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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) 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 February 25, 2010, to March 10, 2010. The last biweekly notice was published on March 9, 2010 (75 FR 10823).

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rulemaking and Directives Branch (RDB), 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 RDB 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, 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 O1F21, 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 Start Printed Page 13787at 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 hearing.docket@nrc.gov, or by telephone at (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 (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 MSHD.Resource@nrc.gov, or by a toll-free call at (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 Start Printed Page 13788reason 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, Public File Area O1F21, 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 pdr.resource@nrc.gov.

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 amendment request: January 27, 2010.

Description of amendment request: The proposed amendment would revise Technical Specifications (TS) Section 3.6.2.2.a to incorporate an expanded range of eductor flow rates for the containment spray additive system. These changes are supported by the use of a new chemical model and new boric acid equilibrium data, revised sump hydrogen-ion concentration (pH) limits, and changes to the containment spray additive tank concentration and volume limits. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

Response: No.

The proposed change provides revised requirements for an expanded range of eductor flow rates using a new chemical model and new boric acid equilibrium data, revised sump pH limits, and changes to CSAT concentration and volume limits. This ensures that the Spray Additive System remains operable within the TS requirements or appropriate actions be taken. The proposed changes do not affect the automatic shutdown capability of the reactor protection system and no accident analyses are impacted by the proposed changes.

Expanding the range of acceptable values of eductor flow rate does not increase the probability of occurrence of any accident. Analyzed events are initiated by the failure of plant structures, systems or components. The containment spray additive system is not considered as an initiator of any analyzed accident. The proposed changes ensure that the spray additive system and the associated containment spray system can perform the accident mitigation functions required during a LOCA [loss-of-coolant accident] or MSLB [main steam line break] event.

The proposed change does not have a detrimental impact on the integrity of any plant structure, system or component that initiates an analyzed event and will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Furthermore, this action does not affect the initiating frequency of a LOCA or MSLB event.

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.

As described above, the proposed change provides revised requirements for an expanded range of eductor flow rates using a new chemical model and new boric acid equilibrium data, revised sump pH limits, and changes to CSAT concentration and volume limits. These proposed changes ensure that the spray additive system and the associated containment spray system can perform the required accident mitigation functions during a LOCA or MSLB event. There are no other types of accidents that can be postulated that would require the use of the spray additive system or the associated containment spray system for mitigation.

The proposed changes do not introduce any new association between the spray additive system and any radioactive system, including the RCS [reactor coolant system].

Emergency operation of the spray additive system, or postulated failures of the spray additive system, cannot initiate any type of accident. No new accident initiators are introduced by the proposed requirements and no new failure modes are created that would cause 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.

The Bases of TS 3.6.2.2 state that the operability of the Spray Additive System ensures that sufficient NaOH [sodium hydroxide] is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between 7.0 and 11.0 for the solution that is recirculated within containment after a LOCA. The spray additive system adds NaOH to the containment spray water being supplied from the refueling water storage tank (RWST) to adjust the pH of the containment spray and containment recirculation sump solutions. This pH range minimizes both the evolution of iodine and the effect of chloride and caustic stress corrosion on mechanical systems and components. The proposed range of flow rate from the RWST through each eductor ensures that the original margin of safety is maintained through acceptable pH control following a LOCA or MSLB event. The initial conditions of the accident analyses are preserved and the consequences of previously analyzed accidents are unaffected.

Therefore, operation of the facility in accordance with the proposed amendment would 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: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Acting Branch Chief: Douglas A. Broaddus (Acting).Start Printed Page 13789

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3 (Oconee 1, 2, and 3), Oconee County, South Carolina; Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2), Mecklenburg County, North Carolina; Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2 (Catawba 1 and 2), York County, South Carolina

Date of amendment request: December 15, 2009.

Description of amendment request: The proposed amendments would revise the Technical Specifications to replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity. The noble gas activity would be based on DOSE EQUIVALENT XE-133 and would take into account only the noble gas activity in the primary coolant. The changes are consistent with nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-490.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of no significant hazards. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's analysis of the no significant hazards consideration is presented below:

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

Reactor coolant specific activity is not an initiator for any accident previously evaluated. The completion time when primary coolant gross activity is not within limit is not an initiator for any accident previously evaluated. The current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. As a result, the proposed change does not significantly increase the probability of an accident. The proposed change will limit primary coolant noble gases to concentrations consistent with the licensee's current accident analyses for Catawba 1 and 2, McGuire 1 and 2 and Oconee 1, 2, and 3. The proposed change to the completion time has no impact on the consequences of any design-basis accident since the consequences of an accident during the extended completion time are the same as the consequences of an accident during the completion time. 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.

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

The proposed change in specific activity limits does not alter any physical part of the plant nor does it affect any plant operating parameter.

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

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

The proposed change revises the limits on noble gas radioactivity in the primary coolant. The proposed change is consistent with the assumptions in the licensee's safety analysis and will ensure the monitored values protect the initial assumptions in the safety analysis.

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

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Ms. Lisa F. Vaughn, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church Street, EC07H, Charlotte, NC 28202.

NRC Branch Chief: Gloria Kulesa.

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: December 3, 2009.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) to incorporate Standard Technical Specification 3.1.8 “Scram Discharge Volume (SDV) Vent and Drain Valves” and associated Bases of NUREG-1433, Revision 3, “Standard Technical Specifications General Electric Plants, BWR/4,” modified to account for plant specific design details.

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 operation of Vermont Yankee Nuclear Power Station (VY) in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed amendment does not impact the operability of any structure, system or component that affects the probability of an accident or that supports mitigation of an accident previously evaluated. The proposed amendment does not affect reactor operations or accident analysis and has no radiological consequences. The operability requirements for accident mitigation systems remain consistent with the licensing and design basis. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The operation of VY in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

3. The operation of VY in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.

The proposed change ensures that the safety functions of the SDV vent and drain valves are fulfilled. The isolation function is maintained by valves in the vent and drain lines and by the required action to isolate the affected line. The ability to vent and drain the SDVs is maintained through administrative controls. In addition, the reactor protection system ensures that an SDV will not be filled to the point that it has insufficient volume to accept a full scram. Maintaining the safety functions related to isolation of the SDV and insertion of control rods ensures that the proposed change does not involve a significant reduction in the margin of safety. The proposed amendment does not change the design or function of any component or system. The proposed amendment does not impact any safety limits, safety settings or safety margins. 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.Start Printed Page 13790

NRC Branch Chief: Nancy Salgado.

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

Date of amendment request: December 14, 2009.

Description of amendment request: The proposed amendments would change the design basis and Final Safety Analysis Report Update (FSARU) to allow use of a damping value of 5 percent of critical damping for the structural dynamic qualification of the control rod drive mechanism (CRDM) pressure housings on the replacement reactor vessel head for the design earthquake (DE), double design earthquake (DDE), Hosgri earthquake (HE), and loss-of-coolant accident (LOCA) loading conditions.

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?

The proposed change revises the design basis and Final Safety Analysis Report Update (FSARU) to reflect a damping value of 5 percent of critical damping for the structural dynamic qualification of the control rod drive mechanism (CRDM) pressure housings for the replacement reactor vessel head for the design earthquake (DE), double design earthquake (DDE), Hosgri earthquake (HE), and loss of coolant accident (LOCA). The 5 percent damping value has been accepted by the NRC staff at several other plants with equivalent CRDMs and seismic support structures.

The damping value is an element of the structural dynamic analysis performed to confirm the CRDMs' ability to function under a postulated seismic disturbance or LOCA while maintaining resulting stresses under ASME Code [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] Section III allowable values. Because the ASME Code requirements continue to be met, this proposed change to the damping value could not result in an increase in the probability or consequences of an accident previously evaluated.

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

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

The proposed change revises the design basis and FSARU to reflect a damping value of 5 percent of critical damping for the structural dynamic qualification of the CRDM pressure housings for the replacement reactor vessel head for the DE, DDE, HE, and LOCA. The 5 percent damping value has been accepted by the NRC staff at several other plants with equivalent CRDMs and seismic support structures and is a conservative value based on the testing performed by the OEM [original equipment manufacturer].

The damping value is an element of the structural dynamic analysis performed to confirm the CRDMs' ability to function under a postulated seismic disturbance or LOCA while maintaining resulting stresses under ASME Code Section III allowable values. Because the ASME Code requirements continue to be met, this proposed change to the damping value could not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

The proposed change revises the design basis and FSARU to reflect a damping value of 5 percent of critical damping for the structural dynamic qualification of the CRDM pressure housings for the replacement reactor vessel head for the DE, DDE, HE, and LOCA. The 5 percent damping value for CRDMs has been accepted by the NRC staff at several other plants with equivalent CRDMs and seismic support structures.

The damping value is an element of a structural dynamic analysis performed to confirm the CRDMs' ability to function under a postulated seismic disturbance or LOCA while maintaining resulting stresses under ASME Code, Section III, allowable values. The margin of safety is maintained by meeting the ASME Code requirements.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: December 29, 2009.

Description of amendment request: The proposed amendments would revise the licensing basis as described in the Final Safety Analysis Report Update (FSARU) to discuss the conformance of the delayed access offsite power circuit (the 500-kV delayed access circuit) to the General Design Criterion 17 requirement that each of the offsite power circuits be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. The proposed amendment will also add information related to reactor coolant pump seal performance during and after (1) a loss of seal injection (with continued thermal barrier cooling); (2) a loss of thermal barrier cooling (with continued seal injection); and (3) a loss of all seal cooling (both thermal barrier cooling and seal injection).

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 amendments would revise the licensing basis as described in the Final Safety Analysis Report Update (FSARU) to discuss the conformance of the delayed access offsite alternating current (ac) power circuit (the 500-kV delayed access circuit) to the General Design Criterion (GDC) 17 requirement that “each of the offsite power circuits be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded.” It would also add information related to reactor coolant pump (RCP) seal performance during and after (1) a loss of seal injection (with continued thermal barrier cooling); (2) a loss of thermal barrier cooling (with continued seal injection); and (3) a loss of all seal cooling (both thermal barrier cooling and seal injection).

PG&E Calculation STA-274 demonstrates that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded following a loss of the 230-kV immediate access offsite power circuit and all onsite emergency ac power supplies until the 500-kV delayed access circuit can be aligned for backfeed. Alignment of the 500-kV delayed offsite circuit to backfeed, implementing RCP seal coping strategy actions to limit maximum RCP seal leakage to 21 gpm [gallons per minute] per pump, and restoring reactor coolant system (RCS) makeup flow to stabilize the plant can be completed within approximately 54 minutes to assure that specified acceptable fuel design limits and Start Printed Page 13791design conditions of the reactor coolant pressure boundary are not exceeded.

The proposed changes will not add any accident initiators, or adversely affect how the plant safety-related structures, systems, or components (SSCs) are operated, maintained, modified, tested, or inspected. There is no increase in the probability of a GDC 17 loss of all ac event occurring, and since the same applicable GDC 17 acceptance criteria continue to be met with the increased RCP seal leakage, there is no change in the consequences associated with this event.

Therefore, the proposed 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 accident from any accident previously evaluated?

Response: No.

The RCP Seal coping strategy implemented in response to Westinghouse Technical Bulletin TB-04-22, Revision 1, ensures that RCP seal integrity is maintained following a loss of all seal cooling associated with the GDC 17 loss of all ac event. PG&E Calculation STA-274 demonstrates that the GDC 17 requirements for a delayed offsite ac power source are met for up to a one-hour time period for the operators to complete the necessary actions associated with establishing the 500-kV backfeed, implementing the RCP seal coping strategy to limit maximum RCS seal leakage to 21 gpm per pump, and restoring RCS makeup flow. This proposed change provides assurance that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. The proposed change does not introduce new equipment that could create a new or different kind of accident, and no new equipment failure modes are created. As a result, no new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this proposed amendment.

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

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

Response: No.

The implementation of the RCP seal coping strategy ensures that RCP seal leakage is limited to 21 gpm per pump following a loss of all seal cooling such that there is no impact or reduction in the margin of safety associated with the GDC 17 loss of all ac event. The analysis associated with the change supports the ability to align the 500-kV delayed access circuit, implement the RCP seal coping strategy actions, and restore RCS makeup flow in sufficient time following a loss of all onsite ac power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. The proposed amendment would not alter the way any safety-related SSC functions and would not alter the way the plant is operated. The amendment demonstrates that the 500-kV backfeed, isolation of RCP seal cooling, and restoration of RCS makeup flow can be reliably completed within 54 minutes, and that there is considerable margin to the GDC 17 acceptance criteria for the 500-kV backfeed as a delayed offsite ac power source. The proposed amendment would not introduce any new uncertainties or change any existing uncertainties associated with any safety limit. Since the proposed amendment would have no impact on the structural integrity of the fuel cladding or reactor coolant pressure boundary, and maintains the RCP seal leakage within controllable limits, there is no impact on the containment structure. Based on the above considerations, the proposed amendment would not degrade the ability to safely shut down the plant in the event of a loss of all ac power.

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

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

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

NRC Branch Chief: Michael T. Markley.

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

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

Description of amendment request: The proposed amendments would revise 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 would add 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.

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 intermediate range neutron flux trip must be operable in Mode 1 below the P-10 setpoint and in Mode 2 when there is a potential for an uncontrolled rod withdrawal accident during reactor startup. Above the P-10 setpoint, the power range neutron flux high setpoint trip and the power range neutron flux high positive rate trip provide core protection for a rod withdrawal accident. The intermediate range channels have no protection function above the P-10 setpoint. The proposed change does not affect the design of structures, systems, or components (SSCs) credited in accident or transient analyses, the operational characteristics or function of SSCs, the interfaces between credited SSCs and other plant systems, or the reliability of SSCs. The proposed change does not impact the initiating frequency of any UFSAR accident or transient previously evaluated. In addition, the proposed change does not impact the capability of credited SSCs to perform their required safety functions. Thus, eliminating the requirement to apply Specification 3.0.3 provisions when two intermediate range channels are inoperable in Mode 1 with the thermal power above the P-10 setpoint 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 intermediate range neutron flux trip must be operable in Mode 1 below the P-10 setpoint and in Mode 2 when there is a potential for an uncontrolled rod withdrawal accident during reactor startup. Above the P-10 setpoint, the power range neutron flux high setpoint trip and the power range neutron flux high positive rate trip provide core protection for a rod withdrawal accident. The intermediate range channels have no protection function above the P-10 setpoint. The proposed change does not involve a change in design, configuration, or method of operation of the plant. The proposed change does not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. The capability of credited SSCs to perform their required function will not be affected by the proposed change. In addition, the proposed change does not affect the interaction of plant SSCs with other plant SSCs whose failure or malfunction can initiate an accident or transient. As such, no new failure modes are being introduced. Thus, eliminating the requirement to apply Specification 3.0.3 provisions when two intermediate range channels are inoperable in Mode 1 with the thermal power above the P-10 setpoint does not create the possibility of a new or different kind of accident.

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

Response: No.

The proposed change resolves an oversight regarding the operability requirements for the Start Printed Page 13792intermediate range neutron flux channels. Currently, Specification 3.0.3 provisions apply when two intermediate range neutron flux channels are declared inoperable in Mode 1 when thermal power is above the P-10 setpoint. Above the P-10 setpoint, the power range neutron flux trip and the power range neutron flux high positive rate trip provide core protection for a rod withdrawal accident. The intermediate range channels have no protection function above the P-10 setpoint. The proposed change does not change the conditions, operating configurations, or minimum amount of operating equipment assumed in the safety analyses for accident or transient mitigation. The proposed change does not alter the plant design, including instrument setpoints, nor does it alter the assumptions contained in the safety analyses. 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 impact the redundancy or availability of SSCs required to accident or transient mitigation, or the ability of the plant to cope with design basis events. In addition, no changes are proposed in the manner in which the credited SSCs provide plant protection or which create new modes of plant operation. Thus, eliminating the requirement to apply Specification 3.0.3 provisions when two intermediate range channels are inoperable in Mode 1 with thermal power above the P-10 setpoint 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 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: Douglas A. Broaddus (Acting).

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

Date of amendment request: January 28, 2010.

Description of amendment request: The proposed amendment will revise the Limiting Condition for Operation (LCO) of Technical Specification (TS) 3.6.3, “Containment Isolation Valves,” for Wolf Creek Generating Station. A note will be added to LCO 3.6.3 to allow the reactor coolant pump (RCP) seal injection valves to be considered OPERABLE with the valves open and power removed.

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

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

Response: No.

This change affects the RCP seal cooling and the containment isolation system. The change allows the removal of power to the four RCP seal injection valves such that they will not close in response to a spurious signal. A spurious closure of one or more of the seal injection valves could lead to a loss of coolant from the RCP seal. Allowance for removal of power to the valve reduces the probability of this event. The RCP seal performance depends on the design, flow rates, pressures and temperatures. There are no changes to the RCP seal design, nor to the seal cooling flow rates, pressures or temperatures.

Therefore, the consequences of a loss of coolant from the RCP seal are not impacted.

The seal injection valves are containment isolation valves. The system design for RCP seal cooling does not require automatic closure of the seal injection valves or closure of the valve within a specified time frame. The design of the system is such that the cooling water pressure passing through these valves is higher than the operating pressure of the reactor coolant system. The cooling water is needed to prevent a loss of coolant from the pump seals and the cooling water is assured because it is provided by the safety related charging pumps. In addition, a check valve is installed inside the containment on each seal injection line to provide a second containment isolation valve on the line. The seal injection valves fail as-is upon loss of electrical power and are not designed to change position following an accident. The seal injection valves are remote manual valves that can be operated from the control room based on plant procedures. These valves are not modeled as containment isolation valves in any accident analysis. A failure in the open position has no consequence due to the normal inflow of the seal injection water.

Therefore, this change will not increase the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment does not change the method by which any safety related plant system, subsystem, or component performs its specified safety function. The proposed changes will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. Plant procedures will still provide for the appropriate closure of the seal injection valves when restoring seal injection. The proposed changes will not alter any assumptions made in the safety analyses regarding limits on RCP seal injection flow.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety related system as a result of this amendment. The proposed amendment will not alter the design or performance of the 7300 Process Protection System, Nuclear Instrumentation System, or Solid State Protection System used in the plant protection systems.

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

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

Response: No.

The proposed change does not affect the acceptance criteria for any analyzed event. There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection function. Removing power from the RCP seal injection valves during normal operation does not impact the assumed ECCS [emergency core cooling system] flow that would be available for injection into the RCS following an accident.

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: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

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.Start Printed Page 13793

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

Date of amendment request: September 28, 2009.

Brief description of amendment request: The proposed amendment would support application of optimized weld overlays or full structural weld overlays. Applying these weld overlays on the reactor coolant pump suction and discharge nozzle dissimilar metal welds requires an update to the DBNPS leak-before-break evaluation.

Date of publication of individual notice in Federal Register: February 22, 2010 (75 FR 7628)

Expiration date of individual notice: March 24, 2010 (Public comments) and April 22, 2010 (Hearing requests).

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

Date of amendment request: April 17, 2009, as supplemented by letter dated January 19, 2010.

Description of amendment request: On July 14, 2009, the Nuclear Regulatory Commission published a Notice of Consideration of Issuance, Proposed No Significant Hazards Consideration Determination, and Opportunity for Hearing in the Federal Register (74 FR 34048) for a proposed amendment that would change the legal name of the licensee and owner from “FPL Energy Point Beach, LLC” to “NextEra Energy Point Beach, LLC.”

On January 19, 2010, the licensee submitted a supplement which expanded the original scope of work. The proposed revisions would correct an administrative error within a License Condition contained in Appendix C of the Renewed Facility Operating Licenses. The correction changes “FPLE Group Capital” to the appropriately titled “FPL Group Capital.”

Date of publication of individual notice in Federal Register: March 3, 2010 (75 FR 9616)

Expiration date of individual notice: May 3, 2010, 60 days from publication of the individual notice.

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 pdr.resource@nrc.gov.

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: February 19, 2009, as supplemented by letters dated December 22, 2009, and February 23, 2010.

Brief description of amendment: The amendments revised the Technical Specifications (TSs) to relocate the reactor coolant system pressure and temperature (P/T) limits and the low temperature overpressure protection (LTOP) enable temperatures to a licensee-controlled document outside of the TSs. The P/T limits and LTOP enable temperatures will be specified in a Pressure and Temperature Limits Report (PTLR) that will be located in the PVNGS Technical Requirements Manual and administratively controlled by a new TS 5.6.9. The proposed changes are in accordance with the guidance in NRC Generic Letter 96-03, “Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits,” dated January 31, 1996.

Date of issuance: February 25, 2010.

Effective date: As of the date of issuance and shall be implemented within 150 days from the date of issuance.

Amendment No.: Unit 1-178; Unit 2-178; Unit 3-178.

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

Date of initial notice in Federal Register: May 19, 2009 (74 FR 23442). The supplemental letters dated December 22, 2009, and February 23, 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 February 25, 2010.

No significant hazards consideration comments received: No.

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of application for amendments: August 18, 2009, as supplemented on December 7, 2009.

Brief Description of amendments: The proposed license amendments revised Technical Specification 3.3.1.1, “Reactor Protection System (RPS) Instrumentation,” Surveillance Requirement 3.3.1.1.8, to increase the frequency interval between local power range monitor calibrations from 1100 megawatt-days per metric ton average core exposure (i.e., equivalent to approximately 907 effective full-power hours (EFPH)) to 2000 EFPH.

Date of issuance: February 24, 2010.Start Printed Page 13794

Effective date: Date of issuance, to be implemented prior to start-up from the 2010 refueling outage (RFO) for Unit 1, and prior to start-up from the 2011 RFO for Unit 2.

Amendment Nos.: 254 and 282.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 1, 2009 (74 FR 62833). The supplement letter dated December 7, 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 amendments is contained in a Safety Evaluation dated February 24, 2010.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 27, 2009.

Description of amendment request: This amendment request would change the Technical Specifications to provide revised values for the Safety Limit Minimum Critical Power Ratio for both single and dual recirculation loop operation.

Date of Issuance: March 8, 2010.

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

Amendment No.: 243.

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

Date of initial notice in Federal Register: January 5, 2010 (75 FR 461).

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated March 8, 2010.

No significant hazards consideration comments received: No.

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 application for amendments: February 16, 2009.

Brief description of amendments: To remove the structural integrity requirements contained in TS 3/4.4.10, and its associated Bases from the Technical Specifications. Also relocate the reactor coolant pump (RCP) motor flywheel inspection requirements from Surveillance Requirement (SR) 4.4.10 to SR 4.0.5 and revises the RCP motor flywheel inspection frequency from the currently approved 10-year inspection interval, to an interval not to exceed 20 years.

Date of issuance: February 23, 2010.

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

Amendment Nos: 242 and 328.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 21, 2009 (74 FR 18255).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 23, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 24, 2008, as supplemented by letters dated September 19, 2008, April 14, May 22, August 7, August 27, November 20, 2009, and February 2, 2010.

Brief description of amendments: These amendments revise the Point Beach Nuclear Plant licensing basis and Technical Specifications (TS) to reflect a revision to the spent fuel pool (SFP) criticality analysis methodology. The changes to TS 3.7.12, “Spent Fuel Pool Storage,” and 4.3.1, “Criticality,” imposes new storage requirements reflecting the new SFP criticality analysis.

Date of issuance: March 5, 2010.

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

Amendment Nos.: 236, 240.

Renewed Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Technical Specifications/License.

Date of initial notice in Federal Register: December 9, 2008 (73 FR 74759).

The September 19, 2008, April 14, May 22, August 7, August 27, November 20, 2009, and February 2, 2010, supplements contained clarifying information and did not change the staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 5, 2010.

No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: February 11, 2009, as supplemented by letter dated February 1, 2010.

Brief description of amendments: The amendments (1) revise the operating licenses, Technical Specifications (TSs), and Appendix B, Environmental Protection Plan (Non Radiological), to change the plant name and its associated acronym from Comanche Peak Steam Electric Station (CPSES) to Comanche Peak Nuclear Power Plant (CPNPP); (2) remove the Table of Contents from the TSs to licensee control in accordance with plant administrative procedures; (3) delete TSs 3.2.1.1, 3.2.3.1, 5.5.9.1, and 5.6.10 and several footnotes from Tables 3.3.1-1, 3.3.2-1, and TS 3.4.10, since these TSs and footnotes are no longer applicable to the operation of CPSES, Units 1 and 2; (4) delete several topical reports from the list of approved analytical methods used to determine core operating limits in TS 5.6.5 which were no longer in use, since these topical reports have been replaced by standard Westinghouse methods and Westinghouse methods have been approved for use at CPSES, Units 1 and 2, under a separate amendment request; (5) make editorial corrections; and (6) reprint and reissue the TSs in their entirety due to adoption of FrameMaker software in place of Microsoft Word software.

Date of issuance: February 26, 2010.

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

Amendment Nos.: Unit 1-150; Unit 2-150.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revise the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: April 7, 2009 (74 FR 15772). The supplemental letter dated February 1, 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 on April 7, 2009 (74 FR 15772).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 26, 2010.Start Printed Page 13795

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 5, 2009, as supplemented by letters dated April 13 and September 23, 2009.

Brief description of amendments: The amendments revise the Technical Specifications Surveillance Requirement (SR) 3.8.1.8 Frequency to allow the use of the SR 3.0.2 interval extension (1.25 times the interval specified in the Frequency).

Date of issuance: March 1, 2010.

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

Amendment Nos.: 194, 183.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: May 19, 2009 (74 FR 23448). The supplemental letters contained clarifying information and 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 March 1, 2010.

No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: December 19, 2008, as supplemented by letters dated January 22, July 24, and November 23, 2009.

Brief description of amendment: The amendment revises Technical Specifications (TSs) to (1) correct an error in TS Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation,” Function 1.a, to reflect correct CONDITIONS for applicable Modes 1, 2, 3, and 4, (2) revise TS Limiting Condition for Operation 3.3.4 degraded voltage relay and loss of voltage relay Limiting Safety System Setting values to reflect the revised analysis, and (3) revise the load requirement of Surveillance Requirement 3.8.1.3 to reflect values supported by the diesel generator loading analysis.

Date of issuance: March 10, 2010.

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

Amendment No.: 109.

Renewed Facility Operating License No. DPR-18: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: April 7, 2009 (74 FR 15775).

The supplemental letters dated July 24, 2009, and November 23, 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.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 10, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 17, 2008.

Brief description of amendment: The amendments revised Technical Specifications (TSs) 1.1, “Definitions,” and 3.4.16, “RCS Specific Activity,” and Surveillance Requirements 3.4.16.1 through 3.4.16.3. The amendments replaced the current TS 3.4.16 limit on reactor coolant system (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit is based on a new dose equivalent Xe-133 definition that would replace the current E-Bar average disintegration energy definition. The amendments are adopting TS Task Force (TSTF)-490.

Date of issuance: March 3, 2010.

Effective date: This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

Amendment Nos.: 258 and 239.

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

Date of initial notice in Federal Register: February 10, 2009 (74 FR 6669). The supplements dated January 26, May 26, and November 23, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 3, 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, Unit Nos. 1 and 2, Surry County, Virginia

Date of application for amendments: April 13, 2009.

Brief Description of amendments: These amendments revised the technical specifications (TSs). The proposed change revised TS Table 3.7.1, Operator Action 3.b, and provides direction for the actions to be taken if the operating condition of fewer than the required minimum channels for the neutron flux intermediate range occurs between 7 percent and 11 percent of rated power.

Date of issuance: February 26, 2010.

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

Amendment Nos.: 268 and 267.

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: July 14, 2009 (74 FR 34049).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 26, 2010.

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.Start Printed Page 13796

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.

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.resource@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.resource@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 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 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 requestor/petitioner 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.Start Printed Page 13797

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 requestor/petitioner seeks to adopt the contention of another sponsoring requestor/petitioner, the requestor/petitioner who seeks to adopt the contention must either agree that the sponsoring requestor/petitioner shall act as the representative with respect to that contention, or jointly designate with the sponsoring requestor/petitioner 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 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 hearing.docket@nrc.gov, or by telephone at (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 (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 MSHD.Resource@nrc.gov, or by a toll-free call at (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://Start Printed Page 13798ehd.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.

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

Date of amendment request: March 3, 2010, as supplemented by letter dated March 4, 2010.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” Condition J, Required Action J.1, and associated Note for the start of the motor-driven auxiliary feedwater pumps on the trip of all main feedwater (MFW) pumps. Wolf Creek Nuclear Operating Corporation has determined that the design and normal operation of the MFW pumps at Wolf Creek Generating Station could result in a condition that does not conform to TS Table 3.3.2-1, Function 6.g and the proposed TS changes are needed to address this condition.

Date of issuance: March 5, 2010.

Effective date: The license amendment is effective as of its date of issuance and shall be implemented within 10 days of the date of issuance.

Amendment No.: 187.

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

Public comments requested as to proposed no significant hazards consideration (NSHC): No.

The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated March 5, 2010.

Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Michael T. Markley.

Start Signature

Dated at Rockville, Maryland this 12th day of March 2010.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

End Signature End Preamble

Footnotes

1.  To the extent that the applications contain 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 and discuss the need for a protective order.

Back to Citation

[FR Doc. 2010-6052 Filed 3-22-10; 8:45 am]

BILLING CODE 7590-01-P