E7-6632. 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 staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from March 16, 2007 to March 29, 2007. The last biweekly notice was published on March 27, 2007 (72 FR 14303).

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

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

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.

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

Written comments may be submitted by mail to the Chief, Rulemaking, Directives and Editing Branch, 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 delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. 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 persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide 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 within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

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

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

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed 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; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, [e-mail]; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at [phone], verification number is [phone]. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to [phone] or by e-mail to [e-mail]. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 [phone], [phone] or by e-mail to [e-mail].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois

Date of amendment request: November 13, 2006.

Description of amendment request: The proposed amendment changes the technical specification (TS) testing frequency for the surveillance requirement (SR) in TS 3.1.4, “Control Rod Scram Times.” The proposed change revises the test frequency of SR 3.1.4.2, control rod scram time testing, from “120 days cumulative operation in MODE 1” to “200 days cumulative operation in Mode 1.”

AmerGen has reviewed the proposed no significant hazards consideration determination published in the Federal Register on August 23, 2004 (69 FR 51864), as part of the consolidated line item improvement process (CLIIP) and has concluded that the proposed determination presented in the notice is applicable to Clinton Power Station, Unit No. 1.

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

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

Response: No.

The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The frequency of surveillance testing is not an initiator of any accident previously evaluated. The frequency of surveillance testing does not affect the ability to mitigate any accident previously evaluated, as the tested component is still required to be operable. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change does not result in any new or different modes of plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change continues to test the control rod scram time to ensure the assumptions in the safety analysis are protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, the proposed change presents no 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.

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

Attorney for licensee: Mr. Bradley J. Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Russell Gibbs.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of amendment request: January 26, 2007.

Description of amendment request: The proposed amendment would revise technical specifications (TS) requirements for unavailable barriers by adding limiting condition for operation (LCO) 3.0.9. This would establish conditions under which TS systems would remain operable when required physical barriers are not capable of providing their related support function. Also, the proposed amendment would make editorial changes to LCO 3.0.8 to be consistent with the terminology in LCO 3.0.9.

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 by a reference to a generic analysis published in the Federal Register on October 3, 2006 (71 FR 58444), which is presented below:

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

Response: No.

The proposed change allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an unavailable barrier if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident while relying on the allowance provided by proposed LCO 3.0.9 are no different than the consequences of an accident while relying on the TS required actions in effect without the allowance provided by proposed LCO 3.0.9. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to an unavailable barrier, if risk is assessed and managed, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

Response: No.

The proposed change allows a delay time for entering a supported system TS when the inoperability is due solely to an unavailable barrier, if risk is assessed and managed. The postulated initiating events which may require a functional barrier are limited to those with low frequencies of occurrence, and the overall TS system safety function would still be available for the majority of anticipated challenges. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A bounding risk assessment was performed to justify the proposed TS changes. This application of LCO 3.0.9 is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The net change to the margin of safety is insignificant as indicated by the anticipated low levels of associated risk (ICCDP and ICLERP) as shown in Table 1 of Section 3.1.1 in the Safety Evaluation.

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

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

Attorney for licensee: David G. Pettinari, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.

NRC Branch Chief: L. Raghavan.

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

Date of amendment request: January 18, 2007.

Description of amendment request: The proposed amendment would revise the expiration limit for the reactor coolant system Pressure/Temperature (P/T) limit graphs in Technical Specifications (TS); revise the adjusted reference temperature for the reactor vessel; and revise the Low Temperature Overpressure Protection (LTOP) arming temperature value specified in TSs. It would also make editorial changes in the use of inequality signs in TSs associated with the LTOP arming temperature in order to make them consistent.

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

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

Response: No

The proposed change does not affect the accident initiators or mitigation assumptions associated with any of the accidents previously evaluated. Operating restrictions on pressure-temperature conditions for the reactor pressure vessel provide assurance that reactor vessel integrity will be maintained under accident or transient conditions. The proposed change uses approved criteria and analysis methods to update the time period for which the current operating limits remain valid.

The LTOP system performs an automatic function by opening relief valves if reactor coolant system pressure reaches a temperature-dependent limit. The proposed change includes establishing a more restrictive temperature limit for when this system must be in service, to reflect the material condition of the reactor vessel at the new EFPY limit proposed for the pressure-temperature graphs. The mitigation function and capability of the LTOP system is not being changed by this request.

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 new accident initiators being introduced by this proposed change. The proposed change does not involve installation of new plant equipment, modification of existing equipment, or changes in the way that plant equipment is operated. Pressure-temperature operating limits depicted by graphs in the technical specifications will not be changed and will continue to be used by plant operators. A change in the LTOP system arming temperature will assure that the graphs remain valid for the proposed new operating period of 27.2 EFPY [effective full power years].

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

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

Response: No

Operating limits on pressure and temperature conditions for the reactor coolant system (RCS) are important to assure that the RCS pressure boundary stresses are within analyzed limits. Margins of safety are inherent in the analysis methods, assumptions, and limits specified in regulations and guidance documents. The proposed change is based on NRC-accepted methods, assumptions and limits and maintains the required margin of safety.

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: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Acting Branch Chief: Douglas V. Pickett.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 (IP2) and 3 (IP3), Westchester County, New York

Date of amendment request: March 13, 2007.

Description of amendment request: The amendment would revise License Condition 2.K for IP2 and License Condition 2.H for IP3, which require the implementation and maintenance of an approved Fire Protection Program for each unit.

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

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

Response: No

The proposed changes are strictly an administrative relocation of the specific fire protection SER [safety evaluation report] references and do not modify any requirements of the fire protection programs.

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.

The proposed changes are strictly an administrative relocation of the specific fire protection SER references and do not modify any requirements of the fire protection programs.

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

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

Response: No.

The proposed changes are strictly an administrative relocation of the specific fire protection SER references and do not modify any requirements of the fire protection programs.

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: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Acting Branch Chief: Douglas V. Pickett.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: March 22, 2007.

Description of amendment request: The proposed amendment will revise the test acceptance criteria specified in Technical Specification Surveillance Requirement (SR) 3.8.1.10 for the diesel generator endurance 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed change revises the acceptance criteria to be applied to an existing surveillance test of the facility emergency diesel generators (DGs). Performing a surveillance test is not an accident initiator and does not increase the probability of an accident occurring. The proposed new acceptance criteria will assure that the DGs are capable of carrying the peak electrical loading assumed in the various existing safety analyses which take credit for the operation of the DGs. Establishing acceptance criteria that bound existing analyses validates the related assumption used in those analyses regarding the capability of equipment to mitigate accident conditions. Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. The proposed change revises the test acceptance criteria for a specific performance test conducted on the existing DGs. The proposed change does not involve installation of new equipment or modification of existing equipment, so no new equipment failure modes are introduced. The proposed revision to the DG surveillance test acceptance criteria also is not a change to the way that the equipment or facility is operated and no new accident initiators are created. 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 change involve a significant reduction in a margin of safety?

No. The conduct of performance tests on safety-related plant equipment is a means of assuring that the equipment is capable of maintaining the margin of safety established in the safety analyses for the facility. The proposed change in the DG technical specification surveillance test acceptance criteria is consistent with values assumed in existing safety analyses and is consistent with the design rating of the DGs. Therefore the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Acting Branch Chief: Douglas V. Pickett.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: February 15, 2007.

Description of amendment request: The proposed changes would revise Technical Specification (TS) 3.10.1 to expand its scope to include provisions for reactor coolant temperature excursions greater than 212 °F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4, which is defined to be reactor coolant temperature less than or equal to 212 °F.

This change was proposed by the industry's TS Task Force (TSTF) and is designated TSTF-484. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 21, 2006 (71 FR 48561), on possible amendments concerning TSTF-484, including a model safety evaluation and model no significant hazards (NSHC) determination, using the consolidated line item improvement process (CLIIP). The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on October 27, 2006 (71 FR 63050). The licensee affirmed the applicability of the following NSHC determination in its application dated February 15, 2007.

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

Criterion 1: The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. 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 or eliminate any existing 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 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.

Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an inservice leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, the NRC staff concludes that the proposed change presents no 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.

Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Acting Branch Chief: Douglas V. Pickett.

Exelon Generation Company, LLC (EGC), Docket Nos. 50-373 and 50-374, LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

Date of amendment request: November 17, 2006.

Description of amendment request: The proposed amendments would replace references to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) with a reference to the ASME Code of Operation and Maintenance of Nuclear Power Plants (OM Code) in Technical Specification (TS) 5.5.7, “Inservice Testing Program [IST].” These proposed changes are consistent with the implementation of the LSCS, Units 1 and 2 third 10-year IST program in accordance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, “Codes and standards,” paragraph (f), “Inservice testing requirements.” The third 10-year interval for LSCS, Units 1 and 2 is scheduled to start on October 12, 2007.

In addition to the replacement of the references, EGC is also adding provisions in TS 5.5.7, item b, to only apply Surveillance Requirement (SR) 3.0.2 to those inservice testing frequencies of two years or less. These proposed changes are based on TS Task Force (TSTF) Traveler No. 479-A (TSTF-479-A), Revision 0, “Changes to Reflect Revision of 10 CFR 50.55a,” as modified by TSTF-497, Revision 0, “Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less” and approved by the NRC in December 6, 2005, and October 4, 2006.

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

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

Response: No.

The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to conform to the requirements of 10 CFR 50.55a, “Codes and standards,” paragraph (f) regarding the inservice testing of pumps and calves for the Third 10-year Interval. The current TS reference the [American Society of Mechanical Engineers] ASME Boiler and Pressure Vessel Code, Section XI, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes would reference the ASME OM Code, which is consistent with 10 CFR 50.55a, paragraph (f), “Inservice testing requirements,” and approved for use by the NRC. In addition, provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only applied to those inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by this license amendment request.

The proposed changes are administrative in nature, do not affect any accident initiators, do not affect the ability of LSCS to successfully respond to previously evaluated accidents and do not affect radiological assumptions used in the evaluations. Thus, the radiological consequences of any accident previously evaluated are not increased.

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

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

Response: No.

The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to conform to the requirements of 10 CFR 50.55a(f) regarding the inservice testing of pumps and valves for the Third 10-year Interval. The current TS reference the ASME Boiler and Pressure Vessel Code, Section XI, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes would reference the ASME OM Code, which is consistent with the 10 CFR 50.55a(f) and approved for use by the NRC. In addition, provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only applied to those inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by this license amendment request.

The proposed changes to TS Section 5.5.7 do not affect the performance of any LSCS structure, system, or component credited with mitigating any accident previously evaluated and do not introduce any new modes of system operation or failure mechanisms.

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

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

Response: No.

The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to conform to the requirements of 10 CFR 50.55a(f) regarding the inservice testing of pumps and valves for the Third 10-Year Interval. The current TS reference the ASME Boiler and Pressure Vessel Code, Section XI, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes would reference the ASME OM Code, which is consistent with the 10 CFR 50.55a(f) and approved for use by the NRC. In addition, provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only applied to those inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by this license amendment request.

The proposed changes do not modify the safety limits setpoints at which protective actions are initiated and do not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.

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 requested amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Bradley J. Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Russell Gibbs.

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: June 2, 2006.

Description of amendment request: The proposed amendments incorporates revised 10 CFR Part 20 requirements for Limerick Generating Station Units 1 and 2 technical specifications (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. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Updating the Technical Specifications (TS) to be consistent with 10 CFR Part 20 has no impact on plant structures, systems, or components, does not affect any accident initiators, and does not change any safety analysis. Therefore, the proposed changes do not involve an 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.

Updating the TS to be consistent with 10 CFR Part 20 will not change any equipment, require new equipment to be installed, or change the way current equipment operates. No credible new failure mechanisms, malfunctions, or accident initiators are created by the proposed changes.

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

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

Response: No.

Updating the TS to be consistent with 10 CFR Part 20 does not adversely affect existing plant safety margins or the reliability of equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed changes. 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: Mr. Brad Fewell, Assistant General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 19348.

NRC Branch Chief: Harold K. Chernoff.

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

Date of amendment request: February 12, 2007.

Description of amendment request: The proposed license amendment would revise Technical Specification (TS) Limiting Condition for Operation 3.9.4, “Containment Penetrations”, to allow penetrations included under TS 3.9.4(c) to be opened during core alterations or movement of irradiated fuel, under administrative controls. This change is based on the TS Task Force Traveler No. 312-A, Revision 1.

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 would allow containment penetrations identified under Technical Specification 3.9.4(c) to remain open during fuel movement and core alterations. These penetrations are normally closed during this time period to prevent the release of radioactive material in the event of a Fuel Handling Accident inside containment. These penetrations are not initiators of any accident. The probability of a Fuel Handling Accident is unaffected by the status of these penetrations.

The Fuel Handling Accident analyses demonstrate that the maximum offsite dose is well [within] the acceptance limits specified in SRP [Standard Review Plan] 15.7.4, and the control room dose is within the acceptance criteria specified in GDC [General Design Criterion] 19. Furthermore, the existing analysis results are independent of the containment release path, and therefore are unaffected by the proposed change.

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

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

Response: No.

The proposed change does not involve the addition or modification of any plant equipment. Also, the proposed change will not alter the design, configuration, or method of operation of the plant beyond the standard functional capabilities of the equipment. The proposed change involves a Technical Specification change that will allow containment penetrations identified under Technical Specification 3.9.4(c) to remain open during fuel movement and core alterations. Open penetrations are not accident initiators, and will not create the possibility of a new kind of accident. Administrative controls will be implemented to ensure the capability to close the affected containment penetrations in the event of a Fuel Handling Accident.

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

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

Response: No.

The proposed change has the potential to slightly increase the post-Fuel Handling Accident dose at the site boundary and in the control room. However, the existing analyses take no credit for containment of the release, so that the existing analysis results will remain bounding.

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

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GO-18, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Russell Gibbs.

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

Date of amendment request: January 19, 2007.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) 5.5.9, “Diesel Fuel Oil Testing Program,” by relocating a reference to a specific American Society for Testing and Materials (ASTM) international standard for fuel oil testing to licensee-controlled documents, and by adding an alternate criteria to the “clear and bright” acceptance test for new fuel oil, per the consolidated line item improvement process (CLIIP).

The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on February 22, 2006 (71 FR 9179), on possible amendments concerning the CLIIP, including a model safety evaluation and a model no significant hazards consideration determination. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 21, 2006 (71 FR 20735), as part of the CLIIP.

In its application dated January 19, 2007, the licensee affirmed the applicability of the following determination.

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

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

Response: No.

The proposed changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. Requirements to perform testing in accordance with applicable ASTM standards are retained in the TS as are requirements to perform surveillances of both new and stored diesel fuel oil. Future changes to the licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, “Changes, tests and experiments,” to ensure that such changes do not result in more than a minimal increase in the probability or consequences of an accident previously evaluated. In addition, the “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to recognize more rigorous testing of water and sediment content. Relocating the specific ASTM standard references from the TS to a licensee-controlled document and allowing a water and sediment content test to be performed to establish the acceptability of new fuel oil will not affect nor degrade the ability of the emergency diesel generators (DGs) to perform their specified safety function. Fuel oil quality will continue to meet ASTM requirements.

The proposed changes do 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 do not adversely affect the ability of structures, systems, and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

Therefore, the changes do not involve a significant increase in the probability or consequences of any 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 proposed changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. In addition, the “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. 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. The requirements retained in the TS continue to require testing of the diesel fuel oil to ensure the proper functioning of the DGs.

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

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

Response: No.

The proposed changes relocate the specific ASTM standard references from the Administrative Controls Section of TS to a licensee-controlled document. Instituting the proposed changes will continue to ensure the use of applicable ASTM standards to evaluate the quality of both new and stored fuel oil designated for use in the emergency DGs.

Changes to the licensee-controlled document are performed in accordance with the provisions of 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that diesel fuel oil testing is conducted such that there is no significant reduction in a margin of safety.

The “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The margin of safety provided by the DGs is unaffected by the proposed changes since there continue to be TS requirements to ensure fuel oil is of the appropriate quality for emergency DG use. The proposed changes provide the flexibility needed to improve fuel oil sampling and analysis methodologies while maintaining sufficient controls to preserve the current margins of safety.

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

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Russell A. Gibbs

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

Date of amendment request: December 14, 2006, as supplemented by letter dated March 14, 2007.

Description of amendment request: The proposed amendment would modify the technical specification (TS) requirements for inoperable snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The changes are consistent with NRC approved Industry/Technical Specification Task Force (TSTF) standard TS change TSTF-372, Revision 4.

The proposed amendment includes an administrative change to LCO 3.0.1 that will clarify that LCO 3.0.7 allows specified TS requirements to be suspended during physics tests performed in accordance with TSs 3.1.8 and 3.1.9. This administrative change will make the CR-3 TSs more consistent with the standard TSs and with TSTF-372, Revision 4.

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed the applicability of the model NSHC determination in its application dated April 26, 2006.

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

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

Response: No.

The proposed change allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. Entrance into Actions or delaying entrance into Actions is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on the delay time allowed before declaring a TS supported system inoperable and taking its Conditions and Required Actions are no different than the consequences of an accident under the same plant conditions while relying on the existing TS supported system Conditions and Required Actions. Therefore, the consequences of an accident previously evaluated are not significantly increased by this change. Therefore, 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 proposed change allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operations. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform its required safety function. The proposed change restores an allowance in the pre-ISTS [improved Standard Technical Specifications] conversion TS that was unintentionally eliminated by the conversion. The pre-ISTS TSs were considered to provide an adequate margin of safety for plant operation, as does the post-ISTS conversion TS. Therefore, this change does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: Thomas H. Boyce.

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

Date of amendment request: March 16, 2007.

Description of amendment request: The proposed amendment would add new Technical Specification (TS) requirements for the response times associated with a steam generator feedwater pump (SGFP) trip and feedwater isolation valve (FIV) closure. The amendment would also revise the TS requirements for the containment fan cooler unit (CFCU) cooling water flow rate. These changes are associated with a revised containment response analysis that credits a SGFP trip and FIV closure (on a feedwater regulator valve failure) to reduce the mass/energy release to the containment during a main steam line break (MSLB). The containment analysis also credits a reduced heat removal capability for the CFCUs, allowing a reduction in the required service water (SW) flow to the CFCUs.

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

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

Response: No.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change establishes response time requirements for feedwater isolation and reduced CFCU flow rates to support containment analyses to accommodate reduced CFCU heat removal capacity. The changes in analysis input assumptions affect plant response to an accident and are not accident initiators; therefore, they have no bearing on the probability of an accident. The Salem FSAR [Final Safety Analysis Report] Chapter 15 accidents which are impacted by a change in the CFCU modeling parameters are LOCA [loss-of-coolant accident] and MSLB mass and energy release Containment analyses. The consequences of these postulated accidents are shown to be acceptable using assumptions consistent with the proposed changes.

For the LOCA transients, the containment cooling systems are considered for three aspects: core response, containment response and dose. The core response is most limiting when the containment conditions minimize back pressure since this increases the blowdown and reduces the effectiveness of the ECCS [emergency core cooling system]. The LOCA core response (10 CFR 50.46 [Section 50.46 of Title 10 of the Code of Federal Regulations]—PCT [peak cladding temperature]) is conservatively biased to minimize the containment backpressure such that any safety injection effectiveness is minimized (the core becomes the highest resistance flow path). Thus, any reduction in the accident capability of the CFCUs has no bearing on the LOCA core response.

The bounding containment integrity analyses are the LBLOCA [large-break LOCA] and the MSLB Inside Containment events. The containment integrity analysis relies on two heat removal paths to maintain containment pressure and temperature conditions. The CFCU air-to-water heat exchangers reject containment energy to the SW System and the Containment Spray System removes containment energy by using spray droplet direct contact heat exchange to transfer the energy from the containment ambient to the containment sump, where it is transferred out of containment via the RHR [residual heat removal] heat exchanger and CCW [component cooling water]/SW Systems. Containment integrity analyses for both LOCA and MSLB, using input assumptions consistent with the proposed changes, show that containment integrity is maintained with reduced CFCU heat removal capacity.

The potential dose impacts due to reduced CFCU heat removal capacity are bounded as the design basis assumptions concerning the number of operating CFCUs (three of five), and the thermal-hydraulic transient operation of the Containment Spray System are unchanged. The Salem design basis only credits Containment Spray iodine removal effectiveness during the LOCA injection and recirculation phases based on a single failure of an entire ESF [engineered safety features] train. This assumption results in 3 of 5 CFCUs being available to ensure adequate mixing of the containment ambient air as well as operation of a single Containment Spray Train, which controls containment spray droplet size and pH, as described in UFSAR [updated FSAR] Section 6.2.3. As a further conservatism, the current LOCA Alternate Source Term (AST) analysis (Calculation S-C-ZZ-MDC-1945, an interim revision of which was sent to the NRC [Nuclear Regulatory Commission] staff for review via letter dated September 16, 2004) only credits two CFCUs for mixing. The Containment Building and Auxiliary Building leakage rates are unaffected by the revised containment analysis as the peak containment pressure and temperatures are less than the design basis values described in the Salem UFSAR. Therefore, there is no impact on offsite dose rates due to the reduced CFCU heat removal capacity.

One other high energy line break for consideration is the rupture of a feedwater line break. From a containment response aspect, this event is bounded by the MSLB event, so it is not explicitly analyzed (or even discussed in the Salem UFSAR).

A review of the Salem design basis for AST dose calculations shows that the revised Containment Integrity Analysis, WCAP-16503, does not challenge any of the assumptions that are part of the AST design basis.

Section 6.2 of the UFSAR indicates that the Appendix J Type A containment leak rate test pressure is based on the containment design pressure of 47.0 psig, not the calculated accident pressure. Since the design pressure value bounds the peak pressure calculated in WCAP-16503 and is not being changed, the Appendix J testing requirements are not impacted.

Thus, in conclusion, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change modifies response time requirements for feedwater isolation, and reduces CFCU flow rates and heat removal requirements consistent with the new containment analysis.

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

Response: No.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes support revised containment analysis to accommodate the reduced CFCU heat removal capacity.

The response time-related changes impose new surveillance acceptance criteria to existing plant equipment that actuates to isolate feedwater following a safety injection signal. There is no change in actuation logic associated with the addition of response time criteria; therefore no new accident sequences would result from the imposition of response time test criteria to existing plant equipment.

The reduction in minimum service water system flow to the CFCUs is supported by analyses demonstrating acceptable system performance and containment integrity following a demand for system operation. The post-accident conditions resulting from the proposed reduction in flow do not adversely impact the environmental qualification of equipment, such that no new consequential failures are introduced to any design basis accident scenario. CFCU operation with the proposed reduction in minimum required accident flow would not result in the progression of any design basis event into a previously unanalyzed accident. Therefore, no new accident scenarios are created from the CFCU flowrate reduction.

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

Response: No.

The proposed change does not involve a significant reduction in the margin of safety. The revised containment analyses accommodate reduced CFCU heat removal capacity using input assumptions consistent with the proposed changes.

The proposed change involves the addition of feedwater isolation response time surveillance criteria and reduction in minimum service water system flows to CFCUs. These changes affect input to the analyses of mass/energy releases and containment response to a design basis main steam line break or loss of coolant accident. The analyses, consistent with the proposed changes, demonstrate that the acceptance criteria continue to be met, and the post-accident conditions do not adversely affect containment integrity or otherwise challenge any safety limit. The margin of safety with respect to containment pressure is preserved by demonstrating that the calculated pressures do not exceed the design limit of 47 psig.

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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Branch Chief: Harold K. Chernoff.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: December 19, 2006.

Brief description of amendments: The amendments requested would revise Technical Specifications (TS) requirement 3.7.5, “Auxiliary Feedwater (AFW) System,” TS 3.8.1, “AC Sources—Operating,” TS 3.8.9, “Distribution Systems—Operating,” and TS Example 1.3-3.

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. D[o] the proposed change[s] involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes eliminate certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the revised Completion Time are no different than the consequences of the same accident during the existing Completion Times. As a result, the consequences of an accident previously evaluated are not affected by this change. The proposed changes do not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes are consistent with the safety analysis assumptions and resultant consequences.

Therefore, the proposed change[s] d[o] not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

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. The changes do not alter any assumptions made in the safety analysis.

Therefore, the proposed change[s] d[o] not create the possibility of a new or different accident from any accident previously evaluated.

3. D[o] the proposed change[s] involve a significant reduction in a margin of safety?

Response: No.

The proposed change to delete the second Completion Time 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 affected by this change. The proposed changes will not result in plant operation in a configuration outside of the design basis.

Therefore, the proposed change[s] d[o] 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: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Branch Chief: David Terao

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: January 18, 2007.

Brief description of amendments: The amendments requested would revise Technical Specifications (TS) requirement 3.8.1, “AC Sources—Operating,” Extension of Completion Times for Diesel Generators.

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 an accident previously evaluated?

Response: No.

The proposed Technical Specification (TS) changes do not significantly increase the probability of occurrence of a previously evaluated accident because the Diesel Generators (DGs) are not initiators of previously evaluated accidents involving a loss of offsite power (LOOP). The proposed changes to the TS Required Actions and Completion Times (CT) do not affect any of the assumptions used in the deterministic or the Probabilistic Safety Assessment (PSA) analysis. Implementation of the proposed changes does not result in a risk significant impact. The onsite AC [alternating current] power sources will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This is demonstrated by showing that the impact on plant safety as measured by the increase in core damage frequency (CDF) is less than 1E-06 per year and the increase in large early release frequency (LERF) is less than 1E-07 per year. In addition, for the CT changes, the incremental conditional core damage probabilities (ICCDP) and incremental conditional large early release probabilities (ICLERP) are less than 5E-07 and 5E-08, respectively. These changes meet the acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, since the onsite AC power sources will continue to perform their functions with high reliability as originally assumed and the increase in risk as measured by ΔCDF, ΔLERF, ICCDP, and ICLERP risk metrics is within the acceptance criteria of existing regulatory guidance, there will not be a significant increase in the consequences of any accidents.

The proposed changes do 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.

The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do 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 are consistent with safety analysis assumptions and resultant consequences.

The proposed TS changes will continue to ensure the DGs perform their function when called upon. Extending the TS CT to 14 days, when an AACPS [alternate AC power source] is available, does not affect the design, the operational characteristics, the function, or the reliability of the DGs. Additionally, the CT extension to 14 days does not affect the interfaces between the DGs and other plant systems. Conversely, in the absence of an AACPS, the DG 72-hour CT will be applied. The availability of the onsite AC power system to perform its accident mitigation function is not affected by the proposed activity and thus there is no impact to the radiological consequences of any accident analysis.

To fully evaluate the effect of the changes to the CT, PSA methods were utilized. The results of this analysis show no significant increase in the CDF and LERF.

The Configuration Risk Management Program (CRMP) in TS 5.5.18 is an administrative program that assesses risk based on plant status. The risk-informed CT will be implemented consistent with the CRMP and approved plant procedures. When utilizing the 14-day extension, requirements of the CRMP per TS 5.5.18 call for the consideration of other measures to mitigate the consequences of an accident occurring while a DG is inoperable. Furthermore, administrative controls will be applied when exercising the 14-day CT extension and are adequate to maintain defense-in-depth and sufficient safety margins.

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 changes do not result in a change in the manner in which the electrical distribution subsystems provide plant protection. The changes to the CT do not change any existing accident scenarios, nor create any new or different accident scenarios.

In addition, the changes do not impose any new or different accident mitigation requirements or eliminate any existing requirements.

The proposed changes are 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 previously evaluated.

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

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. Neither the safety analyses nor the safety analysis acceptance criteria are impacted by these changes. The proposed changes will not result in plant operation in a configuration outside the current design basis. The proposed activities only involve changes to certain TS CTs.

Therefore the proposed change does not involve a 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: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Branch Chief: David Terao.

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

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

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

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

Date of amendment request: March 8, 2007.

Brief description of amendment request: The proposed amendments would revise the McGuire Nuclear Station, Units 1 and 2, Technical Specification 3.5.2.8, and the associated Bases and authorizes changes to the Updated Final Safety Analysis Reports concerning modifications to the emergency core cooling system sump.

Date of publication of individual notice in Federal Register: March 19, 2007.

Expiration date of individual notice: Comments April 18, 2007; Hearing May 18, 2007.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 [phone], [phone] or by e-mail to [e-mail].

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

Date of application for amendment: May 23, 2006, as supplemented by letters dated October 3, 2006, and October 24, 2006.

Brief description of amendment: This amendment revises Technical Specification by modifying the steam generator tube surveillance program at Shearon Harris Nuclear Power Plant, Unit 1.

Date of issuance: March 16, 2007.

Effective date: This amendment is effective as of the date of issuance and shall be implemented within 90 days of issuance.

Amendment No. 124.

Facility Operating License No. NPF-63: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: December 19, 2006 (71 FR 75991). The supplemental letters provided additional information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in the Safety Evaluation dated: March 16, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: July 19, 2006.

Brief description of amendment: The proposed amendment changed the Millstone Power Station, Unit No. 3 (MPS3) reactor core safety limits Technical Specification (TS) and relocated the reactor core safety limit figure to the Core Operating Limits Report in the MPS3 Technical Requirements Manual.

Date of issuance: March 14, 2007

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

Amendment No.: 236

Facility Operating License No. NPF-49: The amendment revised the TSs.

Date of initial notice in Federal Register: August 29, 2006 (71 FR 51227). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 14, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 11, 2006.

Brief description of amendments: (TSTF-372, Rev. 4) The amendments added Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed with an approved Bases Control Program that is consistent with the TS Bases Control Program described in Section 5.5 of the applicable vendor's Standard Technical Specifications. The amendment also made an administrative change, renumbering existing LCO 3.0.8 to LCO 3.0.9.

Date of issuance: March 19, 2007

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

Amendment Nos.: 235, 231

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

Date of initial notice in Federal Register: December 5, 2006 (71 FR 70555). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007.

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: April 11, 2006.

Brief description of amendments: (TSTF-372, Rev. 4) The amendments added Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.8 to allow a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed with an approved Bases Control Program that is consistent with the TS Bases Control Program described in Section 5.5 of the applicable vendor's Standard Technical Specifications. The amendment also made an administrative change, renumbering existing LCO 3.0.8 to LCO 3.0.9.

Date of issuance: March 29, 2007.

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

Amendment Nos.: 238, 220.

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: December 5, 2006 (71 FR 70556). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 29, 2007.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of application for amendment: May 22, 2006, as supplemented by letter dated February 5, 2007.

Brief description of amendment: The amendment revised Technical Specification Surveillance Requirements 3.8.1.11, 3.8.1.12, 3.8.1.16, and 3.8.1.19 to eliminate the specific test-performance mode restrictions for the High-Pressure Core Spray Division 3 diesel generator.

Date of issuance: March 23, 2007.

Effective date: As of the date of issuance and shall be implemented within 45 days from the date of issuance.

Amendment No.: 203.

Facility Operating License No. NPF-21: The amendment revised the Technical Specifications and Facility Operating License.

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40745). The supplemental letter dated February 5, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 23, 2007.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts.

Date of application for amendment: December 27, 2006.

Brief description of amendment: The amendment revised the Technical Specification Limiting Condition for Operation 3.14.A to adopt the Technical Specification Task Force 484, Revision 0, “Use of Technical Specification 3.10.1 for Scram Time Testing Activities.”

Date of issuance: March 26, 2007.

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

Amendment No.: 226.

Facility Operating License No. DPR-35: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: February 20, 2007 (72 FR 7776). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 2007.

No significant hazards consideration comments received: No

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts.

Date of application for amendment: January 15, 2007.

Brief description of amendment: The amendment revised the Technical Specifications (TS) to extend the use of the current pressure-temperature limits as specified in TS Figures 3.6.1, 3.6.2, and 3.6.3 through the end of operating cycle 18.

Date of issuance: March 26, 2007.

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

Amendment No.: 227.

Facility Operating License No. DPR-35: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: February 12, 2007 (72 FR 6609). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 22, 2006.

Brief description of amendment: The amendment revised Technical Specification (TS) requirements for inoperable snubbers by relocating the current TS requirements Limiting Condition for Operation (LCO) 3.6.I and Surveillance Requirement (SR) 4.6.I to the Technical Requirements Manual and adding LCO 3.0.8 to the TSs. The associated TS Bases section has also been relocated.

Date of Issuance: March 26, 2007.

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

Amendment No.: 230.

Facility Operating License No. DPR-28: The amendment revised the License and TSs.

Date of initial notice in Federal Register: June 6, 2006 (71 FR 32604). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated March 26, 2007.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of application for amendment: January 18, 2007.

Brief description of amendment: The amendment revised the description of the control rod assemblies in Grand Gulf Nuclear Station, Unit 1, Technical Specification 4.2.2, “Control Rod Assemblies,” to allow the use of hafnium as an additional type of control material.

Date of issuance: March 16, 2007.

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

Amendment No: 174.

Facility Operating License No. NPF-29: The amendment revises the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: February 13, 2007 (72 FR 6782). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 2007.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: September 26, 2006.

Brief description of amendment: The amendment deleted reference to the containment fan cooler condensate flow switch from Technical Specification 3.4.5.1, “Reactor Coolant System Leakage—Leakage Detection Instrumentation,” and modified or deleted associated actions. The Nuclear Regulatory Commission staff had determined that the remaining leak detection methods provided adequate means for detecting, and to the extent practical, identifying the location of the source of potential reactor coolant leakage.

Date of issuance: March 19, 2007.

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

Amendment No.: 212.

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

Date of initial notice in Federal Register: February 13, 2007 (72 FR 6782). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 19, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 26, 2006, as supplemented on December 26, 2006, and March 14, 2007.

Brief description of amendments: The amendments revised the existing steam generator (SG) tube surveillance program. The changes are modeled after Technical Specifications Task Force (TSTF) traveler TSTF-449, Revision 4, “Steam Generator Tube Integrity,” and the model safety evaluation prepared by the Nuclear Regulatory Commission staff and published in the Federal Register on March 2, 2005 (70 FR 10298). In this regard, the scope of the amendments includes changes to the definition of leakage, changes to the primary-to-secondary leakage requirements, changes to the SG tube surveillance program (SG tube integrity), and changes to the SG reporting requirements.

Date of issuance: March 14, 2007.

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

Amendment Nos.: 298 and 279.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: July 5, 2006 (71 FR 38183).

The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 14, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 7, 2006, as supplemented by letters dated May 30, September 7, December 15, 2006, and January 2, 2007.

Brief description of amendment: The amendment revised Section 4.3, “Fuel Storage,” of the Monticello Nuclear Generating Plant, technical specifications to allow for installation of an additional temporary 8x8 (64-cell) high-density spent fuel storage rack in the spent fuel pool to maintain full core off-load capability.

Date of issuance: March 9, 2007.

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

Amendment No.: 150.

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

Date of initial notice in Federal Register : April 3, 2006 (71 FR 16599).

The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 9, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 23, 2006, as supplemented on December 19, 2006.

Brief description of amendments: The amendments revise Technical Specification (TS) 3.3.4, “Loss of Power (LOP) Diesel Generator (DG) Start and Load Sequence Instrumentation,” and surveillance requirement 3.3.4.3.b to modify the TS title and correct nonconservatisms in the allowable values for the degraded voltage time delay.

Date of issuance: March 21, 2007.

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment Nos.: 225 & 231.

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

Date of initial notice in Federal Register : April 25, 2006 (71 FR 23958).

The December 19, 2006, supplement, contained clarifying information and did not change the staff's initial proposed finding of no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 21, 2007.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of application for amendments: February 16, 2006, supplemented by letters dated July 21, and December 27, 2006.

Brief description of amendments: The amendments consist of changes to the Technical Specifications (TSs) related to steam generator tube integrity. The amendments are modeled after the U.S. Nuclear Regulatory Commission approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, “Steam Generator Tube Integrity,” Revision 4 (ML0510902003).

Date of issuance: March 20, 2007.

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

Amendment Nos.: 177 and 167.

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

Date of initial notice in Federal Register : April 11, 2006 (71 FR 18376)

The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 20, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 30, 2006, as supplemented by letters dated November 22, 2006, and January 11, 2007.

Brief description of amendments: The amendments revised the existing steam generator (SG) tube surveillance program. The changes were modeled after Technical Specification Task Force (TSTF) traveler TSTF-449, Revision 4, “Steam Generator Tube Integrity,” and the model safety evaluation prepared by the U.S. Nuclear Regulatory Commission and published in the Federal Register on March 2, 2005 (70 FR 10298). The scope of the application included changes to the definition of leakage, changes to the primary-to-secondary leakage requirements, changes to the SG tube surveillance program (SG tube integrity), changes to the SG reporting requirements, and associated changes to the Technical Specification Bases.

Date of issuance: March 21, 2007.

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

Amendment Nos.: Unit 1—194; Unit 2—195.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40751). The supplemental letters dated November 22, 2006, and January 11, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 21, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 14, 2006.

Brief description of amendments: The amendments deleted Section 2.G of Facility Operating License Nos. DPR-80 and DPR-82, which require reporting of violations of the requirements of Sections 2.C, 2.E, and 2.F of the operating license. This operating license improvement was made available by the U.S. Nuclear Regulatory Commission on November 4, 2005, as part of the consolidated line item improvement process (CLIIP).

Date of issuance: March 19, 2007.

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

Amendment Nos.: Unit 1-193; Unit 2-194.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: January 3, 2007 (72 FR 154).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of application for amendments: November 18, 2005, as supplemented on November 29, 2006, December 1, 2006, December 15, 2006, January 9, 2007, and March 12, 2007 (PLA-6168 and PLA-6169).

Brief description of amendments: The amendments change the SSES 1 and 2 Technical Specifications (TSs) to implement the Average Power Range Monitor/Rod Block Monitor/TSs/Maximum Load Line Limit Analysis by revising TS 1.1, “Definitions,” TS 5.6.5, “Core Operating Limits Report,” and the surveillance requirement sections of TS 3.3.1.1, “Reactor Protection System Instrumentation,” and TS 3.3.2.1, “Control Rod Block Instrumentation.” The amendments also delete TS 3.2.4, “Average Power Range Monitor Gain and Setpoints,” and its associated references in the TSs. Additionally, the amendments change the method of evaluation for the postulated recirculation line break in the reactor pressure vessel shield annulus region.

Date of issuance: March 23, 2007.

Effective date: As of the date of issuance and to be implemented prior to the startup following the SSES 1 spring 2008 15th refueling outage for Unit 1 and prior to the startup following the SSES 2 spring 2007 13th refueling outage for Unit 2.

Amendment Nos.: 242 and 220.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the TSs and the License.

Date of initial notice in Federal Register: February 14, 2006 (71 FR 7810).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 23, 2007.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of application for amendments: September 7, 2006.

Brief description of amendments: The amendments revise the SSES 1 and 2 Technical Specifications (TSs) Section 5.5.6, “Inservice Testing Program,” and TS 5.5.12, “Primary Containment Leakage Rate Testing Program,” to be consistent with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(f)(4) and 10 CFR 50.55a(g)(4), respectively. The amendments implement TS Task Force (TSTF)-343, Revision 1 and TSTF-479, Revision 0.

Date of issuance: March 19, 2007.

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

Amendment Nos.: 241 and 219.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: December 19, 2006 (71 FR 75997).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania

Date of application for amendment: November 16, 2006, as supplemented on February 15, 2007.

Brief description of amendment: The amendment changes the SSES 2 Technical Specification (TS) Section 2.1.1.2 by revising the Unit 2 Cycle 14 Minimum Critical Power Ratio Safety Limit for two-loop and single-loop operation and the references listed in TS 5.6.5.b.

Date of issuance: March 19, 2007.

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

Amendment No.: 218.

Facility Operating License No. NPF-22: The amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: December 19, 2006 (71 FR 75998).

The supplement dated February 15, 2007, 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 March 19, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 1, 2006.

Brief description of amendments: The amendments eliminate the requirement for a power range neutron flux high negative rate trip and delete the references to this trip in Salem Unit Nos. 1 and 2 Technical Specification (TS) Table 2.2-1, “Reactor Trip System Instrumentation Trip Setpoints,” TS Table 3.3-1, “Reactor Trip System Instrumentation,” TS Table 3.3-2, “Reactor Trip System Instrumentation Response Times,” and TS Table 4.3-1, “Reactor Trip System Instrumentation Surveillance Requirements.” The amendments also incorporate administrative and editorial changes to correct miscellaneous errors in the TSs for Salem Units Nos. 1 and 2.

Date of issuance: March 19, 2007.

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

Amendment Nos.: 278 and 261

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

Date of initial notice in Federal Register : July 18, 2006 (71 FR 40752).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 4, 2006, as supplemented by letter dated February 20, 2007.

Brief description of amendments: The amendments allow the use of blind flanges for containment isolation in the containment purge system supply and exhaust lines, and make corresponding changes to the Technical Specifications (TSs). The amendments also consolidate the containment isolation requirements by moving the requirements of TS 3/4 6.1.7, “Containment Ventilation System,” to TS 3/4 6.3.1 (TS 3/4 6.3 for Unit No. 2), “Containment Isolation Valves.”

Date of issuance: March 19, 2007.

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

Amendment Nos.: 277 and 260.

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

PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

Date of application for amendment: January 18, 2007, as supplemented on February 23, March 9, and March 22, 2007.

Brief description of amendment: The amendment approves a one-time change to the Technical Specifications (TSs) regarding the steam generator (SG) tube inspection and repair required for the portion of the SG tubes passing through the tubesheet region. Specifically, for Salem Unit No. 1 refueling outage 18 (planned for spring 2007) and the subsequent operating cycle, the TS changes limit the required inspection (and repair if degradation is found) to the portions of the SG tubes passing through the upper 17 inches of the approximate 21-inch tubesheet region.

Date of issuance: March 27, 2007.

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

Amendment No.: 279.

Facility Operating License No. DPR-70: The amendment revised the TSs and the License.

Date of initial notice in Federal Register : January 25, 2007 (72 FR 3427).

The letters dated February 23, March 9, and March 22, 2007, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 27, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 28, 2006, as supplemented by letter dated October 24, 2006.

Brief description of amendment: The amendment revises Technical Specification Surveillance Requirement 3.5.1.4 to change the method and frequency for verifying emergency core cooling system accumulator boric acid concentration.

Date of issuance: March 28, 2007.

Effective date: As of the date of issuance to be implemented within 45 days.

Amendment No.: 101.

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

Date of initial notice in Federal Register: April 25, 2006 (71 FR 23960) The October 24, 2006, letter 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 March 28, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: August 22, 2005, as supplemented by letters dated September 18, 2006, October 23, 2006, and February 16, 2007.

Brief description of amendments: These amendments modified Technical Specification (TS) requirements related to control room envelope habitability in TS 3.7.10, “Control Room Emergency Filtration/Pressurization System (CREFS)” and TS Section 5.5, “Administrative Controls—Programs and Manuals.”

Date of issuance: March 26, 2007.

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

Amendment Nos.: 136/136.

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: November 8, 2005 (70 FR 67754). The supplemental letters dated September 18 and October 23, 2006, and February 16, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 31, 2006.

Brief description of amendments: The amendments revised Technical Specification 5.0 entitled, “ADMINISTRATIVE CONTROLS.” Specifically, the change deleted the Vice President, Nuclear Operations, as an alternative to the Plant Manager for certain functions.

Date of Issuance: March 20, 2007.

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

Amendment Nos.: Unit 1-134; Unit 2-134.

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: September 12, 2006 (71 FR 53722).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 20, 2007.

No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station (CPSES), Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: February 21, 2006, as supplemented by letter dated March 19, 2007.

Brief description of amendments: The amendments revise TS 5.6.5 entitled, “Core Operating Limits Report (COLR),” by adding two reports providing Loss-of-Coolant Accident (LOCA) and non-LOCA analysis methodologies for CPSES Unit 1.

Date of issuance: March 26, 2007.

Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance, but no later than the entry into Mode 5 in the restart of Unit 1 from its spring 2007 refueling outage.

Amendment Nos.: 135/135.

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: June 6, 2006 (71 FR 32609). The supplemental letter dated March 19, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 25, 2006, as supplemented by letter dated March 12, 2007.

Brief description of amendment: The amendment revised Technical Specifications 3.1.7, “Rod Position Indication,” 3.2.1, “Heat Flux Hot Channel Factor (F_Q (Z)) (F_Q Methodology),” 3.2.4, “Quadrant Power Tilt Ratio (QPTR),” and 3.3.1, “Reactor Trip System (RTS) Instrumentation,” to allow use of the Westinghouse proprietary computer code, the Best Estimate Analyzer for Core Operations—Nuclear (BEACON). Certain required actions, for when a limiting condition for operation is not met, and certain surveillance requirements are being changed to refer to power distribution measurements or measurement information of the core.

Date of issuance: March 21, 2007.

Effective date: As of its date of issuance and shall be implemented before entry into Mode 2 in the plant restart from the refueling outage scheduled for the spring of 2007. This includes the incorporation of the identified changes to the Final Safety Analysis Report (FSAR) in Attachment 6 of the licensee's application dated May 25, 2006, into the FSAR.

Amendment No.: 182.

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

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40756) The supplemental letter dated March 12, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination published in the Federal Register on July 18, 2006.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 21, 2007.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 3rd day of April 2007.

For the Nuclear Regulatory Commission.

Catherine Haney,

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