69 Fed. Reg. 70712 (Dec. 7, 2004)

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

Federal RegisterDec 7, 2004

69 Fed. Reg. 70712 (Dec. 7, 2004)

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 November 5, 2004, through November 24, 2004. The last biweekly notice was published on November 23, 2004 (69 FRN 68180).

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. (Note: Public access to ADAMS has been temporarily suspended so that security reviews of publicly available documents may be performed and potentially sensitive information removed. Please check the NRC Web site for updates on the resumption of ADAMS access.) 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/. (Note: Public access to ADAMS has been temporarily suspended so that security reviews of publicly available documents may be performed and potentially sensitive information removed. Please check the NRC Web site for updates on the resumption of ADAMS access.) If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Hearingdocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. (Note: Public access to ADAMS has been temporarily suspended so that security reviews of publicly available documents may be performed and potentially sensitive information removed. Please check the NRC Web site for updates on the resumption of ADAMS access.) If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

Date of amendment request: October 20, 2004.

Description of amendment request: The proposed amendment would revise the containment hatch closure requirement in the Technical Specifications (TSs) during fuel handling and refueling operations. Specifically, the requirement of TS 3.8.6 that the containment equipment hatch remain closed with a minimum of 4 bolts securing it in place is replaced with the requirement that the equipment hatch be capable of being closed during fuel handling and refueling operations.

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

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

Response: No.

The proposed change is only related to a postulated fuel handling accident inside the Reactor Building occurring during fuel loading and refueling activities. The proposed change does not increase the probability of a fuel handling accident in that the proposed change deals with the results of such an accident, not the cause of such an accident. The proposed change does not increase the consequences of an accident previously evaluated in that the TMI Unit 1 Alternative Source Term has been previously reviewed and approved by the NRC [Nuclear Regulatory Commission], and this proposed change is consistent with the assumptions of [that] previous analysis. The Alternative Source Term analysis for the Fuel Handling Accident [i]nside the Reactor Building takes no credit for the closure of the containment equipment hatch opening or for a filtered release. Previous analyses of external events were reviewed and the proposed [change does] not affect the conclusions of these analyses. Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not affect nor create a different [kind] of fuel handling accident. The proposed change is consistent with the existing licensing basis accident analysis for a postulated fuel handling accident inside containment during fuel loading and refueling operations. The proposed change does not involve any structure, system, or component relied upon to mitigate any design basis accident. The revised operations are consistent with the fuel handling accident analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

Previously approved analysis demonstrates that the resultant dose consequences are well within the appropriate acceptance criteria. The proposed change is bounded by the previously approved analysis, and thus the margin of safety, as defined by 10 CFR 50.67 and Regulatory Guide 1.183, is maintained. Maintaining the capability to close the containment equipment hatch opening following an evacuation of the containment would further reduce the dose consequences in the event of a fuel handling accident inside containment and provides additional margin to the calculated doses. 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: Thomas S. O'Neill, Associate General Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

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: July 13, 2004.

Description of amendment request: The proposed amendment would revise the fire protection license condition consistent with the guidance provided in Generic Letter 88-12, “Removal of Fire Protection Requirements from Technical Specifications.”

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

The proposed license amendment is an administrative change. The proposed change will revise the fire protection license condition consistent with the guidance provided in Generic Letter 88-12. This revision to the fire protection license condition was to be made at the time the fire protection requirements were relocated from the Technical Specifications to licensee controlled documents. However, this change was not requested, nor granted in License Amendment Request dated December 4, 1996, approved in Amendment Nos. 227 and 201. Therefore, the necessary change was not reflected in the Operating Licenses.

This administrative request does not impact the probability or consequences of an accident previously evaluated. The incorporation of the requested change requires that an evaluation be performed to determine the need for prior NRC approval for changes to the Fire Protection Program. Changes to administrative programs will result from the addition of this condition in the Operating License. However, no changes to the facility or the way it is operated are expected to result from this change.

Therefore, this 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 change is administrative. 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.

The fire protection requirements were removed from the Technical Specifications in accordance with Generic Letter 88-12 in Amendment Nos. 227 and 201, with the exception of the change to the Operating License's fire protection license condition. The proposed administrative change will require an evaluation be performed to determine the need for prior NRC approval for changes to the Fire Protection Program. No margin of safety is impacted by the proposed administrative change.

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

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.

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

Date of amendment request: September 7, 2004.

Description of amendment request: The proposed changes would allow performance of testing for nozzle containment spray blockage to be based on the occurrence of activities that could cause nozzle blockage rather than a fixed periodic basis. Currently, the testing for nozzle blockage is performed every 10 years and Dominion Nuclear Connecticut, Inc. (DNC) proposes to change this frequency to “following maintenance that could cause nozzle blockage”. In addition, specific details limiting the testing method to an air or smoke test that are currently part of the surveillance requirements would be removed. The Technical Specification Bases section would be updated with applicable spray nozzle testing information and will be expanded to include visual inspection.

Criterion 1: Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The spray nozzles and the associated containment spray systems are designed to perform accident mitigation functions only. The QSS [quench spray system] and RSS [recirculation spray system] and associated components are not considered as initiators of any analyzed accidents. The proposed change does not modify any plant equipment and only changes the frequency for performance of a surveillance test which does not impact any failure modes that could lead to an accident. Removing the testing details from the surveillance does not change the ability of the spray nozzles to function as assumed and therefore there is no affect on the consequence of any accident. Also the proposed change does not impact the capability of the QSS and RSS to perform accident mitigation functions and therefore does not impact the consequences of an accident. Based on this discussion, the proposed amendment does not increase the probability or consequence of an accident previously evaluated.

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

Response: No.

The QSS and RSS are not being physically modified and there is no impact on the capability of the systems to perform accident mitigation functions. No system setpoints are being modified and no changes are being made to the method in which borated water is delivered to the spray nozzles. The testing requirements imposed by this proposed change to check for nozzle blockage following activities that could cause nozzle blockage do not introduce new failure modes for the system. By removing the testing details from the surveillance requirement, additional flexibility in the testing methodology is allowed for verifying the nozzles are unobstructed and assists in ensuring operability of the systems. 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.

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

Response: No.

The proposed change does not change or introduce any new setpoints at which mitigating functions are initiated. No changes to the design parameters of the QSS and RSS are being proposed. No changes in system operation are being proposed by this change that would impact an established safety margin. The proposed change modifies the frequency for verification of nozzle operability in such a way that continued high confidence exists for the containment spray systems to functions as designed. In addition, removing specific testing details from the surveillance does not affect the ability of the spray nozzles to function as designed. Therefore, based on the above, the proposed amendment does not involve a significant reduction in a margin of safety.

In summary, DNC concludes that the proposed amendment does not represent a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.

NRC Section Chief: Daniel S. Collins, Acting Section Chief.

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

Date of amendment request: November 1, 2004.

Description of amendment request: The proposed amendment would allow the use of a new gantry crane as part of the cask handling system in the fuel storage building (FSB) for moving spent fuel casks up to 110 tons into and out of the spent fuel pit.

1. The proposed amendment would allow the use of the new [IP2] FSB gantry crane for loads up to 110 tons, and the new crane will prevent the load from being dropped given a single malfunction or failure of a portion of the crane. The handling of a loaded spent fuel cask is below the maximum load that the crane is designed to handle.

This change does not increase the probability of an accident previously evaluated because the probability of a load drop is eliminated. The new crane system is designed in accordance with NUREG-0554 and Ederer's Generic Licensing Topical Report EDR-1 (NP)-A, that if a portion of the crane lifting devices malfunctions or fails, the load will move a limited distance downward prior to backup restraints becoming engaged. The change does not increase the consequences of an accident.

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

2. The process for transporting a cask with the new crane is limited from the FSB truck bay floor to the cask pit area of the spent fuel pool. Once a cask is loaded with spent fuel, it is lifted from the spent fuel pit, and lowered into the truck bay. The cask is never carried over spent fuel in the spent fuel pit.

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

3. The [IP2] FSB gantry crane has been installed to comply with the single-failure-proof requirements of NUREG-0554 and NRC-approved Ederer Topical Report EDR-1, Revision 3, dated October 8, 1982. The installation provides additional load carrying capability up to 110 tons and additional safety features to prevent a cask drop. The safety margins provided by the new crane prevent failure of the crane or any lifting devices associated with it. The implementation of NUREG-0612 general guidelines for the FSB gantry crane provides further assurance that safe load paths, procedures, crane operator training, and crane inspection and maintenance activities will be established to ensure crane operation is performed in a consistently reliable manner.

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

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.

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

Date of amendment request: October 29, 2004.

Description of amendment request: The proposed change to the Technical Specification (TS) assures that sufficient fuel oil inventories are available in the Emergency Diesel Generator (DG) Fuel Oil Storage Tank (FOST) to support the Extended Power Uprate (EPU) consistent with the current licensing basis.

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

Response: No.

The proposed change, which accounts for the fuel oil consumption related to the EPU, will revise the minimum TS volumes associated with the DG FOST. The change continues to assure that each DG can provide on-site power in the event of an accident and thereby assist in the mitigation of the accident.

The proposed change to the five day full load fuel oil volume results in a usable volume 37,000 gallons of fuel oil. The proposed change removes the unusable volume (760 gallons) and other conservatism (240 gallons) that were included in the current TS. The fuel oil volume continues to allow for a runtime of 5 days at full load with the removal of this conservatism.

These changes will not affect the capability of the AC [alternate current] Sources to power the systems required to safely shutdown the plant. The proposed changes are not accident initiators nor do they adversely affect accident initiators or precursors. These changes do not affect the mitigation of any accident nor do they adversely affect structures, systems, or components that are utilized for the mitigation of any analyzed events. The proposed changes will have no affect on the radiological consequences of any accident.

The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in the evaluation of radiological consequences.

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.

Fuel oil is not an accident initiator. Therefore, the possibility of a new or different kind of accident will not be created in relationship to the proposed changes to the TS. No modifications are proposed to the existing fuel oil storage system that would alter the design function or the ability of the DG to perform its safety function.

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

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

Response: No.

The proposed change to the 7-day time dependent fuel oil volume results in an increase in volume to accommodate fuel oil consumption needed to support the EPU.

The reduced volume associated with the 5-day full load volume is equivalent to less than one hour of runtime and does not result in a significant reduction in a margin of safety because the calculational method results in a conservative estimate of the amount of fuel that would be needed during a design bases accident.

The proposed change does not result in a change of the design bases for the DG or its support systems. The system will continue to provide a reliable source of power for safe shutdown of the reactor, assuming the single failure of one of the DGs. Independence, redundancy, and testability are maintained such that the required safety function can be performed by either DG train.

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

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

NRC Section Chief: Michael K. Webb, Acting.

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

Date of amendment request: November 5, 2004.

Description of amendment request: Waterford 3 Technical Specification (TS) currently requires all the Dry Cooling Tower (DCT) fans with cooling coils under the missile grating to be operable during a tornado watch. If one (or more) of these DCT fans is inoperable during a tornado watch, it is required to be restored to operable status within one hour or place the plant in Hot Standby within 6 hours. The purpose of this TS change is to allow the plant to take credit for the DCT fans that are not under the missile grating to meet the fan requirements specified in TS Table 3.7-3. In addition, the proposed change will delete the requirement to monitor ambient temperature conditions when the DCT fan is inoperable on an inoperable train of the Ultimate Heat Sink (UHS).

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

Response: No.

The proposed change will delete the requirement to have all the DCT fans with cooling coils under the missile grating operable during a tornado watch. It has been determined (using tornado missile strike probability methodology—TORMIS) that the probability of damage to the DCT components not under the missile grating (fans, motors, associated conduits, electrical boxes, and cooling coils) is acceptably low. With respect to the probability of occurrence or the consequences of an accident previously analyzed in the FSAR [Final Safety Analysis Report], the possibility of a tornado reaching Waterford 3 and causing damage to plant systems, structures and components, including the DCT fans, is a design basis event considered in the FSAR. The probability of a tornado-generated missile strike on the DCT components was analyzed using the NRC [Nuclear Regulatory Commission] Staff approved probability method TORMIS. TORMIS showed that the change from essentially relying on DCT fans with cooling coils under the missile grating to relying on all operable DCT fans during a tornado watch is acceptable and represents an acceptably low probability of occurrence of tornado generated missile strikes on the DCTs. On this basis, the proposed change is not considered to constitute a significant increase in the probability of occurrence or the consequences of an accident.

The proposed change to TS Action 3.7.4.d eliminates an unnecessary requirement, to determine ambient conditions and verify compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan is inoperable due to its associated train of UHS being inoperable. The determination of ambient temperature conditions and validation of the required number of fans based on the temperature will continue to be required when an UHS fan is inoperable and the associated train of UHS is operable. The UHS fans will not dissipate the required heat load when the associated train of UHS is inoperable, assuming the coincident ambient wet bulb temperature (78 °F) at the historically highest ambient dry bulb temperature (102 °F). This change represents a burden reduction and has no impact on plant safety. This change also does not impact the initiators or mitigation of any design basis event.

The proposed revision to TS Table 3.7-3 ensures consistency with the revisions to the TS Actions. This change is administrative and has no impact on the initiators or the mitigation of accidents previously evaluated.

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 will delete the requirement to have all the DCT fans with cooling coils under the missile grating operable during a tornado watch. It has been determined that the probability of damage to the DCT components not under the missile grating is acceptably low. A tornado at Waterford 3 is a design basis event considered in the FSAR. Therefore, the change will not contribute to the possibility of or be the initiator for any new or different kind of accident, or occur coincident with any of the design basis accidents in the FSAR. The low probability threshold established for tornado missile damage to system components is consistent with these assumptions.

The proposed revision to TS Table 3.7-3 ensures consistency with the revisions to the TS Actions. This is an administrative change.

The above changes also do not have any impact on plant systems nor do they have any impact on the way plant systems are operated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident.

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

Response: No.

The proposed changes do not involve a significant reduction in a margin of safety. The existing licensing basis for Waterford 3 with respect to the design basis event of a tornado reaching the plant, generating missiles, and directing them toward the DCT components is to provide positive missile barriers. The basis for the proposed change recognizes that there is a low probability, below an established acceptance limit, that a tornado missile will strike DCT components. The change from essentially relying on DCT fans with cooling coils under the missile grating to relying on all operable DCT fans during a tornado watch is acceptable and represents an acceptably low probability of occurrence of tornado generated missile strikes on the DCTs. Therefore, this change is not considered to constitute a significant decrease in the margin of safety.

The proposed change to TS Action 3.7.4.d eliminates an unnecessary requirement, to determine ambient conditions and verify compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan is inoperable due to its associated train of UHS being inoperable. The determination of ambient temperature conditions will continue to be required when an UHS fan is inoperable with the associated train of UHS operable. When the UHS is not available, the fans cannot dissipate the required heat load, assuming the coincident ambient wet bulb temperature (78 °F) at the historically highest ambient dry bulb temperature (102 °F). Therefore, it is not necessary to monitor ambient temperature and ensure the fan requirements of TS Table 3.7-3 are met when the UHS train is inoperable. This change represents an operational burden reduction and has no impact on plant safety.

The proposed revision to TS Table 3.7-3 ensures consistency with the revisions to the TS Actions. These changes are administrative and have no impact on the operation of the plant, mitigation of analyzed events, or plant safety,

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

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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

NRC Section Chief: Michael K. Webb, Acting.

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

Date of application for amendment request: February 27, 2004, as supplemented by letter dated October 11, 2004.

Description of amendment request: The proposed amendments would revise the Dresden Nuclear Power Station and Quad Cities Nuclear Power Station technical specifications (TS) to add the Oscillation Power Range Monitor (OPRM) instrumentation to the TS.

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

This proposed change has no impact on any of the existing neutron monitoring functions. It activates the OPRM scram function and updates the TS to add the OPRM-related functions.

Activation of the OPRM scram function will replace the current methods that require operators to insert an immediate manual reactor scram in the reactor operating region where thermal hydraulic instabilities could potentially occur. While this region will continue to be avoided during normal operation, certain transients, such as a reduction in reactor recirculation flow, could place the reactor in this region. Operation in this region, with the OPRM instrumentation scram function activated would no longer require an immediate manual scram and thus may potentially cause a marginal increase in the probability of occurrence of an instability event. This potential increase in probability is acceptable because the OPRM function will automatically detect the instability condition and initiate a reactor scram before the Minimum Critical Power Ratio (MCPR) Safety Limit is reached. Consequences of the potential instability event are reduced because of the more reliable automatic detection and suppression of an instability event, and the elimination of dependence on the manual operator actions. Operators will continue to monitor for indications of thermal hydraulic instability when the reactor is operating in regions of potential instability as a backup to the OPRM instrumentation.

The potential for spurious reactor scrams has been evaluated. Operating experience with the OPRM has not resulted in the generation of any spurious reactor scram signals.

Therefore, the proposed changes do not involve a significant increase in the probability of 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 replace procedural actions that were established to avoid operating conditions where reactor instabilities might occur with an NRC approved automatic detect and suppress function (i.e., OPRM).

Potential failures in the OPRM trip function could result in either failure to take the required mitigating action or an unintended reactor scram. These are the same potential effects of failure of the operator to take the appropriate action under the current procedural actions. The effects of failure of the OPRM equipment are limited to reduced or failed mitigation, but such failure cannot cause an instability event or other type of accident.

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 OPRM trip function is being implemented to automate the detection and subsequent suppression of an instability event prior to exceeding the MCPR Safety Limit. The OPRM trip provides a trip output of the same type as currently used for the APRM [Average Power Range Monitor]. Its failure modes and types are identical to those for the present APRM output. Since the MCPR Safety Limit will not be exceeded as a result of an instability event following implementation of the OPRM trip function, it is concluded that the proposed change does not reduce the margin of safety.

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

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.

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

Date of amendment request: October 4, 2004.

Description of amendment request: The proposed change requests approval to apply the Westinghouse best-estimate loss-of-coolant accident (BELOCA) analysis methodology to Beaver Valley Power Station, Unit Nos. 1 and 2, and requests an amendment of the related Technical Specifications. The BELOCA methodology has previously been approved on a generic basis by the NRC as presented in Topical Report WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), “Code Qualification Document for Best-Estimate LOCA Analysis,” March 1998.

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

No. No physical changes are required as a result of implementing best-estimate large break loss of coolant accident (LOCA) methodology and associated Technical Specification changes. The plant conditions used in the analysis are bounded by the design conditions for all equipment in the plant. Therefore, there will be no increase in the probability of a LOCA. The consequences of a LOCA are not being increased, since it is shown that the emergency core cooling system is designed so that its calculated cooling performance conforms to the criteria contained in 10 CFR 50.46, Paragraph b. No other accident is potentially affected by this change.

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

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

No. There are no physical changes being made to the Beaver Valley Power Station units. No new modes of plant operation are being introduced. The parameters used in the analysis are within the design limits of the existing plant equipment. All plant systems will perform as designed during the response to a potential accident.

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

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

No. It has been shown that the methodology used in the analysis would more realistically describe the expected behavior of plant systems during a postulated LOCA. Uncertainties have been accounted for as required by 10 CFR 50.46. A sufficient number of LOCAs with different break sizes, different locations and other variations in properties are analyzed to provide assurance that the most severe postulated LOCAs are addressed. It has been shown by analysis that there is a high probability that all criteria contained in 10 CFR 50.46, Paragraph b are met.

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

Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: October 15, 2004.

Description of amendment request: The licensee proposed to revise Section 1.7, regarding the definition of “Instrument Channel Calibration,” of the Technical Specifications by incorporating the additional guidance for instrument channels containing resistance temperature detector (RTD) and thermocouple sensors provided by the “Standard Technical Specifications, General Electric Plants, BWR [Boiling-Water Reactor]/4 Specifications,” NUREG-1433, Revision 3. The revised definition would permit in place qualitative assessment of the RTDs and thermocouples, and to allow a signal to be injected downstream of the sensor for the purpose of calibrating the remainder of the channel.

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

Response: No.

The proposed change revises the definition of Instrument Channel Calibration to allow RTD and thermocouple sensors to be qualitatively assessed with the remainder of the channel being calibrated normally. Instrument channel calibration is not an initiator of any accident previously evaluated. Furthermore, the proposed change will not affect the ability of the channel being calibrated to respond as assumed in any accident previously evaluated. The qualitative evaluation of sensor behavior for non-adjustable sensors will provide an accurate indication of sensor operation and will assure that portion of the channel is operating properly, ensuring that the consequences of an accident will remain as previously evaluated. Therefore, the proposed Technical Specification changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change revises the definition of Instrument Channel Calibration to allow RTD and thermocouple sensors to be qualitatively assessed with the remainder of the channel being calibrated as at present. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change also does not adversely affect the operation or operability of existing plant equipment. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed change involves the definition of Instrument Channel Calibration to allow RTD and thermocouple sensors to be qualitatively assessed with the remainder of the channel being calibrated normally. The proposed change to the Instrument Channel Calibration definition does not alter the ability of a channel to respond as designed or as assumed in the safety analyses. Therefore, this change does not involve a significant reduction in a margin of safety.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: October 22, 2004.

Description of amendment request: The licensee proposed to revise the Technical Specifications (TSs), Section 5.0, “Design Features,” by relocating the information to the Updated Final Safety Analysis Report (UFSAR). Specifically, the amendment would relocate these Sections: 5.3, “Reactor Vessel,” 5.4, “Containment,” and 5.6, “Seismic Design.” The licensee stated that such information does not meet the criteria of 10 CFR 50.36(c)(4) for inclusion in the TSs. The information to be relocated to the UFSAR already exists in the UFSAR, and will continue to be controlled by 10 CFR 50.59 and 10 CFR 50.71(e).

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

Response: No.

The proposed change relocates certain design details from the TS to the UFSAR, where the information already exists. The UFSAR is maintained in accordance with 10 CFR 50.71(e). Any future change to these design details as described in the UFSAR will be evaluated per the requirements of 10 CFR 50.59 to assure that the change does not result in more than a minimal increase in the probability or consequences of an accident previously evaluated.

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

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

Response: No.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not impose or eliminate any requirements, and adequate control of the information will be maintained in accordance with applicable regulatory requirements.

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

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

Response: No.

The proposed change has no impact on any analysis assumptions. The design details that are being removed from the TS already exist in the UFSAR. Any future change to these design details described in the UFSAR will be evaluated per the requirements of 10 CFR 50.59 to assure that the change does not result in a design basis limit [or] a fission product barrier being exceeded or altered.

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

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: October 14, 2004.

Description of amendment request: The proposed changes correct administrative errors in Technical Specifications 3.10.i and 6.9.a.4.A.

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

NMC [Nuclear Management Company, the licensee] Response for Proposed Change to TS 3.10.i: No. The NMC has reviewed the proposed change in accordance with the provisions of 10 CFR 50.92 to show no significant hazards exist. This change is being proposed to correct an administrative error that currently exists within the KNPP [Kewaunee Nuclear Power Plant] Technical Specifications; therefore it would not have an affect on the probability of an accident previously evaluated.

NMC Response for Proposed Change to TS 6.9.a.4.A: No. The NMC has reviewed the proposed change in accordance with the provisions of 10 CFR 50.92 to show no significant hazards exist. This change is being proposed to correct an administrative error that currently exists within the KNPP Technical Specifications; therefore it would not have an affect on the probability 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?

NMC Response for Proposed Change to TS 3.10.i: No. The proposed change does not alter plant configuration, operating setpoints, or overall plant performance. Therefore, the proposed change would not create the possibility of a new or different kind of accident from any accident previously evaluated.

NMC Response for Proposed Change to TS 6.9.a.4.A: No. The proposed change does not alter plant configuration, operating setpoints, or overall plant performance. Therefore, the proposed change would not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

NMC Response for Proposed Change to TS 3.10.i: No. The proposed change does not involve a significant reduction in a margin of safety. Inclusion of the omitted word “and” in TS 3.10.i will enhance the margin of safety.

NMC Response for Proposed Change to 6.9.a.4.A: No. The proposed change does not involve a significant reduction in a margin of safety. Correction of the references in TS Section 6.9.a.4.A will enhance the margin of safety.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, P.O. Box 1497, Madison, WI 53701-1497.

NRC Section Chief: L. Raghavan.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power Plant, Unit 3, Humboldt County, California

Date of amendment request: July 9, 2004.

Description of amendment request: The Humboldt Bay Power Plant (HBPP), Unit 3, is a decommissioning nuclear power plant that was permanently shutdown in July 1976. In December of 2003, Pacific Gas and Electric (PG&E or the licensee) applied for a license to store its spent fuel in an onsite dry cask independent spent fuel storage installation (ISFSI). Moving the spent fuel to an ISFSI would permit the licensee to begin significant decommissioning activities. The licensee has chosen to use a Holtec HI-STAR HB spent fuel cask handling system involving a spent fuel multipurpose canister and overpack. To facilitate spent fuel transfer from the HBPP spent fuel pool to the ISFSI, the licensee will also need to install a new crane that can be used to lift the cask handling system loaded with spent fuel assemblies. The licensee states it will be able to satisfy the applicable guidance of NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants,” and NUREG-0554. “Single-Failure Proof Cranes for Nuclear Power Plants,” in performing the necessary movement of the HBPP spent fuel to dry cask storage. The licensee has requested a license amendment that approves the use of the crane and associated changes to the HBPP Defueled Safety Analysis Report (DSAR) along with analyses, design, and procedural changes required to implement transfer of the spent fuel from the spent fuel pool to the ISFSI.

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

No. With the HI-STAR HB System and the associated design and handling procedures, all cask drops and other events, which could damage other spent fuel, have been precluded through the robust handling systems, and mechanical arrangement that preclude crane movement over spent fuel, meeting the guidelines of NUREG-0612. Revisions of the HBPP procedures implementing the control of heavy loads ensures that PG&E will meet the NUREG-0612 guidelines and will protect the fuel storage locations and the new HI-STAR HB System loading/unloading activities. As a result of this design approach, a cask-handling accident that results in a significant offsite radiological release is not considered credible as demonstrated by the probabilistic evaluation that was performed using the guidelines of NUREG-0612 Appendix B and updated information from NUREG-1774 [“A Survey of Crane Operating Experience at U.S. Nuclear Power Plants from 1968 through 2002.”]

Other HBPP licensing-basis events, such as the drop of a spent fuel assembly, have not been affected by these changes and remain bounding events for potential radiological consequences.

The proposed design of the dry cask system, the handling system, and associated procedural controls provide assurance that: (1) Operational errors and mishandling events, and (2) support system malfunctions will not result in an increase in the probability or consequence of an accident previously analyzed.

The proposed changes to use the Holtec HI-STAR HB system have been evaluated for seismic events and tornado missile impacts and it has been determined that these changes will not result in an increase in the probability or consequences of an accident previously evaluated. The Fire Protection Program will ensure that the combustible materials are properly controlled such that the total combustibles meet the current program commitments. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident.

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

No. The engineering design measures and the handling procedures preclude the possibility of new or different kinds of accidents. Damage to 10 CFR 50 structures, systems, and components from the cask handling and associated activities, and events resulting from possible damage to contained fuel have been considered. Both the types of accidents and the results remain within the envelope of existing HBPP DSAR licensing basis analyses, as demonstrated by the PG&E and Holtec analyses.

The rupture of multipurpose canister (MPC) dewatering, forced helium dehydration or related closure system lines or the malfunction of equipment during cask handling operations resulting in radiological consequences are bounded by the HBPP DSAR fuel-handling accident analysis.

Other design considerations, such as spent fuel pool (SFP) thermal, water chemistry and clarity, criticality, and structural, were evaluated and determined not to introduce the possibility of a new or different kind of accident from any previously evaluated.

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

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

No. With the Holtec HI-STAR HB System, and the associated design and handling procedures, cask drops and other events have been precluded through robust load handling systems, providing defense-in-depth as described in NUREG-0612. Cask tipovers, while not considered credible, are shown to be below the 60g limit, preventing damage to the contained fuel assemblies (and associated structures), and meeting the analysis guidelines of NUREG-0612. As the existing licensing basis assumes a nonmechanistic drop damaging the SFP and all fuel, the result of this design approach with the minimization of drops and the associated structural challenges assure the margin of safety has been maintained.

Other HBPP licensing-basis events, such as the drop of a spent fuel assembly, have not been affected by these changes and remain bounding events. Revision of HBPP procedures implementing the control of heavy loads to incorporate the additional restrictions on heavy loads movement will not affect the procedures or methodology used and will, therefore, not affect margins.

Adverse effects from seismic events and/or cask drops or tipovers have been evaluated, assuring that the fuel, MPC, and overpack remain within their design bases. Since design basis criteria are fully satisfied, there is no impact on the margin of safety.

The Fire Protection Program will continue to ensure that the combustible materials are properly controlled such that the total combustibles meet the current program commitments. Thus, there are no significant reductions in margin of safety associated with these changes.

Other design considerations, such as SFP thermal, water chemistry, criticality, and structural, were evaluated and determined to not involve a reduction in a margin of safety.

Based on the above evaluations, the licensee concludes that the activities associated with the above changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92 and accordingly, a finding by the NRC of no significant hazards consideration is justified.

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

NRC Section Chief: Claudia Craig.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of amendment request: September 8, 2004.

Description of amendment request: The proposed amendment would revise Technical Specification 3.1.8, “Scram Discharge Volume (SDV) Vent and Drain Valves,” to allow a vent or drain line with one inoperable valve to be isolated instead of requiring the valve to be restored to Operable status within 7 days.

The NRC staff issued a notice of opportunity for comment in the Federal Register on February 24, 2003 (68 FR 8637), on possible amendments to revise the action for one or more SDV vent or drain lines with an inoperable valve, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 15, 2003 (68 FR 18294). The licensee affirmed the applicability of the model NSHC determination in its application dated September 8, 2004.

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

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

A change is proposed to allow the affected SDV vent and drain line to be isolated when there are one or more SDV vent or drain lines with one valve inoperable instead of requiring the valve to be restored to operable status within 7 days. With one SDV vent or drain valve inoperable in one or more lines, the isolation function would be maintained since the redundant valve in the affected line would perform its safety function of isolating the SDV. Following the completion of the required action, the isolation function is fulfilled since the associated line is isolated. The ability to vent and drain the SDV is maintained and controlled through administrative controls. This requirement assures the reactor protection system is not adversely affected by the inoperable valves. With the safety functions of the valves being maintained, the probability or consequences of an accident previously evaluated are not significantly increased.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Thus, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed change ensures that the safety functions of the SDV vent and drain valves are fulfilled. The isolation function is maintained by redundant valves and by the required action to isolate the affected line. The ability to vent and drain the SDV is maintained through administrative controls. In addition, the reactor protection system will prevent filling of the SDV to the point that it has insufficient volume to accept a full scram. Maintaining the safety functions related to isolation of the SDV and insertion of control rods ensures that the proposed change does not involve a significant reduction in the margin of safety.

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.

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

Date of amendment request: September 22, 2004.

Description of amendment request: The proposed amendment would extend the validity of the reactor pressure vessel (RPV) pressure-temperature (P-T) limit curves from May 1, 2005, to May 1, 2006.

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

Response: No.

The evaluation for the Unit 2 P-T limit curves for 32 EFPYs [effective full-power years] was performed using the approved methodologies of 10 CFR [Part] 50 Appendix G and Code Case-640. The curves generated from these methods were approved as Amendment 174 (Ref. 1) and are currently in the Unit 2 TS. These curves ensure the P-T limits will not be exceeded during any phase of reactor operation. Resolution of the current industry issues related to fluence calculation methodology required PPL to limit applicability of the curves to May 1, 2005 for Unit 2. The proposed change does not alter any of the technical information shown on the present P-T curves. The change extends the expiration date for one year while maintaining the total accumulated exposure well below the 32 EFPY maximum exposure lifetime limit. Therefore, there is no increase in the probability or consequences of any previously evaluated accident as a result of this change.

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 involves changing the expiration date on the Unit 2 P-T limit curves. The change does not affect the present operating margin in the P-T limit curves for inservice leakage and hydrostatic pressure testing, non-nuclear heatup and cooldown, and criticality. Operation in accordance with the present P-T curves, developed in accordance with the provisions of ASME Code [American Society of Mechanical Engineers Boiler and Pressure Vessel Code], Section XI, Appendix G; 10 CFR [Part] 50 Appendix G, and ASME Code Case-640 provides adequate protection against a non-ductile-type fracture of the RPV. This proposed change does not create the possibility of any new or different [kind] of accident. The change extends the expiration date of the present P-T curves and does not result in any new or unanalyzed operation of any system or piece of equipment important to safety.

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

Response: No.

The technical information contained in the present P-T curves approved by Amendment 174 (Ref. 1) is not affected by this change. Extending the expiration date of the curves from May 1, 2005 to May 1, 2006 will not reduce the margin of safety to RPV brittle fracture.

Since the Unit 2 P-T curves have a maximum lifetime exposure of 32 EFPYs and the anticipated exposure by May 1, 2006 will be well below the maximum value, the margin of safety is not reduced as the result of this change in expiration date. Resolution of the current industry issues related to fluence calculation methodology requires PPL to limit applicability of the Unit 2 P-T curves to May 1, 2006.

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.

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

Date of amendment request: September 23, 2004.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to require automatic starting of the auxiliary feedwater (AFW) pumps upon trip of the Turbine Driven Main Feedwater (TDMFW) pumps only when one or more of TDMFW pumps are operating.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (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. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The design basis events which impose AFW safety function requirements are loss of normal main feedwater, main feedline or main steamline break, loss of offsite power, loss of coolant accident, and small break loss of coolant accident. These accident evaluations assume actuation of AFW occurring due to low-low steam generator level or a safety injection signal. These signals are required safety related features unlike start-up of the AFW pumps due to the trip of both TDMFW pumps which is an anticipatory function and not required for either transient or accident analyses. Requiring this function only when the TDMFW pumps are running will not impact any previously evaluated design basis events. 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?

No. This TS change involves the automatic start of the AFW pumps when the TDMFW Pumps trip. This change involves a function that is not a safety related feature and, therefore, is not credited in either transient or accident analyses. Since this change only affects the point at which this trip function needs to be operable and does not affect the function that actuates AFW due to low-low steam generator level or a safety injection signal, it will not be an initiator to a new or different kind of accident from any accident previously evaluated.

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

No. This TS change involves the automatic start of the AFW pumps when the TDMFW pumps trip which is not a safety related plant function. This change does not change any values or limits involved in a safety related function. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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 application request: October 27, 2004.

Description of amendment request: The amendment would revise Technical Specification (TS) 3.7.3, “Main Feedwater Isolation Valves (MFIVs),” to add the main feedwater regulating valves (MFRVs) and the associated MFRV bypass valves (MFRVBVs). In addition, the allowed outage time, or completion time, for inoperable MFIVs would be extended.

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

The proposed changes add the MFRVs and MFRVBVs to TS 3.7.3 and extend the Completion Time for one or more MFIVs inoperable from 4 hours to 72 hours. Extending the Completion Time is not an accident initiator and thus does not change the probability that an accident will occur. However, it could potentially affect the consequences of an accident if an accident occurred during the extended unavailability of the inoperable MFIV. The increase in time that the MFIV is unavailable is small and the probability of an event occurring during this time period which would require isolation of the MFW [main feedwater] flow paths is low. Moreover, the redundancy provided by the MFRVs and MFRVBVs, which have the same actuation signals and closure time requirements as the MFIVs, provides adequate assurance that automatic feedwater isolation will occur if called upon.

Therefore, the proposed changes do 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.

Closure of the MFIVs is required to mitigate the consequences of the Main Steam Line Break and Main Feedwater Line Break accidents. The MFRVs and MFRVBVs provide a diverse backup to this function. [The extended Completion Time for inoperable MFIVs is not an accident initiator.] The proposed changes do 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 changes do not revise any Technical Specification [Safety] Limit or accident analysis assumption. Therefore, [they do] not involve a reduction in a margin of safety.

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

NRC Section Chief: Robert A. Gramm.

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

Date of application request: October 27, 2004.

Description of amendment request: The amendment would delete or revise license conditions in the operating license for the Callaway Plant because the requirements are either obsolete or adequately described elsewhere. The amendment would also revise Technical Specification Tables 5.5.9-2, “Steam Generator Tube Inspection,” and 5.5.9-3, “Steam Generator Repaired Tube Inspection,” to delete the requirement to notify the NRC pursuant to 10 CFR 50.72(b)(2) if the steam generator tube inspection results in a C-3 classification because reporting requirements are given in the regulations.

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

This request involves administrative changes only. The changes consist of duplicates or overly burdensome reporting requirements or the deletion of completed items required by [the TSs or] conditions from the original issuance of Operating License NPF-30 [for the Callaway Plant]. No actual plant equipment or accident analyses will be affected by the proposed changes. Therefore, the proposed changes do 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.

This request involves administrative changes only. The changes consist of duplicates or overly burdensome reporting requirements or the deletion of completed items required by [the TSs or] conditions from the original issuance of Operating License NPF-30. No actual plant equipment or accident analyses will be affected by the proposed change[s] and no failure modes not bounded by previously evaluated accidents will be created. Therefore, the proposed changes do not create 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.

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel and fuel cladding, Reactor Coolant System pressure boundary, and containment structure [pressure boundary]) to limit the level of radiation dose to the public. This request involves administrative changes only.

No actual plant equipment or accident analyses will be affected by the proposed change[s]. The changes consist of duplicates or overly burdensome reporting requirements or the deletion of completed items required by [the TSs or] conditions from the original issuance of Operating License NPF-30. Additionally, the proposed changes will not relax any criteria used to establish safety limits, will not relax any safety system settings, or will not relax the bases for any limiting conditions of operation. Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

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

NRC Section Chief: Robert A. Gramm.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: September 15, 2004.

Description of amendment request: The proposed changes will change the Administrative Controls Section of the Technical Specifications (TS) in order to incorporate title changes, change the location where the plant-specific titles and TS titles are correlated, and relocate the unit staff requirements to the Quality Assurance Program. These proposed changes will support the implementation of proposed Virginia Electric and Power Company Topical Report DOM-QA-1, “Nuclear Facility Quality Assurance Program Description,” currently under NRC staff review. In addition, these proposed TS changes eliminate the descriptions of the onsite and offsite safety review organizations.

1. Operation of Surry Units 1 and 2 in accordance with the proposed license amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change is administrative in nature and does not affect plant systems, structures or components (SSCs) or plant operation during normal or accident conditions. The proposed change only affects the designated titles of personnel, rewords or relocates requirements within TS or deletes requirements that are either not required to be part of TS or are already required by regulation. The change also relocates the detailed description of the onsite and offsite safety review organizations and non-licensed personnel qualification requirements to the Quality Assurance Program. Therefore, this change has no bearing on the probability of an accident. The management organizational structure and safety and operational reviews have not changed and, therefore, do not impact the ability of operating procedures or administrative controls to prevent or mitigate a previously evaluated accident. As such, this change does not alter the conclusions of the existing safety analyses and therefore does not alter the consequences of an accident previously evaluated.

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

The proposed administrative change continues to ensure that adequate management oversight exists at the plant in accordance with the existing Technical Specifications. The proposed change only affects the designated titles of personnel, rewords or relocates requirements within TS or deletes requirements that are either not required to be part of TS or are already required per regulation. The change also relocates the detailed description of the onsite and offsite safety review organizations and non-licensed personnel qualification requirements to the Quality Assurance Program. Therefore this change does not impact plant SSCs or plant operation and therefore does not create the possibility of an accident of a different type than evaluated previously. The management organizational structure and safety and operational reviews have not changed. Therefore, there is no change in the method of plant operation, operation review or system design review. There are no new or different accident scenarios, accident initiators, nor failure mechanisms that will be introduced due to this change.

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

The proposed change only affects the designated titles of personnel, rewords or relocates requirements within TS or deletes requirements that are either not required to be part of TS or are already required per regulation. The change also relocates the detailed description of the onsite and offsite safety review organizations and non-licensed personnel qualification requirements to the Quality Assurance Program. Consequently, this change does not impact plant design, plant operation or any safety margin and, therefore, does not significantly reduce a margin of safety.

This evaluation concludes that the proposed amendments to the Surry Units 1 and 2 Technical Specifications do not involve a significant increase in the probability or consequences of a previously evaluated accident, do not create the possibility of a new or different kind of accident and do not involve a significant reduction in a margin of safety.

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

NRC Section Chief: Mary Jane Ross-Lee (Acting).

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: July 23, 2004.

Description of amendment request: The amendment would revise Technical Specification 3.6.3, “Containment Isolation Valves,” by (1) Adding the abbreviation “(CIV)” for containment isolation valve in Condition A of the Actions for the Limiting Condition for Operation; (2) deleting the Note and revising Condition A to be for only one penetration flow path with one CIV inoperable; (3) revising the completion time for Required Condition A.1 from 4 hours to as much as 7 days depending on the category of the inoperable CIV; and (4) revising Condition C to be for two or more penetration flow paths with one CIV inoperable. The proposed amendment is based on Topical Report WCAP-15791-P, “Risk-Informed Evaluation of Extensions to Containment Isolation Valve Completion Times.”

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes to the Completion Times do not change the response of the plant to any accidents and have an insignificant impact on the reliability of the containment isolation valves. The containment isolation valves will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This is demonstrated by showing that the impact on plant safety as measured by the large early release frequency (LERF) and incremental conditional large early release probabilities (ICLERP) is acceptable. These changes are consistent with the acceptance criteria in [the risk-informed] Regulatory Guides 1.174 and 1.177. Therefore, since the containment isolation valves will continue to perform their [safety] functions with high reliability as originally assumed and the increase in risk as measured by LERF and ICLERP is acceptable, there will not be a significant increase in the consequences of any accidents.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended [safety] function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes are consistent with the safety analysis assumptions and resultant consequences [in Chapter 15, “Accident Analysis,” of the Updated Final Safety Analysis Report (USAR) for the plant].

Therefore, it is concluded that this change does not increase the probability of occurrence of a malfunction of equipment important to safety.

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

The proposed changes do not result in a change in the manner in which the containment isolation valves provide plant protection. There are no design changes associated with the proposed changes. The changes to Completion Times do not change any existing accident scenarios, nor create any new or different accident scenarios.

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

Therefore, the possibility of a new or different malfunction of safety related equipment is not created.

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

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The calculated impact on risk is insignificant and is consistent with the acceptance criteria contained in Regulatory Guides 1.174 and 1.177.

Therefore, it is concluded that this change does not involve a significant reduction in the margin of safety.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Robert Gramm.

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

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

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

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

Date of application for amendments: October 29, 2004.

Brief description of amendments: Provide a one-time change to Function 4a, “Reactor Coolant System (RCS) Hot Leg Temperature Indication,” of Technical Specification Table 3.3.4-1. This would allow continued operation until the next refueling outage (spring of 2005) with one out of four RCS hot leg temperature indications inoperable in the Auxiliary Control Room.

Date of publication of individual notice in the Federal Register: November 5, 2004 (69 FR 64596).

Expiration date of individual notice: November 19, 2004.

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, 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 NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov. (Note: Public access to ADAMS has been temporarily suspended so that security reviews of publicly available documents may be performed and potentially sensitive information removed. Please check the NRC Web site for updates on the resumption of ADAMS access.)

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

Date of application for amendment: August 27, 2004, as supplemented by letters dated October 11 and 19, 2004.

Brief description of amendment: The amendment revised the Technical Specifications, Section 2.1.A, changing the safety limit minimum critical power ratio value from 1.09 to 1.10 for both four-or five-recirculation-loop operation, and from 1.10 to 1.12 for three-recirculation-loop operation.

Date of Issuance: November 16, 2004.

Effective date: November 16, 2004, and shall be implemented within 60 days of issuance.

Amendment No.: 252.

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

Date of initial notice in Federal Register: September 14, 2004 (69 FR 55467). The October 11 and 19, 2004, letters provided clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated November 16, 2004.

No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, et. al., Docket No. 50-219, Oyster Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

Date of application for amendments: March 23, 2004, as supplemented June 16, 2004.

Brief description of amendments: The amendments relocate the Independent Onsite Safety Review Group requirements from the Administrative Controls in Section 6 of the Technical Specifications to the Exelon Generation Company, LLC (EGC)/AmerGen Energy Company, LLC (AmerGen) Quality Assurance Topical Report (QATR) at TMI-1 and OCNGS. In addition, administrative corrections are included, which update references to the EGC/AmerGen QATR, which has replaced the OCNGS and TMI-1 Operational Quality Assurance Plans.

Date of issuance: November 8, 2004.

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

Amendment Nos.: 251 and 252.

Facility Operating License Nos. DPR-16 and DPR-50: Amendments revised the Technical Specifications.

Date of initial notices in Federal Register: May 11, 2004 (69 FR 26186).

The supplement dated June 16, 2004, 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 determinations.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 8, 2004.

No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

Date of application for amendment: April 23, 2004.

Brief description of amendment: The amendment deletes Technical Specification Section 6.16, “Post-Accident Sampling Programs NUREG 0737 (II.B.3, II-F.1.2),” and the related requirements to maintain a Post Accident Sampling System.

Date of issuance: November 22, 2004.

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

Amendment No.: 253.

Facility Operating License No. DPR-50. Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 31, 2004 (69 FR 26187)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 22, 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: March 12, 2004, and supplemented by letters dated June 16 and September 2, 2004.

Brief description of amendments: The amendments modify the LaSalle Technical Specifications (TS) to eliminate selected response time testing requirements associated with Reactor Protection System instrumentation and Primary Containment Isolation instrumentation for Main Steam Line Isolation functions. Specifically, the changes revise the response time testing requirements for TS Section 3.3.1.1, “Reactor Protection System (RPS) Instrumentation,” Reactor Vessel Steam Dome Pressure—High function and TS Section 3.3.6.1, “Primary Containment Isolation Instrumentation,” Reactor Vessel Water Level—Low Low Low, Level 1 and Main Steam Line Pressure—Low functions.

Date of issuance: November 19, 2004.

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

Amendment Nos.: 169, 155.

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

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19569).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 19, 2004.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of application for amendment: December 23, 2003.

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

Date of issuance: November 10, 2004.

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

Amendment No.: 219.

Facility Operating License No. DPR-20. Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: September 16, 2004 (69 FR 55844).

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

No significant hazards consideration comments received: No.

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

Date of amendment request: May 14, 2004.

Brief description of amendment: The amendment relocates the requirements of Technical Specification 3.3(1)a, “Reactor Coolant System and Other Components Subject to ASME XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance” and TS 3.4, “Reactor Coolant System Integrity Testing,” to the Updated Safety Analysis Report (USAR). Requirements in TS 3.3(1)a were related to inservice inspection of ASME Class 1, 2, and 3 components and requirements in TS 3.4 were related to reactor coolant system integrity testing.

Date of issuance: November 8, 2004.

Effective date: November 8, 2004, and shall be implemented within 120 days from the date of issuance.

Amendment No.: 230.

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

Date of initial notice in Federal Register: June 22, 2004 (69 FR 34703)

The Commission's related evaluation of the amendment is contained in a safety evaluation dated November 8, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 22, 2004, as supplemented on September 27, 2004.

Brief description of amendments: The amendments revise the frequency associated with Surveillance Requirement (SR) 3.3.8.1.4, which directs the performance of the logic system functional test, from once every 18 months to once every 24 months. The amendments change the SRs in Hatch, Units 1 and 2 Technical Specifications.

Date of issuance: November 22, 2004.

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

Amendment Nos.: 243/186.

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

Date of initial notice in Federal Register: August 3, 2004 (69 FR 46592).

The supplement dated September 27, 2004, provided clarifying information that did not change the scope of the June 22, 2004, application nor the initial proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

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

Date of amendment request: October 29, 2004, as supplemented November 5, 2004.

Description of amendment request: The amendment provides a one-time change to Function 4a, “Reactor Coolant System (RCS) Hot Leg Temperature Indication,” of Technical Specification (TS) Table 3.3.4-1 to allow continued operations until the next refueling outage with one out of four RCS Hot Leg Temperature Indications inoperable in the Auxiliary Control Room.

Date of Issuance: November 19, 2004.

Effective date: As of the date of issuance and shall be implemented immediately upon receipt.

Amendment No.: 53.

Facility Operating License No. (NPF-90): Amendment revised the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. On November 5, 2004, the Commission issued a notice (69 FR 64596) that included the staff's proposed determination that the amendment request involves no significant hazards consideration (NSHC). 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 November 19, 2004, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment. The supplement of November 5, 2004, is within the scope of that notice, and did not change the proposed no significant hazards consideration.

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 November 19, 2004.

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.

Dated at Rockville, Maryland, this 29th day of November, 2004.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

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

[FR Doc. 04-26606 Filed 12-6-04; 8:45 am]

BILLING CODE 7590-01-P