Skip to Content
Notice

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

 

Table of Contents Back to Top

Background Back to Top

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 to publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, June 25, 2004, through July 8, 2004. The last biweekly notice was published on July 6, 2004 (69 FRN 40668).

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

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

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

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

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

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

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the 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. 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.

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

Date of amendment request: May 25, 2004.

Description of amendment request: The proposed amendments would revise the licensing basis in the Updated Final Safety Analysis Report to support installation of a passive low-pressure injection (LPI) cross connect inside containment for Unit 3. The proposed changes would revise the licensing basis for selected portions of the core flood and LPI piping to allow exclusion of the dynamic effects associated with a postulated rupture of that piping by application of leak-before-break technology. Similar amendments were approved for Unit 1 by NRC letter dated September 29, 2003, and for Unit 2 by NRC letter dated February 5, 2004.

The proposed amendments would also delete technical specifications (TSs) which will no longer apply when the LPI cross connect modification has been implemented.

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) Involve a significant increase in the probability or consequences of an accident previously evaluated: The proposed License Amendment Request (LAR) modifies the Unit 3 licensing basis to allow the dynamic effects associated with postulated pipe rupture of selected portions of the Unit 3 Low Pressure Injection (LPI)/Core Flood (CF) piping to be excluded from the design basis. The proposed LAR also removes Technical Specifications that are no longer applicable due to the completion of the LPI cross connect modification on all three Oconee Units. The proposed design allowances for these selected portions of piping continue to allow the LPI system design to meet General Design Criteria (GDC) 4 requirements related to environmental and dynamic effects. The proposed LAR will continue to ensure that ONS [Oconee Nuclear Station] can meet design basis requirements associated with the LPI safety function. The addition of the crossover line will enhance the ability of the control room operator to mitigate the consequences of specific events for which LPI is credited. Therefore, the proposed LAR does not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from any kind of accident previously evaluated: The proposed LAR modifies the Unit 3 licensing basis to allow the dynamic effects associated with postulated pipe rupture of selected portions of Unit 3 LPI/CF piping to be excluded from the design basis and removes TS requirements that are no longer applicable due to the completion of the LPI cross connect modification on all three Oconee Units. The proposed design allowances for these selected portions of piping continue to allow the LPI system design to meet GDC 4 requirements related to environmental and dynamic effects. The systems affected by the changes are used to mitigate the consequences of an accident that has already occurred. The proposed licensing basis change does not affect the mitigating function of these systems. Consequently, these changes do not alter the nature of events postulated in the Safety Analysis Report nor do they introduce any unique precursor mechanisms. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Involve a significant reduction in a margin of safety: The proposed licensing basis and TS changes do not unfavorably affect any plant safety limits, set points, or design parameters. The changes also do not unfavorably affect the fuel, fuel cladding, RCS [Reactor Coolant System], or containment integrity. Therefore, the proposed changes, which add new design allowances associated with the passive LPI cross connect modification and remove obsolete TS requirements, do not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: Stephanie M. Coffin (Acting).

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

Date of amendment request: June 4, 2004.

Description of amendment request: The proposed amendment would revise the safety limit values in Technical Specification (TS) 2.1.1.2 for the minimum critical power ratio (MCPR) for both single and two recirculation loop operation.

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

1. The operation of JAFNPP in accordance with the proposed amendment, will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The basis of the Safety Limit Minimum Critical Power Ratio (SLMCPR) is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. The new SLMCPR values preserve the existing margin to transition boiling and probability of fuel damage is not increased. The derivation of the revised SLMCPR for JAFNPP for incorporation into the Technical Specifications, and its use to determine plant and cycle-specific thermal limits, have been performed using NRC approved methods. These plant-specific calculations are performed each operating cycle and if necessary, will require future changes to these values based upon revised core designs. The revised SLMCPR values do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient.

Based on the above, JAFNPP has concluded that the proposed change will not result in a significant increase in the probability or consequences of an accident previously evaluated.

2. The operation of JAFNPP in accordance with the proposed amendment, will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes result only from a specific analysis for the JAFNPP core reload design. These changes do not involve any new or different methods for operating the facility. No new initiating events or transients result from these changes.

Based on the above, JAFNPP has concluded that the proposed change will not create the possibility of a new or different kind of accident from those previously evaluated.

3. The operation of JAFNPP in accordance with the proposed amendment, will not involve a significant reduction in a margin of safety.

The new SLMCPR is calculated using NRC approved methods with plant and cycle specific parameters for the current core design. The SLMCPR value remains high enough to ensure that greater than 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. The operating MCPR limit is set appropriately above the safety limit value to ensure adequate margin when the cycle specific transients are evaluated. Accordingly, the margin of safety is maintained with the revised values.

As a result, JAFNPP has determined that the proposed change will not result in a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New York, New York 10019.

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: June 17, 2004.

Description of amendment request: The amendment will (1) modify Technical Specifications (TSs) 5.3.1, Fuel Assemblies, to allow a limited number of lead test assemblies (LTAs) and limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, (2) include ZIRLO TM as an acceptable fuel rod cladding which is consistent with 10 CFR 50.46, (3) relocate some of the information in TS 5.3.1 to TS 5.6.1, (4) change TS 6.9.1.11.1 to allow the use of the Westinghouse Nuclear Physics code package and to incorporate the methodology used to support ZIRLO TM cladding material, and (5) delete the Index from the TSs.

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

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

Response: No.

TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality

The proposed change allows the use of a limited number of lead test assemblies; the use of limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods; and the use of methods required for the implementation of ZIRLO TM clad fuel rods. Inasmuch as the revision identifies codes previously approved by the NRC [Nuclear Regulatory Commission] for CE [Combustion Engineering] cores, the amendment is administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident.

The proposed change in part represents a relocation of a portion of the information previously located in the TSs design features section to the FSAR [Final Safety Analysis Report], which is controlled under 10 CFR 50.59, “Changes, Tests, and Experiments.” This change is administrative in nature because the design requirements for the facility remain the same.

The proposed change does not remove or modify any of the design requirements for the facility or affect any accident initiators, conditions or assumption[s] for an accident previously evaluated.

TS 6.9.1.11, Core Operating Limits Report COLR

The proposed amendment identifies a change in the nuclear physics codes used to confirm the values of selected cycle-specific reactor physics parameter limits and includes minor editorial changes which do not alter the intent of stated requirements. The proposed change also allows the use of methods required for the implementation of ZIRLO TM clad fuel rods. Inasmuch as the proposed change identifies codes previously approved by the NRC for CE cores, the amendment is administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident. Parameter limits specified in the site specific COLR are not changed from the values presently required by TSs. Future changes to the calculated values of such limits may only be made using NRC approved methodologies, must be consistent with all applicable safety analysis limits, and are controlled by the 10 CFR 50.59 process. Assumptions used for accident initiators and/or safety analysis acceptance criteria are not changed by this change.

Index

The proposed change is administrative in nature and does not affect any system or component functional requirements. This change does not affect the operation of the plant or affect any component that is used to mitigate the consequences of any accident.

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

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

Response: No.

TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality

The proposed change allows the use of methods required for the implementation of ZIRLO TM clad fuel rods. Inasmuch as the revision identifies codes previously approved by the NRC for CE cores, the amendment is administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident.

In addition, the proposed change allows the use of a limited number of lead test assemblies. The proposed change is administrative in nature. Prior to the use of lead test assemblies, fuel designs will be analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases to assure no new or different kind of accident from any accident previously evaluated will be created.

And finally the proposed change allows the relocation of a portion of the information previously located in the TSs design features section to the FSAR. This change is administrative in nature and does not create a new or different type of accident than previously evaluated because the design requirements for the facility remain the same.

The proposed change does not remove or modify any of the design requirements for the facility or affect any accident initiators, conditions or assumption[s] for an accident previously evaluated.

TS 6.9.1.11, Core Operating Limits Report COLR

The proposed change identifies a change in the Nuclear Physics codes used to confirm the values of selected cycle-specific reactor physics parameter limits contained in the COLR. The proposed change also allows the use of methodologies required for the implementation of ZIRLO TM clad fuel rods. Neither of these changes results in a change [to] the physical plant or the modes of operation defined in the facility license.

Index

The proposed change is administrative in nature and does not affect any system or component functional requirements. This change does not affect the operation of the plant or affect any component that is used to mitigate the consequences of any accident.

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

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

Response: No.

TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality

The proposed change allows the use of methods required for the implementation of ZIRLO TM clad fuel rods. Inasmuch as the revision identifies codes previously approved by the NRC for CE cores, the amendment is administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident.

In addition, the proposed change allows the use of a limited number of lead test assemblies. The proposed change is administrative in nature. Prior to the use of lead test assemblies, fuel designs will be analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to ensure compliance with any safety analysis acceptance criteria.

And finally the proposed change allows the relocation of a portion of the information previously located in the TSs design features section to the FSAR. This change is administrative in nature and does not create a new or different type of accident than previously evaluated because the design requirements for the facility remain the same.

The proposed change does not remove or modify any of the design requirements for the facility or affect any accident initiators, conditions or assumption[s] for an accident previously evaluated.

TS 6.9.1.11, Core Operating Limits Report COLR

The individual specifications continue to require operation of the plant within the bounds of the limits specified in COLR. Benchmarking has shown that uncertainties for the Westinghouse Physics code system (ANC/PHOENIX-P) yields are essentially the same or less than those obtained for the current ROCS/DIT methodology. Future changes to the values of these limits by the licensee may only be developed using NRC approved methodologies, remaining consistent with all applicable plant safety analysis limits addressed in the Safety Analysis Report, which are controlled by the 10 CFR 50.59 process. The relocation of the supplement numbers, revision numbers, and approval dates related to the analytical methods listed in the COLR does not affect the margin of safety. The analysis will continue to be performed using NRC approved methodology. Safety analysis acceptance criteria are not being altered by this change.

Index

The proposed change is administrative in nature and does not affect any system or component functional requirements. Safety analysis acceptance criteria are not being altered by this change.

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

NRC Section Chief: Robert A. Gramm.

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

Date of application for amendments: June 15, 2004.

Description of amendment request: The proposed amendment would allow the licensee to conduct the monthly diesel surveillance test, the diesel full-load rejection test, the diesel 24-hour run test and the diesel hot restart test at the higher load of 2800 kW.

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

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

Response: No.

The proposed revisions to Technical Specification [TS] Surveillance Requirements SR 3.8.1.3 (the monthly diesel surveillance test), SR 3.8.1.10 (the diesel full-load rejection test), SR 3.8.1.14.b (the diesel 24-hour run test), and SR 3.8.1.15 (the diesel hot restart test) to permit these tests to be conducted at the higher load value of 2800 kW do not involve any physical change to any EDG [emergency diesel generator] equipment. The Operator using existing EDG load controls will adjust the EDG to carry the increased load during surveillance testing.

The EDGs are designed to provide a reliable source of AC electrical power in the event of an accident coincident with a loss of offsite power. The failure of an EDG itself is not considered an accident evaluated in the UFSAR [Updated Final Safety Analysis Report]. This proposed loading change does not affect the current accident initiators or precursors that could lead to a previously evaluated accident.

The failure of a single EDG to perform when required to mitigate the consequences of an accident has already been considered as a subsequent single failure in the current plant safety analyses. The proposed change to increase the allowable load range does not alter the EDG design features, post-accident operation, or accident analysis assumptions which could affect the ability of the EDGs to mitigate the consequences of a previously evaluated accident. Current EDG testing requirements, e.g., starting, timing, and post accident sequencing and loading will continue to ensure reliable EDG operation and are not being changed in this request.

Since the EDG TS surveillance test load is the only parameter involved in this request, the proposed changes will not increase the likelihood of the malfunction of another system, structure, or component that has been assumed as an accident initiator or credited in the mitigation of an accident.

Based on the above discussion, the proposed TS 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 EDGs are designed to provide a reliable source of AC electrical power in the event of an accident coincident with a loss of offsite power. No change in the ability of the EDGs to perform their design function is involved. Instrumentation setpoints, starting, sequencing, and post-accident loading functions associated with the EDGs are not affected by the proposed changes. No modifications to the EDGs are required to implement the proposed TS changes. Therefore, no new failure mechanism, malfunction, or accident initiator is considered credible.

Additionally, the proposed TS changes do not affect the other plant design, hardware, system operation, or procedures. Therefore, based on the above discussion, the above TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The underlying purpose of the four (4) diesel generators is to ensure an available source of onsite power to the ESF [engineered safety feature] systems. This change does [sic] will not impact this underlying purpose. As discussed above, this change may result in a slight increase in engine wear due to the ability to operate at the higher load, but this increased wear is bounded by the existing 24 month maintenance inspection program. The OEM [original equipment manufacturer] has stated that the change to increase the allowable load value still remains well within the EDG 2000-hour rating, and the increased rate of wear is within the acceptable limits of the current maintenance program.

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

The NRC staff has found that, because the EDGs will continue to be operated within the bounds of the current maintenance program, there is no significant increase in the probability of an EDG failure; therefore, there is no significant increase in the probability or consequences of an accident previously evaluated. The NRC staff further finds that, because there is no significant increase in a failure of an EDG to perform its function, the proposed change does not create the possibility of an accident not previously evaluated.

The NRC staff has reviewed the licensee's analysis and, based on this review and the staff's own findings above, 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 and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: James W. Clifford.

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

Date of amendment request: June 2, 2004.

Description of amendment request: The proposed amendment would revise the BVPS-1 and 2 Technical Specifications to allow operation with atmospheric containment designs.

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

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

Response: No. The Beaver Valley Power Station (BVPS) containments are designed to withstand the internal pressure and temperature resulting from a loss of coolant accident (LOCA), main steamline break (MSLB), feedwater line break, and a control rod ejection accident (CREA). Each of these accidents has been previously analyzed with the results provided in the Updated Final Safety Analysis Report (UFSAR) except the feedwater line break. This accident is not analyzed because the MSLB is more limiting. The affect on containment pressure and temperature due to a CREA is bounded by a LOCA, since a CREA is modeled after a small break LOCA. The probability of occurrence for these accidents is independent of the type of containment. Additionally the supporting plant modifications will not increase the probability of an accident because they perform an accident mitigation function and are not accident initiators. Therefore a change from sub-atmospheric to an atmospheric containment will not increase the probability of these accidents.

For accident conditions, the proposed changes will potentially impact the reported dose consequences of the LOCA and CREA for both BVPS units. The radiological consequences of these and the remaining design basis accidents are not adversely impacted by the proposed changes because they are within the current BVPS licensing and design basis.

From a containment integrity viewpoint, the limiting DBA [design-basis accident] presently is the MSLB for Unit 1 and the LOCA for Unit 2. Following the conversion to an atmospheric containment the limiting DBA will be the LOCA for both units. The revised containment integrity analysis demonstrates that with the installation of the supporting plant modifications that the pressures and temperatures associated with the applicable design basis accidents identified above are within the existing containment design limits.

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 design basis accidents, which could be adversely affected by the proposed changes, have been reanalyzed. These [re]analyses demonstrate that all acceptance criteria have been satisfied. The revised containment integrity analysis demonstrates that the containment will not be subjected to temperatures or pressures that are beyond its design limits. Converting to an atmospheric containment will not result in any new or different kind of accidents because no new accident initiators will be introduced.

The affects of the supporting plant modifications and the proposed Technical Specification changes on plant structures, systems and components (SSC) have been evaluated and it has been verified that the capability of the SSCs to perform their design functions will be retained following approval of the proposed Technical Specification changes and installation of the supporting plant modifications.

Changes to instrumentation setpoints, surveillance requirements, installation of the supporting plant modifications, and the elimination of certain operability requirements will not create the possibility of a new or different type of accident since these changes would not result in significant changes to the manner in which the affected equipment is operated during normal plant operations.

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 margin of safety attributed to the containment involves both the pressures and temperatures the containment is subjected to following a DBA, and the on-site and offsite dose consequences associated with normal and post DBA operations.

The revised containment analyses demonstrates that, following a DBA; containment peak pressure and temperature will not exceed the containment's design limits and that the containment pressure will not decrease to below 8 psia following the intentional or inadvertent actuation of the quench spray system. Since the containment design limits are not exceeded, the existing margin of safety between these limits and the containment failure limits is not reduced.

Since the current radiological analyses impacted by the containment conversion are conservatively based on atmospheric operation, it is concluded that the existing dose consequence margin of safety will not be impacted when the BVPS units are operated with an atmospheric containment.

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

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

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

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: April 26, 2004.

Description of amendment request: The proposed amendments would revise Technical Specifications Limiting Conditions for Operation (LCO) 3.7.9, “Ultimate Heat Sink (UHS)” to allow the UHS to remain OPERABLE with three of four fans operating under certain environmental conditions.

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

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

No. The revised requirements will maintain OPERABILITY while allowing maintenance on one fan when ambient wet-bulb temperature is 63 °F or lower. Modifying the condition when one NSCW [nuclear service cooling water] tower is impacted is more restrictive. The UHS is not an initiator to any analyzed accident sequence. Operation in accordance with the proposed TS will continue to ensure that the UHS remains capable of performing its safety function and that all analyzed accidents will continue to be mitigated as previously analyzed. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No. The proposed changes do not introduce any new equipment, create new failure modes for existing equipment, or create any new limiting single failures. Plant operation will not be altered, and all safety functions previously addressed in accident analyses will continue to be performed. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

No. The proposed changes will not adversely affect operation of plant equipment-principally the UHS and the equipment supported by it. Modifying the condition where one NSCW tower is impacted is more restrictive and enhances the margin of safety. Therefore, the proposed changes do not involve a significant reduction in any margin of safety.

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

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216.

NRC Section Chief: Stephanie M. Coffin (Acting).

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: June 21, 2004.

Description of amendment request: The proposed one-time (per unit) change revises the steam generator (SG) inservice inspection frequency requirements in Technical Specification (TS) 4.4.5.3a for Unit 1 immediately after the tenth refueling outage for Unit 1 (1RE10) and for Unit 2 immediately after refueling outage 2RE10. The change would allow a 78-month inspection interval after one inspection resulting in C-1 classification, rather than a 40-month interval after two consecutive inspections resulting in C-1 classification.

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

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

Response: No.

There is no direct increase in SG leakage because the proposed change does not alter the plant design. The scope of inspections performed during 1RE10 and 2RE10, the first refueling outage following SG replacement, exceeded the combined TS requirements for the first two refueling outages after replacement. That is, more tubes were inspected than were required by TS. Currently, neither Unit 1 nor Unit 2 has an active SG damage mechanism and will meet the current industry examination guidelines without performing inspections during the next 78 months. The Condition Monitoring Assessment after 1RE10 and 2RE10 demonstrated that all performance criteria were met during these outages. The Operational Assessment shows that all performance criteria will be met over the proposed operating period.

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 not alter any plant design basis or postulated accident resulting from potential SG tube degradation. The scope of inspections performed during 1RE10 and 2RE10, the first refueling outage for each unit following SG replacement, significantly exceeded the combined TS requirements for the scope of the first two refueling outages after SG replacement. The inspections already performed exceed those required by the current TS over the proposed 78-month period.

The proposed change does not affect the design of the SGs, the method of operation, or reactor coolant chemistry controls. No new equipment is being introduced and installed and equipment is not being operated in a new or different manner. The proposed change involves a one-time extension of the SG tube inservice inspection interval, and therefore will not give rise to new failure modes. In addition, the proposed change does not impact any other plant system or components.

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.

Steam generator tube integrity is a function of design, environment, and current physical condition. Extending the SG tube inservice inspection interval to 78 months will not alter the function or design of the SGs. Inspections conducted prior to placing the SGs into service (pre-service inspections) and inspection during the first refueling outages following SG replacement demonstrate that the SGs do not have fabrication damage or an active damage mechanism. The scope of those inspections significantly exceeded those required by the TS. These inspection results were comparable to similar inspection results for the same model of RSGs [replacement steam generators] installed at other plants, and subsequent inspections at those plants yielded results that support this extension request. The improved design of the RSGs also provides reasonable assurance that significant tube degradation is not likely to occur over the proposed operating period.

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

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

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.

NRC Section Chief: Robert A. Gramm.

Notice of Issuance of Amendments to Facility Operating Licenses Back to Top

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.

Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power Station, Unit No. 1, New London County, Connecticut

Date of amendment request: September 18, 2003.

Brief description of amendment: The amendment revises Technical Specification 4.2, “Fuel Storage,” to eliminate all credit for Boraflex as a neutron absorber, reduce the number of fuel assemblies allowed to be stored in the spent fuel pool (SFP), change the required SFPk eff and eliminate design features requirements of new fuel storage.

Date of issuance: June 29, 2004.

Effective date: June 29, 2004, and shall be implemented within 60 days from the date of issuance.

Amendment No.: 113.

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

Date of initial notice in Federal Register: December 9, 2003 (68 FR 68659). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 29, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 15, 2003.

Brief description of amendments: The amendments added a new Technical Specification (TS) 3.9.7, “Unborated Water Source isolation Valves,” and revised TS 3.9.2, “Nuclear Instrumentation,” to delete the requirement for Boron Dilution Mitigation System automatic valve actuations and makeup water pump trip during Mode 6 and to agree with the wording of NUREG-1431, “Standard Technical Specifications Westinghouse Plants,” Revision 2. The licensee proposed these changes to provide configuration control of the dilution valves during Mode 6 to preclude the possibility of a boron dilution event and to provide an opportunity to conduct maintenance on the volume control tank valves, refueling water storage tank valves, and their respective power supplies.

Date of issuance: June 21, 2004.

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

Amendment Nos.: 215 and 209.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the TSs.

Date of initial notice in Federal Register: March 16, 2004 (69 FR 12366).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 21, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 25, 2003.

Brief description of amendments: The amendments are administrative in nature and incorporate several editorial changes.

Date of issuance: June 21, 2004.

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

Amendment Nos.: 222 and 204.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19565).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 21, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 3, 2004.

Brief description of amendments: The amendments revised the Technical Specifications administrative controls requirements regarding the reactor coolant pump flywheel inspection program to increase the inspection interval from 10 years to 20 years.

Date of issuance: July 2, 2004.

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

Amendment Nos.: 240 and 221.

Facility Operating License Nos. DPR-26 and DPR-64: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19566).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 2, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 27, 2003, as supplemented December 15, 2003, and February 27, 2004.

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

Date of issuance: June 25, 2004.

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

Amendment Nos.: 281 and 265.

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

Date of initial notice in Federal Register: October 14, 2003 (68 FR 59217).

The supplemental letters dated December 15, 2003, and February 27, 2004, provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 25, 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: June 3, 2003, as supplemented by letters dated October 6, 2003, January 15, and February 13, 2004.

Brief description of amendment: The amendment revises the operating license and technical specifications to increase the licensed rated power by 1.4 percent from 2530 megawatts thermal (MWt) to 2565.4 MWt using measurement uncertainty recapture.

Date of issuance: June 23, 2004.

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

Amendment No.: 215.

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

Date of initial notice in Federal Register: July 8, 2003 (68 FR 40714).

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

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 23, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 20, 2004.

Brief description of amendments: The amendments revised the Technical Specification requirements for Shift Technical Advisor coverage.

Date of issuance: June 28, 2004.

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

Amendment Nos.: 132 and 111.

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

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19574).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 28, 2004.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 6, 2003, as supplemented by letter dated February 24, 2004.

Brief description of amendments: The amendments revise the Technical Specifications (TSs) adopting the TS Task Force (TSTF) Standard TS Change Traveler TSTF-360, Revision 1, “DC Electrical Rewrite.” Specifically, the amendments revise the TS 3.8.4, “DC Sources-Operating,” TS 3.8.5, “DC Sources-Shutdown,” TS 3.8.6, “Battery Cell Parameters,” and TS 5.5.19, “Battery Monitoring and Maintenance Program.”

Date of issuance: July 1, 2004.

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

Amendment Nos.: 113 and 113.

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

Date of initial notice in Federal Register: July 8, 2003 (68 FR 40721). The February 24, 2004, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 1, 2004.

No significant hazards consideration comments received: No.

Dated in Rockville, Maryland, this 12th day of July 2004.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

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

[FR Doc. 04-16157 Filed 7-19-04; 8:45 am]

BILLING CODE 7590-01-P

Site Feedback