2011-218. 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 December 16 to December 29, 2010. The last biweekly notice was published on December 28, 2010 (75 FR 81667).

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.

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://ehd.nrc.gov/​EHD_​Proceeding/​home.asp,, 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].

Calvert Cliffs Nuclear Power Plant, LLC, Docket No. 50-318, Calvert Cliffs Nuclear Power Plant, Unit 2, Calvert County, Maryland

Date of amendment request: October 4, 2010.

Description of amendment request: The proposed amendment revises Calvert Cliffs Technical Specification 5.5.16, “Containment Leakage Rate Testing Program” to allow a one-time extension of the Type A Integrated Leakage Rate test interval for no more than 5 years.

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?

No.

This proposed one-time extension of the Type A test interval from 10 years to 15 years does not increase the probability of an accident since there are no design or operating changes involved and the test is not an accident initiator. The proposed extension of the test interval does not involve a significant increase in the consequences of an accident since research documented in NUREG-1493 has found that, generically, fewer than 3% of the potential containment leak paths are not identified by Types B and C testing. Calvert Cliffs, through testing and containment inspections, also provides a high degree of assurance that the Containment will not degrade in a manner detectable only by a Type A test. Inspections required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code are performed to identify containment degradation that could affect leak tightness.

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

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

No.

This proposed one-time extension of the Type A test interval from 10 years to 15 years does not involve any design or operational changes that could lead to a new or different kind of accident from any accident previously evaluated. The test itself is not changing and will be performed after a longer interval. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

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 amendment involve a significant reduction in a margin of safety?

No.

The proposed one-time extension of the Type A test interval from 10 years to 15 years does not involve a significant reduction in the margin of safety of the containment's ability to maintain its integrity during a design basis accident. The generic study of the increase in the Type A test interval, NUREG-1493, concluded there is an imperceptible increase in the plant risk associated with extending the test interval out to 20 years. Further, the extended test interval would have a minimal effect on this risk since Types B and C testing detect 97% of potential leakage paths. For the requested change in the Calvert Cliffs Integrated Leakage Rate Test interval, it was determined that the risk contribution of leakage will increase 0.07% (based on change in offsite dose). This change is considered very small and does not represent a significant reduction in the margin of safety.

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

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

Attorney for licensee: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: November 8, 2010.

Description of amendment request: The proposed amendment would revise Technical Specifications (TS) to eliminate provisions allowing the High Pressure Coolant Injection (HPCI) system and the Reactor Core Isolation Cooling (RCIC) system to be aligned to the suppression pool when required instrument channels are inoperable. In this configuration, the HPCI and RICI systems would not be capable of mitigating some plant events. Also, an administrative change to the TS Table of Contents is proposed.

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 does not significantly increase the probability of an accident since it does not involve a change to any plant equipment that initiates a plant accident. The proposed amendment is more restrictive than the current TS in that it no longer allows the HPCI and RCIC systems to be aligned to the suppression pool when required instrument channels are inoperable. The change requires HPCI and RCIC to be declared inoperable within one hour when the associated trip functions are not operable. The change also updates the TS Table of Contents. The HPCI system is credited to mitigate small break loss-of-coolant accidents and the RCIC System is not credited for accident mitigation. The proposed change ensures the systems are aligned consistent with station analysis assumptions. 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 any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. The proposed amendment is more restrictive than the current technical specifications in that it no longer allows the HPCI and RCIC systems to be aligned to the suppression pool when required instrument channels are inoperable. The change requires HPCI and RCIC to be declared inoperable within one hour when the associated trip functions are not operable. The change also updates the TS Table of Contents. No new or different types of equipment will be installed and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety 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 is more restrictive than the current technical specifications in that it no longer allows the HPCI and RCIC systems to be aligned to the suppression pool when required instrument channels are inoperable. This ensures that safety margins established in station safety analysis are maintained. The proposed amendment does not involve a physical modification of the plant and does not change the design or function of any component or system. The proposed amendment is more restrictive than the current TS in that it no longer allows the HPCI and RCIC systems to be aligned to the suppression pool when required instrument channels are inoperable. The change requires the HPCI and RCIC systems to be declared inoperable within one hour when the associated trip functions are not operable. The change also updates the TS Table of Contents. This ensures analyzed safety margins are maintained. Therefore, operation of VY in accordance with the proposed amendment will not involve a significant reduction in the 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.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: June 25, 2010.

Description of amendment request: The amendment would revise the Oyster Creek Nuclear Generating Station Technical Specifications (TSs) governing actions to be taken if a single emergency diesel generator (EDG) is inoperable. Specifically, the proposed amendment would remove the requirement to test the other EDG daily. Instead, the licensee would be required to either test the other EDG once or determine that it is not inoperable due to a common cause failure.

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

1. [The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.]

The proposed changes are associated with the testing requirements of the two Emergency Diesel Generators (EDGs). The changes will eliminate unnecessary EDG testing requirements that contribute to potential mechanical degradation of the EDGs. The changes are based on the NRC guidance and recommendations provided in Generic Letter (GL) 93-05, “Line-Item Technical Specifications Improvement to Reduce Surveillance Requirements for Testing During Power Operation,” and GL 94-01, “Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators,” and are consistent with NUREG-1433, “Standard Technical Specifications, General Electric Plants, BWR/4.” These proposed changes implement a recommendation promulgated in NUREG-1366, “Improvements To Technical Specifications Surveillance Requirements” to curtail daily testing of remaining operable diesel generator[s] when one of the required diesel generators is inoperable except for when a valid concern (e.g., potential for common cause failure) is posed.

The probability of an accident is not increased by these changes because the EDGs are not initiators of any design basis event. Additionally, the proposed changes do not involve any physical changes to plant systems, structures, or components (SSC[s]), or the manner in which these SSC[s] are maintained [ ]. The surveillance testing required for the limiting condition for operation for one EDG inoperable will be eliminated for the operable EDG when the inoperability is not due to a common cause failure. The EDG reliability will thereby be potentially increased by reducing the stresses on the EDG caused by unnecessary testing while maintaining the requirement to perform a single test if a common cause failure potentially exists. The consequences of an accident will not be increased because the proposed changes to the EDG surveillance requirements will continue to provide a high degree of assurance that their operability is maintained.

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

2. [The proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.]

The proposed changes do not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Accordingly, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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

3. [The proposed changes do not involve a significant reduction in the margin of safety.]

The proposed changes modify the EDG accelerated testing requirements, are consistent with NRC guidance, and [potentially] improve EDG reliability. There are no changes being made to the current periodic surveillance requirements. The proposed changes do not impact the assumptions of any design basis accident, and do not alter assumptions relative to the mitigation of an accident or transient event.

Testing the operable EDG every day for the duration of the inoperable EDG inspection (i.e., 7 days) may be too excessive and may lead to degradation of the EDG and possibly result in [the] potential for unnecessary shutdowns. By reducing the possibility of degradation from this excessive testing, the margin of safety is [not significantly affected.]

The NRC staff has reviewed the licensee's analysis and, based on this review, and with the changes noted above in square brackets, 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. J. Bradley Fewell, Associate General Counsel, Exelon Generation Company LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Harold Chernoff.

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

Date of amendment request: February 26, 2010.

Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) by expanding the scope of the steam generator (SG) tubesheet inspections using the F* inspection methodology to the SG cold-leg tubesheet region for BVPS-2.

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

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

No. The proposed change modifies the BVPS-2 Technical Specifications to expand the scope of steam generator [SG] tubesheet inspections using the F* inspection methodology to the SG cold-leg tubesheet region based on WCAP-16385-P, Revision 1. Of the various accidents previously evaluated in the BVPS-2 Updated Final Safety Analysis Report (UFSAR), the proposed change only affects the SG tube rupture (SGTR) event evaluation and the postulated steam line break (SLB) accident evaluation. Loss-of-coolant accident (LOCA) conditions cause a compressive axial load to act on the tube. Therefore, since the LOCA tends to force the tube into the tubesheet rather than pull it out, it is not a factor in this amendment request. Another faulted load consideration is a safe shutdown earthquake (SSE); however, the seismic analysis of Model 51M SGs has shown that axial loading of the tubes is negligible during an SSE.

For the SGTR event, the required structural margins of the steam generator tubes will be maintained by the presence of the tubesheet. Tube rupture is precluded for cracks in the tube expansion region due to the constraint provided by the tubesheet. Therefore, Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR [pressurized-water reactor] Steam Generator Tubes,” margins against burst are maintained for both normal and postulated accident conditions.

The F* length supplies the necessary resistive force to preclude pullout loads under both normal operating and accident conditions. The contact pressure results from the tube expansion process used during manufacturing and from the differential pressure between the primary and secondary side. The proposed changes do not affect other systems, structures, components or operational features. Therefore, the proposed change results in no significant increase in the probability of the occurrence of an SGTR or SLB accident.

The consequences of an SGTR event are affected by the primary-to-secondary leakage flow during the event. Primary-to-secondary leakage flow through a postulated broken tube is not affected by the proposed change since the tubesheet enhances the tube integrity in the region of the expansion by precluding tube deformation beyond its initial expanded outside diameter. The resistance to both tube rupture and collapse is strengthened by the tubesheet in that region. At normal operating pressures, leakage from primary water stress corrosion cracking (PWSCC) below the F* distance is limited by both the tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint. Consequently, negligible normal operating leakage is expected from cracks within the tubesheet region.

SLB leakage is limited by leakage flow restrictions resulting from the crack and tube-to-tubesheet contact pressures that provide a restricted leakage path above the indications and also limit the degree of crack face opening compared to free span indications. The total leakage (i.e., the combined leakage for all such tubes) meets the industry performance criterion, plus the combined leakage developed by any other alternate repair criteria, and will be maintained below the maximum allowable SLB leak rate limit, such that off-site doses are maintained less than 10 CFR [Title 10 of the Code of Federal Regulation] [Part] 100 guideline values and the limits evaluated in the BVPS-2 UFSAR.

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

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

No. The proposed changes do not introduce any changes or mechanisms that create the possibility of a new or different kind of accident. Tube bundle integrity will continue to be maintained for all plant conditions upon implementation of the F* methodology to the cold-leg tubesheet region.

The proposed changes do not introduce any new equipment or any change to existing equipment. No new effects on existing equipment are created nor are any new malfunctions introduced.

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

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

No. The proposed changes maintain the required structural margins of the SG tubes for both normal and accident conditions. NRC Regulatory Guide (RG) 1.121 is used as the basis in the development of the F* methodology for determining that SG tube integrity considerations are maintained within acceptable limits. Regulatory Guide 1.121 describes a method acceptable to the NRC staff for meeting General Design Criteria 14, 15, 31, and 32. Regulatory Guide 1.121 describes the limiting safe conditions of tube wall degradation beyond which tubes with unacceptable cracking, as established by inservice inspection, should be removed from service or repaired. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.

For primarily axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. WCAP-16385-P, Revision 1, defines a length, F*, of degradation-free expanded tubing that provides the necessary resistance to tube pullout due to the pressure-induced forces (with applicable safety factors applied). Expansion of the application of the F* criteria to the cold-leg tubesheet region will preclude unacceptable primary-to-secondary leakage during all plant conditions. The methodology for determining leakage provides for large margins between calculated and actual leakage values in the F* criteria.

Plugging of the steam generator tubes reduces the reactor coolant flow margin for core cooling. Expansion of the F* methodology to the cold-leg tubesheet region at BVPS-2 will result in maintaining the margin of flow that may have otherwise been reduced by tube plugging.

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

The Nuclear Regulatory Commission (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, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: October 21, 2010.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) 2.1.1, “Reactor Core SLs,” by incorporating revised safety limit minimum critical power ratio (SLMCPR) values resulting from a plant-specific analysis performed for PNPP Cycle 14 core.

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 SLMCPR value will continue to ensure that during normal operation and abnormal operational transients, at 99.9 percent of all fuel rods in the core do not experience transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. The proposed TS changes do not involve any modifications or operational changes to system, structures, or components (SSC). The proposed TS changes do not affect any postulated accident precursors, do not affect any accident mitigating systems, and do no introduce any new accident initiation mechanisms. 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 any new modes of operation, any changes to setpoints, or any plant modifications. The proposed SLMCPR values do not result in the creation of any new precursors to an accident. Therefore, the proposed TS changes do not create the possibility of an accident of a different kind than previously evaluated.

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

Response: No.

The proposed SLMCPR value will continue to ensure that during normal operation and abnormal operational transients, at 99.9 percent of all fuel rods in the core do not experience transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. The proposed TS changes do involve modifications or operational changes that could adversely affect the function or performance of a SSC. The proposed TS changes do not affect any postulated accident precursors, do not affect any accident mitigating systems, and do not introduce any new accident initiation mechanisms. Therefore, the proposed TS changes do not involve a significant reduction in margin of safety.

The U.S. Nuclear Regulatory Commission (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.

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

Date of amendment request: August 5, 2010.

Description of amendment request: The proposed amendment would modify the Callaway Plant, Unit 1, Technical Specifications (TS) by relocating specific surveillance frequencies to a licensee-controlled program with the guidance of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.”

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

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

Response: No.

The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program [(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 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 assumptions and current plant operating practice.

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

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

Response: No.

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

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

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

Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: November 15, 2010.

Description of amendment request: The proposed amendments would delete license conditions that impose specific requirements for the decommissioning trust agreement. In lieu of the license conditions, ZionSolutions will directly implement the requirements of 10 CFR 50.75(h)(1) through (h)(3). ZionSolutions will provide a revised trust agreement as required by 10 CFR 50.75(h)(1)(iii) within 60 days of NRC approval of this proposal. The licensee has stated that the trust agreement will conform with 10 CFR 50.75(h) and ZionSolutions will take no action under the existing trust agreement in the interim that would be inconsistent with the provisions of the regulation.

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?

Response: No.

The proposed amendments alter the requirements for the decommissioning trust fund. These revisions of the financial assurance requirements do not involve any changes to any structures, systems or components (SSCs) or any method of operation, maintenance or testing. The proposed amendments will continue to provide assurance that adequate decommissioning funding is maintained. Changes to the terms of the trust fund will not alter previously evaluated Defueled Safety Analysis Report (DSAR) design basis accident assumptions, add any accident initiators, or affect the function of the plant SSCs as to how they are operated, maintained, modified, tested, or inspected.

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 evaluated?

Response: No.

Implementation of the proposed changes to decommissioning trust fund requirements will have no impact upon the design function of any SSC. Modifying the precise language of the administrative controls on the fund in the trust agreement does not result in the need for any new or different DSAR design basis accident analyses. 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 the proposed amendments.

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?

Response: No.

The margin of safety is associated with the confidence in the ability of the fission product barriers to limit the level of radiation to the public. The proposed amendments would not alter any SSC functions and would not alter the way the plant is operated. The amendments do not alter the way in which financial assurance for decommissioning is achieved. The proposed amendments would not introduce any new uncertainties associated with any safety limit. The proposed amendments would have no impact upon the structural integrity of the fuel cladding or any other barrier to fission product release. There would be no reduction in the effectiveness 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 the margin of safety.

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

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

NRC Branch Chief: Bruce Watson.

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, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 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].

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3, Maricopa County, Arizona

Date of application for amendment: April 8, 2010.

Brief description of amendment: The amendments deleted redundant reporting and operational restriction provisions from Technical Specification (TS) Section 2.2, “Safety Limit Violations,” consistent with Technical Specification Task Force (TSTF) change traveler TSTF-5-A, Revision 1, “Delete Safety Limit Violation Notification Requirements,” and replaced plant-specific titles with generic titles in TS Section 5.2.1, “Onsite and Offsite Organizations,” consistent with TSTF-65-A, Revision 1, “Use of Generic Titles for Utility Positions.”

Date of issuance: December 29, 2010.

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

Amendment No.: Unit 1—183; Unit 2—183; Unit 3—183.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revised the Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: July 27, 2010 (75 FR 44022).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 29, 2010.

No significant hazards consideration comments received: No.

Carolina Power and Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of application for amendment: July 21, 2009, as supplemented March 3 and July 28, 2010.

Brief description of amendment: The amendment revises Technical Specification (TS) Section 6.9.1.6 to add NRC approved Topical Report (TR) EMF-2310(P)(A), “SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors,” to the Core Operating Limits Report methodologies list. This change will allow the use of thermal-hydraulic analysis code S-RELAP5 for Final Safety Analysis Report (FSAR) Chapter 15 non-loss-of-coolant accident (LOCA) transients in the HNP safety analyses. TR EMF-2310(P)(A), Revision 0, was approved by the NRC on May 11, 2001, for the application of the S-RELAP5 thermal-hydraulic analysis computer code to FSAR Chapter 15 non-LOCA transients. EMF-2310(P)(A), Revision 1, approved by the NRC on May 19, 2004, updated Section 5.6 of the TR.

Date of issuance: December 23, 2010.

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

Amendment No.: 135.

Renewed Facility Operating License No. NPF-63: The amendment revises the TSs and facility operating license.

Date of initial notice in Federal Register: November 10, 2009 (74 FR 58060). The supplements dated March 3, and July 28, 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 December 23, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 14, 2009, as supplemented by letters dated September 8, 2010, and October 28, 2010.

Brief description of amendments: The amendments revised the Technical Specifications by revising Surveillance Requirements 3.8.4.3 and 3.8.4.6. These TS SRs address battery connection resistance values.

Date of issuance: December 20, 2010.

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

Amendment Nos.: 262, 258.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register: August 10, 2010 (75 FR 48375). The supplements dated September 8, 2010, and October 28, 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 the amendments is contained in a Safety Evaluation dated December 20, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 14, 2009, as supplemented by letters dated September 8, 2010, and October 28, 2010.

Brief description of amendments: The amendments revised the Technical Specifications by revising Surveillance Requirements 3.8.4.2 and 3.8.4.5. These TS SRs address battery connection resistance values.

Date of issuance: December 20, 2010.

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

Amendment Nos.: 260, 240.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the licenses and the technical specifications.

Date of initial notice in Federal Register: August 10, 2010 (75 FR 48375). The supplements dated September 8, 2010, and October 28, 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 the amendments is contained in a Safety Evaluation dated December 20, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 16, 2009, as supplemented by letters dated April 26 and October 25, 2010.

Brief description of amendment: The amendments revise Technical Specifications Section 5.6.5, “Core Operating Limits Report,” to replace the existing reference for the large break loss-of-coolant accident (LOCA) analysis methodology with a reference to WCAP-16009-P-A, Revision 0, “Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method,” January 2005.

Date of issuance: December 21, 2010.

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

Amendment Nos.: Braidwood Unit 1—164; Braidwood Unit 2—164; Byron Unit No. 1—170; and Byron Unit No. 2—170.

Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: The amendments revise the TSs and Licenses.

Date of initial notice in Federal Register: February 23, 2010 (75 FR 8141). The supplemental letters dated April 26, and October 25, 2010, 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 amendments is contained in a Safety Evaluation dated December 21, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 14, 2009, as supplemented on July 30, 2010.

Brief description of amendments: Amendment modifies Technical Specification (TS) 3/4 .4.10 “Structural Integrity,” in Unit 1 (TS 3/4.4.11 in Unit 2), TS 3.3.3.8, “Accident Monitoring Instrumentation,” in Unit 1 (TS 3.3.3.6 in Unit 2), TS 6.4.1, “Training,” in Units 1 and 2, and several administrative changes in the TSs for both units . The changes delete the Structural Integrity TS, update Accident Monitoring Instrumentation requirements and make various administrative TS changes.

Date of Issuance: December 28, 2010.

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

Amendment Nos.: 210, 159.

Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the TSs.

Date of initial notice in Federal Register: April 20, 2010 (75 FR 20638). The supplement dated July 30, 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 amendments is contained in a Safety Evaluation dated December 28, 2010.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-354, 50-272 and 50-311, Hope Creek Generating Station and Salem Nuclear Generating Station, Unit 1 and 2, Salem County, New Jersey

Date of application for amendments: March 25, 2010.

Brief description of amendments: The amendments revise the Technical Specifications (TSs) associated with reactor coolant system (RCS) structural integrity requirements for Hope Creek Generating Station (HCGS) and Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. Specifically, the amendments revise the TSs to: (1) Delete the RCS structural integrity requirements contained in HCGS TS 3/4.4.8, Salem Unit 1 TS 3/4.4.10, and Salem Unit 2 TS 3/4.4.11; (2) relocate the augmented inservice inspection requirements for the reactor coolant pump flywheel, currently contained in Salem Unit 1 surveillance requirement (SR) 4.4.10.1.1 and Salem Unit 2 SR 4.4.11.1, to a new program in TS 6.8.4.k; and (3) delete the augmented inservice inspection program requirements for the steam generator channel heads currently contained in Salem Unit 1 SR 4.4.10.1.2.

Date of issuance: December 15, 2010.

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

Amendment Nos.: 186, 298 and 281.

Facility Operating License Nos. NPF-57, DPR-70 and DPR-75: The amendments revised the TSs and the Licenses.

Date of initial notice in Federal Register: June 15, 2010 (75 FR 33843).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 15, 2010.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendment: January 26, 2010 (TS 09-05).

Brief description of amendment: The amendments revised the Technical Specification (TS) Table 3.3-1, “Reactor Trip System Instrumentation,” Functional Unit 5, “Intermediate Range, Neutron Flux,” to resolve an oversight regarding the operability requirements for the intermediate range neutron flux channels. The amendments added an action to TS Table 3.3-1 to define that the provisions of Specification 3.0.3 are not applicable above 10 percent of thermal rated power with the number of operable intermediate range neutron flux channels two less than the minimum channels operable requirement.

Date of issuance: December 21, 2010.

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

Amendment Nos.: 328, 321.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: March 23, 2010 (75 FR 13791).

The Commission's related evaluation of the amendment is contained in a safety evaluation dated December 21, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 10, 2010.

Brief Description of amendments: These amendments revise the Technical Specifications 5.2.1, “Fuel Assemblies,” to add Optimized ZIRLO^TM as an acceptable fuel rod cladding material. In addition, the amendments propose adding the Westinghouse topical report for Optimized ZIRLO^TM to the analytical methods used to determine the core operating limits listed in TS 6.2.C.

Date of issuance: December 22, 2010.

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

Amendment Nos.: 271, 270.

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 27, 2010 (75 FR 52781).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 22, 2010.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 4, 2009, as supplemented by letters dated March 25 and November 17, 2010.

Brief description of amendment: The amendment revised the approved fire protection program as described in the Wolf Creek Generating Station (WCGS) Updated Safety Analysis Report (USAR). Specifically, a deviation from certain technical requirements of Title 10 of the Code of Federal Regulations (10 CFR), part 50, appendix R, section III.G.2, as documented in Appendix 9.5E of the WCGS USAR, was requested regarding the use of operator manual actions in lieu of meeting circuit separation protection criteria. Table 3-1 of the submittal dated March 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090771269), identified the proposed feasible and reliable operator manual actions requested for permanent approval and Table 3-2 of the submittal identified the proposed feasible operator manual actions requested for approval on an interim basis. The interim operator actions will be eliminated with the implementation of associated design change package. The amendment also revised license condition 2.C.(5)(a) to include the deviation approved by the amendment request.

Date of issuance: December 16, 2010.

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

Amendment No.: 191.

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

Date of initial notice in Federal Register: April 21, 2009 (75 FR 18258). The supplemental letters dated March 25 and November 17, 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 December 16, 2010.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 16, 2009, as supplemented by letter dated August 26, 2010.

Brief description of amendment: The amendment revised the battery acceptance criteria in Technical Specification 3.8.4, “DC [Direct Current] Sources—Operating,” Surveillance Requirements (SRs) 3.8.4.2 and 3.8.4.5. Specifically, the amendment modified SR 3.8.4.2 and SR 3.8.4.5 by providing limits for inter-cell, inter-tier/inter-bank/terminal, and field jumper connections for 60-cell, 59-cell, and 58-cell configurations.

Date of issuance: December 20, 2010.

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

Amendment No.: 192.

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

Date of initial notice in Federal Register: April 6, 2010 (75 FR 17448). The supplemental letter dated August 26, 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 December 20, 2010.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 30th day of December 2010.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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