2011-3721. 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 January 27, 2011, to February 10, 2011. The last biweekly notice was published on February 8, 2011 (76 FR 6830).

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), Section 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. 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, 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 [phone], to request (1) a digital 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 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 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 EIE, 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 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 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. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, 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. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: January 13, 2011.

Description of amendment request: The proposed amendment would modify the Facility Operating License (FOL) by deleting references to specific Safety Evaluation Reports (SER), Technical Specification (TS) Amendments, and Exemptions from License Condition 2.C(3), Fire Protection, and replacing them with the words “as supplemented.” This is an administrative amendment to the FOL.

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. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed FOL change is administrative and does not involve a plant or design function change. It has no effect on reactor operation or accident analyses, and thus, the proposed FOL change does not increase the probability or consequence of an accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed FOL change is administrative and does not involve a plant or design function change. Because the proposed amendment would not change the design, configuration, or method of operation of the plant, it would not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed FOL change is administrative and does not involve a plant or design function change. No design or safety margin is involved. Therefore, the proposed change does not involve a reduction in any 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: Nancy L. Salgado.

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

Date of amendment request: December 21, 2010.

Description of amendment request: The proposed amendment would revise Technical Specifications (TS) Section 3.6.A “Pressure and Temperature Limitation” to reflect the pressure and temperature (P-T) limits for the reactor coolant system through, approximately the end of the prospective 20-year renewed license period, depending on the plant capacity factor.

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

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

Response: No.

The proposed change revises the period of applicability of the P-T limits. The technical bases for the new period of applicability have been previously reviewed and approved by the NRC as discussed in the submittal. Because the applicable regulatory requirements continue to be met, the change does not significantly increase the probability of any accident previously evaluated. The proposed change provides the same assurance of RPV integrity as previously provided.

The change will require that the reactor pressure vessel and interfacing coolant system continue to be operated within their design, operational or testing limits. Also, the change will not alter any assumptions previously made in evaluating the radiological consequences of accidents.

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.

The proposed change does not involve a modification of the design of plant structures, systems, or components. The change will not impact the manner in which the plant is operated and will not degrade the reliability of structures, systems, or components important to safety as equipment protection features will not be deleted or modified, equipment redundancy or independence will not be reduced, supporting system performance will not be affected and no severe testing of equipment will be imposed. 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 accident previously evaluated.

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

Response: No.

Appendix G to 10 CFR 50 describes the conditions that require pressure-temperature (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 50 and applicable regulatory requirements ensures that adequate margins of safety are maintained and there is no significant 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 in 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, operation of VY in accordance with the proposed amendment will not involve a significant reduction in a margin to 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., 400 Hamilton Avenue, White Plains, NY 10601.

NRC Branch Chief: Nancy Salgado.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio

Date of amendment request: December 15, 2010.

Description of amendment request: The proposed amendment would modify the requirements for testing control rod scram times following fuel movement within the reactor pressure vessel by incorporating Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) change traveler TSTF-222-A, Revision 1.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (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 control rod drive system is not an initiator to any accident sequence analyzed in the PNPP Updated Final Safety Analysis Report (USAR), including Appendix 15C, “Anticipated Transients Without Scram (ATWS).” The proposed TS changes improve existing surveillance requirements by eliminating unnecessary control rob scram time testing, while continuing to provide adequate assurance of control rod performance for those control rods in core cells in which fuel is moved or replaced, or control rod maintenance was performed.

Historically, testing results indicate the control rod drive system is highly reliable. Since the fall 1996 implementation of Improved Technical Specifications, during 6036 control rod tests covering all 177 control rods, only 7 control rod tests (0.12 percent) yielded results slower than the required insertion time limit, and no control rods were inoperable as a result of scream time testing. All seven slow insertion time test results have been attributed to control rod scream solenoid pilot valves (SSPVs). These seven slow tests occurred prior to May 1999, and prior to a control rod SSPV upgrade program during which all 177 SSPV's were replaced.

As such, this type of change does not affect initiators of analyzed events and does not affect the mitigation of any accidents or transients.

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

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

Response: No.

The proposed TS changes do not involve a physical alteration of the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There are no setpoints affected by the changes at which protective or mitigative actions are initiated. The changes will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the USAR. This change does not alter assumptions made in the safety analysis and licensing basis. As such, no new failures modes are being introduced. Accordingly, the proposed changes do not create any new credible failure mechanisms, malfunction, or accident initiators not previously considered in PNPP design and licensing basis.

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

Response: No.

Margin of safety is related to the ability of the fission product barriers to perform their design functions during and following accident conditions. These barriers include the fuel cladding, the reactor coolant system, and the containment. This request does not involve a change to the fuel cladding, the reactor coolant system, or the containment.

The proposed TS changes associated with TSTF-222-1 modify current frequency requirements for scram time testing control rods following refueling outages and for control rod requiring testing due to work activities. Scram times for control rods not affected by fuel movement or control rod maintenance remain unaffected.

The proposed TS changes have no affect on any safety analysis assumptions or methods of performing safety analyses. The changes do not adversely affect system design or operational requirements, and the equipment continues to be tested in a manner and at a frequency necessary to provide confidence that the equipment can perform its intended safety functions.

Therefore, the proposed TS 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: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop. A-GO-15, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Robert D. Carlson.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-440, Perry Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio

Date of amendment request: December 15, 2010

Description of amendment request: The proposed amendment would revise the required testing frequency of Surveillance Requirement (SR) 3.1.4.2 from “120 days cumulative operation in MODE 1” to “200 days cumulative operation in MODE 1” by incorporating Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) change traveler TSTF-460, Revision 0.

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

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

Response: No.

The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The frequency of surveillance testing is not an initiator of any accident previously evaluated. The frequency of surveillance testing does not affect the ability to mitigate any accident previously evaluated, as the tested component is still required to be operable.

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

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

Response: No.

The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change does not result in any new or different modes of plant operation.

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

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

Response: No.

The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change continues to test the control rod scram time to ensure the assumptions in the safety analysis are protected.

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: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Robert D. Carlson.

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

Date of amendment request: October 15, 2010.

Description of amendment request: The proposed amendment would revise Operating License No. DPR-49 by modifying the License to delete the parent guarantee License Condition.

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 is an administrative change deleting the parent guarantee License Condition, as well as other minor editorial changes in format. Deletion of this License Condition does not involve any modifications to the safety-related structures, systems or components (SSCs). Deletion of this License Condition will not alter previously evaluated Final Safety Analysis Report (FSAR) design basis accident analysis assumptions, add any accident initiators, or affect the function of the plant safety-related SSCs as to how they are operated, maintained, modified, tested, or inspected. 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.

The proposed amendment only deletes the parent guarantee License Condition and makes other minor editorial changes. Deletion of this License Condition does not result in the need for any new or different FSAR design basis accident analysis. It does not introduce new equipment that could create a new or different kind of accident, and no new equipment failure modes are created. As a result, no new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed amendment. Therefore, the proposed amendment does not create a possibility for an accident of a new or different type than those previously evaluated.

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

Response: No.

The margin of safety is associated with the confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant pressure boundary, and containment structure) to limit the level of radiation to the public. The proposed amendment would not alter the way any safety-related SSC functions and would not alter the way the plant is operated. The amendment only involves deletion of the parent guarantee License Condition and minor editorial changes. The proposed amendment would not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. The proposed amendment would have no impact on the structural integrity of the fuel cladding, reactor coolant pressure boundary, or containment structure. Based on the above considerations, the proposed amendment would not degrade the confidence in the ability of the fission product barriers to limit the level of radiation to the public. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Ms. Marjan Mashhadi, Florida Power & Light Company, 801 Pennsylvania Avenue, NW., Suite 220, Washington, DC 20004.

NRC Branch Chief: Robert J. Pascarelli.

Indiana Michigan Power Company (the licensee), Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1 (DCCNP-1), Berrien County, Michigan

Date of amendment request: December 16, 2010.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 4.2.1, adding Optimized ZIRLO ^TM fuel rods to the fuel matrix in addition to Zircaloy or ZIRLO fuel rods that are currently in use. The proposed amendment would also add a Westinghouse topical report regarding Optimized ZIRLO ^TM as reference 8 in TS 5.6.5.b, which lists the analytical methods used to determine the core operating limits.

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 would allow the use of Optimized ZIRLO ^TM clad nuclear fuel in the reactors. The NRC approved topical report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A “Optimized ZIRLO ^TM,” prepared by Westinghouse Electric Company LLC (Westinghouse), addresses Optimized ZIRLO ^TM and demonstrates that Optimized ZIRLO ^TM has essentially the same properties as currently licensed ZIRLO ^TM. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO ^TM fuel cladding has been shown to meet all 10 CFR 50.46 acceptance criteria and, therefore, will not increase the consequences of an accident.

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.

Use of Optimized ZIRLO ^TM clad fuel will not result in changes in the operation or configuration of the facility. Topical Report WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material properties of Optimized ZIRLO ^TM are similar to those of standard ZIRLO ^TM. Therefore, Optimized ZIRLO ^TM fuel rod cladding will perform similarly to those fabricated from standard ZIRLO ^TM, thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident.

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

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

Response: No.

The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the Optimized ZIRLO ^TM are not significantly different from those of standard ZIRLO ^TM. Optimized ZIRLO ^TM is expected to perform similarly to standard ZIRLO ^TM for all normal operating and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference in Optimized ZIRLO ^TM material properties relative to standard ZIRLO ^TM could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO ^TM properties will be performed prior to the use of fuel assemblies with fuel rods containing Optimized ZIRLO ^TM. These LOCA analyses will demonstrate that the acceptance criteria of 10 CFR 50.46 will be satisfied when Optimized ZIRLO ^TM fuel rod cladding is implemented.

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: James M. Petro, Jr., Senior Nuclear Counsel, Indiana Michigan Power Company, One Cook Place, Bridgman, MI 49106.

NRC Branch Chief: Robert J. Pascarelli.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine Mile Point Nuclear Station Unit 1 (NMP1), Oswego County, New York

Date of amendment request: September 29, 2010.

Description of amendment request: The proposed amendment would revise the NMP1 Technical Specifications (TSs) Section 3/4.1.5, “Solenoid-Actuated Pressure Relief Valves (Automatic Depressurization System),” and 3/4.2.9, “Pressure Relief Systems—Solenoid-Actuated Pressure Relief Valves (Overpressurization),” to provide for an alternative means of testing the main steam electromatic relief valves (ERVs). Specifically, the proposed amendment would revise TS Surveillance Requirements (SRs) 4.1.5.a and 4.2.9.b to verify each ERV actuator strokes when manually actuated at least once each operating cycle. The functional testing requirements for the ERVs would be described in the Inservice Testing (IST) Program and controlled pursuant to TS Administrative Controls Section 6.5.4, “Inservice Testing Program.” The proposed change would allow demonstration of the capability of the valves to perform their safety function without requiring the ERVs to be cycled with reactor steam pressure while installed in the plant.

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 revises the TS Surveillance Requirements (SRs) to provide for an alternative means of testing the main steam ERVs. The ERVs perform automatic depressurization system (ADS) and overpressurization relief mode safety functions to mitigate the consequences of a small break loss of coolant accident (SBLOCA) and other accidents and transients. The ERVs are not considered an initiator for any accident previously evaluated except for the stuck-open ERV event, which is evaluated in Section XV-B.3.11 of the NMP1 Updated Final Safety Analysis Report (UFSAR). The proposed amendment would allow demonstration of the capability of the valves to perform their safety function through a series of tests, inspections, and maintenance activities without requiring the ERVs to be cycled with reactor steam pressure while installed in the plant, thereby eliminating the possibility of a stuck-open ERV event due to testing. Thus, the proposed amendment does not increase the probability of a stuck-open ERV event. The testing methodology, comprehensive inspections and preventive maintenance, and associated programmatic controls will provide an equivalent level of assurance that the ERVs are capable of performing their intended accident mitigation safety functions and, as such, will have no effect on the types or amounts of radiation released or the predicted offsite doses in the event of an accident. Accordingly, the proposed amendment does not alter the initial conditions, assumptions, or conclusions of any accident analysis.

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

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

Response: No.

The proposed amendment does not affect the assumed accident performance of the ERVs, or of any plant structure, system, or component previously evaluated. The proposed amendment does not involve the installation of new equipment, and installed equipment is not being operated in a new or different manner. The proposed amendment provides for an alternative means of testing the ERVs that does not involve 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 ERVs are capable of performing their accident mitigation safety functions. No setpoints are being changed that would alter the dynamic response of plant equipment. As such, the proposed amendment will not introduce any new failure modes.

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 provides for an alternative means of testing the ERVs, in that the testing requirements will be satisfied by a combination of required testing in accordance with the Inservice Testing Program (controlled in accordance with TS administrative controls) and the revised TS SRs. The proposed changes will provide a complete verification of the functional capability of the ERVs 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 ERVs 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 ERV leakage resulting from testing the ERVs 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: Carey W. Fleming, Senior Counsel, Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, Baltimore, MD 21202.

NRC Branch Chief: Nancy L. Salgado.

Northern States Power Company—Minnesota, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: February 3, 2011.

Description of amendment request: The proposed amendments would revise the Technical Specification (TS) 3.8.1, “AC Sources—Operating”, Surveillance Requirement 3.8.1.10 footnote, which concerns battery charger modifications to be installed during or prior to the Unit 1 2011 refueling outage. The proposed change will allow use of different battery charger modifications to those considered when the footnote was added to the TS.

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.

This license amendment request proposes to revise the footnote to the emergency diesel generator Technical Specification surveillance requirement for loss of offsite power with safety injection actuation. The proposed footnote revision removes some specific requirements for battery charger modifications but will continue to assure that the applicable emergency diesel generator and its associated battery charger perform their required safety functions.

The emergency diesel generators and their associated battery chargers are not accident initiators and therefore, these changes do not involve a significant increase [in] the probability of an accident.

The proposed changes to the Technical Specification footnote will assure that the emergency diesel generator and the associated battery charger continue to perform their required safety function. Since the emergency diesel generator and the associated battery charger will provide required electrical power as assumed in the accident analyses, the results of the previous accident analyses are not changed and the changes proposed in this license amendment request do not involve a significant increase in the consequences of an accident.

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

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

Response: No.

This license amendment request proposes to revise the footnote to the emergency diesel generator Technical Specification surveillance requirement for loss of offsite power with safety injection actuation. The proposed footnote revision removes some specific requirements for battery charger modifications but will continue to assure that the applicable emergency diesel generator and its associated battery charger perform their required safety functions.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. The proposed change does involve modification of plant battery chargers, however, failures of battery chargers has been previously considered and bounded by assuming one safety related train of equipment fails. The modified battery chargers do not create new failure modes or mechanisms and no new accident precursors are generated. Surveillance testing requirements for the emergency diesel generator and battery charger will continue to demonstrate that the Limiting Conditions for Operation are met and the system components are functional.

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

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

Response: No.

This license amendment request proposes to revise the footnote to the emergency diesel generator Technical Specification surveillance requirement for loss of offsite power with safety injection actuation. The proposed footnote revision removes some specific requirements for battery charger modifications but will continue to assure that the applicable emergency diesel generator and its associated battery charger perform their required safety functions.

The proposed Technical Specification footnote change does not affect the availability, operability, or performance of safety-related systems and components: The affected emergency diesel generator and its associated battery will continue to perform their safety functions. The ability of operable structures, systems, and components to perform their designated safety function is unaffected by this proposed change. The proposed change does not involve a significant reduction in a margin of safety because the proposed footnote changes do not reduce the margin of safety that exists in the present Technical Specifications or Updated Safety Analysis Report. The operability requirements of the Technical Specifications are consistent with the initial condition assumptions of the safety analyses and the surveillance testing requirements will continue to demonstrate the operability of the emergency diesel generator.

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 requests involve no significant hazards consideration.

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

NRC Branch Chief: Robert J. Pascarell.

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

Date of amendment request: November 10, 2010.

Description of amendment request: The change to the PPL Susquehanna, LLC (PPL) Unit 1 and Unit 2 Technical Specification (TS) Surveillance Requirement (SR) 3.4.3.1 “Safety/Relief Valves (S/RVs)” proposes a new safety function lift setpoint lower tolerance for the S/RVs. The proposed change will revise the lower tolerances from −3% to −5%. This change would be limited to the lower tolerances and does not affect the upper tolerances. This change only applies to the lower as-found tolerance and not to the as-left tolerance, which will remain unchanged at ±1% of the safety lift setpoint. The as-found tolerances are used for determining past operability and to increase sample sizes for S/RV testing should the upper tolerance be exceeded. There will be no revision to the actual setpoints of the valves installed in the plant due to this change.

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

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

Response: No.

This change has no influence on the probability or consequences of any accident previously evaluated. The lower setpoint tolerance change does not affect the operation of the valves and it does not change the as-left setpoint tolerance. The change only affects the lower tolerance for opening the valve and does not change the upper tolerance, which is the limit that protects from overpressurization.

The proposed action does not involve physical changes to the valves, nor does it change the safety function of the valves. The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions and no changes to existing structures, systems, or components.

The proposed action does not change any other behavior or operation of any S/RVs, and, therefore, has no significant impact on reactor operation. It also has no significant impact on response to any perturbation of reactor operation including transients and accidents previously analyzed in the Final Safety Analysis Report (FSAR).

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

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

Response: No.

The proposed lower setpoint tolerance change only affects the criteria to determine when an as-found S/RV test is considered to be acceptable. This change does not affect the criteria for the upper setpoint tolerance.

The proposed lower setpoint tolerance change does not adversely affect the operation of any safety-related components or equipment. Since the proposed action does not involve hardware changes, significant changes to the operation of any systems or components, nor change to existing structures, systems, or components, there is no possibility that a new or different kind of accident is created.

The proposed change does not involve physical changes to the S/RVs, nor does it change the safety function of the S/RVs. The proposed change does not require any physical change or alteration of any existing plant equipment. No new or different equipment is being installed, and installed equipment is not being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated. This change does not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alterations in the procedures that ensure the plant remains within analyzed limits are being proposed, and no changes are being made to the procedures relied upon to respond to an off-normal event as described in the FSAR. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis.

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

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

Response: No.

The proposed lower setpoint tolerance change only affects the criteria to determine when an as-found S/RV test is considered to be acceptable. This change does not affect the criteria for the upper setpoint tolerance. The TS setpoints for the S/RVs are not changed. The as-left setpoint tolerances are not changed by this proposed change.

The margin of safety is established through the design of the plant structures, systems, and components, 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 significantly impact the condition or performance of structures, systems, and components relied upon for accident mitigation.

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

NRC Branch Chief : Nancy L. Salgado.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia, and Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Unit 1 and 2, Appling County, Georgia

Date of amendment request: December 16, 2010.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) Section 2.0 “Safety Limits.” Specifically, the removal of the requirement to report a Safety Limit Violation, that is redundant to existing regulations, Title 10 of the Code of Federal Regulations (10 CFR), Part 50.36(c)(8) “Written Reports.” The proposed change is described in Technical Specification Task Force Traveler TSTF-5-A, Revision 1, “Delete Safety Limit Violation Notification Requirements,” (Agencywide Documents Access and Management System (ADAMS) Accession No. ML052010227), and was described in the Notice of Availability published in the Federal Register (FR) on November 4, 2005 (70 FR 67202). The proposed changes are consistent with the NRC-approved TSTF-5-A, Revision 1.

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 to remove the duplicative safety limit reporting, notification, and restart constraint requirements from the TS does not affect the plant or operation of the plant. The change simply removes duplicative information from the TS that is covered in the NRC regulations. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to remove the duplicative safety limit reporting, notification, and restart constraint requirements from the TS does not introduce any new accident scenarios, failure mechanisms, or limiting single failures. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safety-related system or component and does not challenge the performance or integrity of any safety related system. This change is considered an administrative action to remove duplicative reporting, notification, and restart constraint requirements. 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 changes are administrative and do not involve any reduction in a margin of safety. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed change has no adverse effect on any safety-related system or component and 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.

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) Public Electronic Reading Room on the Internet at the NRC Web site, 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].

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water Reactor, Vernon County, Wisconsin

Date of application for amendment: July 28, 2009, and supplemented August 7, 2009, May 19, 2010, and August 12, 2010.

Brief description of amendment: The amendment revises the La Crosse Boiling Water Rector (LACBWR) Technical Specifications, in support of the dry cask storage project at LACBWR.

Date of issuance: January 25, 2011.

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

Amendment No.: 71.

Facility Operating License No. DPR-7: This amendment revises the Technical Specifications.

Date of initial notice in Federal Register: October 6, 2009 (74 FR 51326).

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

No significant hazards consideration comments received: No.

Dominion Electric Kewaunee, Inc. Docket No. 50-305, Kewaunee Power Station (KPS), Kewaunee County, Wisconsin

Date of application for amendment: August 24, 2009 (Agencywide Documents and Management System (ADAMS) Accession No. ML092440398), as supplemented by letters dated October 22, 2009 (ADAMS Accession No. ML093070092), April 13, 2010 (ADAMS Accession Nos. ML101060517 and ML101040090), May 12, 2010 (ADAMS Accession No. ML101380399), July 1, 2010 (ADAMS Accession No. ML101890404), July 16, 2010 (ADAMS Accession No. ML102370370), August 18, 2010 (ADAMS Accession No. ML102371064), September 7, 2010 (ADAMS Accession No. ML102730383), September 8, 2010 (ADAMS Accession No. ML102580700), October 15, 2010 (ADAMS Accession No. ML102920037), and December 2, 2010 (ADAMS Accession No. ML103400328).

Brief description of amendment: This amendment converts the current technical specifications (CTSs) to the improved TSs (ITSs) and relocates certain requirements to other licensee-controlled documents. The ITSs are based on NUREG-1431, Rev. 3.0, “Standard Technical Specifications, Westinghouse Plants,” Revision 3.0; “NRC Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors,” dated July 22, 1993 (58 FR 39132); and 10 CFR 50.36, “Technical Specifications.” Technical Specification Task Force changes were also incorporated. The purpose of the conversion is to provide clearer and more readily understandable requirements in the TSs for KPS to ensure safe operation. In addition, the amendment includes a number of issues that were considered beyond the scope of NUREG-1431.

Date of issuance: February 2, 2011.

Effective date: As of the date of issuance and shall be implemented on or before February 23, 2011.

Amendment No.: 207.

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

Date of initial notice in Federal Register: December 15, 2009 (74 FR 66384). The supplements provided, contained clarifying information and did not expand the scope of the application as originally noticed.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 2, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 27, 2010.

Brief description of amendment:

The amendment revises Section 2.E. of the Palisades Nuclear Plant (PNP) Renewed Facility Operating License to remove the name of the former operator of the plant in the title of the PNP physical security plan and replace it with Entergy Nuclear. The change also removes the security plan revision number and the date the plan was submitted to the Nuclear Regulatory Commission.

Date of issuance: January 25, 2011.

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

Amendment No.: 241.

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

Public comments requested as to proposed no significant hazards considerations (NSHC): The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received.

Date of initial notice in Federal Register: November 18, 2010 (75 FR 70708), followed by the repeat biweekly notice in the Federal Register on January 25, 2011 (76 FR 4389).

The Commission's related evaluation of the amendment, state consultation, and final NSHC determination are contained in a Safety Evaluation dated January 25, 2011.

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.

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

Date of application for amendment: January 24, 2010, as supplemented by letters dated September 7 and November 4, 2010.

Brief description of amendment: This amendment request would revise the Technical Specifications (TSs) Section 1.0, Definitions, TS Section 3.6, Primary System Boundary Specifications 3.6.A, and TS Programs and Manuals Section 5.5, to include reference to the Pressure and Temperature Limits Report (PTLR). The proposed PTLR would include revised 43 effective full-power years pressure-temperature curves, neutron fluence, and adjusted reference temperature values.

Date of issuance: January 26, 2011.

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

Amendment No.: 234.

Facility Operating License No. DPR-35: The amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: April 6, 2010 (75 FR 17443). The supplemental letters dated September 7 and November 4, 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.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated January 26, 2011.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 13, 2010 as supplemented by letter dated. February 2, 2011.

Description of amendment request: The amendment would revise Technical Specification (TS) to update the Table of Contents and the Applicability and Objective portions of TS 4.12 as a result of changes made by License Amendment Nos. 230 and 239 and to revise wording in TS 3.7.A.8. The changes are considered administrative in nature and do not materially change any technical requirement.

Date of Issuance: February 9, 2011.

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

Amendment No.: 245.

Facility Operating License No. DPR-28: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register : June 29, 2010 (75 FR 37474). The supplement letter dated February 2, 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 this amendment is contained in a Safety Evaluation dated February 9, 2011.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 20, 2010.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.7.1.2, “Emergency Feedwater System,” Limiting Condition for Operation (LCO) 3/4.7.1.2, “Emergency Feedwater,” to clarify the acceptability of transitioning from Mode 4, Hot Shutdown, to Mode 3, Hot Standby, with the turbine-driven emergency feedwater (EFW) pump inoperable but available. The amendment granted an exception to TS LCO 3.0.4 and Surveillance Requirement 4.0.4 allowing entry into operational Mode 3 with TS LCO equipment, the turbine-driven EFW pump, associated with a shutdown action inoperable.

Date of issuance: January 31, 2011.

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

Amendment No.: 230.

Facility Operating License No. NPF-38: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register : September 21, 2010 (75 FR 57523).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2011.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: February 4, 2010 as supplemented by letters dated September 15, 2010, October 6, 2010, and December 13, 2010.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) 3.3.6.1, “Primary Containment Isolation Instrumentation,” “Table 3.3.6.1-1, “Primary Containment Isolation Instrumentation,” Function 6.a “Shutdown Cooling System Isolation, Recirculation Line Water Temperature—High,” to enable implementation with reactor pressure-based isolation instrumentation, for the Dresden Nuclear Power Station, Units 2 and 3.

Date of issuance: February 7, 2011.

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

Amendment Nos.: 236/229.

Facility Operating License Nos. DPR-19 and DPR-25: The amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register : April 20, 2010 (75 FR 20635).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 7, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendments: Dated October 5, 2009 as supplemented by letters dated June 10, November 23, December 14, and December 22, 2010, and January 11, 24, and 28, 2011.

Brief description of amendments: The proposed amendment would revise Technical Specification (TS) 4.3.1, “Criticality,” to address a non-conservative TS. The proposed change addresses the Boraflex degradation issue in the LSCS Unit 2 spent fuel storage racks by revising TS Section 4.3.1 to allow the use of NETCO-SNAP-IN® rack inserts in LSCS Unit 2 spent fuel storage rack cells as a replacement for the neutron absorbing properties of the existing Boraflex panels.

Date of issuance: January 28, 2011.

Effective date: As of the date of issuance and shall be implemented within 120 days after the end of Unit 2 refueling outage 13.

Amendment Nos.: 199 and 186.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Technical Specifications and License.

Date of initial notice in Federal Register : January 5, 2010 (75 FR 463). The June 10, November 23, December 14, and December 22, 2010, and January 11, 24, and 28, 2011, submittals contained clarifying information and did not change the NRC 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 January 28, 2011.

No significant hazards consideration comments received: No.

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

Date of amendment request: April 15, 2009, as supplemented by letters dated December 18, 2009, October 8, 2010 and January 10, 2011.

Brief description of amendment request: The amendment request and proposed exemption request were to incorporate a new methodology for the development of Reactor Coolant System (RCS) pressure-temperature limits into Technical Specification (TS) 5.6.4, “Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).” The amendment also requested a revision to the period of validity of the analysis for the low temperature overpressure protection (LTOP) system contained in Operating License Condition 2.C(3)(d). An associated revision to the Technical Specification Basis 3.4.12 “Low Temperature Overpressure Protection (LTOP)” supports the change to the operating license condition.

Date of issuance: January 28, 2011.

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

Amendment No.: 282.

Facility Operating License No. NPF-3: The amendment revised the TS and license.

Date of initial notice in Federal Register : June 16, 2009 (72 FR 28577). The supplemental letters contained clarifying information, did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 9, 2009.

Brief description of amendment: The amendment changes the NMP2 Technical Specification (TS) 3.8.4, “DC Sources—Operating,” to remove the Mode restrictions for performance of TS Surveillance Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division 3 direct current (DC) electrical power subsystem battery. The Division 3 DC electrical power subsystem feeds emergency DC loads associated with the high-pressure core spray (HPCS) system. These surveillances verify that the battery capacity is adequate for the battery to perform its required functions. The amendment removes these Mode restrictions for the Division 3 battery, thereby allowing performance of the SRs during Mode 1, 2, or 3 in conjunction with scheduled HPCS system outages.

Date of issuance: January 31, 2011.

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

Amendment No.: 136.

Renewed Facility Operating License No. NPF-069: The amendment revises the License and TSs.

Date of initial notice in Federal Register: April 6, 2010 (75 FR 17444).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2011.

No significant hazards consideration comments received: No.

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 application for amendments: February 2, 2010.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) Table 3.3.1-1 “Reactor Trip System Instrumentation [RTS],” Function 3, “Power Range Neutron Flux High Positive Rate.” Specifically, the revision added surveillance requirement 3.3.1.15 to verify the RTS response time is within limits.

Date of issuance: February 7, 2011.

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

Amendment Nos.: 159 and 141.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the licenses and the TSs.

Date of initial notice in Federal Register: May 4, 2010 (75 FR 23817). The supplement dated October 29, 2010, 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.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 7, 2011.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 30, 2010.

Brief description of amendment: The amendments revised the North Anna Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.”

Date of issuance: January 31, 2011.

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

Amendment Nos.: 262 and 243.

Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments changed the licenses and the technical specifications.

Date of initial notice in Federal Register: May 18, 2010 (75 FR 27833). The supplements dated August 30, 2010, and January 18, 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 January 31, 2011.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 10th day of February 2011.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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