2011-11804. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

I. 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 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 21, 2011 to May 4, 2011. The last biweekly notice was published on May 3, 2011 (76 FR 24926).

Notice of Consideration of Issuance of Amendments to Facility Operating 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; 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.

Written comments may be submitted by mail to the Chief, Rules, Announcements and Directives Branch (RADB), TWB-05-B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be faxed to the RADB at 301-492-3446. Documents may be examined, and/or copied for a fee, 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.

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. 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 Commission's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, 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 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 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 e-mail 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 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 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 e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRC 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 Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by e-mail 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 Commission's PDR, located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible from the ADAMS Library online 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 PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to [e-mail].

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

Date of amendment request: March 7, 2011.

Description of amendment request: The proposed amendment would add an applicability period of 42.1 effective full power years (EFPY) to TS LCO 3.4.3, figures 3.4.3-1 and 3.4.3-2 which contain the pressure-temperature (P/T) limit curves for primary coolant system (PCS) heatup and cooldown, and limiting condition for operation (LCO) 3.4.12 figure 3.4.12-1, which contains the low temperature overpressure protection (LTOP) setpoint limit curve.

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.

No changes are being made to the existing pressure-temperature (P/T) limit curves in TS Limiting Condition for Operation (LCO) 3.4.3 Figures 3.4.3-1 and 3.4.3-2 and the low temperature overpressure (LTOP) setpoint limit curve in LCO 3.4.12 Figure 3.4.12-1. The P/T limits curves and the LTOP setpoint limit curve are only being revised to add the applicability period of 42.1 effective full power years. This applicability period has been verified to be conservative for operation through the expiration of the operating license on March 24, 2031.

The changes to the TS figures are applicable to normal plant operations and do not influence the probability of occurrence or safety analysis considerations for design basis accidents. Consequently, there will be no change to the probability or consequences of accidents previously evaluated. Operating the facility in accordance with the P/T limit and LTOP setpoint limit curves ensures that stresses caused by the thermal gradient through the RV beltline material remain bounded by the stress analyses. The proposed amendment does not involve operation of required structures, systems, or components in a manner or configuration different than previously recognized or evaluated. No radiological barriers are affected by the change.

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

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

Response: No.

No changes are being made to the existing P/T limit curves in TS Figures 3.4.3-1 and 3.4.3-2 and or in the existing LTOP setpoint limit curves in TS Figure 3.4.12-1. The TS figures are only being changed to add the applicability period of 42.1 effective full power years for the P/T limits and LTOP setpoint limit curves. Adding the applicability periods to the TS figures will not create the possibility of any new or different kind of accidents.

The change does not involve a modification of plant structures, systems, or components. The change will not affect the manner in which the plant is operated and will not degrade the reliability of structures, systems, or components. Equipment protection features will not be deleted or modified, equipment redundancy or independence will not be reduced, and supporting system performance will not be affected. No new failure modes or mechanisms will be introduced as a result of this proposed change.

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

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

Response: No.

Appendix G to 10 CFR Part 50 describes the conditions that require P/T limits and provides the general bases for these limits. Operating limits based on the criteria of Appendix G, as defined by applicable regulations, codes and standards, provide reasonable assurance that non-ductile or rapidly propagating failure will not occur. The P/T limits are prescribed for all plant modes to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the reactor coolant pressure boundary. Calculation of P/T limits in accordance with the criteria of Appendix G to 10 CFR Part 50 and applicable regulatory requirements ensures that adequate margins of safety are maintained and there is no significant reduction in a margin of safety.

No change is being made to the existing P/T limit curves or LTOP setpoint curve. Only the applicability period associated with the P/T Limits and LTOP setpoints is being extended. Since the P/T limits and LTOP setpoint limits remain unchanged there is no reduction in a margin of safety.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There is no change or impact on any safety analysis assumption or on any other parameter affecting the course of an accident analysis supporting the basis of any Technical Specification. The proposed change does not involve any increase in calculated off-site dose consequences.

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

NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: April 4, 2011.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to define a new time limit for restoring inoperable reactor coolant system (RCS) leakage detection instrumentation to operable status; establish alternate methods of monitoring RCS leakage when one or more required monitors are inoperable and make conforming TS Bases changes. These changes are consistent with NRC-approved Revision 3 to Technical Specification Task Force (TSTF) Standard Technical Specification (STS) Change Traveler TSTF-514, “Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation.”

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

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

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable Reactor Coolant System (RCS) leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. The monitoring of RCS leakage is not a precursor to any accident previously evaluated. The monitoring of RCS leakage is not used to mitigate the consequences of any accident previously evaluated.

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

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

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

Therefore, it is concluded that 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 clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. Reducing the amount of time the plant is allowed to operate with only the drywell atmospheric gaseous radiation monitor operable increases the margin of safety by increasing the likelihood that an increase in RCS leakage will be detected before it potentially results in gross failure.

Therefore, it is concluded that 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 requested amendments involve no significant hazards consideration.

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

NRC Branch Chief: Robert D. Carlson.

Indiana Michigan Power Company (the Licensee), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2 (DCCNP-1), Berrien County, Michigan

Date of amendment request: March 18, 2011.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TS), removing the specific isolation time for the main steam and main feedwater isolation valves (MSIVs) from Surveillance Requirements 3.7.2.1, 3.7.3.1, and 3.7.3.2. These changes were previously approved generically by the NRC staff and are tracked as Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-491.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee incorporated by reference the no significant hazards consideration (NSHC) analysis endorsed by the NRC staff in a December 29, 2006, Federal Register notice (71 FR 78472) and which was published in an October 5, 2006, Federal Register notice (71 FR 58884). The October 5, 2006, NSHC analysis is reproduced 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 relocating main steam and main feedwater valve isolation times to the Licensee Controlled Document that is referenced in the Bases. The proposed change is described in Technical Specification Task Force (TSTF) Standard TS Change Traveler TSTF-491 related to relocating the main steam and main feedwater valves isolation times to the Licensee Controlled Document that is referenced in the Bases and replacing the isolation time with the phase “within limits.”

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed changes relocate the main steam and main feedwater isolation valve times to the Licensee Controlled Document that is referenced in the Bases. The requirements to perform the testing of these isolation valves are retained in the TS. Future changes to the Bases or licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, “Changes, test and experiments,” to ensure that such changes do not result in more than minimal increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and the amounts of radioactive effluent that may be released, nor significantly increase individual or cumulative occupation/public radiation exposures.

Therefore, the changes do not involve a significant increase in the probability or consequences of any 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 changes relocate the main steam and main feedwater valve isolation times to the Licensee Controlled Document that is referenced in the Bases. In addition, the valve isolation times are replaced in the TS with the phase “within limits.” The changes do not involve a physical altering of the plant (i.e., no new or different type of equipment will be installed) or a change in methods governing normal plant operation. The requirements in the TS continue to require testing of the main steam and main feedwater isolation valves to ensure the proper functioning of these isolation valves.

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

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

The proposed changes relocate the main steam and main feedwater valve isolation times to the Licensee Controlled Document that is referenced in the Bases. In addition, the valve isolation times are replaced in the TS with the phase “within limits.” Instituting the proposed changes will continue to ensure the testing of main steam and main feedwater isolation valves. Changes to the Bases or license controlled document are performed in accordance with 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that main steam and feedwater isolation valve testing is conducted such that there is no significant reduction in the margin of safety.

The margin of safety provided by the isolation valves is unaffected by the proposed changes since there continue to be TS requirements to ensure the testing of main steam and main feedwater isolation valves. The proposed changes maintain sufficient controls to preserve the current margins of safety.

The NRC staff has reviewed the above 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: James M. Petro, Jr., Senior Nuclear Counsel, Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: Robert J. Pascarelli.

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

Date of amendment request: February 23, 2011.

Description of amendment request: The proposed amendment would adopt an approved change to the standard technical specifications (TSs) for General Electric Plants, BWR/4 (NUREG-1433), to allow relocation of specific TS surveillance frequencies to a licensee controlled program. The proposed change is described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (Rev. 3) (ADAMS Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control-RITSTF (Risk-Informed TSTF) Initiative 5b and was described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).

The proposed change is consistent with NRC-approved Industry/Technical Specification Task Force Traveler, TSTF-425, Rev. 3, “Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b.” The proposed change relocates surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). This change is applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,” (ADAMS Accession No. ML071360456).

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 any accident previously evaluated?

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new program—the SFCP. 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 (TS) 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 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 assumption 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, NextEra Energy Duane Arnold will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance with the 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: Ms. Marjan Mashhadi, 801 Pennsylvania Avenue, NW., Suite 220 Washington, DC 20004.

NRC Branch Chief: Robert J. Pascarelli.

NextEra Energy Point Beach, LLC (the Licensee), Docket Nos. 50-266 and 50-301, Point Beach

Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks, Manitowac County, Wisconsin.

Date of amendment request: March 23, 2011.

Description of amendment request: The proposed amendment consists of replacing non-conservative values for five operating limits in the Technical Specifications with more conservative values that incorporate measurement uncertainty. Additionally, one of the operating limits will replace a volume expressed in cubic feet with a volume expressed in tank percent level to allow the plant operators a direct verification of the technical specification limit based on instrument readings.

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

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

Response: No.

The proposed changes clarify the requirements for five plant operating limits by incorporating measurement uncertainties in the Technical Specification values to ensure the parameters remain within the ranges assumed in the accident analysis. The parameters are not accident initiators. Therefore, the proposed change will not increase the probability of an accident previously evaluated. Maintaining the parameters within the ranges specified in the Technical Specifications ensures that the systems will respond as assumed to mitigate the accidents previously evaluated. Therefore, the proposed change will not increase the consequences of an accident previously evaluated.

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

2. 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 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. The change does not alter assumptions made in the safety analysis, but ensures that plant operating parameters will be maintained as assumed in the accident analysis. The proposed change is consistent with the accident analysis assumptions.

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 amendment clarifies the requirements for plant operating limits by incorporating instrument uncertainties to ensure the parameters remain within the initial operating limits or ranges assumed in the accident analysis. No change is made to the accident analysis assumptions.

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

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

Attorney for licensee: William Blair, Senior Attorney, NextEra Energy Point Beach, LLC, P. O. Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief: Robert J. Pascarelli.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit 1 and 2, San Luis Obispo County, California

Date of amendment request: February 17, 2011, as supplemented on April 21, 2011.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) 3.7.1, “Main Steam Safety Valves (MSSVs),” Table 3.7.1-1, “Maximum Allowable Power Range Neutron Flux High Setpoint With Inoperable MSSVs,” and the Bases section for the MSSVs. This license amendment request proposed to remove a one-time note listed in TS Table 3.7.1-1, specific to Diablo Canyon Power Plant, Unit No. 2 for Cycle 15, that is no longer applicable or needed. This license amendment request also proposes to revise the TS Bases B 3.7.1 to reflect a new analysis methodology for establishing the reduced Power Range Neutron Flux High setpoint for one inoperable MSSV as listed in TS Table 3.7.1-1. The supplement dated April 21, 2011, proposes to revise the Final Safety Analysis Report Update (FSARU) Sections 15.2.7.3 and 15.2.16 to reflect the proposed changes to the TS Bases. The supplement provided additional information that clarified the application and did not expand the scope of the February 17, 2011, 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, which is presented below:

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

This License Amendment Request (LAR) proposes to remove a one-time Unit 2 Cycle 15 Limiting Condition for Operation (LCO) exemption that is no longer applicable and revise the safety analysis performed in support of Technical Specification (TS) 3.7.1, “Main Steam Safety Valves (MSSVs),” Table 3.7.1-1, “Maximum Allowable Power Range Neutron Flux High Setpoint with Inoperable MSSVs” for one inoperable MSSV. The revised safety analysis resolves a nonconforming condition associated with the TS 3.7.1 Bases and re-establishes that the Power Range Neutron Flux High setpoint of 87 percent Rated Thermal Power (RTP) continues to provide adequate protection for one inoperable MSSV on each steam lead.

The Power Range Neutron Flux High setpoint TS value does not initiate an accident. Technician adjustments to lower the Power Range Neutron Flux High setpoint could cause a reactor trip; however, this action is already a TS requirement. There has been no change in the TS setpoint value from the current value or in the requirement for a technician to adjust the setpoints downward when MSSVs become inoperable.

Therefore, this proposed change will not increase the probability of a reactor trip.

The revised TS B 3.7.1 safety analyses establishes that the current Power Range Neutron Flux High setpoint of 87 percent with one inoperable MSSV on each loop will ensure the remaining MSSVs will continue to prevent overpressure of the main steam leads and steam generators, and remove adequate heat from the RCS [reactor coolant system].

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

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

The revised safety analysis which credits the Class 1 Over Temperature Delta Temperature (OTDT) reactor trip and the Power Range Neutron Flux High setpoint TS value with one inoperable MSSV do not initiate an accident and do not change the method by which any safety-related system performs its function.

The proposed change does not result in plant operation outside the limits previously considered, nor allow the progression of transients or accidents in a manner different than previously considered.

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

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

The proposed change and revised safety analysis demonstrate that all applicable Reactor Coolant System (RCS) and steam generator (SG) pressure boundary acceptance criteria are satisfied, and re-establish that the existing Power Range Neutron Flux High setpoint TS value for one inoperable MSSV remains conservatively bounding.

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

With the proposed change, the MSSVs will prevent SG pressure from exceeding 110 percent of SG design pressure in accordance with the American Society of Mechanical Engineers code. The conclusions for the Final Safety Analysis Report accident analyses are unaffected by the change, remain valid, and provide margin.

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 requests involve no significant hazards consideration.

Attorney for licensee: Jennifer Post, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: March 14, 2011.

Description of amendment request: The proposed amendments would revise the licenses and the Technical Specifications regarding Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level, specifically, to allow one RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is operable and in operation. The proposed change is described in Technical Specification Task Force Traveler TSTF-361-A, Revision 2, “Allow standby SDC/RHR/DHR [shut down cooling/residual heat removal/decay heat removal] loop to [be] inoperable to support testing,” approved for use by the Nuclear Regulatory Commission in a letter dated October 31, 2000 (Agencywide Documents Access and Management System, Accession No. ML003775261).

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 adds an LCO [Limiting Condition for Operations] Note to LCO 3.9.6, “RHR and Coolant Circulation-Low Water Level,” to allow one RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is Operable and in operation. An inoperable RHR train is not an initiator to any accident previously evaluated. The RHR trains are not credited with mitigating any accident previously evaluated in Mode 6. As a result, the consequences of any accident previously evaluated are 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 change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve 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.

Therefore, this 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 adds an LCO Note to LCO 3.9.6, “RHR and Coolant Circulation-Low Water Level,” to allow one RHR loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is Operable and in operation. This allowance currently appears in Specification 3.4.7 and 3.4.8 and the conditions under which the Note would be applied in Specification 3.9.6 are not significantly different from those specifications. The Note is needed in LCO 3.9.6 to provide the flexibility to perform surveillance testing while ensuring that there is reasonable time for operators to respond to and mitigate any expected failures.

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: Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216.

NRC Branch Chief: Gloria Kulesa.

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

Date of amendment request: July 12, 2010.

Description of amendment request: The proposed revision is an administrative change that: (1) Corrects an error in TS 3.12.E.5, (2) deletes duplicative requirements in TS 3.12.E.2 and TS 3.12.E.4, (3) relocates the shutdown margin value in TS 3.12 and the TS 3.12 Basis to the Core Operating Limits Report (COLR), and (4) expands the TS 6.2 list of parameters defined in the COLR.

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

No. The proposed change is administrative in nature. The proposed LAR does not involve a physical change to any structures, systems, or components (SSCs) at Surry Power Station; nor does it change any of the previously evaluated accidents in the Updated Final Safety Analysis Report (UFSAR).

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

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

No. The proposed change is administrative in nature. The proposed change does not involve a physical change to any SSCs, and there is no impact on their design function. The proposed change does not affect initiators of analyzed events.

Therefore, the proposed change does not introduce any new failures that could 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?

No. The proposed change is administrative in nature. Margin of safety is established through the design of plant SSCs, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not impact the condition or performance of SSCs relied upon for accident mitigation or any safety analysis assumptions.

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

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

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

NRC Branch Chief: Gloria Kulesa.

Notice of Issuance of Amendments to Facility Operating 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.

Notice of Consideration of Issuance of Amendment to Facility Operating License, 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 Commission's Public Document Room (PDR), located at One White Flint North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible from the Agencywide Documents Access and Management System (ADAMS) online at 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 Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to [e-mail].

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

Date of application for amendment: April 13, 2010, as supplemented by a letter dated January 18, 2011.

Brief description of amendment: The licensee proposed to revise section 3.1.a.1.C, “Reactor Coolant Pumps,” section 3.1.a.3, “Pressurizer Safety Valves,” and section 3.1.b, “Heatup and Normal Cooldown Limit Curves for Normal Operation,” of the Technical Specifications (TS), as described in its application of April 13, 2010. After conversion of the TS to Improved Technical Specifications (ITS), the affected information was contained in ITS section 3.4.3, “Reactor Coolant System (RCS) Pressure and Temperature (P-T, or equivalently P/T) Limits”, ITS section 3.4.5, “RCS Loops—MODE 3”, ITS section 3.4.6, “RCS Loops—MODE 4”, ITS section 3.4.10, “Pressurizer Safety Valves”, ITS 3.4.12, “Low Temperature Overpressure Protection (LTOP) System”, and ITS section 3.5.2, “ECCS—Operating,” as described in the licensee's supplement of January 18, 2011. Specifically, the proposed amendment would replace the heatup and cooldown pressure-temperature (P-T) limit curves with new ones, and specify a higher LTOP enabling temperature. The supplement also provided additional restrictions on RCS mass addition until the reactor coolant system cold leg temperature exceeded 356 °F, consistent with Improved Standard Technical Specifications.

Date of issuance: April 29, 2011.

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

Amendment No.: 208.

Renewed Facility Operating License No. DPR-43: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 29, 2010 (75 FR 37473). The supplement dated January 18, 2011, provided additional information that clarified the application, did not expand the scope of the application, and did not change the Commission's proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 29, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 31, 2010, as supplemented by letters dated June 23, June 24, August 9, and September 16, 2010.

Brief description of amendment: The amendment modified the requirements of the Technical Specification definitions, requirements, and terminology related to the use of an Alternate Source Term (AST) associated with accident offsite and control room dose consequences. In addition, implementation of the AST supports adoption of the control room envelope habitability controls in accordance with NRC-approved Technical Specification Task Force (TSTF)-448, Revision 3, “Control Room Habitability.”

Date of issuance: April 26, 2011.

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

Amendment No.: 293.

Renewed Facility Operating License No. NPF-6: Amendment revised the Technical Specifications/license.

Date of initial notice in Federal Register: June 29, 2010 (75 FR 37475). The supplemental letters dated June 23, June 24, August 9, and September 16, 2010, 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 April 26, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 8, 2010.

Brief description of amendment: The amendment revised the Technical Specifications to be consistent with the NRC-approved Technical Specifications Task Force (TSTF) change traveler TSTF-493, “Clarify Application of Setpoint Methodology for LSSS [Limiting Safety System Setting] Functions,” Revision 4, Option A. Under Option A, two surveillance notes are added to TS Table 3.3.5.1-1, “Emergency Core Cooling System Instrumentation,” Function 3.d, “Condensate Storage Tank Level—Low,” and to TS Table 3.3.5.2-1, “Reactor Core Isolation Cooling System Instrumentation,” Function 3, “Condensate Storage Tank Level—Low,” for the suction swap from the condensate storage tank (CST) to the suppression pool function for the high pressure core spray and reactor core isolation cooling function, respectively. Specifically, surveillance notes would be added to surveillance requirements that require verifying trip setpoint setting values (i.e., channel calibration and trip unit calibration). The amendment completes a commitment made by the licensee to address an unresolved issue associated with TS Amendment No. 181 for the CST level-low setpoint change approved by the NRC in its letter dated February 25, 2009 (ADAMS Accession No. ML090290209).

Date of issuance: April 27, 2011.

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

Amendment No.: 185.

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

Date of initial notice in Federal Register: December 28, 2010 (75 FR 81670).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 27, 2011.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit 2 (BVPS-2), Beaver County, Pennsylvania

Date of application for amendment: April 9, 2009, as supplemented by letters dated June 15, 2009, January 18, 2010, March 18, 2010, May 3, 2010, May 21, 2010, June 1, 2010, August 9, 2010, October 7, 2010, October 18, 2010, January 5, 2011, February 18, 2011, March 18, 2011, and March 21, 2011.

Brief description of amendment: The amendment modified Technical Specifications (TSs) to support the replacement of existing Boraflex neutron absorber fuel storage racks in the BVPS-2 spent fuel pool with new high density, Metamic neutron absorber fuel storage racks, which will increase the total storage locations from 1,088 to 1,690.

Date of issuance: April 29, 2011.

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

Amendment No: 173.

Facility Operating License No. NPF-73: The amendment revised the License and the TSs.

Date of initial notice in Federal Register: March 11, 2010 (75 FR 11566). The supplements dated June 15, 2009, January 18, 2010, March 18, 2010, May 3, 2010, May 21, 2010, June 1, 2010, August 9, 2010, October 7, 2010, October 18, 2010, January 5, 2011, February 18, 2011, March 18, 2011, and March 21, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 29, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 7, 2009, as supplemented by letters dated October 17, 2008; April 8, June 17 (2 letters), August 24, September 11, September 25, October 9, November 13, November 20 (2 letters), November 21 (2 letters), December 8, December 16, December 21, and December 22 of 2009; January 7, January 8, January 13, January 22, January 29, February 11, February 12, February 25, March 3, March 24, March 25, April 15, April 21, April 22, April 26, April 28 (2 letters), April 29, April 30, May 6, May 13, May 14, May 20, June 10 (2 letters), June 11, June 14, June 24, July 8 (2 letters), July 15 (2 letters), July 21, July 23, July 27, July 28, July 29, August 2, August 6, August 9 (2 letters), August 12, August 23, August 24 (2 letters), August 26, September 1, September 8, September 9, September 14, September 21, September 27, September 28 (3 letters), October 1, October 12, October 14, October 15, October 28, November 1, November 4, November 12 (2 letters), November 15, November 30, December 1, December 7, December 10 (2 letters), December 13, December 15, December 21 (2 letters), and December 30 of 2010; January 7 (2 letters), January 11, January 13, January 21, February 22, March 2, and March 4 of 2011.

Brief description of amendments: The proposed amendments would increase the licensed core power level for PBNP Units 1 and 2 from 1540 megawatts thermal (MWt) to 1800 MWt. The increase in core thermal power will be approximately 17 percent over the current licensed thermal power level and is defined as an Extended Power Uprate (EPU). The proposed amendments would change the Renewed Facility Operating Licenses, the Technical Specifications (TSs) and licensing bases to support operation at the increased core thermal power level, including changes to the maximum licensed reactor core thermal power, reactor core safety limits, Constant Axial Offset Control (CAOC) operating strategy, Reactor Protection System (RPS) and Engineered Safety Feature Actuation System (ESFAS) Limited Safety System Settings (LSSSs) and diesel generator (DG) start loss of voltage time delays. Additional TS changes include Reactor Coolant System (RCS) flow rate, pressurizer operating level, pressurizer safety valve settings, accumulator and refueling water storage tank boron concentrations, main steam safety valve maximum allowable power level and lift settings, new Main Feedwater Isolation Valves (MFIVs), and Core Operating Limits Report (COLR) references.

The review of the EPU LAR will include the changes to the HELB methodology to verify compliance with the licensing basis and acceptability for EPU conditions. The HELB evaluations have been re-evaluated at EPU conditions using the following: (1) Implementation of NRC Generic Letter (GL) 87-11, “Relaxation in Arbitrary Intermediate Pipe Rupture Requirements,” dated June 19, 1987, and Branch Technical Position MEB 3-1, “Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment,” Revision 2, dated June 1987, (2) mass and energy released from a HELB, (3) compartment pressurization transient evaluation following a HELB event, (4) jet impingement from streams following a HELB event, and (5) operator response time evaluation.

Date of issuance: May 3, 2011.

Effective date: Unit 1—As of the date of issuance and shall be implemented prior to Unit 1 startup from the Fall 2011 refueling outage. Unit 2—As of the date of issuance and shall be implemented prior to startup from the Spring 2011 refueling outage.

Amendment Nos.: 241, 245.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revise the License, Appendix C, and the Technical Specifications.

Date of initial notice in Federal Register: November 17, 2010 (75 FR 70305).

The supplemental letters contained clarifying information and did not change the staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 3, 2011.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 28, 2010.

Description of amendment request: This amendment revises the Seabrook Technical Specifications (TSs) by deleting TS 3/4.8.4.2, “Containment Penetration Conductor Overcurrent Protective Devices and Protective Devices for Class 1E Sources Connected to Non-Class 1E Circuits,” and relocates the information to the Seabrook Technical Requirements Manual.

Date of issuance: April 29, 2011.

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

Amendment No.: 125.

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

Date of initial notice in Federal Register: November 2, 2010 (75 FR 67403).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 29, 2011.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 28, 2010.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.7.7, “Control Room Makeup and Cleanup Filtration System,” to add shutdown actions if the required actions for an inoperable control room envelope (CRE) boundary were not met. The amendments also added a note to the required action for an inoperable CRE boundary to clarify that the boundary is not a required system, subsystem, train, component, or device that depends on a diesel generator as a source of emergency power.

Date of issuance: April 25, 2011.

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

Amendment Nos.: Unit 1—195; Unit 2—183.

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

Date of initial notice in Federal Register: September 21, 2010 (75 FR 57529).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 25, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 30, 2010, as supplemented by letters dated August 23, 2010, and March 4, 2011.

Brief Description of amendments: The amendments revised the Technical Specifications by relocating specific surveillance frequency requirements to a licensee-controlled document using a risk-informed justification.

Date of issuance: April 29, 2011.

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

Amendment Nos.: Unit 1—273 and Unit 2—272.

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: August 10, 2010 (75 FR 48377).

The supplements dated August 23, 2010, and March 4, 2011, 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 29, 2011.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 5th day of May 2011.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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