2010-11564. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC) is publishing this regular biweekly notice. The Act requires that 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 April 22 to May 5, 2010. The last biweekly notice was published on May 4, 2010 (75 FR 23808).

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 basis for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail at [e-mail], or by telephone at [phone], to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.html. System requirements for accessing the E-Submittal server are detailed in NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through EIE, users will be required to install a Web browser plug-in from the NRC Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the agency's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html, by e-mail at [e-mail], or by a toll-free call at [phone]. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/​EHD_​Proceeding/​home.asp, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).

For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, 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 [phone], [phone], or by e-mail to [e-mail].

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

Date of amendment request: March 29, 2010.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) 5.5.7, “Reactor Coolant Pump Flywheel Inspection Program,” by extending the reactor coolant pump (RCP) motor flywheel inspection interval for certain RCP motors from the currently-approved 10-year inspection interval to an interval not to exceed 20 years. The availability of this TS revision was announced in the Federal Register on October 22, 2003 (68 FR 60422) as part of the consolidated line item improvement process. In its application, the licensee affirmed the applicability of the model no significant hazards consideration determination, as published in the Federal Register on June 24, 2003 (68 FR 37590).

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

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

The proposed change to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's [Nuclear Regulatory Commission's] acceptance guidelines contained in RG 1.174 [Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis”] (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits; or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the analysis adopted by the licensee 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 amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Stephen J. Campbell.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: March 19, 2010.

Description of amendment request: This submittal requests changes to extend the Technical Specification (TS) allowed outage time (AOT) for the Unit 1 and Unit 2 Suppression Pool Cooling (SPC) mode of the Residual Heat Removal (RHR) system, the Residual Heat Removal Service Water (RHRSW) system, the Emergency Service Water (ESW) system, and the A.C. Sources-Operating (Emergency Diesel Generators) from 72 hours to seven (7) days in order to allow for repairs of the RHRSW system piping. Specifically, the proposal adds a footnote to the affected TS limiting conditions for operation to indicate that the 72-hour AOT for the affected system may be extended once per calendar year, for one unit only, for a period of up to 7 days to allow for repairs of one RHRSW subsystem piping with the opposite unit shutdown, reactor vessel head removed and reactor cavity flooded, and other specific compensatory measures in effect.

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

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

Response: No.

The proposed TS changes will not increase the probability of an accident since they will only extend the time period that one RHRSW subsystem, one loop of SPC, one ESW loop and two Emergency Diesel Generators (EDGs) can be out of service. The extension of the time duration that one RHRSW, one ESW loop and two EDGs are out of service has no direct physical impact on the plant. The proposed inoperable RHRSW subsystem, ESW loop and two EDGs are normally in a standby mode while the unit is in [Operational Condition] OPCON 1 or 2 and are not directly supporting plant operation. Therefore, they can have no impact on the plant that would make an accident more likely to occur due to their inoperability.

During transients or events which require these subsystems to be operating, there is sufficient capacity in the operable loops/subsystems and available[,] but inoperable[,] equipment to support plant operation or shutdown. Therefore, failures that are accident initiators will not occur more frequently than previously postulated as a result of the proposed changes.

In addition, the consequences of an accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR) will not be increased. With one RHRSW subsystem inoperable, one SPC loop, one ESW loop and two EDGs inoperable but verified available prior to entering the proposed configuration, a known quantity of equipment is inoperable. Based on the support functions of the RHRSW system, a review of the plant was performed to determine the impacts that the inoperable RHRSW subsystem would have on other systems. The impacts were identified for each system and it was determined whether there were any adverse effects on the systems. It was then determined how the adverse effects would impact each system's design basis and overall plant safety. The consequences of any postulated accidents occurring on Unit 1 or Unit 2 during these AOT extensions was found to be bounded by the previous analyses as described in the UFSAR. Since the inoperable ESW loop, selected emergency core cooling system (ECCS) pumps and EDGs will be verified available prior to entering the proposed configuration, they would have no impact on other systems.

The minimum equipment required to mitigate the consequences of an accident and/or safely shut down the plant will be operable or available. Therefore, by extending certain AOTs and extending the assumptions concerning the combinations of events for the longer duration of each extended AOT, Exelon concludes that at least the minimum equipment required to mitigate the consequences of an accident and/or safely shut down the plant will still be operable or available during the extended AOT.

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

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

Response: No.

The proposed TS changes will not create the possibility of a different type of accident since they will only extend the time period that one RHRSW subsystem and one loop of SPC can be out of service, and one ESW loop and two EDGs can be inoperable, but verified available, prior to entering the proposed configuration. The extension of the time duration that one RHRSW subsystem and one SPC loop is out of service, and one ESW loop and two EDGs are inoperable, but verified available, prior to entering the proposed configuration has no direct physical impact on the plant and does not create any new accident initiators. The systems involved are accident mitigation systems. All of the possible impacts that the inoperable equipment may have on its supported systems were previously analyzed in the UFSAR and are the basis for the present TS Action statements and AOTs. The impact of inoperable support systems for a given time duration was previously evaluated and any accident initiators created by the inoperable systems was evaluated. The lengthening of the time duration does not create any additional accident initiators for the plant.

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

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

Response: No.

The present RHRSW, SPC, ESW and EDG AOT limits were set to ensure that sufficient safety-related equipment is available for response to all accident conditions and that sufficient decay heat removal capability is available for a loss of coolant accident (LOCA) coincident with a loss of offsite power (LOOP) on one unit and simultaneous safe shutdown of the other unit. A slight reduction in the margin of safety is incurred during the proposed extended AOT due to the increased risk that an event could occur in a 7-day period versus a 72-hour period. This increased risk is judged to be minimal due to the low probability of an event occurring during the extended AOT and based on the following discussion of minimum ECCS/decay heat removal requirements.

The inoperable ESW loop, selected ECCS pumps and EDGs will be verified available prior to entering the proposed configuration; therefore, extension of the AOT will have no effect on the minimum ECCS equipment available or margin of safety.

The reduction in the margin of safety from the extension of the RHRSW, SPC, ESW and EDG AOT limits is not significant since the remaining operable ECCS equipment is adequate to mitigate the consequences of any accident. This conclusion is based on the information contained in General Electric Company documents NEDO-24708A, “Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors,” Revision 1, dated December 1980, and NEDC[-]3093P-A, “BWR Owner's Group Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Activation Instrumentation),” dated December 1988. These documents describe the minimum requirements to successfully terminate a transient or LOCA initiating event (with scram), assuming multiple failures with realistic conditions, and were used to justify certain TS AOTs per UFSAR Sections 6.3.1.1.2.o and 6.3.3.1. The minimum requirements for short-term response to an accident would be either one Low Pressure Coolant Injection (LPCI) pump or one Core Spray subsystem in conjunction with Automatic Depressurization System (ADS), or the High Pressure Coolant Injection (HPCI) system, which would be adequate to re-flood the vessel and maintain core cooling sufficient to preclude fuel damage. For long-term response, the minimum requirements would be one loop of RHR for decay heat removal, along with another low-pressure ECCS subsystem. These minimum requirements will be met since implementation of the proposed TS changes will require the operability or availability of HPCI, ADS, two LPCI subsystems (or one LPCI subsystem and one RHR subsystem during decay heat removal) and one Core Spray subsystem be maintained during the 7-day period. Operations personnel are fully qualified by normal periodic training to respond to and mitigate a Design Basis Accident, including the actions needed to ensure decay heat removal while LGS Unit 1 and Unit 2 are in the operational configurations described within this submittal. Accordingly, procedures are already in place that address safe plant shutdown and decay heat removal for situations applicable to those in the proposed AOTs.

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

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

Attorney for licensee: J. Bradley Fewell, Esquire, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Harold K. Chernoff.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: March 24, 2010.

Description of amendment request: The proposed amendment would modify the Three Mile Island, Unit 1 (TMI-1) Technical Specifications (TSs) by relocating specific surveillance frequencies to a new licensee-controlled program called the Surveillance Frequency Control Program. This change incorporates the adoption of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,” Revision (Rev.) 1. A description of the Surveillance Frequency Control Program will be added to the TMI-1 TSs.

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

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

Response: No.

The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program [SFCP]. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

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

Response: No.

No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

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

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

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

Attorney for licensee: J. Bradley Fewell, Esquire, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Harold K. Chernoff.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket No. 50-277, Peach Bottom Atomic Power Station (PBAPS), Unit 2, York and Lancaster Counties, Pennsylvania

Date of amendment request: August 28, 2009, as supplemented by letter dated February 25, 2010.

Description of amendment request: The proposed change would modify the PBAPS Unit 2 Technical Specification (TS) Section 5.5.12 to reflect a one-time extension of the Type A containment Integrated Leak Rate Test (ILRT) to no later than October 2015. The proposed TS revision would allow a one-time extension of 5 years to the 10-year frequency of the performance-based leakage rate testing program for the PBAPS Unit 2 containment Type A ILRT test.

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 involves a one-time extension of the Primary Containment ILRT interval from 10 years to 15 years. The proposed change does not involve a physical change to the plant [* * *]. The Primary Containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed change.

Continued containment integrity is assured by the established programs for local leak rate testing and inservice/containment inspections, which are unaffected by the proposed change. As documented in NUREG-1493, “Performance-Based Containment Leak-Test Program,” dated September 1995, industry experience has shown that local leak rate tests (Type B and C) have identified the vast majority of containment leakage paths, and that ILRTs detect only a small fraction of containment leakage pathways.

The potential consequences of the proposed change have been quantified by analyzing the changes in risk that would result from extending the ILRT interval from 10 years to 15 years. Increasing the ILRT interval to 15 years for this one-time change is considered to be insignificant since it represents a very small change to the PBAPS, Unit 2 risk profile. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. PBAPS, Unit 2 has determined that the increase in conditional containment failure probability due to the proposed change is very small. Therefore, it is concluded that the proposed one-time extension of the Primary Containment ILRT interval from 10 years to 15 years does not significantly increase the consequences of an accident previously evaluated.

Based on the above discussion, it is concluded that 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 involves a one-time extension of the Primary Containment ILRT interval. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed)[* * *].

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 one-time extension of the Primary Containment ILRT interval does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the 10 CFR 50 Appendix J testing program plan, as defined in the Technical Specifications, exist to ensure that the degree of Primary Containment structural integrity and leak-tightness that is considered in the plant safety analyses is maintained. The overall containment leakage rate limit specified by the Technical Specifications is maintained, and Type B and C containment leakage tests will continue to be performed at the frequency currently required by the TS.

Containment inspections performed in accordance with [the * * *] plant programs [described above] serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by an ILRT. Furthermore, a risk assessment using the current PBAPS, Unit 2 Probabilistic Risk Assessment internal events model concluded that extending the ILRT test interval from 10 years to 15 years results in a very small change to the PBAPS, Unit 2 risk profile.

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 with the NRC staff changes noted in square brackets above, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. J. Bradley Fewell, Associate General Counsel, Exelon Generation Company LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Harold K. Chernoff.

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

Date of amendment request: February 16, 2010.

Description of amendment request: To revise the licensing bases by removing two technical specifications (TSs) that restrict movements of heavy loads over the spent fuel pools.

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:

TS 3/4.9.7, Crane Travel-Spent Fuel Storage Areas (reviewed for both units)

FPL has evaluated whether or not a significant hazards consideration is involved with removing the TS 3/4.9.7, “Crane Travel—Spent Fuel Storage Areas,” from the Turkey Point Units 3 and 4 TS by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below:

(1) Would operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The removal of TS 3/4.9.7 will not increase the probability of a fuel handling accident (FHA), as evaluated in Chapter 14.2.1 of the UFSAR [Updated Final Safety Analysis Report], and is considered remote because of the administrative controls and physical limitations imposed on fuel handling operations. The load limit restriction, in conjunction with existing plant documents (for example, Turkey Point heavy load handling procedures) that restrict crane or other heavy load handling operations provide a defense-in-depth approach to handling heavy loads in the spent fuel pool vicinity. The load limitation defined in TS 3/4.9.7 is preserved and will be implemented based on the operation limits and safety margins for the control of heavy loads consistent with NUREG-0612. The TS change does not represent any physical change to the plant systems, structures, or components. Therefore, the systems credited with mitigating the dose consequences of a FHA remain in place. The dose consequences of a fuel handling accident as discussed in Turkey Point UFSAR Chapter 14.2.1 will not increase because of the administrative controls and physical limitations imposed on fuel handling operations which minimize the likelihood of a FHA.

Therefore, facility operation in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) Would operation of the facility in accordance with the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The removal of TS 3/4.9.7 does not represent any physical change to the plant systems, structures, or components. The same operational functions of moving new fuel, spent fuel, or other loads over the spent fuel pool are retained and therefore do not create or increase the possibility of a new or different kind of accident from any accident previously evaluated. Additionally, the load limit of 2000 pounds over the spent fuel pool defined in TS 3/4.9.7 is preserved and implemented in existing plant documents and are established based on the operational limits and safety margins for the control of heavy loads consistent with NUREG-0612. Other measures which preclude the creation of a new or different type of accident include interlocks and physical stops, operator training, and load handling procedures.

Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Would operation of the facility in accordance with the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The removal of TS 3/4.9.7 does not change the operational process of moving loads over the spent fuel pool. There are no changes to any physical plant systems, structures, or components. The spent fuel handling crane has weight sensors that are interlocked to limit the total load. In addition, an in-line weight sensing system is provided for each hoist to limit the lifting load to preclude accidental fuel damage should binding occur. When lifting over spent fuel, the total load is limited to 2000 pounds by current procedures, limit switches and load sensors. Because of these measures, no margin of safety is reduced or compromised.

Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.

Based on the above, FPL concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

TS 3/4.9.12, Handling of Spent Fuel Cask (reviewed for both units)

FPL has evaluated whether or not a significant hazards consideration is involved with the proposed amendment of removing TS 3/4.9.12, “Handling of Spent Fuel Cask,” by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below:

(1) Would operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The removal of TS 3/4.9.12 will not involve a significant increase in the probability or consequences of an accident previously evaluated. The accident evaluated for the existing spent fuel cask handling crane is the drop of a single element cask as cited in UFSAR Section 14.2.1.3, “Cask Drop Accident.” This cask drop accident was analyzed and the radiological dose consequence, as a result of the cask drop, is determined to be within the limits of 10 CFR 100. The current spent fuel cask handling crane at Turkey Point Units 3 and 4 has a single 105/15 ton main/auxiliary hook design capacity and is not designed as single-failure-proof. The new spent fuel cask handling crane will be single-failure-proof meeting all of the requirements of NUREG-0554, “Single Failure Proof Cranes for Nuclear Power Plants” and also NUREG-0612, Section 5.1.6, “Single Failure Proof Handling Systems.” The probability of a cask drop accident using a single-failure-proof crane designed and operated to these NUREG requirements is considered to be extremely small.

The design for the upgrade of the spent fuel cask handling crane is to increase the capacity to 130/25 tons (main/auxiliary hook). All crane components (hoist, bridge, girders, etc.) are designed and fabricated to retain control of and hold the maximum critical load (a planned 32 element spent fuel cask) in the unlikely event of the failure of a single component, coincident with a Design or Maximum earthquake.

The objectives cited in Section 5.1 of NUREG-0612, “Recommended Guidelines,” for the control of heavy loads are satisfied. The probability of a cask drop accident using the new single-failure-proof spent fuel cask crane, as compared to the existing non-single-failure-proof crane, is therefore not increased. The increase of the consequences of an accident previously evaluated is also not increased because the potential for a cask drop by the new upgraded spent fuel cask handling crane is considered to be extremely small.

Further, operational limits, interlocks, procedural and administrative controls, that restrict the handling of heavy loads over fuel stored in the spent fuel pool, provide additional defense-in depth to ensure that a load could not be dropped that would result in dose consequences greater than previously evaluated.

It is concluded that facility operation in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) Would operation of the facility in accordance with the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Operation of the spent fuel cask handling crane after the upgrade to a single-failure-proof design will remain the same as the operation of the existing spent fuel cask handling crane. The distinction is the load that will be lifted.

The new spent fuel cask is a multiple assembly cask, in contrast to a single assembly cask as currently specified for use. The current spent fuel cask handling crane is designed to lift a single element spent fuel cask. The upgraded capacity of the new spent fuel cask handling crane will allow for lifting a cask designed to hold a maximum of 32 spent fuel assemblies. Current operating and administrative procedures that restrict the movement of heavy loads over fuel stored in the spent fuel pool remain in place. The new spent fuel cask handling crane is designed, fabricated and tested to single-failure-proof requirements (NUREG-0554, “Single Failure Proof Cranes for Nuclear Power Plants” and NUREG-0612, Section 5.1.6, “Single Failure Proof Handling Systems”) and will be operated within the procedural and administrative framework as the currently installed spent fuel cask handling crane. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created from the removal of TS 3/4.9.12.

Therefore, it can be concluded that the operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Would operation of the facility in accordance with the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The existing spent fuel cask handling crane is not designed as single-failure-proof in accordance with NUREG-0612. The new spent fuel cask handling crane is designed, and will be fabricated, installed and tested to the single-failure-proof requirements as outlined in NUREG-0612, Section 5.1.6, “Single Failure Proof Handling Systems.” The use of the defense-in-depth approach for the control and handling of heavy loads as cited in Section 5.1 of NUREG-0612, “Recommended Guidelines,” provides assurance that there is a sufficient margin of safety in the handling of heavy loads. Thereby, the removal of TS 3/4.9.12 will not involve a significant reduction in the margin of safety.

Defense-in-depth measures include operational limits, interlocks, procedural and administrative controls, rigging, load paths, testing, training, maintenance and other related considerations. These measures provide assurance that the margin of safety is not reduced in the operation of the facility by meeting all the requirements of NUREG-0612 and NUREG-0554. The specific requirements and FPL compliance with them is documented in the NUREG-0554 Compliance Matrix [Attachment 3 to this application].

The design for the upgrade of the spent fuel cask handling crane is to increase the capacity to 130/25 tons (main/auxiliary hook). The spent fuel cask handling crane has a Main Hoist and Auxiliary Hoist Cable Safety Factor of a minimum 10:1 on nominal breaking strength at 130 tons and 25 tons respectively and is fully compliant with ASME NOG-1 Section 5425.1. The Main Hoist Hook and Auxiliary Hoist Hook Safety Factor have a 10:1 minimum on ultimate strength at 130 tons and 25 tons, respectively.

Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.

Based on the above, FPL concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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 Acting Branch Chief: Douglas A. Broaddus.

Date of amendment request: January 14, 2010.

Description of amendment request: The amendments would revise a number of Technical Specification (TS) requirements, to impose similar restrictions on the movement of non-irradiated fuel assemblies to those currently in place for movement of irradiated fuel assemblies. The additional restrictions will limit the movement of all fuel assemblies over irradiated fuel assemblies in containment or in the fuel storage pool. The affected TS Limiting Conditions for Operation (LCOs) are: LCO 3.3.8, “Containment Purge Isolation Signal (CPIS),” LCO 3.3.9, “Control Room Isolation Signal (CRIS),” LCO 3.7.11, “Control Room Emergency Air Cleanup System (CREACUS),” LCO 3.7.16, “Fuel Storage Pool Water Level,” LCO 3.8.2, “AC Sources—Shutdown,” LCO 3.8.5, “DC Sources—Shutdown,” LCO 3.8.8, “Inverters—Shutdown,” LCO 3.8.10, “Distribution Systems—Shutdown,” LCO 3.9.3, “Containment Penetrations,” and LCO 3.9.6, “Refueling Water Level.”

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

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

Response: No.

This proposed change revises Technical Specifications applicability wording regarding the movement of fuel assemblies in containment and the fuel storage pool at the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 to include the movement of both irradiated and non-irradiated fuel assemblies. The proposed applicability is more comprehensive than the current Applicability.

Expanding the applicability of the relevant Technical Specifications is necessary to account for updated fuel drop analyses which demonstrate that impacted spent fuel assemblies may be damaged. Consequently, movement of non-irradiated fuel assemblies could result in a Fuel Handling Accident that has radiological consequences. Changing the applicability of the relevant Technical Specifications does not affect the probability of a Fuel Handling Accident. The expanded applicability provides assurance that equipment designed to mitigate a Fuel Handling Accident is capable of performing its specified safety function, such that the consequences of an accident are not increased.

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

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

Response: No.

The revised spent fuel drop analyses demonstrate that impacted fuel assemblies may be damaged as the result of a dropped fuel assembly. The existing SONGS Technical Specifications regarding movement of fuel assemblies are not applicable for movement of non-irradiated fuel assemblies. A drop of a non-irradiated fuel assembly that has radiological consequences could occur during periods when equipment that would be required to mitigate those consequences is not required to be OPERABLE in accordance with the existing Technical Specifications.

The proposed changes to the Technical Specifications applicability language regarding the movement of fuel assemblies in containment and the fuel storage pool at SONGS Units 2 and 3 ensure that Limiting Conditions of Operation and appropriate Required Actions for required equipment are in effect during fuel movement. This provides assurance that any Fuel Handling Accident that may occur will remain within the initial assumptions of accident analyses.

Consequently, there is no possibility of a new or different kind of accident due to this change.

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

Response: No.

The proposed Technical Specifications change will not affect protection criterion for plant equipment and will not reduce the margin of safety. By extending the Applicability to the movement of non-irradiated fuel assemblies, the current margin of safety is maintained.

Consequently, there is no significant reduction in a margin of safety due to this change.

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

Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: Michael T. Markley.

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

Date of amendment request: November 25, 2009.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” that would add a new Required Action Q.1 to require restoration of an inoperable Balance of Plant (BOP) ESFAS train to OPERABLE status within 24 hours. In addition, the Completion Times for TS 3.3.2 Required Actions J.1 and O.1 to trip inoperable channels that provide inputs to BOP ESFAS would also be extended to 24 hours. Shutdown track Completion Times to be in MODES 3 and 4 would be increased to reflect longer restoration times. Separate Condition entry for TS Condition J would be restricted to assure that Function 6.g in TS Table 3.3.2-1 will provide a start signal to the motor-driven auxiliary feedwater pumps from one train of BOP ESFAS actuation logic.

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.

Overall protection system performance will remain within the bounds of the previously performed accident analyses since no hardware changes are proposed to the protection systems. The same reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) instrumentation will continue to be used. The protection systems will continue to function in a manner consistent with the plant design basis. There will be no changes to the BOP ESFAS surveillance and operating limits.

The proposed changes will not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits.

The proposed changes do not affect the way in which safety-related systems perform their functions.

All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the FSAR [Final Safety Analysis Report].

The applicable radiological dose acceptance criteria will continue to be met.

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

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

Response: No.

There are no proposed changes in the method by which any safety-related plant SSC performs its 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. The proposed changes will not alter any assumptions made in the safety analyses.

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, Solid State Protection System, BOP ESFAS, MSFIS [main steam/feedwater isolation system], or LSELS [load shedder and emergency load sequencer] used in the plant protection systems.

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.

There will be no effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor (F_Q), nuclear enthalpy rise hot channel factor (FΔH), loss of coolant accident peak cladding temperature (LOCA PCT), peak local power density, or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met.

The proposed changes do not eliminate any surveillances or alter the frequency of surveillances required by the Technical Specifications. No instrument setpoints or system response times are affected. None of the acceptance criteria for any accident analysis will be changed.

The proposed changes will have no impact on the radiological consequences of a design basis accident.

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

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

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

NRC Branch Chief: Michael T. Markley.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia

Date of amendment request: March 30, 2010.

Description of amendment request: The proposed amendments would modify the North Anna Technical Specifications (TSs) by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, “Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies.” The changes are consistent with NRC-approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) change TSTF-425, Revision 3. The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of this TS improvement.

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

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

Response: No.

The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program. Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

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

Response: No.

No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

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

Response: No.

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

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

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

Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.

NRC Branch Chief: Gloria Kulesa.

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.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of application for amendments: June 25, 2008, as supplemented on November 6, 2008, March 9, 2009, June 12, 2009, December 18, 2009, and March 26, 2010.

Brief description of amendment request: The proposed amendment would revise the PBAPS, Units 2 and 3, Technical Specification Section 4.3.1.1.a concerning the spent fuel pool k-infinity value.

Date of publication of individual notice in Federal Register: April 26, 2010 (75 FR 21680).

Expiration date of individual notice: May 26, 2010 (comment request); June 25, 2010 (hearing request).

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 [phone], [phone] or by e-mail to [e-mail].

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, Millstone Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: November 23, 2009, as supplemented by letter dated April 26, 2010.

Brief description of amendment: The license amendment request revises the Millstone Power Station, Unit 3 (MPS3) Technical Specification (TS) 6.8.4.g, “Steam Generator Program,” to exclude a portion of the tubes below the top of the steam generator tubesheet from periodic steam generator tube inspections. This request also removes reference to the previous Cycle 13 interim alternate repair criteria.

Date of issuance: May 3, 2010.

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

Amendment No.: 249.

Renewed Facility Operating License No. NPF-49: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: January 26, 2010 (75 FR 4114). The supplemented dated April 26, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 3, 2010.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 1, 2008.

Brief description of amendments: The amendments correct a non-conservative Technical Specification (TS) Surveillance Requirement by revising McGuire TS 3.8.1.4 to increase the minimum required amount of fuel oil for the Emergency Diesel Generators fuel oil day tank as read on the local fuel gauge used to perform the surveillance.

Date of issuance: May 5, 2010.

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

Amendment Nos.: 254 and 234.

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: May 19, 2009 (74 FR 23442).

The supplements dated July 30, 2009, December 2, 2009, and March 10, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of application for amendments: August 7, 2008, as supplemented on May 7, 2009, and January 19, 2010.

Brief description of amendments: The August 7, 2008, submittal contained several areas of review that are being dispositioned as separate amendment requests. The amendments associated with this notice revise the PBAPS Units 2 and 3 Technical Specifications (TS) to delete the list of emergency diesel generator critical trips from TS Surveillance Requirement (SR) 3.8.1.13 and clarify that the purpose of the SR is to verify that the non-critical trips are bypassed. This TS change adopts Technical Specification Task Force (TSTF) Traveler 400, Revision 1, “Clarify SR on Bypass of DG [diesel generator] Automatic Trips.”

Date of issuance: April 30, 2010.

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

Amendment Nos.: 275 and 279.

Renewed Facility Operating License Nos. DPR-44 and DPR-56: Amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: May 5, 2009 (74 FR 20744).

The supplements dated May 7, 2009, and January 19, 2010, clarified the application, did not expand the scope of the application as originally noticed, and did not change the initial proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: April 2, 2009.

Brief description of amendments: The amendment revised Technical Specification (TS) 3.3.1 entitled, “Reactor Trip System (RTS) Instrumentation” to add Surveillance Requirement 3.3.1.16 to Function 3 of TS Table 3.3.1-1 to verify that the RTS response times are within limits every 18 months on staggered basis. The change is based on a reanalysis of the Rod Cluster Control Assembly Bank Withdrawal at Power event.

Date of issuance: April 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—151; Unit 2-151.

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

Date of initial notice in Federal Register: May 19, 2009 (74 FR 23446).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 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.

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 [phone], [phone] or by e-mail to [e-mail].

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 [phone], [phone], or by e-mail to [e-mail]. 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.

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 [e-mail], or by telephone at [phone], to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals/​apply-certificates.html. System requirements for accessing the E-Submittal server are detailed in NRC's “Guidance for Electronic Submission,” which is available on the agency's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through EIE, users will be required to install a Web browser plug-in from the NRC Web site. Further information on the Web-based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http://www.nrc.gov/​site-help/​e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/​site-help/​e-submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.

A person filing electronically using the agency's adjudicatory E-Filing system may seek assistance by contacting the NRC Meta System Help Desk through the “Contact Us” link located on the NRC Web site at http://www.nrc.gov/​site-help/​e-submittals.html,, by e-mail at [e-mail], or by a toll-free call at [phone]. The NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, excluding government holidays.

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

Documents submitted in adjudicatory proceedings will appear in NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/​EHD_​Proceeding/​home.asp,, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.

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

Date of application for amendment: March 29, 2010, as supplemented by letters dated March 29 and April 26, 2010.

Brief description of amendment: The amendment revised Technical Specification (TS) 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” Condition J under function 6.g in TS Table 3.3.2-1. Function 6.g provides an auxiliary feedwater (AFW) start signal that is provided to the motor-driven AFW pumps in the event of a trip of both turbine-driven main feedwater (MFW) pumps. The licensee determined that the design and normal operation of the MFW pumps could result in a condition that does not conform to TS Table 3.3.2-1, function 6.g. Entry into Limiting Condition for Operation (LCO) 3.0.3 will be required; therefore, the TS change was needed to address this condition. The change to Condition J allows placing the two channels in a tripped condition on one MFW pump when placing the pump into service or removing the pump from service prior to resetting the MFW pump. With the revision to Condition J, the licensee will not require an entry into LCO 3.0.3. Specifically, the changes revised Condition J for ESFAS instrumentation function 6.g to read, “One or more Main Feedwater Pumps trip channel(s) inoperable,” made corresponding changes to Required Action J.1, and placed a Note above Required Actions J.1 and J.2 for consistency with the revised Condition.

Date of issuance: May 5, 2010.

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

Amendment No.: 196.

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

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes (75 FR 19431; April 14, 2010).

The supplemental letters dated March 29 and April 26, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed NSHC determination as published in the Federal Register. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided an opportunity to request a hearing by June 14, 2010, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.

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

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

NRC Branch Chief: Michael T. Markley.

Dated at Rockville, Maryland, this 6th day of May 2010.

For the Nuclear Regulatory Commission.

Joseph G. Giitter,

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

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.