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Notice

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

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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 June 17, 2010, to June 30, 2010. The last biweekly notice was published on June 29, 2010 (75 FR 37471).

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

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

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

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

Written comments may be submitted by mail to the Chief, Rules, Announcements and Directives Branch (RADB), TWB-05-B01M, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be faxed to the RADB at 301-492-3446. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room (PDR), located at One White Flint North, 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 at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to Start Printed Page 39976matters 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. e.t. 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., e.t., Monday through Friday, excluding government holidays.

Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/​EHD_​Proceeding/​home.asp, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in Start Printed Page 39977their 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.

Duke Energy Carolinas, LLC, et al., Docket No. 50-414, Catawba Nuclear Station, Unit 2, York County, South Carolina

Date of amendment request: April 28, 2010.

Description of amendment request: The amendment would revise Technical Specification (TS) 5.5.9 to exclude portions of the Steam Generator (SG) tube from periodic SG tube inspections and plugging or repair. In addition, reporting requirement changes are proposed to TS 5.6.8. This submittal is requesting a one-cycle approval for the Catawba Nuclear Station, Unit 2, End of Cycle 17 Refueling Outage and subsequent Cycle 18 operation.

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:

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

Response: No. The proposed changes to TS 5.5.9, TS 5.6.8, and the Facility Operating License have no significant effect upon accident probabilities or consequences. Of the various accidents previously evaluated, the following are limiting with respect to the proposed changes as discussed in this amendment request:

  • SG Tube Rupture evaluation
  • Steam Line Break/Feed Line Break evaluation
  • Locked Rotor evaluation
  • Control Rod Ejection evaluation

Loss of Coolant Accident conditions cause a compressive axial load to act on the tube. Therefore, since this accident tends to force the tube into the tubesheet rather than pull it out, it is not a factor in this amendment request. Another faulted load consideration is a safe Shutdown Earthquake; however, the seismic analysis of Model D5 SGs (the SGs at Catawba) has shown that axial loading of the tubes is negligible during this event.

At normal operating pressures, leakage from Primary Water Stress Corrosion Cracking (PWSCC) below 16.95 inches from the top of the tubesheet is limited by both the tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint. Consequently, negligible normal operating leakage is expected from cracks within the tubesheet region.

For the SG Tube Rupture event, tube rupture is precluded for cracks in the hydraulic expansion region due to the constraint provided by the tubesheet. Therefore, the margin against tube burst/pullout is maintained during normal and postulated accident conditions and the proposed change does not result in a significant increase in the probability of a tube rupture. SG Tube Rupture consequences are not affected by the primary to secondary leakage flow during the event, as primary to secondary leakage flow through a postulated tube that has been pulled out of the tubesheet is essentially equivalent to that from a severed tube. Therefore, the proposed change does not result in a significant increase in the consequences of a tube rupture.

The probability of a Steam Line Break/Feed Line Break, Locked Rotor, and Control Rod Ejection are not affected by the potential failure of a SG tube, as the failure of a tube is not an initiator for any of these events. In WCAP-17072-P, leakage is modeled as flow through a porous medium via the use of the Darcy equation. The leakage model is used to develop a relationship between operational leakage and leakage at accident conditions that is based on differential pressure across the tubesheet and the viscosity of the fluid. A leak rate ratio was developed to relate the leakage at operating conditions to leakage at accident conditions. The fluid viscosity is based on fluid temperature and it has been shown that for the most limiting accident, the fluid temperature does not exceed the normal operating temperature. Therefore, the viscosity ratio is assumed to be 1.0 and the leak rate ratio is a function of the ratio of the accident differential pressure and the normal operating differential pressure.

The leakage factor of 2.65 for Catawba Unit 2 for a postulated Steam Line Break/Feed Line Break has been calculated as shown in WCAP-17072-P, as supplemented. The leakage factor has been increased to 3.27 per additional Westinghouse analysis specific to Catawba. Therefore, Catawba Unit 2 will apply a factor of 3.27 to the normal operating leakage associated with the tubesheet expansion region in the Condition Monitoring assessment and Operational Assessment. Through application of the limited tubesheet inspection scope, the proposed operating leakage limit provides assurance that excessive leakage (i.e., greater than accident analysis assumptions) will not occur. No leakage factor will be applied to the Locked Rotor or Control Rod Ejection due to their short duration, since the calculated leak rate ratio is less than 1.0. Therefore, the proposed change does not result in a significant increase in the consequences of these accidents.

For the Condition Monitoring assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 3.27 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the Operational Assessment, the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 3.27 and compared to the observed operational leakage.

Based on the above, the performance criteria of NEI 97-06, Revision 2 and RG [Regulatory Guide] 1.121 continue to be met and the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No. The proposed changes to TS 5.5.9, TS 5.6.8, and the Facility Operating License do not introduce any changes or mechanisms that create the possibility of a new or different kind of accident. Tube bundle integrity is expected to be maintained for all plant conditions upon implementation of the one-cycle alternate repair criteria. The proposed change does not introduce any new equipment or any change to existing equipment. No new effects on existing equipment are created nor are any new malfunctions introduced.

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

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

Response: No. The proposed changes to TS 5.5.9, TS 5.6.8, and the Facility Operating License maintain the required structural margins of the SG tubes for both normal and accident conditions. NEI 97-06, Revision 2 and RG 1.121 are used as the basis in the development of the limited tubesheet inspection depth methodology for determining that SG tube integrity considerations are maintained within acceptable limits. RG 1.121 describes a Start Printed Page 39978method acceptable to the NRC staff for meeting GDC [General Design Criteria] 14, 15, 31, and 32 by reducing the probability and consequences of a SG Tube Rupture. RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation, the probability and consequences of a SG Tube Rupture are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the ASME [American Society of Mechanical Engineers] Code [Boiler and Pressure Vessel Code].

For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially oriented cracking, WCAP-17072-P defines a length of degradation-free expanded tubing that provides the necessary resistance to tube pullout due to the pressure-induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary to secondary leakage during all plant conditions. The methodology for determining leakage as described in WCAP-17072-P shows that significant margin exists between an acceptable level of leakage during normal operating conditions that ensures meeting the accident induced leakage assumption and the TS leakage limit.

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

Attorney for licensee: Ms. Kate Nolan, Associate General Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 422 South Church Street, Mail Code—EC07H, P.O. Box 1244, Charlotte, NC 28201-1244.

NRC Branch Chief: Gloria Kulesa.

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: June 25, 2009, as supplemented May 21, 2010.

Description of amendment request: Revise the licensing bases to adopt the alternative source term as allowed in Title 10 of the Code of Federal Regulations, § 50.67.

An application that addressed similar issues was previously submitted on June 25, 2009, and noticed in the Federal Register (FR) on December 29, 2009 (74 FR 68870). Due to certain changes in the specifics stated in the May 21, 2010, supplement, from those proposed in the June 25, 2009, application, this is a renotice that includes those changes. Below is the no significant hazards consideration determination (NSHCD) for the changes in the May 21, 2010 supplemental. The original NSHCD as published in the FR December 29, 2009, still applies to the June 25, 2009 application.

Basis for proposed NSHCD: 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 amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to [technical specification] TS 4.4.8 will only provide for better assurance of required sampling and analysis of the reactor coolant system specific activity during thermal power changes and transient conditions (MODES 1, 2, 3, and 4). This will ensure potential consequences of a [defined-basis accident] DBA are bounded by the approved accident analyses.

The proposed changes to TS 3/4.7.5 itemize the system operability requirements and appropriate actions in the event that those requirements are not satisfied. These actions include actions to be taken during the allowed outage times (AOTs) specified in the actions to bring the system back into compliance with the system operability requirements. The actions also provide for restoration of the inoperable component or in some cases provide for placing and maintaining it in a safe condition until it can be restored. The actions may include compensatory measures that require initiation of mitigating actions involving operator action to manually align and place into service a compensatory filtration unit in the event that the normal filtration train is out-of-service. These compensatory measures are required to be taken within 24 hours compared to the current allowed outage time of 84 hours for system inoperability without any compensatory measures specified. Moreover, consistent with the current Turkey Point TS and TSTF-448 AOTs, manually aligning the compensatory filter within 24 hours to maintain [control room emergency ventilation system] CREVS operability is acceptable in order to ensure control room operations will be protected from analyzed radiological hazards. The other action statements for inoperability of a redundant active component provide for an AOT of 7 days consistent with the Westinghouse Standard Technical Specification. They are based on the low probability of occurrence of a DBA challenging the Control Room Habitability during this time period and the continued capability of the remaining system components to perform the required CREVS safety function.

The proposed changes have no effect on the probability of an accident previously evaluated as they do not affect any accident initiators. The proposed changes have no significant effect on the consequences of an accident previously evaluated as they either provide for better monitoring of plant operating parameters or for compensatory actions to be taken for out-of-service equipment not previously available. Design changes to enhance the system capabilities will be made to the same design and quality standards as the existing CREVS. System modifications required to support these proposed changes are evaluated under the 10 CFR 50.59 program and are enhancements to the mitigation strategies.

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

2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes to TS 4.4.8 will only provide for better assurance of required sampling and analysis of the reactor coolant system specific activity during MODES 1, 2, 3, and 4. The proposed modifications to the plant configuration will be fully qualified to the appropriate design requirements to assure their required function is available for accident mitigation. Additionally, functions of other equipment required for accident mitigation are also not adversely impacted. Design changes to enhance the system capabilities will be made to the same design and quality standards as the existing CREVS. The proposed changes to TS 3/4.7.5 will provide for better specification of system operability requirements and appropriate actions in the event that those requirements are not satisfied. The proposed changes have no effect on accident precursors or initiators and only enhance mitigation capabilities with regard to protecting control room personnel from radiological hazards.

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

3. The proposed amendment does not involve a significant reduction in the margin of safety.

The proposed changes to TS 4.4.8 will only provide for better assurance of required sampling and analysis of the reactor coolant system specific activity during thermal power changes and transient conditions (MODES 1, 2, 3, and 4). No plant system or component design or operational requirements are affected by these changes.

The proposed changes to TS 3/4.7.5 will provide for better specification of system operability requirements and appropriate actions in the event that those requirements are not satisfied. The proposed increase in the specified AOT for inoperability of CREVS components from 84 hours to 7 days is considered insignificant as it is consistent with the Westinghouse Standard Technical Specification and based on the low probability of occurrence of a DBA challenging the Control Room Habitability during this time period and the continued capability of the remaining system components to perform the required CREVS safety function. Moreover, consistent with the current Turkey Point TS and TSTF-448 AOTs, manually aligning the compensatory filter within 24 hours to maintain CREVS operability is an acceptable margin of safety to ensure control room operations will be protected from analyzed radiological hazards. The proposed changes provide for compensatory actions to be taken for out-of-service equipment that were not previously Start Printed Page 39979available and thus enhance existing mitigation capabilities with regard to protecting control room personnel from radiological hazards.

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

Based on the above discussion, FPL has determined that the proposed 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 § 50.92(c) are satisfied. Therefore, 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: Douglas A. Broaddus.

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

Date of amendment request: May 14, 2010.

Description of amendment request: The proposed changes would revise the Seabrook Station Technical Specifications (TSs) governing the Containment Enclosure Emergency Air Cleanup System. Specifically, the proposed change would insert a requirement that if both trains of the system are inoperable, at least one train must be returned to operable status within 24 hours or begin a shutdown of the reactor. Currently, since there are no limiting conditions for operation proscribed actions in the event two trains are inoperable, TS 3.0.3 requires a shutdown within 6 hours.

Basis for proposed NSHC determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of NSHC, 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 does not impact the physical function of plant structures, systems, or components (SSCs) or the manner in which SSCs perform their design function. The proposed changes neither adversely affect accident initiators or precursors, nor alter design assumptions. The proposed changes do not alter or prevent the ability of operable [SSCs] to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits.

This change is a revision to the technical specifications (TS[s]) for the containment enclosure emergency air cleanup system (CEEACS), which is a mitigation system designed to prevent uncontrolled releases of radioactivity into the environment. The change would allow intermittent opening of the containment enclosure boundary under administrative controls. These controls would ensure that the opening will be quickly sealed to maintain the validity of the licensing basis analyses of accident consequences. The proposed change adds a new action requirement that would allow 24 hours to restore the containment enclosure boundary in the event that both trains of the CEEACS are inoperable due to an inoperable containment enclosure boundary. The proposed 24 hour completion time is reasonable based on the low probability of a design basis accident occurring during this time period and the use of preplanned compensatory measures. The CEEACS is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased.

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 will not impact the accident analysis. The changes will not alter the requirements of the CEEACS or its function during accident conditions, and no new or different accidents result from the proposed changes to the TS[s]. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the method of plant operation. The changes do not alter assumptions made in the safety analysis. Therefore, this request does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. The proposed changes do not involve a significant change in the method of plant operation, and no accident analyses will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits, will not relax any safety system settings, and will not relax the bases for any limiting conditions for operation. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design bases. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. Therefore, these 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, and with the changes noted above, it appears that the three standards of §50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC.

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

NRC Branch Chief: Harold K. Chernoff.

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

Date of amendment request: March 18, 2010.

Description of amendment request: The proposed amendment would revise the NMP1 Technical Specifications (TSs) for inoperable snubbers by removing TS 3/4.6.4, “Shock Suppressors (Snubbers),” and would also add a new Limiting Condition for Operation (LCO) 3.0.8.

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

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

Response: No.

The proposed change to relocate TS 3/4.6.4 to station procedures is administrative in nature and does not involve the modification of any plant equipment or affect basic plant operation. Snubber operability and surveillance requirements will be contained in the station procedures to ensure design assumptions for accident mitigation are maintained.

The proposed change to add LCO 3.0.8 allows a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. Entrance into TS actions or delaying entrance into actions is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the current TS required actions in effect without the allowance provided by proposed LCO 3.0.8.

Revision of TS Table of Contents to reflect deletion of TS 3/4.6.4 is administrative in nature and therefore does not involve a significant 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 proposed amendment create the possibility of a new or different kind of Start Printed Page 39980accident from any accident previously evaluated?

Response: No.

The proposed change to relocate TS 3/4.6.4 to station procedures is administrative and does not involve any physical alteration of plant equipment. The proposed change does not change the method by which any safety related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to add LCO 3.0.8 does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TSs when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects.

Revision of TS Table of Contents to reflect deletion of TS 3/4.6.4 is administrative in nature and therefore does not create the possibility of a new or different kind of accident from any previously evaluated.

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

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

Response: No.

The proposed change to relocate TS 3/4.6.4 to station procedures is administrative in nature, does not negate any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the TSs to station procedures.

The proposed change to add LCO 3.0.8 to TSs allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform the required safety function. The proposed change retains an allowance in the current NMPI TSs while upgrading it to be more conservative for snubbers supporting multiple trains or sub-systems of an associated system. The updated TS will continue to provide an adequate margin of safety for plant operation upon incorporation of LCO 3.0.8. The station design and safety analysis assumptions provide margin in the form of redundancy to account for periods of time when system capability is reduced.

Revision of TS Table of Contents to reflect deletion of TS 3/4.6.4 is administrative in nature and therefore does not involve a significant reduction in a margin of safety.

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

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

Attorney for licensee: Carey W. Fleming, Senior Counsel, Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, Baltimore, MD 21202.

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: March 22, 2010.

Description of amendment request: The proposed amendment would revise the NMP1 Technical Specifications (TSs) section 4.3.7 “Containment Spray System,” by modifying the testing frequency for the Surveillance Requirement (SR) 4.3.7.b, “Nozzles,” from “at least once per operating cycle * * * ” to “following maintenance that could result in nozzle blockage.” Additional wording changes would be made to the SR to make it more consistent with the corresponding Standard TS, SR 3.6.1.7.4.

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

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

Response: No.

The proposed change modifies the SR to verify that the Containment Spray System (CSS) drywell and torus spray nozzles are unobstructed after maintenance that could introduce material resulting in nozzle blockage. The requirement to test the headers will be removed as well as the type of test to be used. Since the opening within the pipes is much larger than the nozzles, they are not likely to become obstructed unless the nozzles become obstructed. The spray nozzles and headers are not assumed to be initiators of any previously analyzed accident. Therefore, the proposed change does not increase the probability of any accident previously evaluated. The spray nozzles are used in the accident analyses to mitigate design basis accidents. The revised SR to verify system operability following maintenance is considered adequate to ensure operability of the CSS. Since the system will still be able to perform its accident mitigation function, the consequences of accidents 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.

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

Response: No.

The proposed change revises the SR to verify that the CSS nozzles are unobstructed after maintenance that could result in nozzle blockage. The requirement to test the headers will be removed as well as the type of test to be used. The spray nozzles and headers are not assumed to be initiators of any previously analyzed accident. The change does not introduce a new mode of plant operation and does not involve a physical modification to the plant. The change will not introduce new accident initiators or impact the assumptions made in the safety analysis.

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

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

Response: No.

The proposed change revises the frequency for performance of the SR to verify that the CSS nozzles are unobstructed. The frequency is changed from “once per operating cycle” to “following maintenance that could result in nozzle blockage.” The requirement to test the headers will be removed as well as the type of test to be used. The revised testing requirement, along with the foreign material exclusion program, the normal environmental conditions for the system, and the remote physical location of the spray nozzles, provide assurance that the spray nozzles and headers will remain unobstructed. As the spray nozzles and headers are expected to remain unobstructed and able to perform their post-accident mitigation function, plant safety is not significantly affected.

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

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

Attorney for licensee: Carey W. Fleming, Senior Counsel, Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, Baltimore, MD 21202.

NRC Branch Chief: Nancy L. Salgado.

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

Date of amendment request: March 30, 2010, as supplemented on June 1, 2010.Start Printed Page 39981

Description of amendment request: The proposed amendment would revise the NMP2 Technical Specification (TS) section 3.8.1, “AC Sources—Operating,” to extend the Completion Time (CT) for an inoperable Division 1 or Division 2 diesel generator (DG) from 72 hours to 14 days.

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

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

Response: No.

The proposed TS change to increase the CT for an inoperable Division 1 or Division 2 DG from 72 hours to 14 days does not affect the design, function, operational characteristics, or reliability of the DGs. The DGs are designed to mitigate the consequences of previously evaluated accidents and, as such, are not accident initiators.

Extending the CT for an inoperable DG will not significantly affect the capability of the DGs to perform their accident mitigation safety functions or adversely affect DG or offsite power availability. The consequences of previously evaluated accidents will not be significantly affected since the remaining DGs supporting the redundant Engineered Safety Feature (ESF) systems will continue to be available to perform the accident mitigation functions as designed.

Both a deterministic evaluation and a risk impact assessment were performed to support the proposed DG CT extension. The deterministic evaluation concluded that the defense-in-depth philosophy will be maintained with the proposed DG CT extension. The current TS and 10 CFR 50.65 (Maintenance Rule) programmatic requirements and additional administrative controls provide assurance that a loss of offsite power occurring concurrent with an inoperable DG will not result in a complete loss of function of critical systems. The duration of the proposed DG CT is determined considering that there is a minimal possibility that an accident will occur while a component is removed from service. A risk impact assessment was performed which concluded that the increase in plant risk due to the increased DG CT is small and consistent with the guidance contained in Regulatory Guide 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications.”

Based on the above discussion, it is concluded that the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment does not alter the design, configuration, or method of operation of the plant, and does not alter any safety analysis inputs or assumptions. The proposed extended DG CT will not reduce the number of DGs below the minimum required for safe shutdown or accident mitigation. No new component failure modes, system interactions, or accident responses will be created that could result in a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

The proposed extension of the DG CT remains consistent with codes and standards applicable to the onsite alternating current (AC) sources, except that the extension deviates from the recommendations of Regulatory Guide 1.93, “Availability of Electric Power Sources.” The proposed amendment is justified based on the results of a deterministic evaluation and a risk impact assessment. These demonstrate that the defense-in-depth philosophy will be maintained and the increase in plant risk is small and consistent with the guidance contained in Regulatory Guide 1.177.

The DG reliability and availability are monitored and evaluated with respect to Maintenance Rule performance criteria to assure DG out of service times do not degrade operational safety over time. Furthermore, extension of the DG CT does not affect any safety analysis inputs or assumptions and will not erode the reduction in severe accident risk that was achieved with implementation of the Station Blackout (SBO) rule (10 CFR 50.63). The SBO coping analysis is unaffected by the CT extension since the DGs are not assumed to be available during the coping period. The assumptions used in the coping analysis regarding DG reliability are unaffected since preventive maintenance and testing will continue to be performed to maintain the reliability assumptions.

Based on the above discussion, it is concluded that the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Carey W. Fleming, Senior Counsel, Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, Baltimore, MD 21202

NRC Branch Chief: Nancy L. Salgado.

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.

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: May 29, 2009, as supplemented by letters dated November 14, 2008, December 11, 2008, August 13, 2009, August 28, 2009, October 9, 2009, February 4, 2010, and April 5, 2010.Start Printed Page 39982

Brief description of amendment: The proposed amendment transitions the existing fire protection program to a risk-informed, performance-based program based on National Fire Protection Association Standard 805 (NFPA 805), “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,” 2001 Edition, in accordance with Title 10 of the Code of Federal Regulations, § 50.48(c). NFPA 805 allows the use of performance-based methods, such as fire modeling and fire risk evaluations, to demonstrate compliance with the nuclear safety performance criteria.

Date of issuance: June 28, 2010.

Effective date: Effective as of the date of issuance and shall be implemented within 180 days, contingent upon completion of the items identified in section 2.9 of the associated NRC Safety Evaluation.

Amendment No.: 133.

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 19, 2009 (74 FR 29241). The supplements dated November 14, 2008, December 11, 2008, August 13, 2009, August 28, 2009, October 9, 2009, February 4, 2010, and April 5, 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.

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

The Commission's related evaluation of the amendment and final NSHC determination are contained in a safety evaluation dated June 28, 2010.

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

Date of application for amendments: October 2, 2008, as supplemented by letters dated August 25, 2009, and October 23, 2009.

Brief description of amendments: The amendments revise the Technical Specifications (TSs) associated with the verification of ice condenser door operability and TS surveillance requirements 3.6.13.5 and 3.6.13.6.

Date of issuance: June 28, 2010.

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

Amendment Nos.: 256 and 251.

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

Date of initial notice in Federal Register: March 8, 2010 (75 FR 10513). The supplements dated August 25, 2009, and October 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 amendments is contained in a Safety Evaluation dated June 28, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 2, 2008, as supplemented by letters dated June 18, 2009, July 8, 2009, August 13, 2009, September 8, 2009, November 10, 2009 and March 8, 2010.

Brief description of amendments: The amendments revised the technical specifications to allow manual operation of the containment spray system and to revise the upper and lower limits of the refueling water storage tank.

Date of issuance: June 28, 2010.

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

Amendment Nos.: 257 and 252.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the licenses and the Technical Specifications.

Date of initial notice in Federal Register: April 7, 2009 (74 FR 15767). The supplements dated June 18, 2009, July 8, 2009, August 13, 2009, September 8, 2009, November 10, 2009, and March 8, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 28, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 2, 2008, as supplemented by letters dated August 25, 2009, and October 23, 2009.

Brief description of amendments: The amendments revised the Technical Specifications (TSs) associated with the verification of ice condenser door operability and TS surveillance requirements 3.6.13.5 and 3.6.13.6.

Date of issuance: June 28, 2010.

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

Amendment Nos.: 256 and 236.

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 8, 2010 (75 FR 10508). The supplements dated August 25, 2009, and October 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 amendments is contained in a Safety Evaluation dated June 28, 2010.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: July 23, 2009.

Brief description of amendment: The amendment removed the local refueling water storage tank level indication from Technical Specification Surveillance Requirement 3.5.4.5.

Date of issuance: June 28, 2010.

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

Amendment No.: 241.

Facility Operating License No. DPR-64: The amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: October 6, 2009 (74 FR 51329).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 28, 2010.

No significant hazards consideration comments received: No.Start Printed Page 39983

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

Date of application for amendment: June 30, 2009, as supplemented by letter dated May 24, 2010.

Brief description of amendment: This amendment revises the Surveillance Requirement (SR) regarding the start time tests for the Division 3 Emergency Diesel Generator to provide consistency with existing similar Technical Specification (TS) 3.8.1 “AC Sources—Operating” SRs and the time provided in the licensing basis emergency core cooling system analyses.

Date of issuance: June 30, 2010.

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

Amendment No.: 154.

Facility Operating License No. NPF-58: This amendment revised the TSs and License.

Date of initial notice in Federal Register: November 17, 2009 (74 FR 59261). The supplement dated May 24, 2010 provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 30, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 14, 2009 as supplemented by letters dated October 30, 2009, and March 19, 2010.

Brief description of amendment: The amendment modifies the Operating License, Condition 2.C.(2), Appendix B, Environmental Technical Specifications, Part II, “Non-Radiological Environmental Protection Plan.” The amendment deletes outdated program information and relieves I&M from preparing and submitting unnecessary or duplicative environmental reports.

Date of issuance: June 24, 2010.

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

Amendment Nos.: 312 (CNP-1), 295 (CNP-2).

Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Renewed Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: May 5, 2009 (74 FR 20749).

The supplemental information dated October 30, 2009, and March 19, 2010, contained clarifying information, did not change the scope of January 14, 2009, application or the initial no significant hazards consideration determination, and does 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 June 24, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 18, 2009.

Brief description of amendment: The amendment changes the NMP2 Technical Specifications (TSs) for unavailable barriers by adding Limiting Condition for Operation (LCO) 3.0.9. LCO 3.0.9 establishes conditions under which a supported system would remain operable when required physical barriers are not capable of providing their related support function. The submitted change is consistent with the industry Technical Specifications Task Force (TSTF) Traveler TSTF-427, Revision 2, “Allowance for Non Technical Specification Barrier Degradation on Supported System OPERABILITY.” A notice of the TSTF-427, Revision 2 TS improvement was published in the Federal Register on October 3, 2006 (71 FR 58444) as part of the Consolidated Line Item Improvement Process.

Date of issuance: June 29, 2010.

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

Amendment No.: 135.

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

Date of initial notice in Federal Register: April 6, 2010 (75 FR 17445).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 2, 2009.

Brief description of amendment: The amendment revises the TSs by removing position indication for the relief valves and safety valves from TS 3.6.11, “Accident Monitoring Instrumentation.” The amendment would also correct an editorial error in the title of Table 4.6.11, “Accident Monitoring Instrumentation Surveillance Requirement.”

Date of issuance: June 29, 2010.

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

Amendment No.: 205.

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

Date of initial notice in Federal Register: October 14, 2009 (74 FR 52826).

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

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: November 4, 2008, as supplemented by letters dated August 10, 2009, and March 30, 2010.

Brief description of amendments: The amendments modify the technical specifications (TSs) and facility operating licenses by increasing the 24-month test load for the Unit 1 emergency diesel generators (EDGs) and decrease the 24-month test load for the Unit 2 EDGs in TS Surveillance Requirement 3.8.1.9.

Date of issuance: June 21, 2010.

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

Amendment Nos.: 196, 185.

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

Date of initial notice in Federal Register: January 27, 2009 (74 FR 4774).

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.Start Printed Page 39984

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 21, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 30, 2009.

Brief description of amendment: The amendment relocates the Technical Specification (TS) surveillance requirement for the reactor recirculation system motor-generator set scoop tube stop settings to the Technical Requirements Manual.

Date of issuance: June 28, 2010.

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

Amendment No.: 181.

Facility Operating License No. NPF-57: The amendment revised the TSs and the License.

Date of initial notice in Federal Register: October 6, 2009 (74 FR 51333).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 28, 2010.

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: August 10, 2009.

Brief description of amendments: The amendment modified Technical Specification 3.7.5, “Auxiliary Feedwater (AFW) System,” to allow a 7-day Completion Time for the turbine-driven AFW pump if the inoperability of the pump occurs in MODE 3 following a refueling outage, and if MODE 2 has not been entered. This change is consistent with the U.S. Nuclear Regulatory Commission-approved Technical Specification Task Force (TSTF) traveler, TSTF-340,

Revision 3.

Date of issuance: June 30, 2010.

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

Amendment Nos.: Unit 2-223; Unit 3-216.

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: November 17, 2009 (74 FR 59263).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 30, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 4, 2009.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.7.3, “Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulating Valves (MFRVs) and Main Feedwater Regulating Valve Bypass Valves (MFRVBVs),” so that the Limiting Condition for Operation (LCO) and Applicability more accurately reflect the conditions for when the LCO should be applicable and more effectively provide appropriate exceptions to the Applicability for certain valve configurations. The amendment also changed the title of TS 3.7.3 to “Main Feedwater Isolation Valves (MFIVs), Main Feedwater Regulating Valves (MFRVs), and Main Feedwater Regulating Valve Bypass Valves (MFRVBVs),” and the associated page header to “MFIVs, MFRVs, and MFRVBVs.” In addition, the amendment revised footnotes to TS 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” Table 3.3.2-1, in order to improve application of existing notes and/or incorporate more appropriate notes.

Date of issuance: June 29, 2010.

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

Amendment No.: 198.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: August 25, 2009 (74 FR 42932).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 10, 2009.

Brief description of amendment: The amendment deletes the Technical Specification (TS) requirements for the containment hydrogen recombiners and relaxes the requirements for hydrogen and oxygen monitors. The TS changes support implementation of the revisions to Title 10 of the Code of Federal Regulations (10 CFR) § 50.44, “Combustible gas control for nuclear power reactors,” that became effective on October 16, 2003. The changes are consistent with Revision 1 of the NRC-approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-447, “Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen Monitors.” This operating license improvement was made available by the NRC on September 25, 2003 (68 FR 55416), as part of the consolidated line item improvement process. In addition, the amendment corrected four typographical errors in the TSs.

Date of issuance: June 29, 2010.

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

Amendment No.: 199.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: August 25, 2009 (74 FR 42934).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 1, 2009, as supplemented by letters dated August 27, 2009, and March 4, 2010.

Brief description of amendment: The amendment revised the Limiting Condition for Operation (LCO) Applicability Note for Technical Specification (TS) 3.3.9, “Boron Dilution Mitigation System (BDMS).” The LCO Applicability Note was revised to clarify the situations during which the BDMS signal may be blocked in MODES 2 and 3.

Date of issuance: June 29, 2010.

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

Amendment No.: 200.

Facility Operating License No. NPF-30: The amendment revised the Operating License and Technical Specifications.

Date of initial notice in Federal Register: August 25, 2009 (74 FR Start Printed Page 3998542933). The supplemental letters dated August 27, 2009, and March 4, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

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

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 1st day of July 2010.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

End Signature End Preamble

[FR Doc. 2010-16879 Filed 7-12-10; 8:45 am]

BILLING CODE 7590-01-P