Operating licenses, amendments; no significant hazards considerations; biweekly notices,

[Federal Register: January 4, 2005 (Volume 70, Number 2)]

[Notices]

[Page 398-408]

From the Federal Register Online via GPO Access [wais.access.gpo.gov]

[DOCID:fr04ja05-93]

NUCLEAR REGULATORY COMMISSION

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

1. 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 December 10, 2004, through December 22, 2004. The last biweekly notice was published on December 21, 2004 (69 FR 76486).

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

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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 basis 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,

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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.

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 amendment request: July 20, 2004.

Description of amendment request: The proposed administrative amendment corrects references in Technical Specification (TS) 5.6.7 and in TS Table 3.3.10-1, and deletes reference to hydrogen analyzers which were removed from the TSs by Amendment Nos. 262 and 239, for Unit Nos. 1 and 2, respectively, on March 2, 2004.

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

   Amendment Nos. 262 and 239 were approved and issued by the Nuclear Regulatory Commission (NRC) on March 2, 2004. These amendments removed the requirements for the containment hydrogen recombiners and the hydrogen analyzers as equipment required to control hydrogen in the Containment. The amendments required the hydrogen analyzers to be retained as non-safety-related equipment to record hydrogen concentrations in beyond design-basis accidents. The request to remove hydrogen control from the design basis included a mark-up of proposed Technical Specification changes. However, related changes to Technical Specification Table 3.3.10-1, Technical Specification 5.6.7, and Technical Specification 3.8.1 were not included in the markup. Therefore, we are requesting an administrative change to correct this oversight.

   Since the justification for these changes has been approved in Calvert Cliffs Amendment Nos. 262 and 239, there is no technical or safety issue associated with this request.

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

2. Would not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

   The proposed administrative amendment corrects references in a Technical Specification table and in a Technical Specification, and deletes reference to hydrogen analyzers. Since the justification for these changes has been approved in Calvert Cliffs Amendment Nos. 262 and 239, there is no technical or safety issue associated with this request. This request does not involve a change in the operation of the plant, and no new accident initiation mechanism is created by the proposed change, nor does the change involve a physical alteration of the plant.

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

3. Would not involve a significant reduction in a margin of safety.

   Amendment Nos. 262 and 239 were approved and issued by the Nuclear Regulatory Commission (NRC) on March 2, 2004. These amendments removed the requirements for the containment hydrogen recombiners and the hydrogen analyzers as equipment required to control hydrogen in the Containment. The amendments required the hydrogen analyzers to be retained as non-safety-related equipment to record hydrogen concentrations in beyond design-basis accidents. The request to remove hydrogen control from the design basis included a mark-up of proposed Technical Specification changes. However, related changes to Technical Specification Table 3.3.10-1, Technical Specification 5.6.7, and Technical Specification 3.8.1 were not included in the markup. Therefore, we are requesting an administrative change to correct this oversight.

   Because the hydrogen analyzers were removed from the Technical Specifications by Amendment Nos. 262 and 239, no margin of safety is impacted by the proposed administrative changes.

   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: James M. Petro, Jr., Esquire, Counsel, Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, Baltimore, MD 21202.

   NRC Section Chief: Richard J. Laufer.

   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 amendment request: August 3, 2004.

   Description of amendment request: The proposed amendment would extend the surveillance requirement (SR) 3.3.3.1 test interval for reactor trip circuit breakers from 31 to 92 days and impose a staggered test interval consistent with SR 3.3.3.2.

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

   The reactor trip circuit breakers (RTCB) are part of the Reactor Protective System (RPS). The RPS initiates a reactor trip to protect against violating the core specified acceptable fuel design limits and reactor coolant pressure boundary integrity during anticipated operational occurrences. By opening the RTCBs to trip the reactor, the RPS also assists the engineered safety features systems in mitigating accidents. All of the accident analyses that call for a reactor trip assume that the RTCBs operate and interrupt power to the control element drive mechanisms. The proposed testing interval will result in less wear on the RTCBs and, thereby, increase breaker reliability.

   The RTCBs are accident mitigators and do not affect the probability of an accident.

   Topical Report CE NPSD-951-A shows only one failure up to 1993 in the plants studied. Calvert Cliffs' surveillance records show no failures from 1994 to 2003. This data demonstrates that the consequences of an accident will not be significantly

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   increased by extending the surveillance interval and imposing a staggered test interval.

   Therefore, extending the surveillance interval and imposing a staggered test interval does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Would not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

   There is no change in plant equipment or operation related to this license amendment request. The RTCBs are accident mitigators and extending the surveillance interval and imposing a staggered test interval does not adversely affect their operation.

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

3. Would not involve a significant reduction in [a] margin of safety.

   The margin of safety in this case is the reliance on the RTCBs to open on a signal from the RPS. Extending the surveillance frequency and imposing a staggered test interval results in a test every six weeks as opposed to the current monthly test. The new interval will result in less wear on the RTCBs, thereby improving the margin of safety.

   Therefore, extending the surveillance interval and imposing a staggered test interval will 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: James M. Petro, Jr., Esquire, Counsel, Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, Baltimore, MD 21202.

   NRC Section Chief: Richard J. Laufer.

   Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50- 458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

   Date of amendment request: December 17, 2004.

   Description of amendment request: The proposed change will revise the Technical Specification (TS) requirements for direct current (DC) sources. The current TS only includes Action Statements for an inoperable DC Power subsystem. The proposed change will add a new Action Statement to TS 3.8.4, ``DC Sources--Operating,'' to specifically address an inoperable battery charger.

   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 class 1E direct current (DC) electrical power system including the associated battery chargers are not initiators to any accident sequence analyzed in the Updated Safety Analysis Report (USAR). Operation in accordance with the proposed Technical Specification (TS) ensures that the DC system is capable of performing its function described in the USAR. While power to the non class 1E charger will be lost after a Design Basis Accident (DBA), the Division 1 and 2 batteries have the ability to supply all DBA loads and all other standby loads not automatically tripped on a LOCA [Loss of Coolant Accident] signal for 4 hours and have sufficient capacity to restore normal AC [alternating current] and DC power with the charger inoperable. The actions required to restore the power to the non-class 1E charger are included in the procedures for Station Blackout requiring the use of a non class 1E diesel generator. They allow the impacted DC battery and DC bus to be restored to perform its required function as described in the USAR.

   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 changes do not involve a physical change to the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. These changes will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. Any alterations in procedures will continue to assure that the plant remains within analyzed limits, and no change is being made to the procedures relied upon to respond to an off normal event as described in the USAR. As such, no new failures modes are being introduced.

   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 margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes are acceptable because the operability of the safety related DC systems are unaffected and there is no detrimental impact on any equipment design parameter. The plant will still be capable of operating within assumed conditions. Operations in accordance with the proposed TS ensures that the DC system is capable of performing its function as described in the USAR; therefore, the support of the DC system to the plant response to analyzed events will continue to provide the margins of safety assumed by the analysis. In addition, the DC system is within the scope of 10 CFR 50.65, ``Requirements for monitoring the effectiveness of maintenance at nuclear power plants,'' which will ensure the control of maintenance activities associated with the DC system. This provides sufficient management control of the requirements that assure the batteries are maintained in a highly reliable condition. The non-class 1E battery charger is the same model and has the same ratings as the installed Division 1 and 2 class 1E battery chargers (i.e., same input loading and ampere current capability), and was purchased to Class 1E requirements. In addition, the backup battery charger can be powered from an onsite power source (Station Blackout (SBO) diesel generator) should it be required.

   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: Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005.

   NRC Section Chief: Michael K. Webb, Acting.

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

   Date of amendment request: August 31, 2004.

   Description of amendment request: The proposed amendment would modify the existing Technical Specification (TS) 3.4.1, ``Recirculation Loops Operating,'' associated with single recirculation loop operation by incorporating limits for the linear heat generation rate (LHGR) fuel thermal limit into the limiting condition of operation (LCO). Currently, TS 3.4.1 only contains thermal limits for the minimum critical power ratio and the average planar LHGR. Thermal limits associated with the two recirculation operations are contained in TS 3.2.1, ``Average Planar Linear Heat Generation Rate (APLHGR),'' TS 3.2.2, ``Minimum Critical Power Ratio (MCPR),'' and TS 3.2.3, ``Linear Heat Generation Rate (LHGR).'' The proposed TS change will reflect a consistency with the existing two recirculation loop LCOs by including the same three thermal limits into the single recirculation loop LCO.

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

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

   The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on the LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (AOOs). Additionally, the LHGR limits provide assurance the fuel peak cladding temperature (PCT) during a Loss Of Coolant Accident (LOCA) will not exceed the requirements of 10 CFR 50.46.

   The PNPP [Perry Nuclear Power Plant] Core Monitor previously automatically modified the ``composite'' LOCA/Thermal-Mechanical MAPLHGR [minimum average planar linear heat generation rate] limits for single recirculation loop operation. As a result, the LHGR limit was adjusted for single recirculation loop operation by application of the single recirculation loop operation MAPLHGR multiplier to the ``composite'' MAPLHGR limits. The proposed TS change establishes a TS requirement for LHGR limits to be modified, as specified in the Core Operating Limits Report, during single recirculation loop operation. This TS requirement provides assurance that the fuel design limits will remain satisfied during the time the plant may be in single recirculation loop operation.

   There are no physical modifications being made to any plant system or component, including the fuel.

   The manual versus automatic adjustment of the LHGR limits when in single reactor loop operation is considered a change in the implementation of a core monitoring function. However, since the LHGR limits that will be applied to the core are consistent with the NRC-approved fuel design and LOCA methodologies in use at PNPP, this change in monitoring implementation is not considered significant.

   Therefore, since no significant changes are being made to the plant or its operation, the probability or the consequences of an accident have not increased over those previously evaluated.

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

   There are no physical modifications being made to any plant system or component, including the fuel. The manual versus automatic adjustment of the LHGR limits when in single reactor loop operation is considered a change in the implementation of a core monitoring function. However, since the LHGR limits that will be applied to the core are consistent with the NRC-approved fuel design and LOCA methodologies in use at PNPP, this change in monitoring implementation is not considered significant. The proposed TS change provides assurance that the LHGR limits will be adjusted if the plant enters a condition of single recirculation loop operation, thereby ensuring the fuel design limits remain satisfied.

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

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

   There are no physical modifications being made to any plant system or component, including the fuel. The manual versus automatic adjustment of the LHGR limits when in single reactor loop operation is considered a change in the implementation of a core monitoring function. However, since the LHGR limits that will be applied to the core are consistent with the NRC-approved fuel design and LOCA methodologies in use at PNPP, this change in monitoring implementation is not considered significant. The proposed TS change provides assurance that the LHGR limits will be adjusted if the plant enters a condition of single recirculation loop operation, thereby ensuring the fuel design limits remain satisfied.

   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: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

   NRC Section Chief: Gene Y. Suh.

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

   Date of amendment request: October 25, 2004.

   Description of amendment request: The proposed amendment would revise the required channels per trip system for several instrument functions contained in technical specification tables 3.3.6.1-1 (Primary Containment Isolation Instrumentation), 3.3.6.2-1 (Secondary Containment Isolation Instrumentation), and 3.3.7.1-1 (Control Room Emergency Filter System Instrumentation).

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

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

   Response: No.

   Revising the Required Channels Per Trip System to conform with the Cooper Nuclear Station (CNS) design basis resolves an inconsistency that will not result in any changes to instrumentation configuration, operating practices, or means of testing. Thus, these changes are administrative and have no associated effects on the probability or consequences of previously evaluated accidents.

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

   Response: No.

   The proposed changes represent administrative changes to the Technical Specification controls over the affected instrumentation. Thus, the changes will not create new event initiators or alter plant response to postulated plant events.

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

   Response: No.

   The proposed changes have no effect on the manner in which the affected instruments are configured, operated, or tested. Similarly, there is no relaxation in the application of Technical Specifications to inoperable channels. Thus these proposed changes will not result in 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 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 R. McPhail, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

   NRC Acting Section Chief: Michael K. Webb.

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

   Date of amendment requests: September 23, 2004.

   Description of amendment requests: The proposed amendments would revise Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist,'' to increase the required amount of stored diesel fuel to support use of low-sulfur fuel oil required by the California Air Resources Board (CARB).

   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

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   consideration, which is presented below:

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

   The proposed change revises the minimum amount of stored diesel fuel. The change is required to support the use of California Air Resources Board (CARB) fuel oil and ultra-low sulfur (ULS) fuel oil that is replacing the existing Environmental Protection Agency (EPA) red dyed fuel oil currently used at Diablo Canyon Power Plant (DCPP). Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist,'' requires, as a minimum, a supply of diesel fuel sufficient to support 7-days operation of the diesel generators (DGs) to power the minimum engineered safety feature (ESF) systems required to mitigate a design basis loss-of-coolant accident (LOCA) in one unit and those minimum required systems for a concurrent non-LOCA safe shutdown in the remaining unit (both units initially in Mode 1 operation). TS 3.8.3 Condition A requires storage levels to be restored to within limits within 48 hours if they fall below the 7-day minimum, but remain above minimum limits for a 6-day supply. TS 3.8.3 also provides for tank cleaning on a 10-year frequency. During tank cleaning, TS 3.8.3 requires maintaining at least a 4-day supply.

   Because CARB and ULS fuel oils have a lower heat content than EPA fuel, it was necessary to recalculate the amount of fuel required to supply necessary loads for the required 7-day, 6-day, and 4-day time periods addressed in TS 3.8.3.

   The DGs and associated support systems, such as the fuel oil storage and transfer systems, are designed to mitigate accidents, and are not accident initiators. Revising the minimum volumes of stored fuel in the storage tanks will not result in any increase in the probability of any accident previously evaluated.

   Following implementation of this proposed change, there will be no change in the ability of the DGs to supply post-accident loads for 7 days, or 6 days if in TS 3.8.3 Condition A, or 4 days during tank cleaning. This is identical to the current requirements. Therefore, this change will not result in a significant increase in the consequences of any accident previously evaluated.

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

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

   Following implementation of this change, the DGs will still be able to power the minimum ESF systems required to mitigate a design basis LOCA in one unit and those minimum required systems for a concurrent non-LOCA safe shutdown in the remaining unit (both units initially in Mode 1 operation). The current 7-day, 6-day, and 4-day fuel supply requirements will be maintained. The DGs and associated fuel oil storage systems are not accident initiators, but are designed to mitigate accidents.

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

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

   Following implementation of this change, the DGs will still have sufficient fuel oil supply to power the minimum ESF systems required to mitigate a design basis LOCA in one unit and those minimum required systems for a concurrent non-LOCA safe shutdown in the remaining unit (both units initially in Mode 1 operation). When fuel inventory is below that required to support 7 days of operation, the required actions depend on whether or not a 6-day supply is available, or a 4-day supply is available during tank cleaning. The proposed storage limits will maintain these 7-day, 6-day, and 4-day fuel supply requirements, including current margins, following the change to CARB and ULS fuel oils.

   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 requests involve no significant hazards consideration.

   Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

   NRC Section Chief: Robert A. Gramm.

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

   Date of amendment requests: October 29, 2004.

   Description of amendment requests: The proposed amendments would revise the technical specifications (TS) requirements for handling of irradiated fuel in the containment and fuel building, and certain specifications related to performing core alterations. These changes are based on analysis of the postulated fuel handling and core alteration accidents and transients for the Diablo Canyon Power Plant, Units 1 and 2. The proposed amendment is consistent with the NRC- approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-51, Revision 2, ``Revise containment requirements during handling irradiated fuel and core alterations.'' In addition, editorial corrections to TS 3.1.7, ``Rod Position Indication''; TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation''; TS 3.4.16, ``RCS Specific Activity''; TS 3.7.3, ``Main Feedwater Isolation Valves (MFIVs), Main Feedwater Regulating Valves (MFRVs), MFRV Bypass Valves and Main Feedwater Pump (MFWP) Turbine Stop Valves''; and TS 3.7.13, ``Fuel Handling Building Ventilation System (FHBVS),'' are proposed.

   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. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change involves changes to accident mitigation system requirements. These systems are related to controlling the release of radioactivity to the environment and are not considered to be accident initiators for any previously analyzed accident. The proposed changes do not involve physical modifications to plant equipment, and do not change the operational methods or procedures used for moving irradiated fuel assemblies. As such, there are no accident initiators affected by the proposed amendment. Therefore, the proposed change does not impact the probability of postulated accidents.

   Consistent with the previously approved design basis analysis, the reanalysis of the containment fuel handling accident (FHA) concludes that radiological consequences of the accident at the Exclusion Area Boundary and the Low Population Zone Boundary are unchanged and remain well within the 10 CFR 100.11 limits, as defined by acceptance criteria in NUREG 0800, Section 15.7.4, and within the limits of general design criteria (GDC) 19 of 10 CFR 50, Appendix A. However, per this reanalysis, the calculated 30-day doses in the control room increased from 11.56 rem to 22.31 rem thyroid and from 0.00717 rem to 0.00757 rem whole body. Although these calculated doses increased they remain well within the acceptable limits of GDC 19 of 10 CFR 50, Appendix A, for the control room, which is 30 rem thyroid and 5 rem whole body. As a result, the increase in the doses is not considered to be a significant increase.

   The results of the core alteration events, other than a FHA, remain unchanged from the original design basis, which showed that these events do not result in fuel cladding integrity damage or radioactive releases. Therefore, the proposed changes do not significantly increase the consequences of any previously evaluated accident.

   In addition, the editorial corrections have no affect on the associated components, structures or systems, and their operation or design bases.

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

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

   [[Page 404]]

   The proposed change affects a previously evaluated accident (i.e., FHA). However, the proposed change does not introduce any new modes of plant operation and does not involve physical modifications to the plant. The proposed change does not change how design basis accidents were postulated nor does the proposed change initiate a new kind of accident or failure mode with a unique set of conditions.

   In addition, the editorial corrections have no affect on associated components, structures or systems, and their operation or design bases.

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

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

   The proposed change imposes controls to ensure that during performance of activities that represent situations where radioactive releases are postulated, the radiological consequences are at or below the established licensing limit. Safety margins and analytical conservatisms have been evaluated and are understood. Substantial conservatism is retained to ensure that the analysis adequately bounds all postulated event scenarios. Specifically, the margin of safety for a FHA is the difference between the 10 CFR 100.11 limits and the licensing limit defined by the NUREG-0800, Section 15.7.4. The licensing limit is defined by the NUREG as being ``well within'' the 10 CFR 100.11 limits, with ``well within'' defined as 25 percent of the 10 CFR 100 limits of the FHA. Excess margin is the difference between the postulated doses and the corresponding licensing limit.

   The proposed applicability requirements continue to ensure that the whole-body, thyroid and total effective dose equivalent (TEDE) doses at the exclusion area and low population zone boundaries are at or below the corresponding licensing limit for both the FHA inside containment and in the fuel handling building. In addition, control room doses for both FHAs meet GDC 19 criterion. Although the control room doses as a result of the FHA inside containment reanalysis are somewhat higher then previously approved, they still remain well below the GDC-19 limits, therefore, the proposed change does not involve a significant reduction in a margin of safety.

   The margin of safety for core alteration events other than the FHA remains the same as the original licensing analyses, since the proposed change does not impact the TS requirements for systems needed to prevent or mitigate such core alteration events.

   In addition, the editorial corrections have no affect on associated equipment, components, structures or systems, and their operation or margin of safety.Therefore, the proposed change does 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration.

   Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

   NRC Section Chief: Robert A. Gramm.

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

   Date of amendment request: September 8, 2004.

   Description of amendment request: The proposed amendment would change the Unit 2 Technical Specifications (TSs) by revising the Unit 2 Cycle 13 (U2C13) Minimum Critical Power Ratio (MCPR) Safety Limits in Section 2.1.1.2 and the references listed in Section 5.6.5.b.

   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 to the MCPR Safety Limits does not directly or indirectly affect any plant system, equipment, component, or change the processes used to operate the plant. Further, the U2C13 MCPR Safety Limits are generated using NRC approved methodology and meet the applicable acceptance criteria. In addition, the effects of channel bow were conservatively addressed by increasing the amount of channel bow assumed in the MCPR SL calculation. Thus, this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   Prior to the startup of U2C13, licensing analyses are performed (using NRC approved methodology referenced in Technical Specification Section 5.6.5.b) to determine changes in the critical power ratio as a result of anticipated operational occurrences. These results are added to the MCPR Safety Limit values proposed herein to generate the MCPR operating limits in the U2C13 COLR [core operating limits report]. These limits could be different from those specified in the U2C12 COLR. The COLR operating limits thus assure that the MCPR Safety Limit will not be exceeded during normal operation or anticipated operational occurrences. Postulated accidents are also analyzed to confirm NRC acceptance criteria are met.

   The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U2C13 core operating limits. The use of this approved methodology does not increase the probability or consequences of an accident previously evaluated.

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

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

   Response: No

   The change to the MCPR Safety Limits does not directly or indirectly affect any plant system, equipment, or component and therefore does not affect the failure modes of any of these systems. Thus, the proposed changes do not create the possibility of a previously unevaluated operator or a new single failure.

   The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U2C13 core operating limits. The use of this approved methodology does not create the possibility of a new or different kind of accident.

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

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

   Response: No.

   Since the proposed changes do not alter any plant system, equipment, component, or the processes used to operate the plant, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. The proposed MCPR Safety Limits do not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections, because the MCPR Safety Limits calculated for U2C13 preserve the required margin of safety.

   The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U2C13 core operating limits. This approved methodology is used to demonstrate that all applicable criteria are met, thus, demonstrating that there is no reduction in the margin of safety.

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

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

   Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

   NRC Section Chief: Richard J. Laufer.

   [[Page 405]]

   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 amendment request: August 23, 2004.

   Description of amendment request: The proposed amendments would revise the Surveillance Requirements for Technical Specifications 3.6.1.3, ``Primary Containment Isolation Valves,'' for Hatch Units 1 and 2. The proposed amendments would substitute the requirement for valve seat replacement with a requirement to perform an Appendix J leakage rate test on the valves. Conforming revisions to the Technical Specification Bases B 3.6.1.3, ``Primary Containment Isolation Valves'' are also included.

   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 proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seats. The proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated because the extensive industry operating experience derived from test results has demonstrated that the resilient seat material does not experience aging degradation and cause containment isolation valves to leak. Thus, the valves will perform as assumed in the accident analyses and therefore, this change does not involve a significant increase in the consequences of an accident previously evaluated. Further, these valves are not accident initiators, and therefore, this change does not involve a significant increase in the probability of occurrence of a previously evaluated event.

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

   The proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seats. The proposed change does not involve physical alteration of the plant (no new or different type of equipment will be installed nor changes in methods governing normal plant operation). In particular, it does not require the valves to function in any manner other than that which is currently required. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seats. The proposed change does not involve a significant reduction in margin of safety because it has no effect on any safety analysis bases or assumptions. It does not change the leakage acceptance criteria. Sufficient data has been collected to demonstrate that resilient seats do not experience aging degradation. Deleting the seat replacement requirement will not reduce the margin of safety provided by Technical Specifications.

   For the above reasons, the margin of safety is not reduced by this proposed Technical Specifications change.

   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: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

   NRC Section Chief: John A. Nakoski.

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

   Date of amendment request: December 9, 2004.

   Description of amendment request: The proposed amendment would revise the Watts Bar Updated Final Safety Analysis Report to include an alternate methodology for concrete reinforcement bar splicing. The change in methodology applies to restoration of the concrete Shield Building dome as part of the upcoming steam generator replacement project. The alternate methodology uses a Bar-Lock mechanical splice in lieu of the Cadweld splice used for the original design and construction of the plant.

   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.

   No changes in event classification, as discussed in the UFSAR

   [Updated Final Safety Analysis Report] Chapter 15, will occur due to use of the Bar-Lock couplers.

   The restoration of the temporary concrete construction openings in the Shield Building will utilize Bar-Lock couplers to splice new rebar to the existing rebar. The Shield Building structure limits the release of radioactivity following an accident and protects the systems, structures, and components inside containment from external events. The accidents of interest are those that rely on the Shield Building to limit the release of radioactivity to the environment, and those that result from some external events. The design of the Shield Building is such that it is not postulated to fail and initiate an accident described in the UFSAR.

   The Bar-Lock coupler qualification tests detailed in Topical Report 24370-TR-C-001-A demonstrate that the Bar-Lock coupler meets the ASME [American Society of Mechanical Engineers] strength requirements and is, therefore, acceptable for use in nuclear safety-related applications. Based on these test results, it is concluded that use of the Bar-Lock couplers in restoring the temporary concrete construction openings will not reduce the structural capability of the repaired structure. The Shield Building will continue to perform its design function as described in the WBN UFSAR.

   Therefore, the proposed use of the Bar-Lock couplers will not significantly increase 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 design of the Shield Building is such that it is not postulated to fail and initiate an accident described in the UFSAR. The Bar-Lock couplers are passive devices and as such will not initiate or cause an accident.

   The restoration of the temporary concrete construction openings in the Shield Building will utilize Bar-Lock couplers to splice new rebar to the existing rebar. The Bar-Lock coupler qualification tests detailed in Topical Report 24370-TR-C-001-A demonstrate that the Bar-Lock coupler meets the ASME strength requirements and is, therefore, acceptable for use in nuclear safety-related applications. Based on these test results, it is concluded that use of the Bar-Lock couplers in restoring the temporary concrete construction openings will not reduce the structural capability of the Shield Building. The Shield Building will, therefore, continue to perform its design functions as described in the WBN UFSAR.

   Therefore, the possibility of a new or different accident situation occurring as a result of this condition is not created.

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

   Response: No.

   As indicated in the WBN UFSAR, the structural design of the reinforced concrete Shield Building is in compliance with the proposed ACI-ASME [American Concrete Institute--American Society of Mechanical Engineers] (ACI-359) Code for Concrete Reactor Vessels and Containment, Article CC-3000, as issued for trial use, April 1973, for the loading combinations defined in UFSAR Table 3.8.1-1. Allowable stresses are based on this code with the exception of allowable tangential shear stresses in walls

   [[Page 406]]

   where the ACI 318-71 code is used. The reinforcing steel conforms to the requirements of American Society for Testing Maintenance (ASTM) A 615, Grade 60. The WBN UFSAR states that reinforcing bars were lap spliced and Cadwelded in accordance with ACI 318-7 requirements for strength design.

   The restoration of the temporary concrete construction openings in the Shield Building will utilize Bar-Lock couplers to splice new rebar to the existing rebar. The restoration of the construction openings, including use of the Bar-Lock couplers, will conform to the requirements of ACI-359 (April 1973) and ACI 318. Therefore, following completion of the modification, the Shield Building will continue to comply with ACI-359 (April 1973) and ACI 318 requirements.

   In addition to conforming to ACI-359 (April 1973) and ACI 318 requirements, the Bar-Lock coupler qualification tests detailed in Topical Report 24370-TR-C-001-A demonstrate that the Bar-Lock coupler meets the ASME strength requirements.

   Therefore, a significant reduction in the margin to safety is not created by this modification.

   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.

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

   Date of amendment request: October 27, 2004.

   Description of amendment request: The requested change will delete Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.'' The Table of Contents will also be revised to reflect the deletions.

   The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated October 27, 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:

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 letter 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.

   Based upon the reasoning presented above, the requested change does not involve a significant hazards consideration.

   Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

   NRC Section Chief: Robert Gramm.

   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.

   Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck Plant, Middlesex County, Connecticut

   Date of amendment request: January 9, 2004.

   Brief description of amendment: The amendment revises Technical Specifications to incorporate Technical Specification Task Force (TSTF) travelers 152, 258, and 308 to reflect changes due to revision of Part 20 of Title 10 of the Code of Federal Regulations, and TSTF 65 to reflect the use of generic titles

   Date of issuance: December 17, 2004.

   [[Page 407]]

   Effective date: The license amendment shall be implemented within 90 days of its effective date.

   Amendment No.: 200.

   Facility Operating License No. DPR-61: The amendment revises the Technical Specifications.

   Date of initial notice in Federal Register: March 30, 2004 (69 FR 16616).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation Report, dated December 17, 2004.

   No significant hazards consideration comments received: No.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation Report, dated December 17, 2004.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: June 22, 2004.

   Brief description of amendment: The amendment deletes the post- accident monitoring instrumentation requirements to maintain the primary containment hydrogen and oxygen monitors from the Technical Specifications.

   Date of issuance: December 8, 2004.

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

   Amendment No.: 280.

   Facility Operating License No. DPR-59: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 31, 2004 (69 FR 53103). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 8, 2004.

   No significant hazards consideration comments received: No.

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

   Date of application for amendments: January 31, 2003, and supplemented by letters dated July 7 and November 15, 2004.

   Brief description of amendments: The amendments provide new pressure-temperature (P-T) limits for the technical specifications that are valid to 20 effective full power years for each unit. The changes to the P-T curves are based, in part, on the American Society of Mechanical Engineers Code Case -640, ``Alternative Reference Fracture Toughness for Development of P-T Limit Curves Section XI, Division 1,'' which was reviewed and approved by NRC staff for use by the LaSalle County Station in a letter dated November 8, 2000.

   Date of issuance: December 10, 2004.

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

   Amendment Nos.: 170, 156.

   Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: April 1, 2003 (68 FR 15759).

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

   No significant hazards consideration comments received: No.

   Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

   Date of application for amendment: June 25, 2004.

   Brief description of amendment: The amendment revises the Technical Specifications (TSs) to reduce the temperature at which shutdown and control rod drop tests are performed from greater than or equal to 541 degrees Fahrenheit to greater than or equal to 500 degrees Fahrenheit. Additionally, the amendment makes format changes to improve the TS page appearance.

   Date of issuance: December 20, 2004.

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

   Amendment No.: 284.

   Facility Operating License No. DPR-58: Amendment revises the Technical Specifications.

   Date of initial notice in Federal Register: August 3, 2004 (69 FR 46585).

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

   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: August 25, 2003, as supplemented by letters dated October 31, 2003, and March 9, September 28, and November 5, 2004.

   Brief description of amendment: The amendment revises the Technical Specifications (TS) Surveillance Requirement 3.3.2.1.4 and TS Table 3.3.2.1-1 to correct mathematical symbols and use allowable values in the place of analytical limits.

   Date of issuance: December 22, 2004.

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

   Amendment No.: 208.

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

   Date of initial notice in Federal Register: September 30, 2003 (68 FR 56344).

   The supplemental letters dated October 31, 2003, and March 9, September 28, and November 5, 2004, 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 December 22, 2004.

   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: October 5, 2004.

   Brief description of amendment: The amendment deletes technical specification (TS) 6.9.a.2.B (requirement to submit an occupational radiation exposure report), TS 6.9.a.2.C (requirement to report challenges to and failures of pressurizer power operated relief valves and safety valves), and TS 6.9.a.3, ``Monthly Operating Report.''

   Date of issuance: December 22, 2004.

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

   Amendment No.: 179.

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

   Date of initial notice in Federal Register: November 9, 2004 (69 FR 64989).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 22, 2004.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: April 27, 2004, as supplemented by letters dated September 9, 2004, and December 2, 2004.

   [[Page 408]]

   Brief description of amendment: The amendment revised the Safety Limit Minimum Critical Power Ratio values for two recirculation loop and one recirculation loop operation for all fuel types to be used in the core.

   Date of issuance: December 22, 2004.

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

   Amendment No.: 158.

   Facility Operating License No. NPF-57: This amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: June 22, 2004 (69 FR 34704). The September 9, 2004 and December 2, 2004 letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 22, 2004.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: March 31, 2004, as supplemented by letters dated August 9, 2004, and October 20, 2004.

   Brief description of amendment: The amendment created a Technical Specification (TS) for the Oscillation Power Range Monitor system. Additionally, it revised TS 3/4.4.1 to remove Thermal Hydraulic instability-related limiting conditions for operation and required actions.

   Date of issuance: December 22, 2004.

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

   Amendment No.: 159.

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

   Date of initial notice in Federal Register: August 3, 2004 (69 FR 46588). The August 9, 2004, and October 20, 2004 letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 22, 2004.

   No significant hazards consideration comments received: No.

   Dated at Rockville, Maryland, this 27th day of December 2004.

   For the Nuclear Regulatory Commission. James E. Lyons, Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

   [FR Doc. 05-2 Filed 1-3-05; 8:45 am]

   BILLING CODE 7590-01-P