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Notice

Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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Information about this document as published in the Federal Register.

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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 staff) 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 29, 2009, to February 11, 2009. The last biweekly notice was published on February 10, 2009 (74 FR 6662).

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 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, Rulemaking, Directives and Editing Branch, 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. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 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, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Start Printed Page 8282Documents 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 within 60 days, 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 set forth the specific contentions which the petitioner/requestor 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 petitioner/requestor 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 petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor 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 petitioner/requestor to relief. A petitioner/requestor 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, and the Commission has not made a final determination on the issue of no significant hazards consideration, 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, which the NRC promulgated in August 28, 2007 (72 FR 49139). 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 a waiver in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at hearingdocket@nrc.gov, or by calling (301) 415-1677, 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/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms Viewer TM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms Viewer TM is free and is available at http://www.nrc.gov/​site-help/​e-submittals/​install-viewer.html. 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.

Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE 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 EIE 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 may seek assistance through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html or by calling the NRC electronic filing Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays. The electronic filing Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at MSHD.Resource@nrc.gov.

Participants who believe that they have a good cause for not submitting documents electronically must file a motion, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to Start Printed Page 8283submit 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.

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://www.ehd.nrc.gov/​EHD_​Proceeding/​home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a 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.

For further details with respect to this amendment action, see the application for amendment which is available for public inspection at the Commission's 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 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 pdr@nrc.gov.

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

Date of amendment request: September 29, 2008, as supplemented by letter dated January 16, 2009.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) Sections 5.6.1.3.a and 5.6.1.3.b to incorporate the results of a new criticality analysis. Specifically the TSs would be revised to add new requirements for the Boiling Water Reactor (BWR) spent fuel storage racks containing Boraflex in Spent Fuel Pools A and B. The requirements for the BWR spent fuel racks as currently contained in TS 5.6.1.3 would be revised to specify applicability to the spent fuel storage racks containing Boral in Spent Fuel Pool B.

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 activity changes the design basis of the BWR Boraflex storage racks, but does not make physical changes to the facility. The change to TS Section 5.6.1.3 (BWR Storage Racks in Pools A and B), which is an update to the administrative controls for maintaining the required boron concentration in the Boraflex BWR spent fuel storage racks located in Pools A and B, does not modify the facility.

The accidents currently analyzed in the FSAR [Final Safety Analysis Report] applicable to the proposed activity are fuel handling accidents. These accidents include dropping a fuel assembly onto the top of a fuel rack or in the space between a rack and the pool wall. These events are caused either by personnel error or equipment malfunction.

Based on the new criticality analysis, revised acceptance criteria are needed to ensure the criticality safety of fuel storage in BWR Boraflex racks in Pools A and B. Similar administrative controls were previously placed on fuel stored in the PWR [Pressurized Water Reactor] Boraflex racks in Pools A and B. These changes will eliminate the dependence on the Boraflex absorber in the BWR storage racks. These changes do not impact the probability of having a fuel handling accident and do not impact the consequences of a fuel handling accident.

Therefore, this amendment 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.

These revised acceptance criteria applicable to the irradiated fuel stored in the BWR Boraflex racks in Pools A and B are being added to TS Section 5.6.1.3.a.

The proposed change does not result in any credible new failure mechanisms, malfunctions or accident initiators not considered in the original design and licensing bases.

Detailed analyses have been performed to ensure a criticality accident in Pools A and B is not a credible event. The events that could lead to a criticality accident are not new. These events include a fuel mispositioning event, a fuel drop event, and a boron dilution event. The proposed changes do not impact the probability of any of these events.

The detailed criticality analyses performed demonstrates that criticality would not occur following any of these events. Even in a more likely event, such as a fuel mispositioning event, the acceptance criteria for keff [the effective multiplication factor] remains less than or equal to 0.95. In the unlikely event that the spent fuel storage pool boron concentration were reduced to zero, keff remains less than 1.0. A criticality accident is considered “not credible” and the proposed action does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Therefore, the proposed change will 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 a margin of safety?

Response: No.

Incorporation of the revised criteria for fuel stored in the BWR Boraflex racks in Pools A and B do not involve a reduction in the margin of safety. The updated fuel storage condition continues to meet keff <0.95 with credit for soluble boron and keff < 1.0 when flooded with unborated water.

The proposed changes for storage of irradiated fuel in BWR Boraflex racks in Start Printed Page 8284Pools A and B continues to provide the controls necessary to ensure a criticality event could not occur in the spent fuel storage pool. The acceptance criteria are consistent with the acceptance criteria specified in 10 CFR 50.68, which provide an acceptable margin of safety with regard to the potential for a criticality event.

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

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

Attorney for licensee: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Branch Chief: Thomas H. Boyce.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: August 28, 2008, as supplemented by letter dated January 19, 2009.

Description of amendment request: The proposed amendment would implement the Technical Specification Task Force Standard Technical Specification Change Traveler 449, Revision 4 inspection requirements for the replacement once through steam generators (OTSGs) that are being installed during the Crystal River Unit 3 Nuclear Generating Plant fall 2009 refueling outage. The replacement OTSGs differ from the existing OTSGs in that the tube material is Alloy 690 thermally treated in the replacements versus Alloy 600 in the existing OTSGs. Additionally, this amendment would remove inspection requirements that are designated for specific damage conditions in the existing OTSGs, remove tube repair techniques approved by the license amendment No. 233, dated May 16, 2007, for the existing OTSGs, and remove inspection and reporting requirements specific to those repair techniques.

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

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

The proposed change for replacement OTSGs continues to implement the current OTSG Program that includes performance criteria which provide reasonable assurance that the replacement OTSG tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cooldown and all anticipated transients included in the design specifications). This change removes repair criteria from the OTSG Program that were approved by previous License Amendments for the existing Steam Generators which are not applicable to the replacement OTSGs. It removes references to use of repairs and reporting of repair results in other Technical Specification sections. This change removes inspection requirements that are designated for specific damage conditions in the existing OTSGs.

The change also revises the inspection interval for 100% inspections of OTSG tubes and the maximum interval for inspection of a single OTSG consistent with Technical Specification Task Force item 449 for the Alloy 690 tube material in the replacement OTSGs. The revised inspection requirements are based on properties and experience with the improved Alloy 690 tube material. The revised inspection requirements will result in the same outcome that OTSG tube integrity will continue to be maintained.

This change continues to implement steam generator performance criteria for tube structural integrity, accident induced leakage, and operational leakage for the replacement OTSGs. Meeting the performance criteria provides reasonable assurance that the replacement OTSG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident. The performance criteria are only a part of the OTSG program required by the existing ITS [Improved Technical Specification]. The program, defined by NEI [Nuclear Energy Institute] 97-06, Steam Generator Program Guidelines, includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring. These features will continue to be implemented as they are currently approved. The proposed changes do not, therefore, significantly increase the probability of an accident previously evaluated.

The consequences of design basis accidents are, in part, functions of the DOSE EQUIVALENT I-131 in the primary coolant and the primary to secondary LEAKAGE rates resulting from an accident. Therefore, limits are included in the plant technical specifications for operational leakage and for DOSE EQUIVALENT I-131 in the primary coolant to ensure the plant is operated within its analyzed condition. The analysis of the limiting design basis accident assumes that the primary to secondary leak rate, after the accident, is 1 gallon per minute with no more than 150 gallons per day in any one SG [steam generator], and that the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the TS [technical specification] values before the accident. The proposed change to the OTSG inspection program does not affect the design of the OTSGs, their method of operation, operational leakage limits, or primary coolant chemistry controls. The proposed change does not adversely impact any other previously evaluated design basis accident. In addition, the proposed changes do not affect the consequences of a Main Steam Line Break, rod ejection, or a reactor coolant pump locked rotor event, or other previously evaluated accident. Therefore, the proposed change does not affect the consequences of a Steam Generator Tube Rupture accident and the probability of such an accident is unchanged.

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

The proposed license amendment does not affect the design of the OTSGs, their method of operation, or primary or secondary coolant chemistry controls. In addition, the proposed amendment does not impact any other plant system or component. The change modifies existing OTSG inspection requirements for 100% inspection intervals, but establishes inspection requirements that are considered equivalent based on properties and experience with improved materials. Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated.

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

The steam generator tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the steam generator tubes are Start Printed Page 8285unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the steam generator tubes isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a steam generator is maintained by ensuring the integrity of its tubes. Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change to the OTSG inspection program does not affect tube design or operating environment. The existing OTSG Program is maintained in this change. The repair criteria that are being removed are specific to the existing OTSGs and are not applicable to the replacement OTSGs. In the case of the roll repair that is being removed, it potentially leads to additional cracking over subsequent operating cycles due to tube cold working during the re-roll. If tube defects are detected that exceed limits in the new generators, then the tube will be removed from service. This is considered a more effective means for removing defects than repairs. For the above reasons, the margin of safety is not changed and overall plant safety will be enhanced by the proposed change to the ITS. Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

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 T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Branch Chief: Thomas H. Boyce.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida

Date of amendment request: November 6, 2008.

Description of amendments request: The proposed change would revise the Crystal River Unit 3 (CR-3) Improved Technical Specifications Surveillance Requirements (SRs); SR 3.8.1.2, SR 3.8.1.6, and SR 3.8.1.10 to restrict the voltage and frequency limits for all Emergency Diesel Generator (EDG) starts. The steady state voltage limits would be revised to be more restrictive (plus or minus 2 percent of the nominal voltage) to accurately reflect the appropriate calculation and the way the plant is operated and tested. The steady state frequency limits would be revised to be more restrictive (plus or minus 1 percent for all EDG starts) to ensure compliance with the plant design bases and the way the plant is operated. These changes would ensure that the EDGs are capable of supplying power, with the correct voltage and frequency, to the required electrical loads.

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

The LAR [license amendment request] proposes to provide more restrictive steady state voltage and frequency limits for the Emergency Diesel Generators (EDGs). The voltage band is going from a range of greater than or equal to 3933 V [volts] but less than or equal to 4400 V, to greater than or equal to 4077 V but less than or equal to 4243 V. The proposed limits are +/−2% [percent] around the nominal safety-related bus voltage of 4160 V. The Frequency Limits are going from a 2% tolerance band to a 1% tolerance band around the nominal frequency of 60 Hz [hertz] (59.4 Hz to 60.6 Hz) for all starts of the EDGs.

The EDGs are a safety-related system that functions to mitigate the impact of an accident with a concurrent loss of offsite power. A loss of offsite power is typically a significant contributor to postulated plant risk and, as such, onsite AC [alternating current] generators have to be maintained available and reliable in the event of a loss of offsite power event. The EDGs are not initiators for any analyzed accident, therefore; the probability for an accident that was previously evaluated is not increased by this change. The revised, voltage and frequency limits will ensure the EDGs will remain capable of performing their design function.

The consequences of an accident refer to the impact on both plant personnel and the public from any radiological release associated with the accident. The EDG supports equipment that is supposed to preclude any radiological release. More restrictive voltage and frequency limits for the output of the EDG restores design margin, and provides assurance that the equipment supplied by the EDG will operate correctly and within the assumed timeframe to perform their mitigating functions.

Until the proposed CR-3 ITS [Improved Technical Specifications] EDG voltage and frequency limits are approved by the NRC, administratively controlled limits have been established in accordance with NRC Administrative Letter 98-10 to ensure all EDG mitigation functions will be performed, per design, in the event of a loss of offsite power. These administrative limits have been determined as acceptable and have been incorporated into the surveillance test procedures under the provisions of 10 CFR 50.59. Periodic testing has been performed with acceptable results. Since EDGs are mitigating components and are not initiators for any analyzed accident, no increased probability of an accident can occur. Since administrative limits will ensure the EDGs will perform as designed, consequences will not be significantly affected.

2. Does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Administrative voltage limits were established using verified design calculations and the guidance of NRC Administrative Letter 98-10. These administrative limits will ensure the EDGs will perform as designed. No new configuration is established by this change. The administrative limits for the EDG frequency were determined to be sufficient to account for measurement and other uncertainties.

The proposed amendment will place the administrative limits into the CR-3 ITS. The more restrictive voltage and frequency limits will provide additional assurance that the EDG can provide the necessary power to supply the required safety-related loads during an analyzed accident.

The proposed ITS voltage and frequency limits restore the EDG capability to those analyzed by engineering calculation. No new configuration is established. Therefore, no new or different kind of accident from any previously evaluated can be created.

3. Does not involve a significant reduction in a margin of safety.

The LAR proposes to provide more restrictive steady state voltage and frequency limits for the EDGs. The change in the acceptance criteria for specific surveillance testing provides assurance that the EDGs will be capable of performing their design function. Previous test history has shown that the new limits are well within the Start Printed Page 8286capability of the EDGs and are repeatable. The “as-left” settings for voltage and frequency will be adjusted such that they remain within a tight band and this ensures that the “as-found” settings will be in an acceptable tolerance band.

The proposed ITS limits on voltage and frequency will ensure that the EDG will be able to perform all design functions assumed in the accident analyses. Administrative limits are in place to ensure these parameters remain within analyzed limits. As such, 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 T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, NC 27602.

NRC Branch Chief: Thomas H. Boyce.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

Date of amendment request: September 26, 2008.

Description of amendment request: The amendments would revise the Technical Specifications to adopt Nuclear Regulatory Commission (NRC)-approved Revision 3 to Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF-448, “Control Room Envelope Habitability.” The proposed amendments include changes to the TS requirements related to control room envelope (CRE) habitability in TS 3/4.7.5, “Control Room Emergency Ventilation System (CREVS),” and TS Section 6.8, “Administrative Controls—Procedures and Programs.” In addition, the improvements to TSTF-448, Revision 3 as recommended in TSTF-508, Revision 0, “Revise Control Room Envelope Habitability Actions to Address Lessons Learned from TSTF-448 Implementation,” have been incorporated as appropriate.

The NRC staff published a notice of opportunity for comment in the Federal Register on October 17, 2006 (71 FR 61075), on possible amendments adopting TSTF-448, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on January 17, 2007 (72 FR 2022). The licensee affirmed the applicability of the following NSHC determination in its application dated September 26, 2008.

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

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

The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility. The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change revises the TS for the CRE emergency ventilation system, which is a mitigation system designed to minimize unfiltered air leakage into the CRE and to filter the CRE atmosphere to protect the CRE occupants in the event of accidents previously analyzed. An important part of the CRE emergency ventilation system is the CRE boundary. The CRE emergency ventilation system is not an initiator or precursor to any accident previously evaluated.

Therefore, the probability of any accident previously evaluated is not increased. Performing tests to verify the operability of the CRE boundary and implementing a program to assess and maintain CRE habitability ensure that the CRE emergency ventilation system is capable of adequately mitigating radiological consequences to CRE occupants during accident conditions, and that the CRE emergency ventilation system will perform as assumed in the consequence analyses of design basis accidents. Thus, the consequences of any accident previously evaluated are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not impact the accident analysis. The proposed change does not alter the required mitigation capability of the CRE emergency ventilation system, or its functioning during accident conditions as assumed in the licensing basis analyses of design basis accident radiological consequences to CRE occupants. No new or different accidents result from performing the new surveillance or following the new program. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. The proposed change does not alter any safety analysis assumptions and is consistent with current plant operating practice. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the 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. The proposed change does not affect safety analysis acceptance criteria. The proposed change will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory measures. The proposed change does not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Branch Chief: Thomas H. Boyce.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: January 5, 2009.

Description of amendment request: The proposed amendment would modify Technical Specifications (TS) requirements for mode change limitations in accordance with Revision 9 of Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) change TSTF-359, “Increase Flexibility in Mode Restraints.”

In a Federal Register notice dated August 2, 2002 (67 FR 50475), the NRC staff issued a notice of opportunity to comment on a model safety evaluation and model no significant hazards consideration (NSHC) determination for proposed license amendments adopting TSTF-359 using the consolidated line item improvement process (CLIIP).

In a Federal Register notice dated April 4, 2003 (68 FR 16579), the NRC staff issued a notice of availability of a model application for proposed license amendments adopting TSTF-359 using the CLIIP. The notice also included a revised model safety evaluation and a Start Printed Page 8287model NSHC determination. In its application dated January 5, 2009, the licensee affirmed the applicability of the model NSHC determination which is presented below.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC is presented below:

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO [Limiting Condition for Operation] 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve the physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS Limiting Conditions for Operation (LCO). The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The new change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

Based upon the reasoning presented above 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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: Harold K. Chernoff.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

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

Date of amendment request: January 5, 2009.

Description of amendment request: The proposed amendments would delete Section 2.F of the Facility Operating License (FOL) for Hope Creek Generating Station (Hope Creek) and Section 2.I of the FOL for Salem Nuclear Generating Station (Salem) Unit No. 2. The FOL sections being deleted require reporting of violations of the requirements in Section 2.C of the respective FOLs. The proposed amendments would also delete Technical Specification (TS) 6.9.3 for Hope Creek, Salem Unit No. 1 and Salem Unit No. 2. These TSs contain a reporting requirement that is duplicative of Nuclear Regulatory Commission (NRC) regulations.

The NRC staff issued a “Notice of Opportunity to Comment on Model Safety Evaluation on Elimination of Typical License Condition Requiring Reporting of Violations of Section 2.C of Operating Licensing Using the Consolidated Line Item Improvement Process,” in the Federal Register on August 29, 2005 (70 FR 51098). The notice included a model safety evaluation (SE) and a model no significant hazards consideration (NSHC) determination. On November 4, 2005, the NRC staff issued a notice in the Federal Register (70 FR 67202) announcing that the model SE and model NSHC determination may be referenced in plant-specific applications to adopt the changes. In its application dated January 5, 2009, the licensee affirmed the applicability of the model NSHC determination which is presented below.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC 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 change involves the deletion of a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not significantly increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change is administrative in that it deletes a reporting requirement. The change does not add new plant equipment, change existing plant equipment, or affect the operating practices of the facility. Therefore, the change does 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 a margin of safety?

Response: No.

The proposed change deletes a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not involve a significant reduction in a margin of safety.

Based on the above, the NRC staff proposes that the change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: Harold K. Chernoff. Start Printed Page 8288

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3, Limestone County, Alabama

Date of amendment request: October 30 and November 20, 2008 (TS-463-T).

Description of amendment request: The BFN requests adoption of an approved change to the Standard Technical Specifications (TSs) for General Electric Plants (NUREG-1433, BWR/4) and plant-specific TSs, that allows: (1) Revising the frequency of Surveillance Requirement (SR) 3.1.3.2, notch testing of fully withdrawn control rod, from “7 days after the control rod is withdrawn and THERMAL POWER is greater than the low-power set point (LPSP) of rod worth minimizer (RWM)” to “31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM,” (2) adding the word “fully” to Limiting Condition for Operation LCO 3.3.1.2, Required Action E.2 to clarify the requirement to fully insert all insertable control rods in core cells containing one or more fuel assemblies when the associated source range monitor instrument is inoperable, and (3) revising Example 1.4-3 in Section 1.4 “Frequency” to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the “SURVEILLANCE” column in addition to the time periods in the “FREQUENCY” column.

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

This change does not affect either the design or operation of the Control Rod Drive Mechanism (CRDM). The affected surveillance and Required Action is not considered to be an initiator of any analyzed event. Revising the frequency for notch testing fully withdrawn control rods will not affect the ability of the control rods to shutdown the reactor if required. Given the extremely reliable nature of the CRDM, as demonstrated through industry operating experience, the proposed monthly notch testing of all withdrawn control rods continues to provide a high level of confidence in control rod operability. Hence, the overall intent of the notch testing surveillances, which is to detect either random stuck control rods or identify generic concerns affecting control rod operability, is not significantly affected by the proposed change. Requiring control rods to be fully inserted when the associated SRM is inoperable is consistent with other similar requirements and will increase the shutdown margin. The clarification of Example 1.4-3 in Section 1.4 “Frequency” is an editorial change made to provide consistency with other TSTF-475, Rev. 1 discussions in Section 1.4. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No

Revising the frequency for notch testing fully withdrawn control rods does not involve physical modification to the plant and does not introduce a new mode of operation. Requiring control rods to be fully inserted will make this action consistent with other similar actions. The clarification of Example 1.4-3 in Section 1.4 “Frequency” is an editorial change made to provide consistency with other discussions in Section 1.4. 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 a margin of safety?

Response: No

The CRDs and CRDMs are extremely reliable systems and, as such, reducing the number of control rod notch tests will not significantly impact the likelihood of detecting a stuck control rod. If a stuck control rod is detected, existing action requirements will ensure prompt action is taken to ensure there is not a generic problem. Other surveillances are routinely performed to ensure that the performance of the control rods in the event of a DBA [design-basis accident] or transient meets the assumptions used in the safety analyses. As such, potential effects of reducing the number of notch tests are far outweighed by the benefit of reducing undue burden on reactor operators and reducing the potential for mispositioning events which accompanies any control rod manipulation. Requiring control rods to be fully inserted instead of partially inserted when the associated SRM is inoperable will increase the margin of safety. The clarification of Example 1.4-3 in Section 1.4 “Frequency” is an editorial change made to provide consistency with other discussions in Section 1.4. 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: Thomas H. Boyce.

Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

Duke Power Company LLC, Docket Nos. 50-414, Catawba Nuclear Station, Unit 2, York County, South Carolina

Date of application for amendments: January 20, 2009.

Brief description of amendments: The proposed amendment would allow a one-time limited duration extension of the Technical Specification (TS) Surveillance (SR) 3.3.1.4 frequency. SR 3.3.1.4 is a Trip Actuating Device Operational Test (TADOT) of the reactor trip breakers (RTBs) and reactor trip bypass breakers.

Date of publication of individual notice in Federal Register : January 28, 2009 (74 FR 4986).

Expiration date of individual notice: 30 days February 27, 2009; 60 days March 30, 2009. Start Printed Page 8289

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 Systems (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 pdr@nrc.gov.

Carolina Power & 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: January 4, 2008.

Brief description of amendment: The amendment establishes more effective and appropriate action, surveillance, and administrative requirements related to ensuring the habitability of the control room envelope in accordance with the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification change traveler TSTF-448, Revision 3, “Control Room Habitability.” This technical specification improvement was initially made available in the Federal Register by the NRC on January 17, 2007 (72 FR 2022).

Date of issuance: January 29, 2009.

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

Amendment No: 128.

Renewed Facility Operating License No. NPF-63: The amendment revises the Technical Specifications and Facility Operating License.

Date of initial notice in Federal Register : May 20, 2008 (73 FR 29161).

The Commission's related evaluation of the amendment is contained in a safety evaluation dated January 29, 2009.

No significant hazards consideration comments received: No.

Carolina Power & 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: April 3, 2008, as supplemented by letters dated December 9, 2008, and January 9, 2009.

Brief description of amendment: The amendment revises Technical Specification Section 5.6.3.b to allow a reconfiguration of the fuel racks in Spent Fuel Pool (SFP) C and allow the use of Metamic as an alternate neutron poison material in the new storage racks for SFP C and D. The amendment: (1) Revises the rack configuration in SFP C to allow the substitution of four previously approved (13 × 13 cell) Boiling Water Reactor racks with an equal number of (9 × 9 cell) Pressurized Water Reactor racks, and (2) authorizes the use of Metamic as an alternate spent fuel rack poison material.

Date of issuance: January 29, 2009.

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

Amendment No: 129.

Renewed Facility Operating License No. NPF-63: The amendment revises the Technical Specifications and Facility Operating License.

Date of initial notice in Federal Register : June 10, 2008 (73 FR 32744). The supplemental letters provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a safety evaluation dated January 29, 2009.

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: January 22, 2008.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) requirements related to control room envelope habitability in accordance with TS Task Force (TSTF) traveler TSTF-448, “Control Room Habitability,” Revision 3. This TS improvement was made available by the Commission on January 17, 2007 (72 FR 2022) as part of the consolidated line item improvement process (CLIIP).

Date of issuance: January 30, 2009.

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

Amendment Nos.: 249 and 229.

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 : March 25, 2008 (73 FR 15784).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 30, 2009.

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 application for amendment: September 22, 2008.

Brief description of amendment: The amendment revised the Technical Specification (TS) to change requirements related to Battery Systems specified in TS Section 3.10 resulting in Start Printed Page 8290removing the Limiting Condition for Operation pertaining to 345 kV switchyard batteries, chargers and associated direct current distribution panel.

Date of Issuance: February 11, 2009.

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

Amendment No.: 234.

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

Date of initial notice in Federal Register : November 18, 2008 (73 FR 68454).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 30, 2009.

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: January 2, 2008, as supplemented by letter dated January 22, 2009.

Brief description of amendment: The amendment revised the actions for inoperable containment isolation valves (CIVs) in Technical Specification 3/4.6.3, “Containment Isolation Valves,” to increase the allowed outage time from 4 hours to 72 hours for inoperable CIVs for penetrations with closed systems inside containment.

Date of issuance: January 30, 2009.

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

Amendment No.: 217.

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

Date of initial notice in Federal Register : January 29, 2008 (73 FR 5219). The supplemental letter dated January 22, 2009, 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 January 30, 2009.

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: February 21, 2008.

Brief description of amendment: The amendments approved revisions to the current licensing basis for Braidwood and Byron associated with the application of an alternative source term (AST) methodology, previously approved by the Nuclear Regulatory Commission staff. Specifically, the amendments approved removing credit for the control room ventilation system recirculation prefilters and reducing the assumed control room unfiltered inleakage in the AST analyses.

Date of issuance: February 5, 2009.

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

Amendment Nos.: Braidwood Unit 1-155; Braidwood Unit 2-155; Byron Unit No. 1-160; and Byron Unit No. 2-160.

Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: The amendments revised the current licensing basis for Braidwood and Byron.

Date of initial notice in Federal Register : June 3, 2008 (73 FR 31720).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 5, 2009.

No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: January 23, 2008.

Brief description of amendment: The proposed amendment would extend the pressure temperature (PT) limit curves and the low temperature overpressure protection (LTOP) setpoints for operation to 55 Effective Full Power Years (EFPYs). The current PT limit curves (and the LTOP setpoints) are applicable to 21.7 EFPYs. The new PT limits and LTOP settings will be applicable to 60 calendar years, which includes the period until the end of the renewed operating license.

Date of Issuance: January 29, 2009.

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

Amendment No.: 154.

Renewed Facility Operating License No. NPF-16: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register : September 9, 2008 (73 FR 52418).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 29, 2009.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: February 6, 2008, as supplemented on September 16 and November 6, 2008.

Brief description of amendment: The amendment approved the installation and use of the General Electric—Hitachi nuclear measurement analysis and control digital Power Range Neutron Monitoring System (PRNMS), and approved changes in the Technical Specifications to reflect use of the PRNMS at Monticello Nuclear Generating Plant.

Date of issuance: January 30, 2009.

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

Amendment No.: 159.

Facility Operating License No. DPR-22. Amendment revised the Technical Specifications and Facility Operating License.

Date of initial notice in Federal Register : March 11, 2008 (73 FR 13025).

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 30, 2009.

No significant hazards consideration comments received: No.

Southern California Edison Company, et. al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of application for amendments: June 27, 2008.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) to adopt Technical Specification Task Force (TSTF) Change Traveler TSTF-487, Revision 1, “Relocate DNB [Departure from Nucleate Boiling] Parameters to the COLR [Core Operating Limits Report].” Specifically, the amendments revised TS 3.4.1 and its associated bases and TS Start Printed Page 82915.7.1.5 to replace the DNB numeric limits in TSs with references to the COLR.

Date of issuance: February 3, 2009.

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

Amendment Nos.: Unit 2-219; Unit 3-212.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : September 23, 2008 (73 FR 54868).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 3, 2009.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 23, 2008.

Brief description of amendments: The amendments revised the actions specified in Technical Specification (TS) 3.6.1.3, “Containment Air Locks,” when limiting condition for operation (LCO) 3.6.1.3 is not met. The amendments allow plant personnel to repair containment air lock components while the plant remains at power and ensure that the containment air locks will continue to meet the requirements of the design basis.

Date of issuance: January 30, 2009.

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

Amendment Nos.: Unit 1-190; Unit 2-178.

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

Date of initial notice in Federal Register : March 25, 2008 (73 FR 15788).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 30, 2009.

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: July 10, 2008, as supplemented by letter dated August 26, 2008.

Brief description of amendment: The amendment modified Technical Specification (TS) 5.5.6 consistent with the Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-419, Revision 0, “Revise PTLR [Pressure and Temperature Limits Report] Definition and References in ISTS [Improved Standard TS] 5.6.6, RCS [Reactor Coolant System] PTLR.” The revised TS 5.6.6 references only the Topical Report (TR) number and title in TS 5.6.6, “Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).” This allows the use of the currently approved TRs to determine the pressure and temperature limits in the PTLR without having to submit an amendment to the Operating License. The change does not alter (1) the U.S. Nuclear Regulatory Commission (NRC) reviewed and approved analytical methods used to determine the pressure and temperature limits or Low Temperature Overpressure Protection System setpoints, or (2) the requirement to use NRC-approved analytical methods to determine the limits or setpoints.

Date of issuance: January 27, 2009.

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

Amendment No.: 180.

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

Date of initial notice in Federal Register : August 26, 2008 (73 FR 50362). The supplemental letter dated August 26, 2008, 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 January 27, 2009.

No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

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 for the amendment 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.

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved. Start Printed Page 8292

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 51.12(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 application for amendment, (2) the amendment to Facility Operating License, 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's (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 pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. 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, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. 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 petitioner/requestor 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 petitioner/requestor 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 petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the 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.[1] 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 petitioner to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Each contention shall be given a separate numeric or alpha designation within one of the following groups:

1. Technical—primarily concerns/issues relating to technical and/or health and safety matters discussed or referenced in the applications.

2. Environmental—primarily concerns/issues relating to matters discussed or referenced in the environmental analysis for the applications.

3. Miscellaneous—does not fall into one of the categories outlined above.

As specified in 10 CFR 2.309, if two or more petitioners/requestors seek to co-sponsor a contention, the petitioners/requestors shall jointly designate a representative who shall have the authority to act for the petitioners/requestors with respect to that contention. If a petitioner/requestor seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/requestor a representative who shall have the authority to act for the petitioners/requestors with respect to that contention.

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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

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 Start Printed Page 8293under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule, which the NRC promulgated in August 28, 2007 (72 FR 49139). The E-Filing process requires participants to submit and serve 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 a waiver in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at HEARINGDOCKET@NRC.GOV, or by calling (301) 415-1677, 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/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRC-issued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms ViewerTM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms ViewerTM is free and is available at http://www.nrc.gov/​site-help/​e-submittals/​install-viewer.html. 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.

Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE 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 EIE 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 may seek assistance through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html or by calling the NRC electronic filing Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays. The electronic filing Help Desk can be contacted by telephone at 1-866-672-7640 or by e-mail at MSHD.Resource@nrc.gov.

Participants who believe that they have a good cause for not submitting documents electronically must file a motion, 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.

Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://www.ehd.nrc.gov/​EHD_​Proceeding/​home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a 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.

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

Date of amendment request: January 23, 2009, as supplemented by letters of January 26, January 30 and February 5, 2009.

Description of amendment request: The amendment revised the KPS facility operating license by modifying the Technical Specifications in Section 3.7.a.7 from “The two underground storage tanks combine to supply at least 35,000 gallons of fuel oil for either diesel generator and the day tanks for each diesel generator contain at least 1,000 gallons of fuel oil” to require each diesel generator's underground storage tank and corresponding day tanks to contain a minimum useable volume of 32,888 gallons.

Date of issuance: February 6, 2009.

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

Amendment No.: 203.

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

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. The Nuclear Regulatory Commission (NRC) staff published a public notice of the proposed amendment, issued a proposed finding of NSHC, and requested that any comments on the proposed NSHC be provided to the NRC staff no later than close of business on February 5, 2009. The notice was published in the “Herald Times Reporter” of Manitowoc, Wisconsin, on January 29, 2009. No comments have been received.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated February 6, 2009. Start Printed Page 8294

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.

NRC Branch Chief: Lois M. James.

Start Signature

Dated at Rockville, Maryland, this 12th day of February 2009.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

End Signature57 End Preamble

Footnotes

1.  To the extent that the application contains attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant's counsel to discuss the need for a protective order.

Back to Citation

[FR Doc. E9-3515 Filed 2-23-09; 8:45 am]

BILLING CODE 7590-01-P