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

Federal RegisterMar 15, 2005

70 Fed. Reg. 12743 (Mar. 15, 2005)

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 February 18, 2005, through March 3, 2005. The last biweekly notice was published on March 1, 2005 (70 FR 9986).

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, Rules and Directives 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, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. 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 (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. 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 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: February 2, 2005.

Description of amendment request: The amendment would revise Tables 3.1.1 and 4.1.1 of the Technical Specifications (TSs) to incorporate the isolation trip setting and the instrumentation surveillance requirements of the reactor water cleanup (RWCU) system high energy line break (HELB) detection and isolation equipment.

The first standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The equipment modification, which is the subject of the proposed amendment, had been installed by the licensee in 1998 using the provisions of 10 CFR 50.59, and the licensee had been performing the surveillance requirement as is now proposed for this amendment. The purpose of the modification was to ensure that the RWCU system can be isolated on an HELB downstream of the RWCU system isolation valves. The proposed addition of the RWCU HELB detection/isolation equipment setpoints and surveillance requirements to the TSs satisfies the 10 CFR 50.36 requirements for limiting conditions for operation (LCOs) and surveillance requirements (SRs) that should be included in the TSs. Thus, the proposed amendment would not alter the physical design or operational procedures associated with any plant structure, system, or component (i.e., the RWCU system will be isolated by existing equipment, in case there is an HELB, in the same way as before the amendment). Consequently, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The second standard requires that operation of the unit in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment does not lead to any changes in the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. The proposed amendment would only add requirements to the TSs for the operability and surveillance testing of the RWCU system HELB detection/isolation equipment. Accordingly, the proposed amendment does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The third standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. The proposed amendment will not affect any margin of safety as defined in the Oyster Creek Nuclear Generating Station Final Safety Analysis Report. The amendment only adds LCOs and SRs to assure that the RWCU system HELB detection/isolation equipment is operable under the plant operating conditions when an RWCU system HELB is possible. The amendment does not change the RWCU system isolation time as compared to original plant design. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: January 27, 2005.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) testing frequency for the surveillance requirement (SR) in TS 3.1.4, “Control Rod Scram Times.” The proposed change would revise 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.”

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in licensing amendment applications in the Federal Register on August 23, 2004 (69 FR 51864). The licensee affirmed the applicability of the model NSHC determination in its application dated January 27, 2005.

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

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

Response: No.

The proposed change 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.

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 Section Chief: Michael L. Marshall, Jr.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: February 14, 2005.

Description of amendment request: The proposed amendments would revise Technical Specification Surveillance Requirement (SR) 3.3.7.1 to extend the frequency of the channel functional test to once every 31 days to once every 92 days.

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated:

The proposed LAR [license amendment request] extends the current 31 day surveillance frequency to a 92 day surveillance frequency. The proposed LAR does not alter the method of operating or configuration for any structure, system, or component. Extension of the surveillance interval will not affect any accident analysis or the plant safety system response to the accident. The extension of the surveillance interval will not affect the ability of ES [engineered safeguards] to actuate Engineered Safeguards Protective System (ESPS) equipment. Therefore, the proposed LAR does not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from any kind of accident previously evaluated:

The proposed change does not necessitate a change in parameters governing plant operation. Consequently, the proposed LAR does not alter the nature of events postulated in the UFSAR [Updated Final Safety Analysis Report] nor does the LAR introduce any unique precursor mechanisms. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Involve a significant reduction in the margin of Safety

The proposed change does not adversely affect any plant safety limits, setpoints, or design parameters. The changes will not adversely affect the fuel, fuel cladding, RCS [reactor coolant system], or containment integrity. The proposed change to the frequency for SR [surveillance requirement] 3.3.7.1 will not impact the operation of the ESPS Digital Automatic Actuation Logic Channels nor the actuation of ESPS equipment. Additionally, the channel functional testing of the ESPS Digital Channels will continue to be performed within an acceptable timeframe following implementation of the proposed change. As such, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP, 1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: December 22, 2004.

Description of amendment request: The requested change would delete Technical Specification (TS) 6.9.1.5, “Occupational Radiation Exposure Report,” and 6.9.1.6, “Monthly Operating Reports,” as described in the Notice of Availability published in the Federal Register on June 23, 2004 (69 FR 35067).

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 eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed 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, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. 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?

Response: No.

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: January 27, 2005.

Description of amendment request: The proposed License Amendment Request (LAR) would allow the licensee to utilize a probabilistic methodology to determine the contribution to main streamline break (MSLB) leakage rates for the once-through steam generator (OTSG) from the tube end crack (TEC) alternate repair criteria (ARC) described in Crystal River Unit 3 (CR-3) Improved Technical Specification (ITS) 5.6.2.10.2.f. This LAR involves no change to the CR-3 ITS.

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

This LAR proposes to change the method to determine the projected MSLB leakage rates for TEC. Potential leakage from OTSG tubes, including leakage contribution from TEC, is bounded by the MSLB evaluation presented in the Final Safety Analysis Report (FSAR). The inspection required by the ARC will continue being performed as required by CR-3 ITS 5.6.2.10. This inspection provides continuous monitoring of tubes with TEC indications remaining in service, and ensures that degradation of new tubes containing TEC indications is detected. The proposed change in method to determine MSLB leakage rates for TEC does not change any accident initiators.

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

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

This LAR proposes to change the method to determine the projected MSLB leakage rates for TEC. The change introduces no new failure modes or accident scenarios. The proposed change does not change the assumptions made in Topical Report BAW-2346P, Revision 0, which demonstrated structural and leakage integrity for all normal operating and accident conditions for CR-3. The design and operational characteristics of the OTSGs are not impacted by the use of a probabilistic methodology to determine MSLB leakage rates.

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

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

This LAR proposes to change the method to determine the projected MSLB leakage rates for TEC. The resulting leakage estimates will be lower than the estimates from the old method. However, the estimates from the proposed method will be more realistic and do not impact the acceptance criteria. The methodology relies on the same accident analyses described in Topical Report BAW-2346P, Revision 0, and License Amendment Request #249, Revision 0, and utilizes the same leakage test data and leakage limit. The FSAR analyzed accident scenarios are not affected by the change and remain bounding. The limits established in CR-3 ITS 3.4.12, and 5.6.2.10.2.f have not been changed. Therefore, the proposed change does not reduce the margin of safety.

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

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

NRC Section Chief: Michael L. Marshall, Jr.

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

Date of amendment request: January 20, 2005.

Description of amendment request: The submittal requests revision to several Technical Specifications (TSs) using eight TS Task Force (TSTF) generic changes. The eight TSTFs (nos. 5, 93, 95, 101, 258, 299, 308, and 361) delete redundant safety limit violation notification requirements; extend the pressurizer heater surveillance frequency from 92 days to 18 months; extend the completion time for reducing the Power Range High trip setpoint from 8 to 72 hours; change the auxiliary feedwater pump test frequency to be consistent with the inservice test program frequency; remove redundant requirements and add other requirements to Section 5.0, Administrative Controls; clarify the requirements regarding the frequency of testing for cumulative and projected dose contributions from radioactive effluents; and add a note to the residual heat removal requirements during Mode 6 low water level operations that allows one required residual heat removal (RHR) loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is operable and in operation.

(1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes revise administrative requirements, actions, action times, surveillance requirements and surveillance frequencies. The revised requirements are not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased by the proposed changes. The Technical Specifications continue to require the systems, structures, and components associated with the revised requirements to be operable. Therefore, any mitigation functions assumed in the accident analyses will continue to be performed. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed amendments do not involve a significant increase in the probability or consequences of any accident previously evaluated.

(2) Operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes do not alter the design or physical configuration of the plant. No changes are being made to the plant that would introduce any new accident causal mechanisms. The proposed changes do not affect any other plant equipment. Therefore, operation of the facility in accordance with the proposed amendments does not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

The proposed changes do not change the design or function of plant equipment. The proposed changes do not significantly reduce the level of assurance that any associated plant equipment will be available to perform its function. The proposed changes provide operating flexibility without significantly affecting plant operation. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: Michael L. Marshall, Jr.

National Aeronautics and Space Administration, Docket No. 50-30, the Plum Brook Test Reactor, Sandusky, Ohio

Date of amendment request: January 14, 2005.

Description of amendment request: The proposed amendment will clarify the license requirements for confirmation of Final Status Survey results prior to backfilling or covering of excavated areas. The amendment will allow performance of the Final Status Survey for an area that has been excavated and allow backfilling of the area without the performance of confirmatory surveys by the NRC. Backfilling without the performance of a confirmatory survey would be allowed only when the NRC Staff has determined that there is appropriate safety or technical justification for backfilling and that, based on the NRC Staff review of the completed Final Status Survey for the affected area, there is reasonable assurance that the Licensee's surveys have demonstrated that the affected area satisfies the unrestricted release criteria.

The proposed amendment to License TR-3 is necessary to assure that, in limited situations, areas excavated for the performance of dismantlement and remediation activities do not result in unnecessary industrial safety hazards. The proposed changes do not involve a significant hazard as shown in the following:

A. The proposed amendment to License TR-3 does not involve a significant increase in the probability or consequences of an accident previously analyzed.

The accident scenarios applicable to the decommissioning of the Plum Brook Reactor Facility are described in section 3.3 of the Decommissioning Plan for the Plum Brook Reactor. The Decommissioning Plan describes postulated events that could result in a release of radioactive materials from the site and analyzes the radiation dose consequences of these events and demonstrates that no adverse public health and safety impacts are expected from these events. Performance of Final Status Surveys is a continuation of decommissioning and is an activity that involves measurements and analysis of residual radioactivity in areas in which decommissioning has already been performed. It is a process used to confirm that radioactivity has been removed to achieve the acceptance criteria specified in 10 CFR 20, Subpart E. These surveys are subjected to NRC review, and in some instances NRC confirmatory surveys. These surveys are performed in areas where other decommissioning activities are already complete and where there is no credible event that could initiate the analyzed accidents. Backfilling an excavated area, or otherwise rendering it inaccessible, will have no impact on other decommissioning activities, or on postulated accidents from other decommissioning activities. Therefore, the proposed amendment will have no affect on the probability or consequences of accidents previously analyzed.

B. The proposed amendment to License TR-3 will not create the possibility of a new or different kind of accident from any accident previously evaluated.

Accidents previously analyzed in the Decommissioning Plan assess different scenarios that could cause the dispersion of radioactive material to the environment. These scenarios arise from dismantlement activities associated with the decommissioning. Excavation of areas to support dismantlement and remediation is an activity that is described in the Decommissioning Plan. Excavated areas will be surveyed in accordance with the Final Status Survey Plan to verify that they have been radiologically remediated to the unrestricted release criteria. These areas may be backfilled without the performance of confirmatory surveys performed by the NRC or their Contractor. However, this will only be permitted after NRC review of the Licensee's survey results and a determination that there is reasonable assurance that no residual radioactivity in excess of the release criteria remains. Therefore, there will be adequate verification by the regulatory agency that there is reasonable assurance that there is no potential for the dispersion of radioactive material to the environment from the backfilled area. The methods and processes used for control of work activities and for control and monitoring of radioactivity will remain the same as those used prior to this amendment. Therefore, no new or different types of accidents are created by this proposed amendment.

C. The proposed amendment to License TR-3 will not involve a significant reduction in a margin of safety.

As discussed previously, the activities that will be performed at the facility are as previously described and evaluated in the accident analyses presented in the Decommissioning Plan. The radiological criteria to be used in applying for termination of the NRC Licenses will remain the same as originally proposed and are consistent with the criteria of 10 CFR 20 Subpart E. The results of Final Status Surveys performed by the Licensee will remain subject to review by the U.S. NRC for adequate implementation of the Final Status Survey Plan. Therefore, the margins of safety applicable to assessing the long term dose to members of the public from exposure to the facility after termination of the license remain unchanged. In addition, since this amendment does not impact any previously reviewed accident analyses as previously discussed, no margins of safety are affected by this proposed amendment.

Attorney for the Licensee: J. William Sikora, Esquire, 21000 Brookpark Road, Mail Stop 500-118, Cleveland, Ohio 44135.

NRC Section Chief: Patrick M. Madden.

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 amendment request: October 15, 2004.

Description of amendment request: The proposed amendment revises Technical Specification 5.5.6, “Reactor Coolant Pump Flywheel Inspection Program,” to extend the allowable inspection interval to 20 years.

The Nuclear Regulatory Commission (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 October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated October 15, 2004.

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

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

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

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

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

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

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

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

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

Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.

NRC Section Chief: L. Raghavan.

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 amendment request: February 1, 2005.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) 3.7.17, “Spent Fuel Pool Storage,” Technical Specification 4.3, “Fuel Storage” and the corresponding TS bases, using revised spent fuel pool (SFP) criticality analysis methodology which takes credit for soluble boron in the spent fuel pool.

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

Response: No.

This license amendment proposes to revise the plant licensing basis by: (1) Replacing the spent fuel pool criticality analyses; and (2) revising the spent fuel storage Technical Specifications 3.7.17, “Spent Fuel Pool Storage” and 4.3, “Fuel Storage” utilizing the proposed analyses. The proposed Technical Specification revisions allow spent fuel to be stored in different configurations.

The proposed changes relate to prevention of criticality accidents in the spent fuel pool. Since the current spent fuel pool criticality analyses and Technical Specifications ensure that a criticality accident does not occur, criticality accidents have not been previously evaluated. Likewise the proposed spent fuel pool criticality analyses and Technical Specifications ensure that a criticality accident does not occur. Thus the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

Events that could cause a criticality accident were evaluated and analyses demonstrated that the current Technical Specification required soluble boron is more than adequate to assure that a criticality accident does not occur. Thus the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

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 licensing basis changes do not involve a change in system operation, or procedures involved with the fuel storage system. It does revise the allowable storage configurations. The proposed changes provide a conservative basis for evaluating spent fuel pool criticality and storage of fuel assemblies in a safe configuration which meets criticality evaluation acceptance criteria. There are no new failure modes or mechanisms created through use of the proposed analyses or proposed Technical Specifications. Use of these licensing basis changes for storage of fuel assemblies does not involve any modification in the operational limits of plant systems. There are no new accident precursors generated with use of these licensing basis changes.

Therefore, the proposed changes do 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 licensing basis change will result in a conservative calculation of the required spent fuel pool soluble boron concentration for the proposed fuel storage configurations. The current Technical Specification required spent fuel pool boron concentration significantly exceeds the proposed criticality analyses required boron concentration. The proposed analyses demonstrate that the criticality analysis acceptance criteria for the proposed fuel storage configurations are met. The proposed analyses utilize industry accepted analysis codes which have been benchmarked for the spent fuel pool criticality analyses proposed for the Prairie Island Nuclear Generating Plant. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.

NRC Section Chief: L. Raghavan.

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

Date of amendment request: October 1, 2004.

Description of amendment request: The proposed amendment would delete requirements from the Technical Specifications (TSs) to maintain hydrogen recombiners and hydrogen and oxygen monitors. A notice of availability for this TS improvement using the consolidated line item improvement process was published in the Federal Register on September 25, 2003 (68 FR 55416). Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2 in 1979. Requirements related to combustible gas control were imposed by order for many facilities and were added to, or included in, the TSs for nuclear power reactors currently licensed to operate. The revised Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors.

The Nuclear Regulatory Commission (NRC) staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated October 1, 2004.

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

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

The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The NRC has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity.

With the elimination of the design-basis LOCA hydrogen release, hydrogen and oxygen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG 1.97 Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen and oxygen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44, the NRC found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. Also, as part of the rulemaking to revise 10 CFR 50.44, the NRC found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment.

The regulatory requirements for the hydrogen and oxygen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, [classification of the oxygen monitors as Category 2,] and removal of the hydrogen and oxygen monitors from TSs will not prevent an accident management strategy through the use of the severe accident management guidelines, the emergency plan, the emergency operating procedures, and site survey monitoring that support modification of emergency plan protective action recommendations.

Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TSs, does not involve a significant increase in the probability or the consequences of any accident previously evaluated.

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

The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TSs, will not result in any failure mode not previously analyzed.

The hydrogen recombiner and hydrogen and oxygen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen and oxygen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building.

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

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

The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TSs, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety.

The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The NRC has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage.

Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors.

Category 2 oxygen monitors are adequate to verify the status of an inerted containment.

Therefore, this change does not involve a significant reduction in the margin of safety. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related oxygen monitors. Removal of hydrogen and oxygen monitoring from TSs will not result in a significant reduction in their functionality, reliability, and availability.

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 Section Chief: Darrell Roberts.

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

Date of amendment request: February 4, 2005.

Description of amendment request: The proposed change will relocate Technical Specification (TS) requirements regarding the Traversing In-core Probe (TIP) System to the Hope Creek Updated Final Safety Analysis Report (UFSAR). Additionally, the associated TS Bases would be deleted. Formatting changes would also be made, as required, in order to incorporate these changes.

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

Response: No.

The proposed change will relocate the requirements from the Technical Specifications (TS) to the Hope Creek Updated Final Safety Analysis Report (UFSAR) a licensee-controlled document. The relocated requirements will be retained in licensee-controlled documents, which will be maintained under the requirements of the provisions of 10 CFR 50.59. Since any changes to licensee-controlled documents are required to be evaluated per 10 CFR 50.59, no increase in the probability or consequences of an accident previously evaluated is allowed.

In addition, the proposed change will not affect any equipment important to safety, in structure or operation. This change will not alter operation of process variables[,] Structures, systems or components as described in the UFSAR or licensing basis. Therefore, the proposed change does not involve a significant increase in the probability or radiological 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 change the design function or operation of any plant equipment. No new failure mechanisms, malfunctions, or accident initiators are being introduced by the proposed changes. 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?

Response: No.

The proposed change will not reduce the margin of safety since they have no impact on safety analysis assumptions. Any future changes to the TIP system requirements will be evaluated under 10CFR50.59. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

NRC Section Chief: Darrell J. Roberts.

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: October 6, 2004.

Brief description of amendments: The proposed amendments will (1) Revise Technical Specification (TS) 3.8.1, “AC Sources-Operating,” to allow surveillance requirement (SR) testing of the onsite standby diesel generators (DGs) during power operation, by removing specific surveillance test MODE restrictions, (2) incorporate changes based on industry approval TSs Task Force (TSTF) standard TS change traveler, TSTF-283, Revision 3, (3) add a new note to TS 3.8.1 Limiting Condition for Operation (LCO) that permits one DG to be connected in parallel with offsite power in order to conduct the required surveillance testing, and (4) delete the expired TS LCO 3.8.1, Required Action A.3 one time 21 day completion time allowance for Startup Transformer XST2 preventive maintenance.

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 focusing on the three standards set forth in 10 CFR 50.92. The licensee's 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 design of plant equipment is not being modified by the proposed changes. In addition, the DGs and their associated emergency loads are accident mitigating features. As such, testing of the diesel generators (DGs) themselves is not associated with any potential accident-initiating mechanism. Therefore, there will be no significant impact on any accident probabilities by the approval of the requested changes.

The changes include an increase in the online time that a DG under test will be paralleled to the grid (for SRs 3.8.1.10 and 3.8.1.14) or unavailable due to testing (per SR 3.8.1.13). However, the overall time that the DG is paralleled in all modes (outage /non-outage) should remain unchanged. As such, the ability of the tested DG to respond to a design basis accident [(DBA)] could be adversely impacted by the proposed changes. However, the impacts are not considered significant based, in part, on the ability of the remaining DG to mitigate a DBA or provide safe shutdown. With regard to SR 3.8.1.10 and SR 3.8.1.14, experience shows that testing per these SRs typically does not perturb the electrical distribution system and share[s] the same electrical configuration alignment as the current monthly surveillance. In addition, operating experience and qualitative evaluation of the probability of the DG or bus loads being adversely affected concurrent with or due to a significant grid disturbance, while the DG is being tested, support the conclusion that the proposed changes do not involve any significant increase in the likelihood of a safety-related bus blackout or damage to plant loads.

The SR changes that are consistent with TSTF-283 have been approved generically and for individual licensees. The on-line tests allowed by the TSTF are only to be performed for the purpose of establishing OPERABILITY. Performance of these SRs during restricted MODES will require an assessment to assure plant safety is maintained or enhanced.

Deletion of expired TS LCO 3.8.1 Required Action A.3 one time 21 day Completion Time allowance for Startup Transformer XST2 preventive maintenance is an administrative change only.

Therefore, the proposed 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 accident from any accident previously evaluated?

Response: No.

The proposed changes would not create any new accidents since no changes are being made to the plant that would introduce any new accident causal mechanisms. Equipment will be operated in the same configuration as currently allowed for other DG SRs that allow testing during at-power operation. Deletion of expired TS LCO 3.8.1 Required Action A.3 one time 21 day Completion Time allowance for Startup Transformer XST2 preventive maintenance is an administrative change only. This license amendment request does not impact any plant systems that are accident initiators; neither does it adversely impact any accident mitigating systems.

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

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

Response: No.

The proposed changes do not involve a significant reduction in the margin of safety. The margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed changes do not directly affect these barriers, nor do they involve any significant adverse impact on the DGs which serve to support these barriers in the event of an accident concurrent with a loss of offsite power. The proposed changes to the testing requirements for the plant DGs do not affect the OPERABILITY requirements for the DGs, as verification of such OPERABILITY will continue to be performed as required (except during different allowed MODES). The changes have an insignificant impact on DG availability, as continued verification of OPERABILITY supports the capability of the DGs to perform their required function of providing emergency power to plant equipment that supports or constitutes the fission product barriers. Only one DG is to be tested at a time, so that the remaining DG will be available to safely shut down the plant if required. Consequently, performance of the fission product barriers will not be impacted by implementation of the proposed amendment.

In addition, the proposed changes involve no changes to setpoints or limits established or assumed by the accident analysis. On this and the above basis, no safety margins will be impacted.

Deletion of expired TS LCO 3.8.1 Required Action A.3 one time 21 day Completion Time allowance for Startup Transformer XST2 preventive maintenance is an administrative change only.

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 that review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments 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 Section Chief: Allen G. Howe.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia; Docket Nos. 50-280 and 50-281, Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia

Date of amendment request: August 30, 2004.

Description of amendment request: The proposed amendment revises the Reactor Coolant Pump (RCP) Flywheel Inspection Programs to extend the allowable inspection interval to 20 years.

The NRC staff issued a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 24, 2003 (68 FR 37590). The notice of availability of the model application was issued on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated August 30, 2004.

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

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

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

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

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

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

Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Section Chief: John A. Nakoski.

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.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania

Date of amendment request: June 1, 2004, as supplemented July 23, 2004, and February 18, 2005.

Description of amendment request: These amendments would lower the BVPS-2 overpressure protection system enable temperature, allow one inoperable residual heat removal loop during surveillance testing, remove the BVPS-1 list of figures and list of tables from the Index of the BVPS-1 Technical Specifications (TSs), and make minor changes to achieve consistency between the units and with the Standard TSs for Westinghouse plants and with some TS Task Force changes.

Date of publication of individual notice in Federal Register : February 25, 2005 (70 FR 9391).

Expiration date of individual notice: March 11, 2005, for comments; April 26, 2005, for hearing.

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 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html . If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: May 20, 2004, as supplemented by letter dated October 19, 2004.

Brief description of amendment: The amendment revised the Technical Specifications, Section 3.2.B.4, regarding control rod operability requirements for inoperable control rods, clarifying the application of the action requirements for inoperable control rods.

Date of issuance: February 25, 2005.

Effective date: February 25, 2005, and shall be implemented within 30 days of issuance.

Amendment No.: 253.

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

Date of initial notice in Federal Register: July 6, 2004 (69 FR 40671).

The October 19, 2004, letter provided clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated February 25, 2005.

No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of application for amendments: July 15, 2004 as supplemented on January 31, 2005.

Brief description of amendments: The amendment adds references to the list of approved core operating limits analytical methods in Technical Specification 5.6.5.b for Calvert Cliffs, Unit Nos. 1 and 2.

Date of issuance: February 24, 2005.

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

Amendment Nos.: 271 and 248.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 29, 2004 (69 FR 78056). The supplement dated January 31, 2005, 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 these amendments is contained in a Safety Evaluation dated February 24, 2005.

No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: December 19, 2003, as supplemented by letter dated January 14, 2004.

Brief description of amendment: The amendment modified Technical Specifications to adopt the provisions of Industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints.” The availability of TSTF-359 for adoption by licensees was announced in the Federal Register on April 4, 2003 (68 FR 16579).

Date of issuance: February 14, 2005.

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

Amendment No.: 203.

Renewed Facility Operating License No. DPR-23: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: February 17, 2004 (69 FR 7519).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 14, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 21, 2004.

Brief description of amendment: This amendment deletes the technical specification requirements associated with hydrogen recombiners, and hydrogen and oxygen monitors.

Date of issuance: March 3, 2005.

Effective date: March 3, 2005, and shall be implemented within 120 days from the date of issuance.

Amendment No.: 189.

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

Date of initial notice in Federal Register: December 21, 2004 (69 FR 76488).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 3, 2005.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 28, 2004.

Description of amendment request: The amendment revised the Seabrook Station, Unit No. 1 Technical Specifications (TSs) by eliminating the requirements associated with hydrogen recombiners and hydrogen monitors.

Date of issuance: February 22, 2005.

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

Amendment No.: 99.

Facility Operating License No. NPF-86: The amendment revised the TSs.

Date of initial notice in Federal Register: September 14, 2004 (69 FR 55471).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 22, 2005.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 6, 2003, as supplemented by letters dated May 5, May 24, July 8, September 13, 2004, and January 13, 2005.

Description of amendment request: The amendment revised the Seabrook Station, Unit No. 1 licensing basis to incorporate a full-scope application of an alternative source term methodology in accordance with 10 CFR 50.67.

Date of issuance: February 24, 2005.

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

Amendment No.: 100.

Facility Operating License No. NPF-86: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 9, 2003 (68 FR 68670).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 24, 2005.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 17, 2004, as supplemented by letters dated March 17, 2004, April 1, 2004, May 26, 2004, September 13, 2004 (2 letters), October 12, 2004, October 28, 2004, December 3, 2004, December 28, 2004, and January 28, 2005.

Description of amendment request: The amendment revised the Seabrook Station, Unit No. 1 operating license to increase the licensed rated power by 5.2 percent from 3411 megawatts thermal to 3587 megawatts thermal.

Date of issuance: February 28, 2005.

Effective date: As of its date of issuance, and shall be implemented within 12 months.

Amendment No.: 101.

Facility Operating License No. NPF-86: The amendment revised the Technical Specifications and the operating license.

Date of initial notice in Federal Register: June 22, 2004 (69 FR 34701).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 28, 2005.

No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: October 25, 2004, as supplemented by letters dated December 29, 2004, and January 26, 2005.

Brief description of amendment: The amendment revises Technical Specification 2.1.1.2 for the dual recirculation loop and single recirculation loop Safety Limit Minimum Critical Power Ratio values to reflect results of a cycle-specific calculation.

Date of issuance: February 1, 2005.

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

Amendment No.: 210.

Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 23, 2004 (69 FR 68183).

The supplements dated December 29, 2004, and January 26, 2005, 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 February 1, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 15, 2004.

Brief description of amendment: The amendment revised Section 1.7, which defines the term “Instrument Channel Calibration,” by adding two new sentences pertaining to calibration of resistance temperature detector or thermocouple sensors.

Date of issuance: February 17, 2005.

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

Amendment No.: 187.

Facility Operating License No. DPR-63: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 7, 2004 (69 FR 70719).

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated February 17, 2005.

No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: September 7, 2004.

Brief description of amendment: The amendment deleted the technical specification requirements to submit a monthly operating report and an annual occupational radiation exposure report.

Date of issuance: February 18, 2005.

Effective date: February 18, 2005, and shall be implemented within 60 days from the date of issuance.

Amendment No.: 231.

Renewed Facility Operating License No. DPR-40: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 26, 2004 (69 FR 62477).

The Commission's related evaluation of the amendment is contained in a safety evaluation dated February 18, 2005.

No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 21, 2004, as supplemented on October 29, and December 16, 2004.

Brief description of amendment: The amendment revised Technical Specifications (TSs) Section 2.3(4), “Emergency Core Cooling System—Trisodium Phosphate (TSP),” regarding the volume and the form of the TSP; and TS Section 3.6(2)d.(i), “Safety Injection and Containment Cooling Systems Tests,” the surveillance requirement for the TSP volume.

Date of issuance: March 1, 2005.

Effective date: March 1, 2005, and shall be implemented within 60 days from the date of issuance.

Amendment No.: 232.

Renewed Facility Operating License No. DPR-40: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: The October 29, and December 16, 2004, 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 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 1, 2005.

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: September 8, 2004, as supplemented by letter dated February 1 and 14, 2005.

Brief description of amendments: The amendment changed the SSES-2 Technical Specifications by revising the Unit 2 Cycle 13 Minimum Critical Power Ratio Safety Limits in Section 2.1.1.2 and the references listed in Section 5.6.5.b.

Date of issuance: February 28, 2005.

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

Amendment No.: 194.

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

Date of initial notice in Federal Register: January 4, 2005 (69 FR 698). The supplements dated February 1 and 14, 2005, 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 February 28, 2005.

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: May 21, 2003, as supplemented December 1, 2003 (two letters), February 16, March 1 and 8, April 22, May 21, July 8 and 14, August 6 and 18, September 10, October 14 and 18, December 3 and 6, 2004, and January 27, 2005.

Brief description of amendment: The amendment revises the Technical Specifications to reflect design modifications to the Control Room Emergency Air Treatment System and elimination of the requirements for the Containment Post Accident Charcoal Filters. The amendment also changes the source term for the Dose Calculation Methodology to the Alternate Source Term, and revises both the reactor coolant dose equivalent I-131 specific activity limit and the containment spray NaOH concentration limit.

Date of issuance: February 25, 2005.

Effective date: As of the date of issuance to be implemented upon completion of the installation and testing of the new Control Room Emergency Air Treatment System.

Amendment No.: 87.

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

Date of initial notice in Federal Register: July 8, 2003 (68 FR 40718). 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 staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 25, 2005.

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 1, 2004.

Brief description of amendment: The amendment modifies Technical Specification requirements to adopt the provisions of Industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints.”

Date of issuance: March 1, 2005.

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

Amendment No.: 88.

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

Date of initial notice in Federal Register: June 22, 2004 (69 FR 34706).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 1, 2005.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of application for amendments: May 21, 2004.

Brief description of amendments: The amendments revised the Technical Specifications to delete the requirements to maintain hydrogen recombiners and hydrogen monitors and oxygen monitors.

Date of issuance: February 23, 2005.

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

Amendment Nos.: 244 and 188.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57992).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 23, 2005.

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: September 10, 2004.

Brief description of amendments: The amendments delete the technical specification requirements to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: March 2, 2005.

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

Amendment Nos.: 115/115.

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

Date of initial notice in Federal Register: October 26, 2004 (69 FR 62479).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 2, 2005.

No significant hazards consideration comments received: No.

Dated in Rockville, Maryland, this 4th day of March 2005.

For the Nuclear Regulatory Commission.

James E. Lyons,

Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

[FR Doc. 05-4792 Filed 3-14-05; 8:45 am]

BILLING CODE 7590-01-P