 TRM is administrative in nature because all requirements will be relocated. Any changes after being relocated to the Waterford 3 TRM will require that the [10 CFR 50.59] process be entered ensuring the public health and safety is maintained. By using the [10 CFR 50.59] process for future changes, the regulatory requirements ensure that no significant reduction in the margin of safety occurs.

In addition, the TS 3.9.6 and TS 3.9.7 requirements do not prevent the design basis accident conditions from occurring and do not limit the severity of the accident. Thus, TS 3.9.6 and TS 3.9.7 relocation will not adversely impact the accident analyses and will not cause 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:*                     Joseph A. Aluise, Associate General Council—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.                

*NRC Branch Chief:*                     Meena K. Khanna.                

#### Exelon Generation Company, LLC (EGC), Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

*Date of amendment request:*                     November 17, 2014. A publicly available version is in ADAMS under Accession No. ML14321A744.                

*Description of amendment request:*                     The proposed amendment would revise the NMP2 Technical Specification (TS) Allowable Value for the Main Steam Line Tunnel Lead Enclosure Temperature-High instrumentation from an ambient temperature dependent (variable setpoint) to ambient temperature independent (constant Allowable Value). The changes would delete Surveillance Requirement (SR) 3.3.6.1.2 and revise the Allowable Value for Function 1.g on Table 3.3.6.1-1, “Primary Containment Isolation 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:                
      ( print page 11477)    
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 involve a significant increase in the probability or consequences of an accident previously evaluated because the performance of any equipment credited in the radiological consequences of an accident is not affected by the change in the leak detection capability.

The Main Steam Line Tunnel Lead Enclosure Temperature—High is provided to detect a steam leak in the lead enclosure and provides diversity to the high flow instrumentation. This function provides a mitigating action for a steam leak in the Main Steam Line Tunnel Lead Enclosure, which could lead to a pipe break. This function does not affect any accident precursors, and the proposed changes do not affect the leak detection capability. Additionally, the proposed changes do not degrade the performance of or increase the challenges to any safety systems assumed to function in the accident analysis.

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

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

Response: No.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed changes do not add or remove equipment and do not physically alter the isolation instrumentation. In addition, the Main Steam Line Tunnel Lead Enclosure LDS [Leak Detection System] is not utilized in a different manner. The proposed changes do not introduce any new accident initiators and new failure modes, nor do they reduce or adversely affect the capabilities of any plant structure, system, or component to perform their safety function. The Main Steam Line Tunnel Lead Enclosure LDS will continue to be operated in the same manner.

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

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

Response: No.

The proposed changes do not involve a significant reduction in a margin of safety because the changes eliminate the temperature setpoint dependency on lead enclosure temperature while maintaining the existing upper AV [Allowable Value] = 175.6 °F, that was previously evaluated and approved. There is no adverse impact on the existing equipment capability as well as associated structures. The increase in the steam leak rate and associated crack size continues to be well below the leak rate associated with critical crack size that leads to pipe break. The proposed changes continue to provide the same level of protection against a main steam line break as the existing setpoint values.

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

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

*Attorney for licensee:*                     J. Bradley Fewell, Senior Vice President, Regulatory Affairs, Nuclear, and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.                

*NRC Branch Chief:*                     Benjamin G. Beasley.                

#### Florida Power and Light Company, et al. (FPL), 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, as supplemented by letters dated December 11, 2014, January 13 and January 28, 2015. Publicly-available in ADAMS under Accession Nos. ML14070A087, ML14349A333, ML15029A497 and ML15042A122.                

*Description of amendment request:*                     The NRC staff has previously made a proposed determination that the amendment request dated February 20, 2014, involves no significant hazards consideration (see 79 FR 42550, July 22, 2014). Subsequently, by letter dated January 28, 2015, the licensee provided additional information that expanded the scope of the amendment request as originally noticed. Accordingly, this notice supersedes the previous notice in its entirety.                

                    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 (Revision 1), “Risk-Informed Technical Specifications 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 Decisionmaking: Technical Specifications” (ADAMS Accession No. ML003740176). The licensee stated that the changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler 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. In the supplement dated January 28, 2015, the licensee requested (1) additional surveillance frequencies be relocated to the licensee-controlled program, (2) editorial changes, (3) administrative deviations from TSTF-425, and (4) other changes resulting from differences between the St. Lucie Plant TSs and the TSs on which TSTF-425 was based.                

*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 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 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                               ( print page 11478)                            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 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 change 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 has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

*NRC Branch Chief:*                     Shana R. Helton.                

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

*Date of amendment request:*                     November 13, 2014. A publicly-available version is in ADAMS under Accession No. ML14337A013.                

*Description of amendment request:*                     The amendment would revise Technical Specification (TS) 3/4.5.2, “ECCS [Emergency Core Cooling System] Subsystems—Tavg                     [average temperature] Greater Than or Equal to 350 °F [degrees Fahrenheit],” to correct non-conservative TS requirements. The licensee also requested editorial changes to the TS.                

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

No. The proposed TS changes involve TS 3.5.2 Action `a', new TS 3.5.2 Action `h', and the provision in SR [Surveillance Requirement] 4.5.2.a to address non-conservative TS requirements. Editorial changes are also proposed for consistency and clarity. These changes do not affect any precursors to any accident previously evaluated and subsequently, will not impact the probability or consequences of an accident previously evaluated. Furthermore, these changes do not adversely affect mitigation equipment or strategies.

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

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

No. The proposed TS changes involve TS 3.5.2 Action `a', new TS 3.5.2 Action `h', and the provision in SR 4.5.2.a to address non-conservative TS requirements. Editorial changes are also proposed for consistency and clarity. The proposed changes provide better assurance that the ECCS systems, subsystems, and components are properly aligned to support safe reactor operation consistent with the licensing basis requirements. The proposed changes do not introduce new modes of plant operation and do not involve physical modifications to the plant (no new or different type of equipment will be installed). There are no changes in the method by which any safety related plant structure, system, or component (SSC) performs its specified safety function. As such, the plant conditions for which the design basis accident analyses were performed remain valid. 

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of the proposed change. There will be no adverse effect or challenges imposed on any SSC as a result of the proposed change. 

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

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

No. Margin of safety is related to confidence in the ability of the fission product barriers to perform their accident mitigation functions. The proposed TS changes involve TS 3.5.2 Action `a', new TS 3.5.2 Action `h', and the provision in SR 4.5.2.a to address non-conservative TS requirements. Editorial changes are also proposed for consistency and clarity. The proposed changes provide better assurance that the ECCS systems, subsystems, and components are properly aligned to support safe reactor operation consistent with the licensing basis requirements. The proposed changes do not physically alter any SSC. There will be no effect on those SSCs necessary to assure the accomplishment of specified functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, loss of cooling accident peak cladding temperature (LOCA PCT), or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. 

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

*Attorney for licensee:*                     William S. Blair, Managing Attorney—Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.                 

*NRC Branch Chief:*                     Shana R. Helton.                 

#### Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan 

*Date of amendment request:*                     February 6, 2015. A publicly-available version is in ADAMS under Accession No. ML15041A069.                 

*Description of amendment request:*                     The proposed amendments would modify the technical specifications requirements for unavailable barriers by adding limiting condition for operation                           ( print page 11479)                        (LCO) 3.0.8. The changes are consistent with the NRC approved Technical Specification Task Force (TSTF) Standard Technical Specification change TSTF-427, “Allowance for Non-Technical Specification Barrier Degradation on Supported System OPERABILITY,” Revision 2.                 

*Basis for proposed no significant hazards consideration determination:*                     As required by 10 CFR 50.91(a), the licensee has affirmed the applicability of the model proposed no significant hazards consideration published on October 3, 2006 (71 FR 58444), “Notice of Availability of the Model Safety Evaluation.” The findings presented in that evaluation are presented below:                 

#### Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated 

The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 

#### Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated 

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to an unavailable barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated. 

#### Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety 

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG [Regulatory Guide] 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.8 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP [incremental conditional core damage probability] and ICLERP [incremental large early release probability]) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation. Therefore, this change does not involve a significant reduction in a margin of safety. 

The NRC staff has reviewed the 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 requests involve no significant hazards consideration. 

*Attorney for licensee:*                         Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, Michigan 49106.                     

*NRC Branch Chief:*                         David L. Pelton.                     

#### PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania 

*Date of amendment requests:*                         October 27, 2014. A publicly-available version is available in ADAMS under Accession No. ML14317A052.                     

*Description of amendment requests:*                         The proposed amendments will modify the Susquehanna technical specifications (TS). Specifically, the proposed amendments will modify the TS by relocating specific surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), with implementation of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies” (ADAMS Accession No. ML071360456). The changes are consistent with NRC-approved TS Task Force (TSTF) Standard TS change TSTF-425, “Relocate Surveillance Frequencies to Licensee Control-Risk Informed Technical Specifications Task Force (RITSTF) Initiative 5b,” Revision 3 (ADAMS Accession No. ML090850642). The                         **Federal Register**                         notice published on July 6, 2009 (74 FR 31996), announced the availability of this TSTF improvement, and included a model no significant hazards consideration and safety evaluation.                     

*Basis for proposed no significant hazards consideration determination:*                         An analysis of the no significant hazards consideration was presented in the TSTF-425. The licensee has affirmed its applicability of the model no significant hazards consideration, which is presented below:                     

1. Does the proposed change involve a significant increase in the probability or consequences of any 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 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. Does the proposed change 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. 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 accident previously evaluated.                               ( print page 11480)    

3. Does the proposed change 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, PPL will perform a risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-10, Rev. 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, the proposed changes do not involve a significant reduction in a margin of safety.

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

*Attorney for licensee:*                     Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.                 

*NRC Branch Chief:*                     Douglas A. Broaddus.                 

#### Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia 

*Date of amendment request:*                     July 18, 2014. A publicly-available version is in ADAMS under Accession Package No. ML14203A124.                 

*Description of amendment request:*                     The licensee requested 23 revisions to the Technical Specifications (TSs). These revisions adopt various previously NRC-approved Technical Specifications Task Force (TSTF) Travelers. A list of the requested revisions is included in Enclosure 1 of the application.                 

*Basis for proposed no significant hazards consideration determination:*                     As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration for each of the 24 changes requested, which is presented below:                 

#### 1: TSTF-2-A, Revision 1, “Relocate the 10 Year Sediment Cleaning of the Fuel Oil Storage Tank to Licensee Control” for TS pages 3.8.3-3 and 3.8.3-4 

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

Response: No. 

The proposed change removes the Surveillance Requirement for performing sediment cleaning of diesel fuel oil storage tanks every 10 years from the Technical Specifications and places it under licensee control. Diesel fuel oil storage tank cleaning is not an initiator of any accident previously evaluated. This change will have no effect on diesel generator fuel oil quality, which is tested in accordance with other Technical Specifications requirements. Removing the diesel fuel oil storage tank sediment cleaning requirements from the Technical Specifications will have no effect on the ability to mitigate an accident. 

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

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

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

Response: No. 

The proposed change removes the requirement to clean sediment from the diesel fuel oil storage tank from the Technical Specifications and places it under licensee control. The margin of safety provided by the fuel oil storage tank sediment cleaning is unaffected by this relocation because the quality of diesel fuel oil is tested in accordance with other Technical Specifications requirements. 

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 2: TSTF-27-A, Revision 3, “Revise SR [Surveillance Requirement] Frequency for Minimum Temperature for Criticality” for TS 3.4.2, TS Page 3.4.2-1 

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

Response: No. 

                        The proposed change revises the Surveillance Frequency for monitoring [reactor coolant system] RCS temperature to ensure the minimum temperature for criticality is met. The Frequency is changed from a 30 minute Frequency when certain conditions are met to a periodic Frequency that it is controlled in accordance with the Surveillance Frequency Control Program. The initial Frequency for this Surveillance will be 12 hours. This will ensure that Tavg                         [average temperature] is logged at appropriate intervals (in addition to strip chart recorders and computer logging of temperature). The measurement of RCS temperature is not an initiator of any accident previously evaluated. The minimum RCS temperature for criticality is not changed. As a result, the mitigation of any accident previously evaluated is not affected.                     

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

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

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

Response: No. 

The proposed change revises the Surveillance Frequency for monitoring RCS temperature to ensure the minimum temperature for criticality is met. The current, condition based Frequency represents a distraction to the control room operator during the critical period of plant startup. RCS temperature is closely monitored by the operator during the approach to criticality, and temperature is recorded on charts and computer logs. Allowing the operator to monitor temperature as needed by the situation and logging RCS temperature at a periodic Frequency that it is controlled in accordance with the Surveillance Frequency Control Program is sufficient to ensure that the LCO [Limiting Condition for Operation] is met while eliminating a diversion of the operator's attention. 

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

                        Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.                               ( print page 11481)    

#### 3: TSTF-28-A, Revision 0, “Delete Unnecessary Action to Measure Gross Specific Activity, TS 3.4.16,” TS page 3.4-16 

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

Response: No. 

The proposed change eliminates Required Action B.1 of Specification 3.4.16, “RCS Specific Activity,” which requires verifying that Dose Equivalent I-131 specific activity is within limits. Determination of Dose Equivalent I-131 is not an initiator of any accident previously evaluated. Determination of Dose Equivalent I-131 has no effect on the mitigation of any accident previously evaluated. 

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

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

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

Response: No. 

The proposed change eliminates a Required Action. The activities performed under the Required Action will still be performed to determine if the LCO is met or the plant will exit the Applicability of the Specification. In either case, the presence of the Required Action does not provide any significant margin of safety. 

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 4: TSTF-45-A, Revision 2, “Exempt Verification of CIVs that are Locked, Sealed or Otherwise Secured,” TS 3.6.3, TS pages 3.6.3-4, 3.6.3-5 

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

Response: No. 

The proposed change exempts containment isolation valves (CIVs) located inside and outside of containment that are locked, sealed, or otherwise secured in position from the periodic verification of valve position required by Surveillance Requirements 3.6.3.3 and 3.6.2.4. The exempted valves are verified to be in the correct position upon being locked, sealed, or secured. Because the valves are in the condition assumed in the accident analysis, the proposed change will not affect the initiators or mitigation of any accident previously evaluated. 

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

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

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

Response: No. 

The proposed change replaces the periodic verification of valve position with verification of valve position followed by locking, sealing, or otherwise securing the valve in position. Periodic verification is also effective in detecting valve mispositioning. However, verification followed by securing the valve in position is effective in preventing valve mispositioning. Therefore, the proposed change does not involve a significant reduction in a margin of safety. 

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 5: TSTF-46-A, Revision 1, “Clarify the CIV Surveillance to Apply Only to Automatic Isolation Valves,” TS 3.6.3, TS page 3.6.3.5 

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

Response: No. 

The proposed change revises the requirements in Technical Specification SR 3.6.3.5, and the associated Bases, to delete the requirement to verify the isolation time of “each power operated” containment isolation valve (CIV) and only require verification of closure time for each “automatic power operated isolation valve.” The closure times for CIVs that do not receive an automatic closure signal are not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The CIVs are used to respond to accidents previously evaluated. Power operated CIVs that do not receive an automatic closure signal are not assumed to close in a specified time. The proposed change does not change how the plant would mitigate an accident previously evaluated. 

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

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

Response: No. 

                        The proposed change does not result in a change in the manner in which the CIVs provide plant protection or introduce any new or different operational conditions. Periodic verification that the closure times for CIVs that receive an automatic closure signal are within the limits established by the accident analysis will continue to be performed under SR 3.6.3.5. The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. There are also no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (                        *i.e.,*                         no new or different type of equipment will be installed).                     

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

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

Response: No. 

The proposed change provides clarification that only CIVs that receive an automatic isolation signal are within the scope of the SR 3.6.3.5. The proposed change does not result in a change in the manner in which the CIVs provide plant protection. Periodic verification that closure times for CIVs that receive an automatic isolation signal are within the limits established by the accident analysis will continue to be performed. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis. 

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 6: TSTF-87-A, Revision 2, “Revise `RTBs [Reactor Trip Breaker] Open' and `CRDM [Control Rod Drive Mechanism] De-energized' Actions to `Incapable of Rod Withdrawal,'” TS 3.4.5, TS Pages 3.4.5-2, 3.4.9-1 

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

                        Response: No.                               ( print page 11482)    

This change revises the Required Actions for LCO 3.4.5, “RCS Loops—Mode 3,” Conditions C.2 and D.1, from “De-energize all control rod drive mechanisms,” to “Place the Rod Control System in a condition incapable of rod withdrawal.” It also revises LCO 3.4.9, “Pressurizer,” Required Action A.1, from requiring Reactor Trip Breakers to be open after reaching MODE 3 to “Place the Rod Control System in a condition incapable of rod withdrawal,” and to require full insertion of all rods. Inadvertent rod withdrawal can be an initiator for design basis accidents or events during certain plant conditions, and therefore must be prevented under those conditions. The proposed Required Actions for LCO 3.4.5 and LCO 3.4.9 satisfy the same intent as the current Required Actions, which is to prevent inadvertent rod withdrawal when an applicable Condition is not met, and is consistent with the assumptions of the accident analysis. As a result, the proposed change does not increase the probability of any accident previously evaluated. The proposed change does not change how the plant would mitigate an accident previously evaluated, as in both the current and proposed requirements, rod withdrawal is prohibited. 

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

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

Response: No. 

                        The proposed change provides less specific, but equivalent, direction on the manner in which inadvertent control rod withdrawal is to be prevented when the Conditions of LCO 3.4.5 and LCO 3.4.9 are not met. Rod withdrawal will continue to be prevented when the applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (                        *i.e.,*                         no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis.                     

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 amendment involve a significant reduction in a margin of safety? 

Response: No. 

The proposed change provides the operational flexibility of allowing alternate, but equivalent, methods of preventing rod withdrawal when the applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis. 

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 7: TSTF-95-A, Revision 0, “Revise Completion Time for Reducing Power Range High trip Setpoint from 8 to 72 Hours,” TS 3.2.1, TS Pages 3.2.1-1 and 3.2.2-1 

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

Response: No. 

The proposed change extends the time allowed to reduce the Power Range Neutron Flux—High trip setpoint when Specification 3.2.1, “Heat Flux Hot Channel Factor,” or Specification 3.2.2, “Nuclear Enthalpy Rise Hot Channel Factor,” are not within their limits. Both specifications require a power reduction followed by a reduction in the Power Range Neutron Flux—High trip setpoint. Because reactor power has been reduced, the reactor core power distribution limits are within the assumptions of the accident analysis. Reducing the Power Range Neutron Flux—High trip setpoints ensures that reactor power is not inadvertently increased. Reducing the Power Range Neutron Flux—High trip setpoints is not an initiator to any accident previously evaluated. The consequences of any accident previously evaluated with the Power Range Neutron Flux—High trip setpoints not reduced are no different under the proposed Completion Time than under the existing Completion Time. 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 change does not involve a physical alteration to the plant (                        *i.e.,*                         no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.                     

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

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

Response: No. 

The proposed change provides additional time before requiring the Power Range Neutron Flux—High trip setpoint be reduced when the reactor core power distribution limits are not met. The manual reduction in reactor power required by the specifications provides the necessary margin of safety for this condition. Reducing the Power Range Neutron Flux—High trip setpoints carries an increased risk of a reactor trip. Delaying the trip setpoint reduction until the power reduction has been completed and the condition is verified will minimize overall plant risk. 

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 8: TSTF-110-A, Revision 2, “Delete SR Frequencies Based on Inoperable Alarms,” TS 3.1, TS pages 3.1.4-3, 3.1.6-3, 3.2.3-1, 3.2.4-4 

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

Response: No. 

The proposed change removes surveillance Frequencies associated with inoperable alarms (rod position deviation monitor, rod insertion limit monitor, AFD [Axial Flux Difference] monitor and QPTR [Quadrant Power Tilt Ratio] alarm) from the Technical Specifications and places the actions in plant administrative procedures. The subject plant alarms are not an initiator of any accident previously evaluated. The subject plant alarms are not used to mitigate any accident previously evaluated, as the control room indications of these parameters are sufficient to alert the operator of an abnormal condition without the alarms. The alarms are not credited in the accident analysis. 

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

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

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

Response: No. 

                        The proposed change removes surveillance Frequencies associated with inoperable alarms (rod position deviation monitor, rod insertion limit monitor, AFD monitor and QPTR alarm) from the Technical Specifications and places the actions in plant administrative procedures. The alarms are not being removed from the plant. The actions to be taken when the alarms are not available are proposed to be controlled under licensee administrative procedures. As a result, plant operation is unaffected by this change and there is no effect on a margin of safety.                               ( print page 11483)    

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 9: TSTF-142-A, Revision 0, “Increase the Completion Time When the Core Reactivity Balance is Not Within Limit,” TS 3.1.2, TS Page 3.1.2-1 

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

Response: No. 

The proposed change extends the Completion Time to take the Required Actions when measured core reactivity is not within the specified limit of the predicted values. The Completion Time to respond to a difference between predicted and measured core reactivity is not an initiator to any accident previously evaluated. The consequences of an accident during the proposed Completion Time are no different from the consequences of an accident during the existing Completion Time. 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 change does not involve a physical alteration to the plant (                        *i.e.,*                         no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.                     

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

Response: No. 

The proposed change provides additional time to investigate and to implement appropriate operating restrictions when measured core reactivity is not within the specified limit of the predicted values. The additional time will not have a significant effect on plant safety due to the conservatisms used in designing the reactor core and performing the safety analyses and the low probability of an accident or transient which would approach the core design limits during the additional time. Therefore, the proposed change does not involve a significant reduction in a margin of safety. 

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 10: TSTF-234-A, Revision 1, “Add Action for More Than One [D]RPI Inoperable,” TS 3.1.7, TS Pages 3.1.7-1 and 3.1.7-2. 

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

Response: No. 

The proposed change provides a Condition and Required Actions for more than one inoperable digital rod position indicator (DRPI) per rod group. The DRPIs are not an initiator of any accident previously evaluated. The DRPIs are one indication used by operators to verify control rod insertion following an accident, however other indications are available. Therefore, allowing a finite period to time to correct more than one inoperable DRPI prior to requiring a plant shutdown will not result in a significant increase in the consequences of any accident previously evaluated. 

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

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

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

Response: No. 

The proposed change provides time to correct the condition of more than one DRPI inoperable in a rod group. Compensatory measures are required to verify that the rods monitored by the inoperable DRPIs are not moved to ensure that there is no effect on core reactivity. Requiring a plant shutdown with inoperable rod position indications introduces plant risk and should not be initiated unless the rod position indication cannot be repaired in a reasonable period of time. As a result, the safety benefit provided by the proposed Condition offsets the small decrease in safety resulting from continued operation with more than one inoperable DRPI. 

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified. 

#### 11: TSTF-245-A, Revision 1, “AFW Train Operable When in Service,” TS 3.7.5, TS Page 3.7.5-3 

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

Response: No. 

The proposed change revises the requirements in Technical Specification 3.7.5, “Auxiliary Feedwater (AFW) System,” to clarify the operability of an AFW train when it is aligned for manual steam generator level control. The AFW System is not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The AFW System is used to respond to accidents previously evaluated. The proposed change does not affect the design of the AFW System, and no physical changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

                        The proposed change does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (                        *i.e.,*                         no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Manual control of AFW level control valves is not an accident initiator.                    

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 amendment involve a significant reduction in a margin of safety?

Responses: No.

                        The proposed change provides the operational flexibility of allowing an AFW train(s) to be considered operable when it is not in the normal standby alignment and is temporarily incapable of automatic initiation, such as during alignment and operation for manual steam generator level control, provided it is capable of being manually realigned to the AFW heat removal mode of operation. The proposed change does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings                               ( print page 11484)                            or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.                    

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 12: TSTF-247-A, Revision 0, “Provide Separate Condition Entry for Each [Power Operated Relief Valve] PORV and Block Valve,” TS 3.4.11, TS Pages 3.4.11-1, 3.4.11-2, 3.4.11-3

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

Response: No.

The proposed change revises the requirements in Technical Specification 3.4.11, “Pressurizer PORVs,” to clarify that separate Condition entry is allowed for each block valve. Additionally, the Actions are modified to no longer require that the PORVs be placed in manual operation when both block valves are inoperable and cannot be restored to operable status within the specified Completion Time. This preserves the overpressure protection capabilities of the PORVs. The pressurizer block valves are used to isolate their respective PORV in the event it is experiencing excessive leakage, and are not an initiator of any design basis accident or event. Therefore the proposed change does not increase the probability of any accident previously evaluated. The PORV and block valves are used to respond to accidents previously evaluated. The proposed change does not affect the design of the PORV and block valves, and no physical changes are made to the plant. The proposed change does not change how the plant would mitigate an accident previously evaluated.

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

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

Response: No.

                        The proposed change does not result in a change in the manner in which the PORV and block valves provide plant protection. The PORVs will continue to provide overpressure protection, and the block valves will continue to provide isolation capability in the event a PORV is experiencing excessive leakage. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (                        *i.e.,*                         no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Operation of the PORV block valves is not an accident initiator.                    

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes provide clarification that separate Condition entry is allowed for each block valve. Additionally, the Actions are modified to no longer require that the PORVs be placed in manual operation when both block valves are inoperable and cannot be restored to operable status within the specified Completion Time. This preserves the overpressure protection capabilities of the PORVs. The proposed change does not result in a change in the manner in which the PORV and block valves provide plant protection. The PORVs will continue to provide overpressure protection, and the block valves will continue to provide isolation capability in the event a PORV is experiencing excessive leakage. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 13: TSTF-248-A, Revision 0, “Revise Shutdown Margin Definition for Stuck Rod Exception,” TS 1.1, TS Page 1.1-6

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

Response: No.

The proposed change modifies the definition of Shutdown Margin to eliminate the requirement to assume the highest worth control rod is fully withdrawn when calculating Shutdown Margin if it can be verified by two independent means that all control rods are inserted. The method for calculating shutdown margin is not an initiator of any accident previously evaluated. If it can be verified by two independent means that all control rods are inserted, the calculated Shutdown Margin without the conservatism of assuming the highest worth control rod is withdrawn is accurate and consistent with the assumptions in the accident analysis. As a result, the mitigation of any accident previously evaluated is not affected.

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

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

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

Response: No.

The proposed change modifies the definition of Shutdown Margin to eliminate the requirement to assume the highest worth control rod is fully withdrawn when calculating Shutdown Margin if it can be verified by two independent means that all control rods are inserted. The additional margin of safety provided by the assumption that the highest worth control rod is fully withdrawn is unnecessary if it can be independently verified that all controls rods are inserted.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 14: TSTF-266-A, Revision 3, “Eliminate the Remote Shutdown System Table of Instrumentation and Controls,” TS 3.3.4, TS Pages 3.3.4-1, 3.3.4-3

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

Response: No.

The proposed change removes the list of Remote Shutdown System instrumentation and controls from the Technical Specifications and places them in the Bases. The Technical Specifications continue to require that the instrumentation and controls be operable. The location of the list of Remote Shutdown System instrumentation and controls is not an initiator to any accident previously evaluated. The proposed change will have no effect on the mitigation of any accident previously evaluated because the instrumentation and controls continue to be required to be operable.

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.                              ( print page 11485)    

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

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

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

Response: No.

The proposed change removes the list of Remote Shutdown System instrumentation and controls from the Technical Specifications and places it in the Bases. The review performed by the NRC when the list of Remote Shutdown System instrumentation and controls is revised will no longer be needed unless the criteria in 10 CFR 50.59 are not met such that prior NRC review is required. The Technical Specification requirement that the Remote Shutdown System be operable, the definition of operability, the requirements of 10 CFR 50.59, and the Technical Specifications Bases Control Program are sufficient to ensure that revision of the list without prior NRC review and approval does not introduce a significant safety risk.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 15: TSTF-272-A, Revision 1, “Refueling Boron Concentration Clarification,” TS 3.9.1, TS Page 3.9.1-1

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

Response: No.

The proposed change modifies the Applicability of Specification 3.9.1, “Boron Concentration,” to clarify that the boron concentration limits are only applicable to the refueling canal and the refueling cavity when those volumes are attached to the Reactor Coolant System (RCS). The boron concentration of water volumes not connected to the RCS are not an initiator of an accident previously evaluated. The ability to mitigate any accident previously evaluated is not affected by the boron concentration of water volumes not connected to the RCS.

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

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

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

Response: No.

The proposed change modifies the Applicability of Specification 3.9.1, “Boron Concentration,” to clarify that the boron concentration limits are only applicable to the refueling canal and the refueling cavity when those volumes are attached to the RCS. Technical Specification SR 3.0.4 requires that Surveillances be met prior to entering the Applicability of a Specification. As a result, the boron concentration of the refueling cavity or the refueling canal must be verified to satisfy the LCO prior to connecting those volumes to the RCS. The margin of safety provided by the refueling boron concentration is not affected by this change as the RCS boron concentration will continue to satisfy the LCO.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 16: TSTF-273-A, Revision 2, “Safety Function Determination Program Clarifications,” TS 5.5.15, TS Page 5.5-15

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

Response: No.

The proposed TS changes add explanatory text to the programmatic description of the Safety Function Determination Program (SFDP) in Specification 5.5.15 to clarify in the requirements that consideration does not have to be made for a loss of power in determining loss of function. The Bases for LCO 3.0.6 is revised to provide clarification of the “appropriate LCO for loss of function,” and that consideration does not have to be made for a loss of power in determining loss of function. The changes are editorial and administrative in nature, and therefore do not increase the probability of any accident previously evaluated. No physical or operational changes are made to the plant. The proposed change does not change how the plant would mitigate an accident previously evaluated.

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

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

Response: No.

                        The proposed changes are editorial and administrative in nature and do not result in a change in the manner in which the plant operates. The loss of function of any specific component will continue to be addressed in its specific TS LCO and plant configuration will be governed by the required actions of those LCOs. The proposed changes are clarifications that do not degrade the availability or capability of safety related equipment, and therefore do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no design changes associated with the proposed changes, and the changes do not involve a physical alteration of the plant (                        *i.e.,*                         no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analysis, and are consistent with the safety analysis assumptions and current plant operating practice. Due to the administrative nature of the changes, they cannot be an accident initiator.                    

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes to TS 5.5.15 are clarifications and are editorial and administrative in nature. No changes are made the LCOs for plant equipment, the time required for the TS Required Actions to be completed, or the out of service time for the components involved. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. 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 a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 17: TSTF-284-A, Revision 3, “Add `Met vs. Perform' to Technical Specification 1.4, Frequency,” TS 1.4, TS 3.4, TS 3.9, TS Pages 1.4-1, 1.4-4, 3.4.11-3, 3.4.12-4 and 3.9.4-2

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

Response: No.

                        The proposed changes insert a discussion paragraph into Specification 1.4, and several new examples are added to facilitate the use and application of SR Notes that utilize the terms “met” and “perform.” The changes also modify SRs in multiple Specifications to appropriately use “met” and “perform” exceptions. The changes are administrative in nature because they provide clarification and correction of existing expectations, and therefore the proposed change does not increase the probability of any accident previously evaluated. No physical or                               ( print page 11486)                            operational changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated.                    

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

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

Response: No.

                        The proposed changes are administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes provide clarification and correction of existing expectations that do not degrade the availability or capability of safety related equipment, and therefore do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no design changes associated with the proposed changes, and the changes do not involve a physical alteration of the plant (                        *i.e.,*                         no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analysis, and are consistent with the safety analysis assumptions and current plant operating practice. Due to the administrative nature of the changes, they cannot be an accident initiator.                    

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes are administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes provide clarification and correction of existing expectations that do not degrade the availability or capability of safety related equipment, or alter their operation. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. 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 a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 18: TSTF-308-A, Revision 1, “Determination of Cumulative and Projected Dose Contributions in RECP [Radioactive Effluent Controls Program],” TS 5.5.4, TS Page 5.5-3

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

Response: No.

The proposed change revises Specification 5.5.4, “Radioactive Effluent Controls Program,” paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are not an assumption in any accident previously evaluated and have no effect on the mitigation of any accident previously evaluated.

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

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

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

Response: No.

The proposed change revises Specification 5.5.4, “Radioactive Effluent Controls Program,” paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are administrative tools to assure compliance with regulatory limits. The proposed change revises the requirement to clarify the intent, thereby improving the administrative control over this process. As a result, any effect on the margin of safety should be minimal.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 19: TSTF-312-A, Revision 1, “Administrative Control of Containment Penetrations,” TS 3.9.4, TS Page 3.9.4-1

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

Response: No.

                        The proposed change would allow containment penetrations to be unisolated under administrative controls during core alterations or movement of irradiated fuel assemblies within containment. The status of containment penetration flow paths (                        *i.e.,*                         open or closed) is not an initiator for any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The proposed change does not affect the design of the primary containment, or alter plant operating practices such that the probability of an accident previously evaluated would be significantly increased. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated, and is bounded by the fuel handling accident (FHA) accident analysis.                    

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.

                        Allowing penetration flow paths to be open is not an initiator for any accident. The proposed change to allow open penetration flow paths will not affect plant safety functions or plant operating practices such that a new or different accident could be created. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (                        *i.e.,*                         no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice.                    

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 amendment involve a significant reduction in a margin of safety?

Response: No.

TS 3.9.4 provides measures to ensure that the dose consequences of a postulated FHA inside containment are minimized. The proposed change to LCO 3.9.4 will allow penetration flow path(s) to be open during refueling operations under administrative control. These administrative controls will can and will be achieved in the event of an FHA inside containment, and will minimize dose consequences. The proposed change is bounded by the existing FHA analysis. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.

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

                        Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.                              ( print page 11487)    

####                         20: TSTF-314-A, Revision 0, “Require Static and Transient FQ                         Measurement,” TS 3.1.4, 3.2.4, TS Pages 3.1.4-2, 3.2.4-1, 3.2.4-3                    

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

Response: No.

The proposed change revises the Required Actions of Specification 3.1.4, “Rod Group Alignment Limits,” and Specification 3.2.4, “Quadrant Power Tilt Ratio,” to require measurement of both the steady state and transient portions of the Heat Flux Hot Channel Factor, FQ(Z). This change will ensure that the hot channel factors are within their limits when the rod alignment limits or quadrant power tilt ratio are not within their limits. The verification of hot channel factors is not an initiator of any accident previously evaluated. The verification that both the steady state and transient portion of FQ(Z) are within their limits will ensure this initial assumption of the accident analysis is met should a previously evaluated accident occur.

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

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

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

Response: No.

                        The proposed change revises the Required Actions in the Specifications for Rod Group Alignment Limits and Quadrant Power Tilt Ratio to require measurement of both the steady state and transient portions of the Heat Flux Hot Channel Factor, FQ                        (Z). This change is a correction that ensures that the plant conditions are as assumed in the accident analysis.                    

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 21: TSTF-340-A, Revision 3, “Allow 7 Day Completion Time for a Turbine—Driven AFW Pump Inoperable,” TS 3.7.5, TS Page 3.7.5-1

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

Response: No.

The proposed change revises Specification 3.7.5, “Auxiliary Feedwater (AFW) System,” to allow a 7 day Completion Time to restore an inoperable AFW turbine-driven pump in Mode 3 immediately following a refueling outage, if Mode 2 has not been entered. An inoperable AFW turbine-driven pump is not an initiator of any accident previously evaluated. The ability of the plant to mitigate an accident is no different while in the extended Completion Time than during the existing Completion Time.

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

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

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

Response: No.

The proposed change revises Specification 3.7.5, “Auxiliary Feedwater (AFW) System,” to allow a 7-day Completion Time to restore an inoperable turbine-driven AFW pump in Mode 3 immediately following a refueling outage if Mode 2 has not been entered. In Mode 3 immediately following a refueling outage, core decay heat is low and the need for AFW is also diminished. The two operable motor driven AFW pumps are available and there are alternate means of decay heat removal if needed. As a result, the risk presented by the extended Completion Time is minimal.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 22: TSTF-343-A, Revision 1, “Containment Structural Integrity,” TS 5.5, TS Page 5.5-16

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

Response: No.

The proposed change revises the Technical Specifications (TS) Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed changes affect the frequency of visual examinations that will be performed for the steel containment liner plate for the purpose of the Containment Leakage Rate Testing Program.

The frequency of visual examinations of the containment and the mode of operation during which those examinations are performed does not affect the initiation of any accident previously evaluated. The use of NRC approved methods and frequencies for performing the inspections will ensure the containment continues to perform the mitigating function assumed for accidents previously evaluated.

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

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

Response: No.

The proposed change revises the TS Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed change affects the frequency of visual examinations that will be performed for the steel containment liner plate for the purpose of the Containment Leakage Rate Testing Program.

                        The proposed changes do not involve a modification to the physical configuration of the plant (                        *i.e.,*                         no new equipment will be installed) or change in the methods governing normal plant operation. The proposed changes will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure.                    

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

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

Response: No.

The proposed changes revise the Technical Specifications (TS) Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed change affects the frequency of visual examinations that will be performed for the steel containment liner plate for the purpose of the Containment Leakage Rate Testing Program. The safety function of the containment as a fission product barrier will be maintained.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

#### 23: TSTF-349-A, Revision 1, “Add Note to LCO 3.9.5 Allowing Shutdown Cooling Loops Removal From Operation,” TS 3.9.6, TS Page 3.9.6-1

                        1. Does the proposed amendment involve a significant increase in the probability or                               ( print page 11488)                            consequences of an accident previously evaluated?                    

Response: No.

The proposed change adds an LCO Note to LCO 3.9.6, “RHR and Coolant Circulation—Low Water Level,” to allow securing the operating train of Residual Heat Removal (RHR) for up to 15 minutes to support switching operating trains. The allowance is restricted to conditions in which core outlet temperature is maintained at least 10 degrees F below the saturation temperature, when there are no draining operations, and when operations that could reduce the reactor coolant system (RCS) boron concentration are prohibited. Securing an RHR train to facilitate the changing of the operating train is not an initiator to any accident previously evaluated. The restrictions on the use of the allowance ensure that an RHR train will not be needed during the 15 minute period to mitigate any accident previously evaluated.

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

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

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

Response: No.

The proposed change adds an LCO Note to LCO 3.9.6, “RHR and Coolant Circulation—Low Water Level,” to allow securing the operating train of RHR to support switching operating trains. The allowance is restricted to conditions in which core outlet temperature is maintained at least 10 degrees F below the saturation temperature, when there are no draining operations, and when operations that could reduce the reactor coolant system (RCS) boron concentration are prohibited. With these restrictions, combined with the short time frame allowed to swap operating RHR trains and the ability to start an operating RHR train if needed, the occurrence of an event that would require immediate operation of an RHR train is extremely remote.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(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 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.                

#### Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power, Unit Nos. 1 and 2, Louisa County, Virginia

*Date of amendment request:*                     February 4, 2015. A publicly-available version is in ADAMS under Accession No. ML15041A667.                

*Description of amendment request:*                     The proposed license amendment requests the changes to the Technical Specification (TS) TS 3.1.7, Rod Position Indication, to provide an additional monitoring option for an inoperable control rod position indicator. Specifically, the proposed changes would allow monitoring of control rod drive mechanism stationary gripper coil voltage every eight hours as an alternative to using the movable in core detectors every eight hours to verify control rod position.                

*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 provides an alternative method for verifying rod position of one rod. The proposed change meets the intent of the current specification in that it ensures verification of position of the rod once every 8 hours. The proposed change provides only an alternative method of monitoring rod position and does not change the assumptions or results of any previously evaluated accident.

Therefore, operation of the facility in accordance with the proposed amendment would 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 provides only an alternative method of determining the position of one rod. No new accident initiators are introduced by the proposed alternative manner of performing rod position verification. The proposed change does not affect the reactor protection system. Hence, no new failure modes are created that would cause a new or different kind of accidents from any accident previously evaluated.

Therefore, operation of the facility in accordance with the proposed amendments 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 basis of TS 3.1.7 states that the operability of the rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. The proposed change does not alter the requirement to determine rod position but provides an alternative method for determining the position of the affected rod. As a result, the initial conditions of the accident analysis are preserved and the consequences of previously analyzed accidents are unaffected.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

Based on the above, Dominion concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(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 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.                

## 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                           ( print page 11489)                        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.

#### Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power Station, Kewaunee County, Wisconsin

*Date of application for amendment:*                     May 29, 2013, as supplemented by letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and January 7, March 13, April 29, and October 6, 2014, and January 15, 2015.                

*Brief description of amendment:*                     The amendment revised the Renewed Facility Operating License and associated Technical Specifications to conform to the permanent shutdown and defueled status of the facility. It also denied a proposal to delete paragraphs 1.B, 1.I, and 1.J of the Kewaunee Operating License.                

*Date of issuance:*                     February 13, 2015.                

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

*Amendment No.:*                     215. A publicly-available version is in ADAMS under Accession No. ML14237A045; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Renewed Facility Operating License No. DPR-43:*                     The amendment revised the renewed facility operating license and Technical Specifications.                

*Date of initial notice in**Federal Register**:*                     August 20, 2013 (78 FR 51224). The supplemental letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and January 7, March 13, April 29, and October 6, 2014, and January 15, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the                     **Federal Register**                    .                

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 13, 2015.

No significant hazards consideration comments received: No.

#### Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

#### Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370 McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

#### Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

*Date of application for amendments:*                     July 21, 2014.                

*Brief description of amendments:*                     The amendment revises the licensed operator training requirements to be consistent with the National Academy for Nuclear Training (NANT) program. Additionally, the amendment makes administrative changes to Technical Specification Sections 5.1, “Responsibility;” 5.2, “Organization;” 5.3, “Unit Staff Qualifications;” 5.5, “Programs and Manuals;” and for Catawba and McGuire, Section 5.7, “High Radiation Area.”                

*Date of issuance:*                     February 12, 2015.                

*Effective date:*                     This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.                

*Amendment Nos.:*                     273, 269, 276, 256, 389, 391, and 390. A publicly-available version is available in ADAMS under Accession No. ML15002A324.                

*Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-17, DPR-38, DPR-47, and DPR-55:*                     Amendments revised the licenses and Technical Specifications.                

*Date of initial notice in**Federal Register**:*                     November 12, 2014 (79 FR 67199).                

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 12, 2015.

No significant hazards consideration comments received: No.

#### Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas

*Date of application for amendment:*                     December 17, 2012, as supplemented by letters dated November 7, and December 4, 2013; January 6, May 22, June 30, August 7, September 24, and December 9, 2014.                

*Brief description of amendment:*                     The amendment authorized the transition of the Arkansas Nuclear One, Unit No. 2, 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:*                     February 18, 2015.                

*Effective date:*                     As of its date of issuance and shall be implemented by 6 months from the date of issuance.                

*Amendment No.:*                     300. A publicly-available version is in ADAMS under Accession No. ML14356A227; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Renewed Facility Operating License No. NPR-6:*                     Amendment revised the License and Technical Specifications.                

*Date of initial notice in**Federal Register**:*                     July 23, 2013 (78 FR 44171). The supplemental letters dated November 7 and December 4, 2013; and January 6, May 22, June 30, August 7, September 24, and December 9, 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 February 18, 2015.

No significant hazards consideration comments received: No.

#### Entergy Nuclear FitzPatrick, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

*Date of amendment request:*                     October 8, 2013, as supplemented by a letter dated November 18, 2014.                

*Brief description of amendment:*                     The amendment modifies the Technical Specifications (TSs) to reduce the reactor steam dome pressure associated                           ( print page 11490)                        with the Reactor Core Safety Limit from 785 psig to 685 psig in TS 2.1.1.1 and TS 2.1.1.2. This change addresses the potential to not meet the pressure/thermal power/minimal critical power ratio TS safety limit during a pressure regulator failure-maximum demand (open) (PRFO) transient. The PRFO transient was reported by General Electric as a notification pursuant to Title 10 of the                     *Code of Federal Regulations,*                     Part 21, “Reporting of Defects and Noncompliance.”                

*Date of issuance:*                     February 9, 2015.                

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

*Amendment No.:*                     309. A publicly-available version is in ADAMS under Accession No. ML15014A277; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Renewed Facility Operating License No. DPR-59:*                     Amendment revised the Renewed Facility Operating License and Technical Specifications.                

*Date of initial notice in**Federal Register**:*                     July 8, 2014 (79 FR 38589). The supplemental letter dated November 18, 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 February 9, 2015.

No significant hazards consideration comments received: No.

#### 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:*                     November 14, 2013, as supplemented by letters dated June 9, 2014, August 6, 2014, and October 9, 2014.                

*Description of amendment request:*                     The amendment eliminates operability requirements for secondary containment when handling sufficiently decayed irradiated fuel or a fuel cask following a minimum of 13 days after the permanent cessation of reactor operation.                

*Date of Issuance:*                     February 12, 2015.                

*Effective date:*                     The license amendment becomes effective 13 days after the licensee's submittal of the certifications, as required by 10 CFR 50.82(a)(1)(i) and (ii).                

*Amendment No.:*                     262. A publicly-available version is in ADAMS under Accession No. ML14304A588; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Facility Operating License No. DPR-28:*                     The amendment revised the Renewed Facility Operating License and Technical Specifications.                

*Date of initial notice in**Federal Register**:*                     September 16, 2014 (79 FR 55511).                

                    The supplemental letters dated June 9, 2014, August 6, 2014, and October 9, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the                     **Federal Register**                    .                

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated February 12, 2015.

No significant hazards consideration comments received: No.

#### NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

*Date of amendment request:*                     June 23, 2014.                

*Brief description of amendment:*                     The amendment revised the Technical Specification (TS) requirements to address NRC Generic Letter (GL) 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” as described in TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”                

*Date of issuance:*                     February 10, 2015.                

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

*Amendment No.:*                     290. A publicly-available version is in ADAMS under Accession No. ML15014A200; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Renewed Facility Operating License No. DPR-49:*                     The amendment revised the Renewed Facility Operating License and Technical Specifications.                

*Date of initial notice in**Federal Register**:*                     September 30, 2014 (79 FR 58820).                

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 10, 2015.

No significant hazards consideration comments received: No

#### NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

*Date of amendment request:*                     June 24, 2014, as supplemented by letter dated December 11, 2014.                

*Brief description of amendment:*                     The amendment revised the Seabrook Technical Specifications (TSs). Specifically, the amendment modifies Seabrook TSs to address U.S. Nuclear Regulatory Commission Generic Letter (GL) 2008-01, “Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,” as described in TSTF-523, Revision 2, “Generic Letter 2008-01, Managing Gas Accumulation.”                

*Date of issuance:*                     February 6, 2015.                

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

*Amendment No.:*                     144. A publicly-available version is in ADAMS under Accession No. ML14345A288; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Facility Operating License No. NPF-86:*                     The amendment revised the License and TS.                

*Date of initial notice in**Federal Register**:*                     September 2, 2014 (79 FR 52066). The supplemental letter dated December 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 February 6, 2015.

*No significant hazards consideration comments received:*                     No.                

#### South Carolina Electric and Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, Fairfield County, South Carolina

*Date of amendment request:*                     November 15, 2011, as supplemented by letters dated November 22, 2011; January 26 and October 10, 2012; February 1, April 1, October 14, and November 26, 2013; January 9, February 25, May 2, May 11, August 14, October 9, and December 11, 2014.                

*Brief description of amendment:*                     The amendment authorizes the transition of the V.C. Summer fire protection program to a risk-informed, performance-based program based on                           ( print page 11491)                        National Fire Protection Association (NFPA) 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition” (NFPA 805), in accordance with 10 CFR 50.48(c).                

*Date of issuance:*                     February 11, 2015.                

*Effective date:*                     This amendment is effective as of its date of issuance and shall be implemented per the December 11, 2014, supplement, Attachment S, Table S-2 “Implementation Items”, requiring full implementation by March 31, 2016.                

*Amendment No.:*                     199. A publicly-available version is in ADAMS under Accession No. ML14287A289; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Renewed Facility Operating License No. NPF-12:*                     Amendment revised the Facility Operating License.                

*Date of initial notice in**Federal Register**:*                     August 14, 2012 (77 FR 48561). The supplemental letters dated November 22, 2011; October 10, 2012; February 1, April 1, October 14, and November 26, 2013; January 9, February 25, May 2, May 11, August 14, October 9, and December 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 February 11, 2015.

No significant hazards consideration comments received: No.

#### Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear Plant (HNP), Unit No. 2, Appling County, Georgia

*Date of amendment request:*                     August 8, 2014, as supplemented by letters dated September 8 and October 24, 2014.                

*Brief description of amendments:*                     The amendment revises the Technical Specification value of the Safety Limit Minimum Critical Power Ratio to support operation in the next fuel cycle.                

*Date of issuance:*                     February 18, 2015.                

*Effective date:*                     As of the date of issuance and shall be implemented prior to reactor startup following the HNP, Unit 2, spring 2015 refueling outage.                

*Amendment No(s).:*                     218. A publicly-available version is in ADAMS under Accession No. ML15020A434; documents related to this amendment are listed in the Safety Evaluation enclosed with the amendment.                

*Renewed Facility Operating License Nos. DPR-57 and NPF-5:*                     Amendment revised the licenses and the Technical Specifications.                

*Date of initial notice in**Federal Register**:*                     January 6, 2015, (80 FR 536).                

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 18, 2015.

No significant hazards consideration comments received: No.

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

*Date of amendment request:*                     July 23, 2013, as supplemented by letters dated May 12 (two letters), May 19, and December 17, 2014.                

*Brief description of amendments:*                     The amendments revised the STP, Units 1 and 2, Fire Protection Program (FPP) related to the alternate shutdown capability. Specifically, it approves the following operator actions in the control room prior to evacuation due to a fire for meeting the alternate shutdown capability, in addition to manually tripping the reactor that is currently credited in the STP, Units 1 and 2, FPP licensing basis:                
    - Initiate main steam line isolation
    - Closing the pressurizer power-operated relief valves block valves
    - Securing all reactor coolant pumps
    - Closing feedwater isolation valves
    - Securing the startup feedwater pump
    - Isolating reactor coolant system letdown
    - Securing the centrifugal charging pumps

In addition, the licensee credits the automatic trip of the main turbine upon the initiation of a manual reactor trip for meeting the alternate shutdown capability.

*Date of issuance:*                     February 13, 2015.                

*Effective date:*                     As of the date of issuance and shall be implemented within 45 days of issuance.                

*Amendment Nos.:*                     Unit 1—203; Unit 2—191. A publicly-available version is in ADAMS under Accession No. ML14339A170; documents related to these amendments are listed in the Safety Evaluation enclosed with the amendments.                

*Facility Operating License Nos. NPF-76 and NPF-80:*                     The amendments revised the Facility Operating Licenses.                

*Date of initial notice in**Federal Register**:*                     October 29, 2013 (78 FR 64546). The supplements dated May 12 (two letters), May 19, and December 17, 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 February 13, 2015.

No significant hazards consideration comments received: No.

#### Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

*Date of amendment request:*                     December 18, 2013, as supplemented by letter dated June 13, 2014.                

*Brief description of amendment:*                     The amendment revised the Technical Specification (TS) 3.4.9, “RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,” Figures 3.4.9-1 through 3.4.9-2. The P/T limits are based on proprietary topical report NEDC-33178P-A, Revision 1, “GE [General Electric] Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves.” NEDO-33178-A, Revision 1 is the non-proprietary version of the NRC-approved topical report.                

*Date of issuance:*                     February 2, 2015.                

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

*Amendment No.:*                     287. A publicly available version is in ADAMS under Accession No. ML14325A501; documents related to this amendment are listed in the Safety Evaluation (SE) enclosed with the amendment.                

*Renewed Facility Operating License No. DPR-33:*                     Amendment revised the TSs and the Operating License.                

*Date of initial notice in**Federal Register**:*                     May 6, 2014 (79 FR 25902). The supplemental letter dated June 13, 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 the SE dated February 2, 2015.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 23rd day of February 2015.
      ( print page 11492)    
For the Nuclear Regulatory Commission.

Michele G. Evans, 

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

[FR Doc. 2015-04298 Filed 3-2-15; 8:45 am]

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