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Notice

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

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Information about this document as published in the Federal Register.

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I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, February 7, 2003, through February 20, 2003. The last biweekly notice was published on February 18, 2003 (68 FR 7810).

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.

Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing 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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By April 3, 2003, 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 2.714,[1] 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 by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, 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 Start Printed Page 10278following factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide 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 to relief. A petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-examine witnesses.

If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to hearingdocket@nrc.gov. 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 because of continuing disruptions in delivery of mail to United States Government offices, 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 filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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.

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

Date of amendment request: January 14, 2003.

Description of amendment request: The proposed amendment would revise the TMI-1 Technical Specification Sections 3.8.9, 3.15.2, and 4.12.2, and the associated Bases to delete the requirements for the Reactor Building Purge Air Treatment System.

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.

This change will delete the existing Technical Specifications 3.15.2 and 4.12.2 and revise Technical Specification 3.8.9. The proposed change does not impact nor change the physical configuration of any system, structure or component, nor does it change the manner in which any system is operated. Any change to the system design will be evaluated in accordance with the requirements of 10 CFR 50.59. Failure of the system will neither initiate any type of accident nor increase the severity of the consequences of an accident previously evaluated. Previously approved analyses of the dose consequences of the accidents described in the TMI Unit 1 UFSAR [Updated Final Safety Analysis Report] are not affected by the proposed change and dose consequences remain below the limits of 10 CFR 50.67 without the operation of the Reactor Building Purge Air Treatment System fan and filter components. The Reactor Building Purge Air Treatment System fan and filter components are not required for mitigation of any accident as described in the TMI Unit 1 UFSAR. Reactor Building purge operations will continue to be conducted in accordance with the existing plant administrative controls, which will ensure the limits of 10 CFR part 50 Appendix I are met.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Start Printed Page 10279

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

Response: No.

This activity will delete sections of the Technical Specifications applicable to the Reactor Building Purge Air Treatment System fan and filter components. The proposed change does not physically alter any system, structure or component. Any change to the system design will be evaluated in accordance with 10 CFR 50.59. The proposed change will not cause the Reactor Building Purge Air Treatment System to operate outside of its existing design basis. There will be no impact to any operational feature of the system or any procedures that control its operation that could result in a new or different failure mode. The design basis of the Reactor Building Purge Air Treatment System as currently described in the TMI Unit 1 UFSAR is not revised.

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 deletion of Technical Specification Sections 3.15.2 and 4.12.2 and the revision of Technical Specification 3.8.9 will not impact the operation of the Reactor Building Purge Air Treatment System. The proposed change will not cause the system to be placed in a configuration outside of its design basis. The proposed change will not reduce the margin of safety of any safety related system. Reactor Building purge operations will continue to be conducted in accordance with existing plant administrative controls, which will ensure the limits of 10 CFR part 50 appendix I are met. The system will continue to be operable in accordance with applicable plant operating procedures.

The system will also continue to be tested and maintained under periodic operations surveillance and the TMI Unit 1 Preventive Maintenance Program.

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: Edward J. Cullen, Jr., Esquire, Vice President, General Counsel and Secretary, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: Richard J. Laufer.

Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear Power Station, Unit 2, Grundy County, Illinois

Date of amendment request: January 31, 2003.

Description of amendment request: The proposed amendments would revise the safety limit minimum critical power ratio for Unit 2 for two loop operation and for single 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. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established consistent with NRC [Nuclear Regulatory Commission] approved methods to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change conservatively establishes the safety limit for the minimum critical power ratio (SLMCPR) for Dresden Nuclear Power Station (DNPS), Unit 2, Cycle 18 such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences (AOOs).

Changing the SLMCPR does not increase the probability of an evaluated accident. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected.

The proposed change revises the SLMCPR to protect the fuel during normal operation as well as during any transients or anticipated operational occurrences. Operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences) is met. Since the operability of plant systems designed to mitigate any consequences of accidents has not changed, the consequences of an accident previously evaluated are not expected to increase.

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?

Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation. The proposed change does not involve any modifications of the plant configuration or allowable modes of operation. The proposed change to the SLMCPR assures that safety criteria are maintained for DNPS, Unit 2, Cycle 18. 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?

The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1% of the rods are expected to be in boiling transition if the MCPR limit is not violated. The proposed change will ensure the appropriate level of fuel protection. Additionally, operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as AOOs) are met.

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

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

Attorney for licensee: Mr. Edward J. Cullen, Deputy General Counsel, Exelon BSC—Legal, 2301 Market Street, Philadelphia, PA 19101.

NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: December 20, 2002.

Description of amendment request: The proposed amendment would make an administrative change to Technical Specification (TS) Sections 6.7, 6.14, and 6.15 by replacing “Station Review Board” to “Plant Operations Review Committee” to be consistent with the name for this type of onsite review committee that is used at other FirstEnergy Nuclear Operating Company plants. Additionally, the proposed amendment would make an administrative change to TS 6.8 to update the version of Regulatory Guide 1.33 referenced in that Section.

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

Start Printed Page 10280

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

Response: No.

The administrative changes do not affect any existing limits, and accident initial conditions, probability, and assumptions remain as previously analyzed. The proposed change to the name of the onsite review committee or the version of the Regulatory Guide will have no significant effect on accident initiation frequency. The proposed changes do not invalidate the assumptions used in evaluating the radiological consequences of any accident. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes are administrative and do not introduce any new or different accident initiators. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed changes are administrative and will not have a significant effect on any margin of safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, FENOC 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.

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

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

NRC Section Chief: Anthony J. Mendiola.

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 amendment requests: January 14, 2003.

Description of amendment requests: The proposed amendments would revise the Technical Specifications (TSs) for the control room emergency ventilation system (CREVS) such that movement of irradiated fuel assemblies will be allowed to commence with one CREVS pressurization train inoperable, provided the appropriate TS Action requirements are 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. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No

Probability of Occurrence of an Accident Previously Evaluated

[Cook Nuclear Plant] CNP TS 3.0.4 requires that TS limiting conditions for operation be met without reliance on the Action statements prior to entering an Applicability condition. The proposed change to the CNP CREVS TS to allow an exception to TS 3.0.4 during movement of irradiated fuel assemblies does not affect any accident initiators or precursors. The CREVS function is purely mitigative. There is no design basis accident that is initiated by a failure of the CREVS function. An exception to TS 3.0.4 will not create any adverse interactions with other systems that could result in initiation of a design basis accident. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased.

Consequences of an Accident Previously Evaluated

The accident consequence that is relevant to the proposed change is the dose to control room personnel from a fuel handling accident. The CNP licensing basis analysis of a fuel handling accident has determined that the dose would be within the applicable limits of GDC 19. The current TS specify actions to be taken if one CREVS pressurization train is inoperable during movement of irradiated fuel assemblies. These actions provide assurance that the CREVS will perform its mitigating function as assumed in the accident analysis. Since the proposed change will continue to require these actions, the fuel handling accident analysis will remain valid. Therefore, the consequences of an accident previously analyzed are not significantly increased.

In summary, the probability of occurrence and the consequences of an accident previously evaluated are not significantly increased.

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 create any new or different accident initiators or precursors. The option to commence movement of irradiated fuel assemblies while relying on the provisions of the Action statement does not affect the manner in which any accident begins. The proposed change does not create any new accident scenarios and does not change the interaction between the CREVS and any other system. Thus, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No

The margin of safety associated with the proposed change is that associated with the applicable control room dose limit specified by GDC 19. The proposed change will continue to require actions that assure the dose to control room personnel determined by the fuel handling accident analysis remains valid. Therefore, the proposed change does not involve a significant reduction in margin of safety.

In summary, based upon the above evaluation, [Indiana Michigan Power] I&M has concluded that the proposed change involves 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.

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

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: L. Raghavan.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: January 14, 2003.

Description of amendment request: The proposed one-time change revises the steam generator inservice inspection frequency requirements in Technical Specification 4.4.5.3.a for V.C. Summer Nuclear Station (VCSNS) immediately after refueling outage RF-12. The change would allow a 58-month maximum inspection interval after two inspections resulting in C-1 classification, rather than a 40-month maximum inspection interval. This change is proposed to eliminate premature/unnecessary steam generator inspections, due to a shortened operating cycle, which will result in significant dose and schedule impacts.

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 Start Printed Page 10281consequences of an accident previously evaluated?

Response: No.

The proposed one-time extension of the Technical Specification inspection interval does not involve changing any structure, system or component or affect plant operations. It is not an initiator of any accident and does not change any FSAR [Final Safety Analysis Report] safety analyses. As such, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

Probability of an Accident

The VCSNS Steam Generator Management Program includes provisions that are more rigorous than existing Technical Specification requirements. The topics addressed by the program include:

  • Steam generator performance criteria, including a reduced operational leakage limit.
  • Steam generator repair criteria and repair methods.
  • Steam generator inspections that include Degradation Assessments, Condition Monitoring Assessments, and Operational Assessments.
  • NDE [nondestructive examination] technique requirements.

The results of the above program requirements demonstrated that all performance requirements were met during Refuel 12.

Consequences of an Accident

The consequences of design basis accidents are, in part, functions of the specific activity in the primary coolant and the primary to secondary leakage rates resulting from an accident. Therefore, limits are included in the Technical Specifications for operational leakage and for specific activity in the reactor coolant to ensure the plant is operated in its analyzed condition.

The VCSNS program requires a 150-gallon per day per steam generator limit for leakage prior to an accident. This limit is a reduction in the current Technical Specification value. The post accident leak rate remains at the same value assumed by the accident analysis (1 gallon per minute). Since the new operational leakage limit is more conservative than the existing value, it will not increase the likelihood or consequences of an accident.

In consideration of the above, past 100% eddy current results after 5.4 EFPY [effective full-power years] of operation, and the current leak free condition of the steam generators, extending the tube inspection frequency does not involve a significant increase in the consequences of a previously evaluated accident.

Summary

The proposed change does not affect the design of the steam generators, their method of operation, or primary coolant chemistry controls. The change does not adversely impact any other previously evaluated design basis accident.

Therefore, the change does not affect the consequences of a SGTR [steam generator tube rupture] or any other accident.

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 one-time extension of the Technical Specification inspection interval does not involve changing any structure, system or component or affect plant operations. It is not an initiator of any accident and does not change any FSAR safety analyses.

Primary to secondary leakage that may be experienced during plant conditions is expected to remain within current accident analysis assumptions.

The proposed change does not affect the design of the steam generators, their method of operation, or primary coolant chemistry controls. In addition, the change does not impact any other plant system or component.

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

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

Response: No.

The steam generator tubes are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system pressure and inventory. As part of the RCS [reactor coolant system] boundary, the tubes are unique in that they are also relied upon as a heat transfer medium between the primary and secondary systems such that heat may be removed from the primary system. Additionally, the steam generator tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of the steam generator is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. Extending the tube inspection frequency will not alter the design function of the steam generators. Previous inspections conducted during Refuel 12 demonstrate that there is no active tube damage mechanism. The improved design of the Model Delta 75 generator also provides reasonable assurance that leakage is not likely to occur over the next operating period.

For the above reasons, the margin of safety is unchanged and overall plant safety will be maintained by the proposed Technical Specification revision.

Pursuant to 10 CFR 50.91, the preceding analyses provide a determination that the proposed Technical Specification change poses no significant hazard as delineated by 10 CFR 50.92.

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 G. Eppink, South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Section Chief: John A. Nakoski.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: January 14, 2003.

Description of amendment request: The proposed change will exclude the Charging/Safety Injection (SI) pumps and the Residual Heat Removal pumps from the requirement to vent emergency core cooling system pump casings located in Technical Specification (TS) Section 4.5.2.b.2, eliminate the 31-day venting surveillance for the SI pumps, and add discussion for this exclusion in the Technical Basis of TS Section B 3/4.5.2.

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

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

Response: No.

The proposed changes to Technical Specification 4.5.2.b.2 and its associated bases do not contribute to the initiation of any accident previously evaluated. Supporting factors are as follows:

  • The safety function of the Charging/SI system, which is related to accident mitigation, has not been altered. Therefore, the probability of an accident is not increased by the exclusion of the Charging/SI system discharge venting requirements.
  • The exclusion of the Charging/SI system venting requirements does not affect the integrity of the Charging/SI system such that its function in the control of radiological consequences is affected. In addition, the exclusion of the Charging/SI system venting requirements does not alter any fission product barrier. The exclusion of the Charging/SI system venting requirements does not change, degrade, or prevent the response of the Charging/SI system to accident scenarios, as described in FSAR [Final Safety Analysis Report] Chapter 15. In addition, the exclusion of the Charging/SI system venting requirements does not alter any assumptions previously made in the radiological consequence evaluations nor affect the mitigation of the radiological consequences of an accident described in the FSAR. Therefore, the consequences of an accident previously evaluated in the FSAR will not be increased.
  • The clarification of the RHR [residual heat removal] pump piping venting does not affect the integrity of the RHR system such that its function in the control of radiological consequences is affected. In addition, the Start Printed Page 10282clarification does not alter any of the fission product barriers. The clarification does not change, degrade, or prevent the response of the RHR system to accident scenarios, as described in FSAR Chapter 15. In addition, the clarification to the RHR pump piping venting does not alter any assumption previously made in the radiological consequences evaluations nor affect the mitigation of the radiological consequences of an accident described in the FSAR. Therefore, the consequences of an accident previously evaluated in the FSAR will not be increased.

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

Response: No.

The proposed changes to Technical Specification 4.5.3.b.2 and its associated bases do not introduce any new accident initiator mechanisms. The clarification of the RHR pump piping venting and the exclusion of the Charging/SI system venting requirements does not cause the initiation of any accident nor create any new credible limiting single failure. The exclusion of the Charging/SI system venting requirements does not result in any event previously deemed incredible being made credible. As such, it does not create the possibility of an accident different than any evaluated in the FSAR.

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

Response: No.

The exclusion of the Charging/SI system venting requirements does not result in a condition where the design, material, and construction standards that were acceptable prior to this change of the Charging/SI or RHR system venting requirements are altered. The proposed changes to Technical Specification 4.5.2.b.2 and its associated bases will have no affect on the availability, operability, or performance of the Charging/SI or RHR systems. Therefore, the clarification of the RHR pump piping venting and the exclusion of the Charging/SI system venting requirements will not reduce the margin of safety, as described in the bases to any technical specification.

Pursuant to 10 CFR 50.91, the preceding analyses provide a determination that the proposed Technical Specifications change poses no significant hazard as delineated by 10 CFR 50.92.

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 G. Eppink, South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: July 25, 2002 as supplemented by letter dated February 5, 2003.

Brief description of amendments: The proposed amendments would change the CPSES Facility Operating Licenses as follows: Section 2.C.(4)(b) would be changed to be consistent with the license conditions stated in the U.S. Nuclear Regulatory Commission (NRC) Order and Safety Evaluation issued December 21, 2001, which approved the direct transfer of ownership interest and operating authority for CPSES to TXU Generation Company LP; Section 2.E, which requires reporting any violations of the requirements contained in Section 2.C of the licenses, would be deleted. Additionally, Technical Specification Table 5.5-2 “Steam Generator Tube Inspection,” Table 5.5-3, “Steam Generator Repaired Tube Inspection for Unit 1 Only,” and Section 5.6.10, “Steam Generator Tube Inspection Report,” would be revised to delete the requirement to notify the NRC pursuant to Section 50.72(b)(2), “Immediate notification requirements for operating nuclear power reactors,” of Title 10 of the Code of Federal Regulations (10 CFR) if the steam generator tube inspection results are in a C-3 classification. The basis for the proposed no significant hazards consideration determination associated with the application was published in the Federal Register on September 3, 2002 (67 FR 56329).

By letter dated February 5, 2003, TXU Generation Company, LP requested that the proposed change to license conditions in Section 2.C.(4)(b) be superseded by the proposed deletion of the license conditions, related to Decommissioning Trusts, specified in Sections 2.C.(4)(a), 2.C.(4)(b), 2.C.(4)(d), 2.C.(4)(e), and 2.C.(6).

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), “Notice for public comment; State consultation,” the licensee has provided its analysis of the issue of no significant hazards consideration, as they relate to the February 5, 2003 supplement, which is presented below:

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

Response: No.

The requested changes delete certain license conditions pertaining to Decommissioning Trust Agreements currently in Sections 2.C.(4)(a), 2.C.(4)(b), 2.C.(4)(d), 2.C.(4)(e), and 2.C.(6) of the CPSES Facility Operating Licenses (NPF-87 and NPF-89). The requested changes are consistent with the NRC's Final Rule for Decommissioning Trust Provisions as published in the Federal Register on December 24, 2002 (67 FR 78332).

The revised regulations of 10 CFR 50.75(h)(4)[, “Reporting and recordkeeping for decommissioning planning,”] state “Unless otherwise determined by the Commission with regard to a specific application, the Commission has determined that any amendment to the license of a utilization facility that does no more than delete specific license conditions relating to the terms and conditions of decommissioning trust agreements involves “no significant hazard[s] consideration'.”

This request involves administrative changes only. 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. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

This request involves administrative changes only to be consistent with the NRC's Final Rule for Decommissioning Trust Provisions as published in the Federal Register (67 FR 78332).

No actual plant equipment or accident analyses will be affected by the proposed change 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. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

This request involves administrative changes only to be consistent with the NRC's Final Rule for Decommissioning Trust Provisions as published in the Federal Register (67 FR 78332).

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) to limit the level of radiation dose to the public.

No actual plant equipment or accident analyses will be affected by the proposed change. Additionally, the proposed changes will not relax any criteria used to establish safety limits, will not relax any safety systems 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.

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), “Issuance of amendment,” are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Start Printed Page 10283

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, 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 email to pdr@nrc.gov.

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

Date of application for amendments: November 13, 2002, as supplemented November 20, 2002.

Brief description of amendments: The amendments delete Technical Specification 5.5.3, “Post Accident Sampling System (PASS),” and thereby eliminate the requirements to have and maintain the PASS at Brunswick Steam Electric Plant, Units 1 and 2.

Date of issuance: February 11, 2003.

Effective date: February 11, 2003, to be implemented within 180 days of issuance.

Amendment Nos.: 226 & 253.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: January 7, 2003 (68 FR 799).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 11, 2003.

No significant hazards consideration comments received: No.

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

Date of amendment request: May 14, 2002, as supplemented by letter dated December 6, 2002.

Brief description of amendment: The amendment revises the technical specification safety function lift setpoint tolerances for the Safety/Relief valves (S/RVs). The changes also allow surveillance of the relief mode of operation of the S/RVs to be performed without physically lifting the disk of a valve off the seat at power.

Date of issuance: February 13, 2003.

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

Amendment No.: 130.

Facility Operating License No. NPF-47: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 25, 2002 (67 FR 42822).

The December 6, 2002, 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 amendment is contained in a Safety Evaluation dated February 13, 2003.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: April 19, 2002, as supplemented by letters dated September 9, 2002 and January 3, 2003.

Brief description of amendments: The amendments revise Technical Specifications (TS) 3.6.6, “Containment Spray and Cooling Systems,” to change the frequency of Surveillance Requirement (SR) 3.6.6.8 from “10 years” to “Following maintenance that could result in nozzle blockage OR Following fluid flow through nozzles.”

Date of issuance: February 20, 2003.

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

Amendment Nos.: 126.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 11, 2002 (67 FR 40023) 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 safety evaluation addresses Braidwood Station Units 1 and 2 only. The NRC staff's evaluation of the Byron Units 1 and 2 will be addressed separately.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 20, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 5, 2002, as supplemented August 13, September 30, October 31, November 13, and November 25, 2002.

Brief description of amendment: The amendment approves an increase in maximum steady-state core power level from 2544 megawatts thermal (MWt) to 2568 MWt, an increase of approximately 0.9 percent.

Date of issuance: December 4, 2002.

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

Amendment No.: 205.

Facility Operating License No. DPR-72: Amendment revises the Facility Start Printed Page 10284Operating License and the Technical Specifications.

Date of initial notice in Federal Register: June 25, 2002 (67 FR 42826). The August 13, September 30, October 31, November 13, and November 25, 2002, supplements contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 4, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 13, 2002.

Brief description of amendment: The amendment revises Improved Technical Specification (ITS) 3.3.8, “Emergency Diesel Generator (EDG) Loss of Power Start (LOPS),” by changing the completion time for required action D.2 from 12 to 36 hours. The amendment also corrects a typographical error in ITS 3.3.8.

Date of issuance: February 11, 2003.

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

Amendment No.: 206.

Facility Operating License No. DPR-72: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 9, 2002 (67 FR 45570).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 11, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 16, 2001, as supplemented by letter dated September 13, 2002.

Brief description of amendments: The amendments revise Technical Specification (TS) 1.1, “Definitions, Dose Equivalent I-131,” and authorize revision of the Final Safety Analysis Report (FSAR) Update to reflect the revised steam generator tube rupture and main steam line break radiological consequences analyses.

Date of issuance: February 20, 2003.

Effective date: February 20, 2003, and shall be implemented in the next periodic update to the FSAR Update.

Amendment Nos.: Unit 1—156; Unit 2—156.

Facility Operating License Nos. DPR-80 and DPR-82: The amendment revised the Technical Specifications and the FSAR Update.

Date of initial notice in Federal Register: January 8, 2002 (67 FR 931). The September 13, 2002, supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 20, 2003.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: March 4, 2002 (TS 01-03).

Brief description of amendments: The amendments revise the SQN Unit 1 and 2 Technical Specifications (TSs) by deleting one definition and modifying several subsections contained in TS Section 6.0, “Administrative Controls.” These changes have been prepared based on existing NRC guidance.

Date of issuance: February 11, 2003.

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

Amendment Nos.: 281 & 272.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revise the TSs.

Date of initial notice in Federal Register: April 16, 2002 (67 FR 18649). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 11, 2003.

No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

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 for the amendment 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.

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual 30-day Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.Start Printed Page 10285

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 application for amendment, (2) the amendment to Facility Operating License, 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 Assess 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 Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By April 3, 2003, 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 2.714,[2] 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, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide 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 to relief. A petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-examine witnesses. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of the continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to hearingdocket@nrc.gov. A copy of the petition for leave to intervene and request for hearing should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and because of continuing disruptions in delivery of mail to United States Government offices, it is requested that copies be transmitted Start Printed Page 10286either 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 filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

Date of application for amendment: January 16, 2003, as supplemented on January 31, 2003.

Brief description of amendment: The amendment modifies Technical Specification 3.1.7 to permit the use of an alternate method of determining rod position for Control Rod H-10 until the end of Cycle 22 or until the next shutdown of sufficient duration, whichever occurs first.

Date of issuance: February 13, 2003.

Effective date: February 13, 2003.

Amendment No. 197.

Facility Operating License No. DPR-23. Amendment revised the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes (68 FR 3556 dated January 24, 2003). The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided for an opportunity to request a hearing by February 24, 2003, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of exigent circumstances, and final determination of NSHC are contained in a Safety Evaluation dated February 13, 2003.

Attorney for licensee: William D. Johnson, Vice President and Corporate Secretary, Carolina Power & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Allen G. Howe.

Start Signature

Dated at Rockville, Maryland, this 21st day of February 2003.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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

End Signature End Preamble

Footnotes

1.  The most recent version of Title 10 of the Code of Federal Regulations, published January 1, 2002, inadvertently omitted the last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2) regarding petitions to intervene and contentions. For the complete, corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29, 2002.

Back to Citation

2.  The most recent version of Title 10 of the Code of Federal Regulations, published January 1, 2002, inadvertently omitted the last sentence of 10 CFR 2.714(d) and paragraph (d)(1) and (d)(2) regarding petitions to intervene and contentions. For the complete, corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April 29, 2002.

Back to Citation

[FR Doc. 03-4623 Filed 3-3-03; 8:45 am]

BILLING CODE 7590-01-P