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Notice

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

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Information about this document as published in the Federal Register.

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I. Background

Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC 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 June 24 to July 7, 2005. The last biweekly notice was published on July 5, 2005 (70 FR 38712).

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 Start Printed Page 41443any 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 email 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 email to pdr@nrc.gov.

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

Date of amendment request: May 27, 2005.

Description of amendment request: The proposed amendment would revise technical specifications (TS) testing frequency for the surveillance requirement (SR) in TS 3.1.4, “Control Rod Scram Times.” Specifically, the proposed change would revise the frequency for 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 Start Printed Page 41444determination in its application dated May 27, 2005. 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.

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 Section Chief: L. Raghavan.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: February 25, 2005.

Description of amendment request: The proposed change would modify the Millstone Power Station, Unit No. 2 Technical Specification (TS) Surveillance Requirement for trisodium phosphate (TSP) to remove the granularity term and chemical detail. In addition, the proposed change will increase the allowed outage time from 48 to 72 hours. 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 [license] amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The TSP stored in containment is designed to buffer the acids expected to be produced after a loss of coolant accident and is credited in the radiological analysis for iodine retention. The type and amount of TSP is not considered to be an initiator of any analyzed accident. The proposed change does not modify any plant equipment and only clarifies language used in a TSP surveillance requirement which does not impact any failure modes that could lead to an accident. Removing the detail for TSP granularity and type from the surveillance and increasing the allowed outage time, does not change the solubility or buffering capability of the TSP. Therefore this change does not impact the consequences of any accident. Based on this discussion, the proposed amendment does not increase the probability or consequence of an accident previously evaluated.

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

Response: No.

The TSP chemical in containment is not being modified in any way by this proposed amendment. There is no impact on the capability of the TSP to increase the sump water pH to 7 or greater after a loss of coolant accident. No parameters of the TSP baskets are being modified and no changes are being made to the method in which borated water is delivered to the sump. The proposed changes to remove the terms “granular” and “dodecahydrate,” and to increase the allowed outage time do not introduce any new failure modes for the containment sump system. Removing the detail from the surveillance requirement will clarify that the intended parameter to be measured is volume. The proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

There is no significant reduction in the established margin of safety posed by the proposed change to remove detail from the TSP surveillance requirement and increase the allowed outage time. The TSP in containment provides the necessary pH control following a loss of coolant accident to assure iodine retention. Consequently iodine concentrations in the containment atmosphere are maintained within the assumptions of the offsite dose calculations. The proposed change does not introduce any new requirements for the TSP chemical used in containment that would impact a margin of safety. The allowed outage time of 72 hours is consistent with other emergency core cooling components which are also required to perform during a loss of coolant accident. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: Darrell J. Roberts.

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

Date of amendment request: April 27, 2005

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) related to the safety-related battery systems. The revision is based on TS Task Force (TSTF) Change Traveler TSTF-360, Revision 1, “Direct Current (DC) Electrical Rewrite,” and would revise TSs for inoperable battery chargers, provide alternative testing criteria for battery charger testing, and revise TSs for battery cell monitoring. 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?

The DC Sources and Battery Cell Parameters are not initiators of any accident sequence analyzed in JAFNPP's Updated Final Safety Analysis Report (UFSAR). As such, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

The initial conditions of the Design Basis Accident (DBA) and transient analyses in JAFNPP's UFSAR assume Engineered Safety Feature (ESF) systems are operable. The DC electrical power distribution system is designed to provide sufficient capacity, capability, redundancy, and reliability to Start Printed Page 41445ensure the availability of necessary power to ESF systems so that the fuel, reactor coolant system, and containment design limits are not exceeded. The operability of the DC electrical power distribution system in accordance with the proposed TS is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the plant. Therefore, the proposed changes do not involve a significant increase in the 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?

The proposed changes do not involve any physical alteration of the JAFNPP. The temporary charger, when placed in service, will be powered from an emergency bus and have appropriate electrical isolation. Installed equipment is not being operated in a new or different manner. There are no setpoints at which protective or mitigative actions are initiated that are affected by the proposed changes. The operability of the DC electrical power distribution system in accordance with the proposed TS is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the plant. These proposed changes will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alteration in the procedures, which ensure the unit remains within analyzed limits, is proposed, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced. The proposed changes do not alter assumptions made in the safety analyses. Therefore, the proposed 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?

The proposed changes will not adversely affect operation of plant equipment. These changes will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The changes associated with the new administrative TS program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical power distribution system will continue to provide adequate power to safety-related loads in accordance with analyses assumptions. 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 Section Chief: Richard J. Laufer.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, and Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: June 15, 2005.

Description of amendment request: Exelon Generation Company, LLC (EGC), plans to transition to Westinghouse SVEA-96 Optima2 fuel at Dresden Nuclear Power Station (DNPS) and Quad Cities Nuclear Power Station (QCNPS) beginning with the QCNPS Unit 2 refueling outage in March 2006. Specifically, EGC requests approval of revisions to Technical Specifications (TSs) Section 3.1.4, “Control Rod Scram Times,” TS Section 4.2.1, sbull I11“Fuel Assemblies,” and TS Section 5.6.5, “Core Operating Limits Report (COLR),” to support this transition. The core reload analyses using the new Westinghouse analytical methods for the affected units may result in the need for additional TS changes to support the transition to SVEA-96 Optima2 fuel, such as a change to the safety limit minimum critical power ratio. These changes, if any, will be submitted to the NRC in a separate license amendment request.

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 has no effect on any accident initiator or precursor previously evaluated and does not change the manner in which the core is operated. The type of fuel is not a precursor to any accident. The new methodologies for determining core operating limits have been validated to ensure that the output accurately models predicted core behavior, and use of the methodologies will be within the ranges previously approved. The new methodologies being referenced will have all been submitted to the NRC, and have either been approved or are currently under NRC review. Those methodologies that are currently under NRC review are scheduled to receive NRC approval prior to the first use of SVEA-96 Optima2 fuel in a reload core at either DNPS or QCNPS.

There is no change in the consequences of an accident previously evaluated. The proposed change in the administratively controlled analytical methods does not affect the ability to successfully respond to previously evaluated accidents and does not affect radiological assumptions used in the evaluations. Source term from SVEA-96 Optima2 fuel will be bounded by the source term assumed in the accident analyses. There is no effect on the type or amount of radiation released, and there is no effect on predicted offsite doses in the event of an accident.

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 affect the performance of any DNPS or QCNPS structure, system, or component credited with mitigating any accident previously evaluated. The use of new analytical methods, which have either been reviewed and approved by the NRC or are currently being reviewed by the NRC, for the design of a core reload will not affect the control parameters governing unit operation or response of plant equipment to transient conditions. The proposed change does not introduce any new modes of system operation or failure mechanisms.

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 to TS 3.1.4 clarifies that analyses for design basis accidents and transients will continue to support the scram times listed in TS Table 3.1.4-1, independent of whether General Electric analyzes the core. The proposed change does not alter the acceptance criteria for control rod scram times. Future core reloads will be analyzed using the NRC-approved methodology for modeling control rod insertion during a scram. The proposed change to TS Section 4.2.1 revises the description of fuel assemblies to envelope the SVEA-96 Optima2 fuel characteristics. The proposed change to TS Section 5.6.5 adds new analytical methods for design an analysis of core reloads to the list of methods currently used to determine the core operating limits. The NRC has either previously approved the analytical methods being added, or is currently reviewing the methods.

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

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

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are Start Printed Page 41446satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Section Chief: Gene Y. Suh.

Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear Plant, Unit 1, Rhea County, Tennessee

Date of amendment request: April 4, 2005.

Description of amendment request: In order to support the steam generator replacement project (SGRP), the proposed amendment would temporarily revise the Operating License to allow the licensee to operate with one of the two recently installed 18-inch diameter penetrations through the Shield Building dome to be opened while the unit is in Modes 1-4. Either of the Shield Building penetrations will be allowed to be opened for a combined total of up to 5 hours a day, 6 days a week while in Modes 1-4 during the portion of the ongoing Cycle 7 operation between receipt of NRC approval and Mode 5 at the start of the Cycle 7 refueling outage. The technical specifications will revert to the pre-amendment requirements prior to entering Mode 4 during startup from the Cycle 7 outage, since work activities related to the SGRP will permanently eliminate these penetrations.

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 bounding transients and accidents (i.e., loss-of-coolant-accident (LOCA), tornado, and earthquake) that are potentially affected by the assumptions associated with the use of one of the Shield Building dome penetrations have been evaluated/analyzed. Weather and seismic related events are determined by regional conditions. Therefore, the probability of a tornado or earthquake is not affected by the use of one of the Shield Building dome penetrations. Failure of the Shield Building or emergency gas treatment system (EGTS) is not an initiator of any of the accidents and transients described in the Updated Final Safety Analysis Report (UFSAR). Therefore, since no initiating event mechanisms are being changed, the use of one of the Shield Building dome penetrations will not result in an increase in the probability of any previously evaluated accident.

The use of one of the Shield Building dome penetrations affects the integrity of the Shield Building and the ability of the EGTS to maintain the annulus at a negative pressure relative to the outside atmosphere such that the function in mitigating the radiological consequences of an accident is affected. TVA's evaluation documents the radiological consequences of a LOCA assuming the open penetration is closed within fifteen minutes and the mission dose an individual may receive during ingress from the Auxiliary Building roof to the Shield Building dome, closure of the steel hatch assembly, and egress from the Shield Building dome. The LOCA radiological consequences with the penetration open for fifteen minutes are higher than those described in the UFSAR, however, the offsite and Control Room doses remain within the limits of 10 CFR [Title 10, Code of Federal Regulations] 100, Reactor Site Criteria, and 10 CFR 50, Appendix A, General Design Criteria (GDC) 19, Control Room, respectively. The calculated mission doses are also less than the limits of GDC 19. Therefore, since the increase in radiological consequences of the previously evaluated LOCA remains bounded by the applicable regulatory limits, the increased consequences are not considered significant.

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

Response: No.

Loss of Shield Building integrity or EGTS failure is not an initiator of any of the accidents and transients described in the UFSAR. A loss of Shield Building integrity during Modes 1-4 puts the plant into a Limiting Condition for Operation (LCO) situation and requires that the plant initiate shutdown within a specified timeframe if Shield Building integrity cannot be restored within the specified timeframe. The steel hatch assembly over each Shield Building dome penetration performs the same function as the concrete it replaces. Similar to a failure of the Shield Building, a failure of the steel hatch assembly will not initiate any of the accidents and transients described in the UFSAR. Postulated failures of the steel hatch assembly are degradation/damage to the seal or damage to the hatch hinges. Like any other Shield Building failure, these postulated steel hatch assembly failures result in a loss of Shield Building integrity and require that the failed component be repaired or replaced within a specified timeframe or that plant shutdown be initiated.

Therefore, a failure of a steel hatch assembly during use of the Shield Building dome penetration will not initiate an accident nor create any new failure mechanisms. The changes do not result in any event previously deemed incredible being made credible. The use of the Shield Building dome penetration is not expected to result in more adverse conditions in the annulus and is not expected to result in any increase in the challenges to safety systems.

Manual action is required to close an open Shield Building dome penetration and to configure the EGTS control loops following the opening and closing of a Shield Building dome penetration such that the EGTS will respond as designed. NRC Information Notice (IN) 97-78, Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times, and ANSI/ANS [American Nuclear Standard Institute/American Nuclear Society]-58.8, Time Response Design Criteria for Safety-related Operator Actions, provide guidance for consideration of safety-related operator actions.

The manual actions implemented as a result of this change can be completed within the guidance and criteria provided in IN 97-78 and ANSI/ANS-58.8. Consequently, the manual actions can be credited in the mitigation of events that require Shield Building integrity. With credit for the manual actions to close an open Shield Building dome penetration and configure the EGTS control loops subsequent to an event, the types of accidents currently evaluated in the UFSAR remains the same.

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 manual actions to close an open Shield Building dome penetration and to configure the EGTS control loops following the opening and closing of a Shield Building dome penetration ensure that the EGTS will respond as designed. Safety-related instrumentation is available to inform operators that a reactor trip has occurred, and dedicated trained individuals will be positioned to close an open Shield Building dome penetration, should an accident occur. The manual actions meet the criteria for safety-related operator actions contained in NRC IN 97-78 and ANSI/ANS-58.8. The use of manual actions maintains the margin of safety by assuring compliance with acceptance limits reviewed and approved by the NRC. The appropriate acceptance criteria for the various analyses and evaluation have been met; therefore, there has not been a reduction in any margin of safety.

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

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Section Chief: Michael L. Marshall, Jr.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Start Printed Page 41447Commission 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: February 4, 2004.

Brief description of amendments: The amendments revise Technical Specification 3.7.1, “Main Steam Safety Valves (MSSVs),” to permit operation in Mode 3 with five to eight inoperable MSSVs (two to five operable MSSVs) per steam generator, increase the Completion Time to reduce the variable overpower trip setpoint when one to four MSSVs per steam generator are inoperable, and make associated editorial changes.

Date of issuance: July 7, 2005.

Effective date: July 7, 2005, and shall be implemented within 90 days of the date of issuance.

Amendment Nos.: Unit 1-155, Unit 2-155, Unit 3 -155.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revise the Technical Specifications.

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

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 7, 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 20, 2004.

Brief description of amendments: The amendments correct references in TS 5.6.7 and TS Table 3.3.10-1, and delete reference to hydrogen analyzers in TS 3.8.1, which were removed from the TSs by Amendment Nos. 262 and 239, for Unit Nos. 1 and 2, respectively, on March 2, 2004.

Date of issuance: July 5, 2005.

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

Amendment Nos.: 274 and 251.

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

Date of initial notice in Federal Register : January 4, 2005 (70 FR 400).

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated July 5, 2005.

No significant hazards consideration comments received: No.

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

Date of amendment request: May 17, 2005.

Description of amendment request: The amendments replace the existing requirement of Technical Specification 3.4.5, “RCS [Reactor Coolant System] Leakage Detection Instrumentation,” Required Action D.1, to enter Limiting Condition for Operation (LCO) 3.0.3 if required leakage detection systems are inoperable with the requirement to be in Mode 3 within 12 hours and Mode 4 within 36 hours.

Date of issuance: June 28, 2005.

Effective date: June 28, 2005.

Amendment Nos.: 237 and 265.

Facility Operating License Nos. 50-325 and 50-324: Amendments revise the technical specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes (70 FR 34161 dated June 13, 2005). The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided an opportunity to request a hearing by August 12, 2005, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated June 28, 2005.

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.

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: October 15, 2004.

Brief description of amendment: This amendment revises Technical Specifications by extending the inspection interval for reactor coolant pump flywheels to 20 years.

Date of issuance: June 21, 2005.

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

Amendment No.: 119.

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

Date of initial notice in Federal Register: March 1, 2005 (70 FR 9988).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 21, 2005.

No significant hazards consideration comments received: No. Start Printed Page 41448

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, Millstone Power Station, Unit Nos. 2 and 3, New London County, Connecticut

Date of application for amendments: September 8, 2004, as supplemented May 23, 2005.

Brief description of amendments: These amendments delete the Technical Specifications associated with hydrogen recombiners and hydrogen monitors.

Date of issuance: June 29, 2005.

Effective date: As of the date of issuance and shall be implemented by December 31, 2005.

Amendment Nos.: 287 and 224.

Facility Operating License Nos. DPR-65 and NPF-49: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 1, 2005 (70 FR 5238). The May 23, 2005 supplement provided clarifying information that did not change the scope of the proposed amendments as described in the original notice of proposed action published in the Federal Register, and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 29, 2005.

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: May 21, 2003, as supplemented on July 23, 2003, and March 31, 2005.

Brief description of amendment: The amendment changes the Technical Specifications (TSs) to extend the surveillance test interval for the reactor protection system (RPS) intermediate range monitor (IRM) functional tests from weekly to 31 days. In addition, the amendment adds instrument check and calibration requirements for the RPS IRM—High Flux function.

Date of Issuance: July 7, 2005.

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

Amendment No.: 225.

Facility Operating License No. DPR-28: Amendment revised the TSs.

Date of initial notice in Federal Register: July 8, 2003 (68 FR 40713). The supplements contained clarifying information only, and did not change the initial no significant hazards consideration determination or expand the scope of the initial Federal Register notice.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated July 7, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 26, 2003, as supplemented December 8, 2004.

Brief description of amendments: These amendments approve modifications to the Fire Protection Program. Specifically, the modifications involve converting the existing automatic carbon dioxide fire suppression systems installed in each of the four emergency diesel generator rooms and the cable spreading room to manual actuation.

Date of issuance: June 24, 2005.

Effective date: As of the date of issuance, to be implemented following completion of fire protection system modifications.

Amendments Nos.: 255 and 258.

Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments approve modifications to the Fire Protection Program.

Date of initial notice in Federal Register: December 9, 2003 (68 FR 68669). The December 8, 2004, letter 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 amendments is contained in a Safety Evaluation dated June 24, 2005.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of application for amendment: October 29, 2004.

Brief description of amendment: The amendment revises Technical Specification 3.1.8, “Scram Discharge Volume (SDV) Vent and Drain Valves,” for the condition of having one or more SDV vent or drain lines with one valve inoperable.

Date of issuance: June 23, 2005.

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

Amendment No.: 259.

Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 1, 2005 (70 FR 5247).

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

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: December 19, 2003, as supplemented February 18, and March 17, 2004.

Brief description of amendment: The amendment conforms the license to reflect the transfer of Operating License No. DPR-43 to Dominion Energy Kewaunee, Inc., as approved by order of the Commission dated June 10, 2004.

Date of issuance: July 5, 2005.

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

Amendment No.: 185.

Facility Operating License No. DPR-43: Amendment revised the Operating License.

Date of initial notice in Federal Register: January 20, 2004 (69 FR 2734). The supplements dated February 18, and March 17, 2004, were within the scope of the initial application as originally noticed.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 10, 2004.

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

Date of application for amendment: December 20, 2004.

Brief description of amendment: The amendment revises the sampling and testing requirements in Technical Specification 5.5.12, “Diesel Fuel Oil Testing Program,” which verify the acceptability of new diesel fuel oil for use, prior to addition to the storage tanks, and to stored fuel oil.

Date of issuance: July 7, 2005.

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

Amendment No.: 91.

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

Date of initial notice in Federal Register: April 12, 2005 (70 FR 19117).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 7, 2005.

No significant hazards consideration comments received: No. Start Printed Page 41449

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of application for amendments: May 27, 2005, as supplemented by letters dated June 7, June 24, and July 1, 2005.

Brief description of amendments: The amendments revise Technical Specification 3.3.7, “DG-Undervoltage Start,” by changing Surveillance Requirement 3.3.7.3.a to lower the allowable values for dropout and pickup of the degraded voltage function.

Date of issuance: July 1, 2005.

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

Amendment Nos.: 196 and 187

Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 14, 2005 (70 FR 34506). The supplemental letters dated June 7, June 24, and July 1, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendments request: May 17, 2005, as supplemented June 13, 2005.

Brief Description of amendments: The amendments revise the Technical Specification Section 3.7, “Plant Systems,” and Section 4.0, “Design Features,” to establish cask storage area boron concentration limits and to restrict the minimum burnup of spent fuel assemblies associated with spent fuel cask loading operations.

Date of issuance: June 29, 2005.

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

Amendment Nos.: 169 and 161.

Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: May 25, 2005 (70 FR 30148). The supplement dated June 13, 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 amendments is contained in a Safety Evaluation dated June 29, 2005.

No significant hazards consideration comments received: No. The NRC staff made a final determination that the amendment involves no significant hazards considerations.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of application for amendments: October 26, 2004

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

Date of issuance: June 24, 2005.

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

Amendment Nos.: 137 and 116.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 18, 2005 (70 FR 2898).

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

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of application for amendment: July 8, 2004, as supplemented on April 15, 2005.

Brief description of amendment: This amendment removes the requirement to maintain an automatic transfer capability for the power supply to the Low Pressure Coolant Injection inboard injection and recirculation pump discharge valves. The amendment also deletes references to Reactor Motor Operator Valve Boards D and E from the Technical Specifications.

Date of issuance: June 20, 2005.

Effective date: The amendment is effective as of the date of issuance.

Amendment No.: 254.

Facility Operating License No. DPR-33: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 2004 (69 FR 64990). The April 15, 2005, letter provided clarifying 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 a Safety Evaluation dated June 20, 2005.

No significant hazards consideration comments received: No.

Start Signature

Dated in Rockville, Maryland, this 11th day of July 2005.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

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

End Signature End Preamble

[FR Doc. E5-3793 Filed 7-18-05; 8:45 am]

BILLING CODE 7590-01-P