2012-10195. Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

Background

Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission 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 April 5, 2012 to April 18, 2012. The last biweekly notice was published on April 17, 2012 (77 FR 22808).

ADDRESSES:

You may access information and comment submissions related to this document, which the NRC possesses and is publicly available, by searching on http://www.regulations.gov under Docket ID NRC-2012-0096.

You may submit comments by the following methods:

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0096. Address questions about NRC dockets to Carol Gallagher; telephone: [phone]; email: [e-mail].
• Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
• Fax comments to: RADB at 301-492-3446.

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

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2012-0096 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and is publicly available, by the following methods:

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0096.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly available documents online in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail]. Documents may be viewed in ADAMS by performing a search on the document date and docket number.
• 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-2012-0096 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

The NRC cautions you not to include identifying or contact information in comment submissions that you do not want to be publicly disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS, and the NRC does not 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 in their comment submissions that they do not want to be publicly disclosed. Your request should state that the NRC will not edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Combined Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Title 10 of the Code of Federal Regulations (10 CFR) 50.92, 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; (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.

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 ”Rules of Practice for Domestic Licensing Proceedings” 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. NRC regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

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

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.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 the 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 agency'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 Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by email at [e-mail], or by a toll-free call at 1-866 672-7640. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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

Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/​ehd/​, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. 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. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737, or by email to [e-mail].

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: January 10, 2012.

Description of amendment request: The proposed amendment would modify Fermi 2 Plant Operating License, Appendix A, Technical Specifications (TS) to revise the Residual Heat Removal (RHR) Suppression Pool Cooling Surveillance Requirement (SR) 3.6.2.3.2, flow requirement from greater than or equal to 10,000 gallons per minute (gpm) to greater than or equal to 9,250 gpm. This change is consistent with the RHR suppression cooling rate associated with RHR heat exchanger minimum thermal performance requirements. Additionally, the proposed license amendment clarifies that SR 3.6.2.3.2 applies only to pumps required for meeting the Limiting Condition of Operation.

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

The proposed TS SR 3.6.2.3.2 minimum flow of greater than or equal to 9,250 gpm is consistent with that assumed for accident extrapolation calculations of measured thermal performance obtained during RHR heat exchanger testing. This testing is performed to periodically demonstrate that the actual heat exchanger thermal performance exceeds that assumed for establishing the maximum post-accident bulk average suppression pool temperature. Therefore, the change in required RHR suppression pool cooling flow will not result in any increase in post-accident suppression pool temperature above that already evaluated for demonstrating adequate Net Pump Suction Head (NPSH) for any Emergency Core Cooling System (ECCS) pump. The change in the applicability of the surveillance to each required RHR pump provides consistency with the design of the system and maintains full capability of each RHR suppression pool cooling subsystem to provide post accident design basis cooling.

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

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

The proposed change revises TS SR 3.6.2.3.2 for RHR suppression pool cooling flow to be consistent with that assumed for evaluating measured heat exchanger thermal performance against the minimum requirements of the plant safety analysis. Changing the applicability of the surveillance to each required RHR pump is consistent with the system design requirement and maintains full capability of each RHR suppression pool cooling subsystem to provide the post accident cooling function. No physical changes are being made to the installed RHR system or the manner in which it is operated. No new or different accident scenarios are created by this change.

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

3. The proposed change does not involve a significant reduction in the margin of safety.

The RHR system has historically been capable of meeting TS SR 3.6.2.3.2. This Surveillance requires demonstration of a system flow, in conjunction with a prescribed RHR heat exchanger capacity that ensures the overall suppression pool cooling capacity meets the requirements of the safety analysis. However, the lack of available operating margin inherent in the design orifices of the RHR suppression pool cooling test return line and identification of a non-conservative bias in the test flow instrument calibration have eroded the flow test margin such that it is possible that the TS SR may not be satisfied in the future even though a large margin is maintained compared to the minimum performance assumed in the containment safety analyses. The proposed change makes the margin between TS SR 3.6.2.3.2 and the performance assumed in the plant safety analyses available as a design and operating margin. This is ensured by establishing a higher level of required heat exchanger performance, where ample margin is available. Heat exchanger testing is conducted in accordance with existing testing standards as prescribed by EPRI TR-107397, Service Water Heat Exchanger Testing Guidelines. The minimum required flow rate necessary to satisfy RHR suppression pool cooling TS SR 3.6.2.3.2 will be documented in the plant design basis with the minimum required flow adjusted upward as necessary to account for instrument uncertainty and bias as well as differences between assumed accident and actual test operating conditions.

The change in the applicability of the surveillance to each required RHR pump is consistent with the design basis of the plant and maintains full capability of the system to provide its safety related cooling function following a design basis accident.

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

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

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

NRC Acting Branch Chief: Shawn A. Williams.

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: January 10, 2012.

Description of amendment request: The proposed amendment would modify Fermi 2 Plant Operating License, Appendix A, Technical Specifications (TS) to modify Surveillance Requirement (SR) 3.4.3.2, in TS 3.4.3, “Safety Relief Valves (SRVs)”, SR 3.5.1.13, in TS 3.5.1, “ECCS-Operating,” and SR 3.6.1.6.1, in TS 3.6.1.6, “Low-Low Set (LLS) Valves.” This proposed amendment replaces the current requirement in these TS SRs to verify the SRV opens when manually actuated with an alternate requirement that verifies the SRV is capable of being opened. The verification of that capability would be satisfied by a series of overlapping tests, performed during a refueling outage, that demonstrate the required functions of successive valve stages.

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

The proposed change does not modify the method of demonstrating the operability of the Safety Relief Valves (SRVs) in both the safety and relief modes of operation. As currently stated in the Technical Specification (TS) Bases “* * * valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation.” The proposed change does modify the method for demonstrating the proper mechanical functioning of the SRVs. The SRVs are required to function in the safety mode to prevent overpressurization of the reactor vessel and reactor coolant system pressure boundary during various analyzed transients, including Main Steam Isolation Valve closure. SRVs associated with the Automatic Depressurization System are also required to function in the relief mode to reduce reactor pressure to permit injection by low pressure Emergency Core Cooling System (ECCS) pumps during certain reactor coolant pipe break accidents. The current testing method demonstrates the proper mechanical functioning of the SRVs in both modes through manual actuation of the SRVs. The proposed new testing method demonstrates both operability and proper mechanical functioning using a series of overlapping tests that demonstrate proper functioning of the SRV and supporting control components. This proposed testing method results in acceptable demonstration of the SRV functions in both the safety and relief modes, and therefore provides assurance that the probability of SRV failure will not increase. None of the accident safety analyses is affected by the requested TS changes. Therefore, the consequences of accidents mitigated by the SRVs will not increase.

Certain SRV malfunctions are included in the UFSAR safety analyses. Specifically, the plant safety analyses include the inadvertent opening of an SRV and a stuck open SRV. By reducing or not actuating the SRVs during plant operation for testing and thus reducing the potential incidence of pilot stage leakage of the SRVs, the proposed testing reduces a contributor to these events.

Based on these considerations, the proposed test method does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change modifies the method of testing of the SRVs, but does not alter the functions or functional capabilities of the SRVs. Testing under the proposed method is performed in offsite test facilities and in the plant during outage periods when the SRV functions are not required. Existing analyses address events involving an SRV inadvertently opening or failing to reclose. Analyses also address the failure of one or more SRVs to open. The proposed change does not introduce any new failure mode, and therefore, does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

3. The proposed change does not involve a significant reduction in the margin of safety.

The proposed amendment provides for an alternative means of testing the SRVs. The proposed changes will provide a complete verification of the functional capability of the SRVs by performing a series of tests, inspections, and maintenance activities without opening the valves with reactor steam while installed in the plant. The alternative testing and associated programmatic controls will provide an equivalent level of assurance that the SRVs are capable of performing their intended accident mitigation safety functions. The proposed amendment does not affect the valve setpoints or adversely affect any other operational criteria assumed for accident mitigation. No changes are proposed that alter the setpoints at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation. Moreover, it is expected that the alternative testing methodology will increase the margin of safety by reducing the potential for SRV leakage resulting from testing the SRVs with reactor steam pressure while installed in the plant.

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

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

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

NRC Acting Branch Chief: Shawn A. Williams.

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 amendment request: September 20, 2011, as supplemented by letter dated November 21, 2011.

Description of amendment request: The proposed amendments would allow revisions to the current licensing basis to allow a measurement uncertainty recapture (MUR) power uprate.

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

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

Response: No.

The proposed amendment changes the rated thermal power from 2568 megawatts thermal (MWt) to 2610 MWt; an increase of approximately 1.64% Rated Thermal Power. Duke Energy's evaluations have shown that all structures, systems and components (SSCs) are capable of performing their design function at the uprated power of 2610 MWt. A review of station accident analyses found that all but two analyses remain bounding at the uprated power of 2610 MWt. These two analyses (High Energy Line Break and Double Main Steam Line Break) were reanalyzed at the higher power level and found to be acceptable.

The radiological consequences of operation at the uprated power conditions have been assessed. The proposed power uprate does not affect release paths, frequency of release, or the analyzed reactor core fission product inventory for any accidents previously evaluated in the Final Safety Analysis Report. Analyses performed to assess the effects of mass and energy releases remain valid. All acceptance criteria for radiological consequences continue to be met at the uprated power level.

As summarized in Sections IV, V, and VI of Enclosure 2, the proposed change does not involve any change to the design or functional requirements of the associated systems. That is, the increased power level neither degrades the performance of, nor increases the challenges to any safety systems assumed to function in the plant safety analysis.

While power level is an input to accident analyses, it is not an initiator of accidents. The proposed change does not affect any accident precursors and does not introduce any accident initiators. The proposed change does not impact the usefulness of the Surveillance Requirements (SRs) in evaluating the operability of required systems and components.

In addition, evaluation of the proposed TS [Technical Specification] change demonstrates that the availability of equipment and systems required to prevent or mitigate the radiological consequences of an accident is not significantly affected. Since the impact on the systems is minimal, it is concluded that the overall impact on the plant safety analysis is negligible.

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

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

Response: No.

A Failure Modes and Effects Analysis of the new system was performed, and the possible effects of failures of the new equipment and the increased power level on the overall plant systems were reviewed. This review found that no new or different accidents were created by the new equipment or the uprated power levels.

No installed equipment is being operated in a different manner. The proposed changes have no significant adverse affect on any safety-related SSCs and do not significantly change the performance or integrity of any safety-related system.

The proposed changes do not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than previously evaluated. The uprated power does not create any new accident initiators. Credible malfunctions are bounded by the current accident analyses of record or recent evaluations demonstrating that applicable criteria are still met with the proposed changes.

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

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

Response: No.

Although the proposed amendment increases the operating power level of the plants, it retains the margin of safety because it is only increasing power by the amount equal to the reduction in uncertainty in the heat balance calculation. The margins of safety associated with the power uprate are those pertaining to core thermal power. These include fuel cladding, reactor coolant system pressure boundary, and containment barriers. Analyses demonstrate that the current design basis continues to be met after the MUR power uprate. Components associated with the reactor coolant system pressure boundary structural integrity, including pressure-temperature limits, vessel fluence, and pressurized thermal shock are bounded by the current analyses. Systems will continue to operate within their design parameters and remain capable of performing their intended safety functions.

The current Oconee safety analyses, and the revised design basis radiological accident dose calculations, bound the power uprate and therefore do not significantly impact margins.

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

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

Attorney for licensee: Lara S. Nichols, Associate General Counsel, Duke Energy Corporation, 526 South Church Street—EC07H, Charlotte, NC 28202.

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: December 8, 2011.

Description of amendment request: The amendment would revise Technical Specifications (TS) 3.8.1; “AC [Alternating Current] Sources—Operating.” Specifically, the amendment would revise TS 3.8.1 and the associated Bases, to expand its scope to include provisions for testing of the automatic transfer function from the station 22 kiloVolt (kV) bus to offsite power for Division III. A new Surveillance Requirement (SR) would be added to ensure availability of offsite power after loss of the station (onsite) 22 kV bus when offsite power remains available. The amendment would also add notes to the Limiting Condition for Operation (LCO) and SR to require this feature when Division III is powered by onsite power. In addition, new ACTIONS would be added to ensure this transfer from onsite to offsite is maintained when a required offsite power source is lost.

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

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

Response: No.

The proposed change revises the Technical Specification Surveillance Requirements to allow power for emergency systems to be supplied from onsite power prior to event initiation. This power supply will be transferred to the current accepted offsite power source if the main generator is no longer available. The proposed Surveillance Requirement is to confirm the automatic transfer function.

The proposed changes do not involve a change in the design requirements of the electrical power systems, including the emergency power systems. The plant will continue to operate within acceptable parameters (electrical loading, etc.) The proposed changes do not change the function of plant equipment, or affect the response of emergency power systems.

The proposed changes do not involve a change in the design basis initiators for loss of offsite power to the emergency power systems. The proposed change utilizes existing components and circuits. The change will add a new surveillance requirement to confirm the design function operation.

The proposed change does not impact other design basis accident initiators or analyzed events or assumed mitigation of accident or transient events.

The proposed change does not involve a change to the consequences of a design basis event as described in the Safety Analysis Report.

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

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

Response: No.

The proposed change revises the Technical Specification surveillance requirements to confirm operation of existing components and circuits. The proposed changes do not involve a change in the design basis initiators for loss of offsite power to the emergency power systems.

The proposed changes do not involve a change in the design requirements of the electrical power systems, including the emergency power systems. The proposed changes do not change the function of plant equipment, nor do they affect the response of emergency power systems.

The proposed changes do not involve a change in the operational limits or physical design of the electrical power systems, particularly the emergency power systems. The proposed changes do not change the design function or operation of plant equipment, nor do they introduce any new failure mechanisms. This change will implement surveillance requirements to confirm the design function operation.

The transfer function components supporting the safety-related buses have been designed to applicable quality standards and design criteria. As such, no new failure modes are being introduced. The plant equipment will continue to respond in accordance with the design and analyses, and no malfunction of a new or different type is being introduced by the proposed changes.

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

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

Response: No.

The proposed change revises the Technical Specification surveillance requirements. The proposed changes do not involve a change in the operational limits or the response of that equipment if it is called upon to operate.

The performance capability of the emergency diesel generators will not be affected. The plant will continue to operate within acceptable parameters (electrical loading, etc.)[.]

In addition, administrative controls will ensure there are adequate administrative controls are in place to ensure the plant configuration remains as evaluated.

The results of the PRA performed to quantitatively assess the risk impact of this change indicate there is a minimal risk impact.

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

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

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

NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit 2, Westchester County, New York

Date of amendment request: January 11, 2012.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) Table 3.3.6-1, “Containment Purge System and Pressure Relief Line Isolation Instrumentation.” The proposed amendment would change the term “ALLOWABLE VALUE” to “TRIP SETPOINT” and revise the current setpoint used for the Containment Purge Systems and Pressure Relief Line isolation. The proposed revision to TS Table 3.3.6-1 will change “≤ 3 x background” to allow the trip setpoint to be as specified in the Offsite Dose Calculation Manual (ODCM).

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 will revise the term “ALLOWABLE VALUE” to “TRIP SETPOINT” and change the setpoint requirements from “≤ 3 [x] background” to allow the allowable value to be as specified in the Offsite Dose Calculation Manual (ODCM). The change to trip setpoint is a correction of an administrative error and will only affect the instrument setting specified. Therefore it does not involve the initiation of an accident or the consequences. The values for the instrument setting are provided for isolating the Containment Purge and Pressure Relief Systems due to increased source terms and are redundant to containment isolation signals. They have no effect on the probability of an accident previously evaluated. The change in the setting will be negligible for purposes of an accident termination. The ODCM limits are based on 10 CFR 20 [Title 10 of the Code of Federal Regulations, Part 20] limits which are substantially below accident analysis release rates. Therefore the change has a minimum effect on the consequences of such accidents.

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

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 will revise the term “ALLOWABLE VALUE” to “TRIP SETPOINT” and change the setpoint requirements. The changes do not affect the system operations, plant operating procedures or affect how the plant is operated. The change does not create the possibility of any equipment failure or effect on other equipment.

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

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

Response: No.

The proposed change will revise the term “ALLOWABLE VALUE” to “TRIP SETPOINT” and change the setpoint requirements. The change to trip setpoint is correcting an administrative error and has no significant affect on the margin of safety. The proposed change involves changes to existing setpoints for automatic isolation of the Containment Purge and Pressure Relief Systems. However, the ability of the systems to isolate remains within current evaluations and therefore does not significantly reduce the safety margin.

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: Mr. William C. Dennis, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue White Plains, NY 10601.

NRC Branch Chief: George Wilson.

Northern States Power Company—Minnesota, Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

Date of amendment request: April 5, 2012.

Description of amendment request: The licensee proposed to revise the MNGP Technical Specifications (TS) 3.3.5.1, “Emergency Core Cooling System (ECCS) Instrumentation.” Specifically, it is proposed to revise the lower allowable value limit for Table 3.3.5.1, Functions 1.e and 2.e, “Reactor Steam Dome Pressure Permissive—Bypass Timer (Pump Permissive).” The licensee has determined that the upper allowable value limit for the Automatic Depressurization System (ADS) bypass timer function provides the operator sufficient time to assess the situation and inhibit ADS actuation if the event does not require rapid reactor depressurization. The lower allowable value limit ADS bypass timer function pertains to providing adequate margin to unwanted pump starts during reactor water level transients and is not credited in the safety analyses.

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 (NSHC). The NRC staff reviewed the licensee's NSHC analysis and prepared its own as follows:

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

Response: No.

The proposed change does not physically impact the plant nor does it impact any design or functional requirements of the Automatic Depressurization System (ADS). The proposed change does not degrade the performance or increase the challenges to any safety systems assumed to function in the accident analysis. There is no change to normal plant operating parameters or accident mitigation performance.

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

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

Response: No.

There are no hardware changes nor are there any changes in the method by which plant systems perform a safety function. This request does not affect the normal method of plant operation. No new equipment is introduced which could create a new or different kind of accident. No new equipment failure modes are created. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced.

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

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

Response: No.

The proposed amendment does not affect the assumptions of the safety analysis or the availability or operability of any plant equipment. There is no reduction in the margin of safety because the criteria for the performance of the ADS are not changed and there are no changes to those plant systems necessary to assure the accomplishment of protection functions.

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

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

Attorney for the licensee: Peter M. Glass, Assistant General Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.

NRC Branch Chief: Istvan Frankl, Acting.

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

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

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

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

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the NRC's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available documents created or received at the NRC are accessible electronically through the Agencywide Documents Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to [e-mail].

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station (DBNPS), Unit 1, Ottawa County, Ohio

Date of application for amendment: May 20, 2011 as supplemented by letter dated February 7, 2012.

Brief description of amendment: This amendment revised Technical Specification (TS) 5.5.8.g to perform the special visual inspections based on a condition rather than a specific frequency. Specifically, TS 5.5.8.g requires visual inspection of the secured internal auxiliary feedwater header (AFWH), header to shroud attachment welds, and external header thermal sleeves of the steam generators (SGs) at DBNPS to be performed during the third period of each 10-year inservice inspection interval (ISI). With the proposed change, if eddy current inspections (required by TS 5.5.8.d.5) identify any SG peripheral rube to secured internal AFWH gaps less than ¼ inches or there is evidence that the header is degrading or has moved, then the TS 5.5.8.g visual inspections shall be performed on the affected SG.

Date of issuance: April 18, 2012.

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

Amendment No.: 285.

Facility Operating License No. NPF-3: Amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: September 20, 2011 (76 FR 58306). The February 7, 2012 supplement 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 finding of no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 18, 2012.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 30, 2011, as supplemented on March 20, 2012.

Brief description of amendment: The proposed amendment changes the NMP2 Updated Safety Analysis Report allowing the use of Modified Alloy 718 material for fabrication of the NMP2 reactor recirculation system jet pump holddown beams.

Date of issuance: April 13, 2012.

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

Amendment No.: 141.

Renewed Facility Operating License No. NPF-069: The amendment revises the License and Updated Safety Analysis Report.

Date of initial notice in Federal Register: February 8, 2012 (77 FR 6601). The supplemental letter dated March 20, 2012, provided additional information that clarified the application and did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 13, 2012.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 28, 2011, as supplemented on February 14 and March 14, 2012.

Brief Description of amendments: These amendments permanently revise the Technical Specifications (TS) 6.4.Q, “Steam Generator (SG) Program,” to exclude portions of the SG tube below the top of the SG tubesheet from periodic inspections. In addition, this amendment request proposes to revise TS 6.6.A.3, “Steam Generator Tube Inspection Report,” to remove references to the previous Unit 1 one-time and Unit 2 temporary alternate repair criteria and provides reporting requirements specific to the permanent alternate repair criteria. This amendment also addressed minor administrative revisions to reinstate the superscript number 1 as the end of the TS 4.13.B.

Date of issuance: April 17, 2012.

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

Amendment Nos.: Unit 1—277 and Unit 2—277.

Renewed Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the licenses and the technical specifications.

Date of initial notice in Federal Register: October 25, 2011 (76 FR 66090). The supplements dated February 14, 2012 and March 14, 2012, 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. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 17, 2012.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 20th day of April 2012.

For the Nuclear Regulatory Commission.

Michele G. Evans,

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