67 Fed. Reg. 40019 (Jun. 11, 2002)

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

Federal RegisterJun 11, 2002

67 Fed. Reg. 40019 (Jun. 11, 2002)

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 May 17, 2002, through May 30, 2002. The last biweekly notice was published on May 28, 2002 (67 FR 36924).

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 NRC's Public Document Room (PDR), located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By July 11, 2002, 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, which is available at the NRC's PDR, located at One White Flint North, 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/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 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.

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, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and 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, 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 PDR Reference staff at 1-800-397-4209, 304-415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: May 7, 2002.

Description of amendment request: The proposed amendment would relocate the Boration System Technical Specification (TS) requirements to the Technical Requirements Manual (TRM). Additional TS changes to retain boron dilution analysis restrictions would be made as a result of the relocation of the Boration System TS requirements to the TRM. The proposed amendment would also revise the TS Limiting Condition for Operation, action requirements, and surveillance requirements associated with the Emergency Core Cooling, Containment Spray and Cooling, and Auxiliary Feedwater Systems. The proposed changes would remove redundant testing requirements that are already addressed by the Inservice Testing Program, which is required pursuant to TS 4.0.5. The proposed changes would also increase the allowed outage time and shutdown time for an inoperable train (subsystem) of the Emergency Core Cooling System, consistent with standard industry guidelines and other Millstone Unit No. 2 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. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The staff's review is presented below:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes will not alter the way any structure, system, or component functions, and will not alter the manner in which the plant is operated. The proposed changes to the TSs do not impact any system or component that could cause an accident. The ability of the equipment associated with the proposed changes to mitigate the design-basis accidents will not be affected. In addition, the design-basis accidents will remain the same postulated events described in the Millstone Unit No. 2 Final Safety Analysis Report, and the consequences of those events will not be affected. Therefore, the proposed changes will not increase the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes will not alter the plant configuration (no new or different type of equipment will be installed) or require any unusual operator actions. The proposed changes will not alter the way any structure, system, or component functions, and will not alter the manner in which the plant is operated. There will be no adverse effect on plant operation or accident mitigation equipment. The response of the plant and the operators following an accident will not be different. In addition, the proposed changes do not introduce any new failure modes. Therefore, the proposed changes 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 changes to the TSs do not impact any system or component that could cause an accident and will not result in any change in the operational characteristics of the associated accident mitigation equipment. The equipment associated with the proposed TS changes will continue to be able to mitigate the design-basis accidents as assumed in the safety analysis. In addition, the proposed changes will not affect equipment design and there are no changes being made to the TS-required safety limits or safety system settings. Therefore, the proposed changes will not result in a reduction in a margin of safety.

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Section Chief: James W. Clifford.

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: April 24, 2002.

Description of amendment request: Entergy Operations, Inc. requests revision of the River Bend Station, Unit 1 licensing basis and Technical Specifications to utilize the alternative accident source term described in NUREG-1465.

This proposed amendment to the River Bend Technical Specifications (TS) revises those specifications affected by the implementation of the alternative source term concepts in accordance with NUREG 1465. In addition, based on the alternative source term, changes are proposed to selected specifications associated with handling irradiated fuel in the primary containment or Fuel Building and CORE ALTERATIONS. The alternative source term changes affect the definitions, and the specifications for the Control Room Fresh Air System, Standby Gas Treatment System, Fuel Building Ventilation System and leakage rates for Primary Containment and the Personnel Airlocks seal air systems.

Entergy Operations, Inc. [Entergy] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below:

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

Response: No.

The alternative source term does not require modification of the facility; rather, once the occurrence of an accident has been postulated the new source term is an input to evaluate the potential consequences. The implementation of the alternative source term has been evaluated in revisions to the analyses of the limiting design basis accidents at River Bend Station. Based on the results of these analyses, it has been demonstrated that, even with the requested Technical Specification changes, the dose consequences of these limiting events are within the regulatory guidance currently approved by the NRC for use with the alternative source term. This guidance is presented in Regulatory Guide 1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, “Radiological Consequences Analyses Using Alternative Source Terms.”

Because the equipment affected by the revised operational conditions is not considered an initiator to any previously analyzed accident, inoperability of the equipment cannot increase the probability of any previously evaluated accident. The proposed requirements bound the conditions of the current design basis fuel handling accident analysis which concludes that the radiological consequences are within the acceptance criteria of NUREG 0800, Section 15.7.4 and General Design Criteria 19. As noted above, with the alternative source term implementation, the acceptance criteria are also being revised. The results of the revised Fuel Handling Accident demonstrate that the dose consequences are within the NRC regulatory guidance. This guidance is presented in Regulatory Guide 1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, “Radiological Consequences Analyses Using Alternative Source Terms.”

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

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

Response: No.

The proposed changes using the alternative source term dose methodology are analytical in nature and do not physically alter the facility or of any equipment within the facility. Similarly, the alternative source term does not create any new initiators or precursors of a new or different kind of accident. The proposed changes to the Technical Specifications, while they revise certain performance requirements, do not involve any physical modifications to the plant.

The proposed changes related to shutdown controls based on the alternative source term do not create the possibility of a new or different kind of accident from any previous analyzed.

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

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

Response: No.

The changes above are associated with the implementation of a new licensing basis for River Bend Station. Approval of the basis change from the original source term in accordance with TID-14844 to the new alternative source term of NUREG-1465 is requested by this submittal. The results of the accident analyses prepared in support of this submittal are subject to revised acceptance criteria. These analyses have been performed using conservative methodologies as outlined in the regulatory guidance and conservatively represent the requested Technical Specification changes. Safety margins and analytical conservatisms have been evaluated and are well understood. The analyzed events have been carefully selected and margin has been retained to ensure that the analyses adequately bound all postulated event scenarios. The dose consequences of these limiting events are within the acceptance criteria also found in the latest regulatory guidance. This guidance is presented in Regulatory Guide 1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, “Radiological Consequences Analyses Using Alternative Source Terms.”

The proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries as well as control room, are within the corresponding regulatory limits. In a similar way, the results of the existing analyses demonstrated that the dose consequences were within the applicable NRC-specified regulatory limit. Specifically, the margin of safety for these accidents is considered to be that provided by meeting the applicable regulatory limit for Alternate Source Term methodologies, which, for most events, is conservatively set at, or below, the 10CFR50.67 limit. With respect to the control room personnel doses, the margin of safety is the difference between the 10CFR100 limits and the regulatory limit defined by 10CFR50, Appendix A, General Design Criterion (GDC) 19.

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

Based on the above, Entergy concludes that the proposed amendment(s) present 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: Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Robert A. Gramm.

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.

Description of amendment request: Entergy Operations, Inc. is proposing that the River Bend Station, Unit 1 Operating License be amended to reflect a 1.7 percent increase in the licensed 100% reactor core thermal power level (an increase in reactor power level from 3,039 megawatts thermal to 3,091 megawatts thermal). These changes result from increased accuracy of the feedwater flow measurement to be achieved by utilizing high accuracy ultrasonic flow measurement instrumentation. The basis for this change is consistent with the revision, issued in June 2000, to appendix K to part 50 of title 10 of the Code of Federal Regulations, allowing operating reactor licensees to use an uncertainty factor of less than 2 percent of rated reactor thermal power in analyses of postulated design basis loss-of-coolant accidents.

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

Response: No.

The comprehensive analytical efforts performed to support the proposed change included a review of the Nuclear Steam Supply System (NSSS) systems and components that could be affected by this change. All systems and components will function as designed, and the applicable performance requirements have been evaluated and found to be acceptable.

The comprehensive analytical efforts performed to support the proposed uprate conditions included a review and evaluation of all components and systems that could be affected by this change. Evaluation of accident analyses confirmed the effects of the proposed uprate are bounded by the current dose analyses. All systems will function as designed, and all performance requirements for these systems have been evaluated for the uprate conditions and found acceptable. Because the integrity of the plant will not be affected by operation at the new power level conditions, it is concluded that all structures, systems, and components required to mitigate a transient remain capable of fulfilling their intended functions. The reduced uncertainty in the flow input to the power calorimetric measurement allows the current safety analyses to be used, with small changes to the core operating limits, to support operation at a core power of 3,091 megawatts thermal (MWt). As such, all Final Safety Analysis Report (FSAR) Chapter 15 accident analyses continue to demonstrate compliance with the relevant event acceptance criteria. Those analyses performed to assess the effects of mass and energy releases remain valid. The source terms used to assess radiological consequences have been reviewed and determined to either bound operation at the new power level condition, or new analyses were performed to verify all acceptance criteria continue to be met.

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.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed changes have no adverse effects on any safety-related system or component and do not challenge the performance or integrity of any safety related system.

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.

Operation at the uprated power condition does not involve a significant reduction in a margin of safety. Analyses of the primary fission product barriers have concluded that all relevant design criteria remain satisfied, both from the standpoint of the integrity of the primary fission product barrier and from the standpoint of compliance with the required acceptance criteria. The calculated loads on all affected structures, systems and components have been shown to remain within design criteria for all design basis event categories. No NRC [U.S. Nuclear Regulatory Commission] acceptance criterion is exceeded.

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

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

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: April 19, 2002.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.6.6 surveillance requirement (SR) to verify each spray nozzle on the containment spray ring headers at the top of containment dome is unobstructed. The current TS 3.6.6.8 requirement is to verify each spray nozzle every 10 years. The proposed requirement is to revise the frequency to “Following maintenance that could result in nozzle blockage OR Following fluid flow through the nozzles.”

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

The proposed change revises the Frequency for Technical Specifications (TS) Surveillance Requirement (SR) 3.6.6.8 for verifying each spray nozzle is unobstructed from “10 years” to “Following maintenance that could result in nozzle blockage OR Following fluid flow through the nozzles.”

Analyzed events are initiated by the failure of plant structures, systems, or components. The Containment Spray (CS) system is not considered as an initiator of any analyzed event. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. No active or passive failure mechanisms that could lead to an accident are affected. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The initial conditions of Design Basis Accident (DBA) and transient analyses in the Byron/Braidwood Stations' UFSAR assume the CS system is operable.

The operability of the CS system in accordance with the proposed TS is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the plant. Since plant safety can be ensured at the proposed Frequency, we are proposing to revise the CS system testing provisions to require nozzle testing only after activities that could result in nozzle blockage, i.e., following maintenance that could result in nozzle blockage or following fluid flow through the nozzles. Nozzle blockage is considered unlikely during periods without maintenance or without fluid flow through the nozzles, since the nozzles are of a passive design and the system is kept in a normally dry state, thus minimizing corrosion susceptibility. In addition, the location of the nozzles at the top of the containment dome limits the possibility of the introduction of foreign material from sources external to the CS system. The proposed Frequency will continue to provide confidence that an unobstructed flow path is available, and will preclude the need for unnecessary testing when no activities have occurred that would introduce debris to the spray ring headers, or when no other active degradation mechanism is present. Operability of the CS system will not be affected. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change does not involve the use or installation of new equipment. Installed equipment is not operated in a new or different manner. No new or different system interactions are created, and no new processes are introduced. The current foreign material exclusion practices have been reviewed and judged sufficient to provide high confidence that debris will not be introduced during times when the CS system boundary is breached. The design of the CS system at Braidwood and Byron Stations precludes borated water from reaching the spray nozzles, except during a CS actuation. 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?

The proposed change does not introduce any new setpoints at which protective or mitigative actions are initiated. No current setpoints are altered by this change. The design and functioning of the CS system is unchanged. Since the system is not susceptible to corrosion induced obstruction nor is the introduction of foreign material from external sources likely, and the design of the CS system at Braidwood and Byron Stations precludes borated water from reaching the spray nozzles except during a CS actuation, the proposed testing Frequency is sufficient to provide high confidence that the CS system will continue to function as designed. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Therefore, based on the above evaluation, we have concluded that the proposed change does not involve any significant hazards consideration.

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

NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: February 15, 2002.

Description of amendment request: Exelon proposed to increase the trip setpoints for Items 3.b and 3.c in Table 3.3.2-2, for the Reactor Water Cleanup System (RWCS) steam leak detection temperature isolation actuation instrumentation in the technical specifications.

Basis for proposed no significant hazards consideration determination: As required by Section 50.91(a) the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below:

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

No. The RWCS is not required for safety purposes nor is it required to operate after a design-basis accident. The RWCS instrumentation and controls are not required for safe operation of the reactor. They provide a means of monitoring parameters and protecting the system. The increase in the isolation setpoint and allowable value for the RWCS pump room high ambient temperature and high differential temperature will not make any physical changes (modification) to the plant equipment. Therefore, the proposed changes to the RWCS setpoints will not increase the probability of an accident previously evaluated.

This license amendment request (LAR) does not increase the consequences of an accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR). This proposed change has no impact on the high-energy line break or loss-of-coolant accident (LOCA) accident analyses. This LAR does not adversely affect mitigating systems, structures or components (SSCs), and does not adversely affect the initial conditions of any accidents. Affected equipment will remain within the limitations of the Environmental Qualification Program. Redundancy and diversity of mitigating systems are unchanged as a result of this LAR. This LAR does not affect onsite or offsite radiological consequences of any accident previously evaluated in the UFSAR.

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

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

No. The increase in the RWCS pump room high ambient temperature and high differential temperature settings proposed by this LAR does not change any SSC. This LAR does not create new operating or failure modes. Existing instruments are not accident initiators in any failure mode and changing settings does not change the instrument's functions. Therefore, this LAR 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?

No. This LAR will allow the plant to operate at higher ambient temperatures in the RWCS pump rooms during normal operation. This change does not create additional heat loads or change the way any of the equipment is operated. No safety-related setpoints are associated with the RWCS system. The RWCS system instrumentation and controls are not required for safe operation of the reactor. They provide a means of monitoring parameters and protecting the system. Therefore, a change to the TSs for RWCS pump room high ambient temperature and high differential temperature limits to the new setpoints is not considered a significant reduction in a margin of safety.

Attorney for licensee: Mr. Edward Cullen, Vice President & General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: James W. Clifford.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: April 24, 2002.

Description of amendment request: The proposed amendment would change the Technical Specifications (TSs) to relocate to the Seabrook Station Technical Requirements (SSTR) Manual, specific pressure, differential pressure and flow values, as well as specific test methods, contained in Surveillance Requirements (SRs) 4.6.2.1, “Containment Spray System,” and 4.7.1.2.1b, “Auxiliary Feedwater System.”

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

The proposed changes to relocate the specific pump pressure and flow criteria TS SRs to the SSTR are administrative in nature and do not adversely affect accident initiators or precursors, or alter the design assumptions, conditions, and configuration of the facility or the manner in which it is operated. The proposed changes do not alter or prevent the ability of structures, systems, or components to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR).

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

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

The proposed changes do not alter the design assumptions, conditions, or configuration of the facility or the manner in which it is operated. The proposed changes have no adverse impact on component or system interactions. Since there are no changes to the design assumptions, parameters, conditions and configuration of the facility, or the manner in which the plant is operated and surveilled, the proposed changes do not create the possibility of a new or different accident from any previously analyzed.

3. The proposed changes do not involve a significant reduction in a margin of safety.

There is no adverse impact on equipment design or operation and there are no changes being made to the TSs themselves that would adversely affect any current margin of safety. The proposed changes are administrative in nature and impose alternative procedural and programmatic controls on these parameter limits.

Therefore, relocation of the specific pump pressure and flow criteria do not involve a significant reduction in a margin of safety.

Attorney for licensee: William J. Quinlan, Esq. Assistant General Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270.

NRC Section Chief: James W. Clifford.

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

Date of amendment requests: April 10, 2002.

Description of amendment requests: The proposed license amendments would revise several of the Required Actions in the DCPP Technical Specifications (TS) that require suspension of operations involving positive reactivity additions or suspension of operations involving reactor coolant system (RCS) boron concentration reductions. In addition, this license amendment request (LAR) proposes to revise several Limiting Condition for Operation (LCO) Notes that preclude reductions in RCS boron concentration when a reactor coolant pump(s) and/or a residual heat removal pump(s) are removed from operation. The proposed changes would allow small, controlled, safe insertions of positive reactivity, but limit the introduction of positive reactivity to ensure that compliance with the required shutdown margin or refueling boron concentration limits will still be satisfied.

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

Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no hardware changes. The reactor trip system instrumentation and reactivity control systems will be unaffected. Protection systems will continue to function in a manner consistent with the plant design basis. All design, material, and construction standards that were applicable prior to the request are maintained.

The probability and consequences of accidents previously evaluated in the Final Safety Analysis Report Update (FSAR) are not adversely affected because the changes to the Required Actions and LCO Notes assure the limits on SDM [shutdown margin] and refueling boron concentration continue to be met, consistent with the analysis assumptions and initial conditions included within the safety analysis and licensing basis. The activities covered by this LAR are routine operating evolutions. The proposed changes do not reduce the capability to borate the RCS.

The equipment and processes used to implement RCS boration or dilution evolutions are unchanged and the equipment and processes are commonly used throughout the applicable modes under consideration. There will be no degradation in the performance of or an increase in the number of challenges imposed on, safety-related equipment assumed to function during an accident. There will be no change to normal plant operating parameters or accident mitigation performance.

The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the FSAR.

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

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

There are no hardware changes or any changes in the method by which any safety-related plant system performs its safety function. This amendment will not affect the normal method of plant operation or change any operating limits. The proposed changes permit the conduct of normal operating evolutions when additional controls over core reactivity are imposed by the TS. The proposed changes do not introduce any new equipment into the plant or alter the manner in which existing equipment will be operated. The changes to operating procedures are minor, with clarifications provided that required limits must continue to be met. No performance requirements or response time limits will be affected. These changes are consistent with assumptions made in the safety analysis and licensing basis regarding limits on SDM and refueling boron concentration.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this LAR. There will be no adverse effect or challenges imposed on any safety-related system as a result of this LAR.

This LAR does not alter the design or performance of the reactor protection system, nuclear instrumentation system, or solid state protection system used in the plant protection systems.

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

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

The proposed changes do not alter the limits on SDM or refueling boron concentration. These limits continue to assure that core parameters remain within the bounds of the accident analysis. The nominal trip setpoints specified in the TS and the safety analysis limits assumed in the transient and accident analyses are unchanged. None of the acceptance criteria for any accident analysis is changed.

The proposed changes do not affect the manner in which safety limits or limiting safety system settings are determined, nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. Also, the proposed changes do not impact the overpower limit, departure from nucleate boiling ratio limits, heat flux hot channel factor (F_Q), nuclear enthalpy rise hot channel factor (FΔH), loss of coolant accident peak cladding temperature, peak local power density, or any other margin of safety. The radiological dose consequence acceptance criteria will continue to be met.

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

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

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

Date of amendment requests: April 15, 2002.

Description of amendment requests: The proposed license amendments would approve changes in the implementation of the DCPP Control of Heavy Loads Program and other analyses, design and procedure changes required to implement a dry cask Independent Spent Fuel Storage Installation (ISFSI) at DCCP.

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

With the Holtec International (Holtec) HI-STORM 100 System and the associated design and handling procedures, most cask drops and other events, which could damage other spent fuel, have been precluded through redundant handling systems, control system upgrades, and mechanical stops/electrical interlocks that preclude crane movement over spent fuel, meeting PG&E's commitments to the guidelines of NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants.” For those remaining cases where a cask drop is still credible, the impact-limiter design ensures the deceleration of the contained spent fuel remains below fuel design limits, preventing damage to the contained fuel assemblies (and associated structures), and meeting the analysis guidance of NUREG-0612. As a result of this design approach, a cask-handling accident that results in a significant offsite radiological release is not considered credible.

Other Diablo Canyon Power Plant (DCPP) licensing-basis events, such as the drop of a spent fuel assembly, have not been affected by these changes and remain bounding events for potential radiological consequences.

Revision of the DCPP Control of Heavy Loads Program ensures that PG&E's commitments to NUREG-0612 guidelines will protect the new fuel storage locations and the new transfer cask/multi-purpose canister (MPC) loading/unloading activities.

The addition of restraint structures and use of impact limiters preclude adverse effects from seismic events and/or cask drops or tipovers, assuring that the fuel, MPC, transfer cask, and other potentially affected 10 CFR 50 structures remain within their design bases. The addition and installation of this equipment will be done after necessary evaluation and analysis is performed, to ensure the equipment does not introduce any unacceptable effect (e.g., seismic interaction).

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

The proposed changes to use the Holtec HI-STORM 100 system have been evaluated for seismic events and tornado missile impacts and it has been determined that these changes will not result in an increase in the probability or consequences of an accident previously evaluated.

The Fire Protection Program will ensure that the combustible materials are properly controlled such that the total combustibles meet the current program commitments.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident.

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

The engineering design measures and the handling procedures preclude the possibility of new or different kinds of accidents. Damage to 10 CFR 50 SSCs [structures, systems and components] from the cask handling and associated activities, and events resulting from possible damage to contained fuel, have been carefully considered in the following safety analyses. Both the types of accidents and the results remain within the envelope of existing analyses, as demonstrated by the PG&E and Holtec analyses.

In Supplement No. 2 to the Safety Evaluation of DCPP (Reference 7.18 [of the April 15, 2002, license amendment request]), the NRC reviewed and accepted Amendment 27 of the original DCPP Final Safety Analysis Report (FSAR) analysis of a cask-drop accident. Amendment 22 to Facility Operating License No. DPR-80 and Amendment 21 to Facility Operating License No. DPR-82 allowed expansion of the spent fuel pool (SFP) storage capacity. In the safety evaluation for these amendments, the NRC reviewed the cask-drop accident and noted that the licensee had proposed administrative controls that would preclude the movement of a spent-fuel shipping cask in an exclusion zone over, and in the vicinity of, stored spent fuel that could result in a cask drop or tipping accident damaging stored spent fuel.

Supplement No. 27 to the Safety Evaluation Report for DCPP Unit 1 (Reference 7.19 [of the April 15, 2002, license amendment request]) and in Supplement No. 31 to the Safety Evaluation Report for Unit 2 (Reference 7.20 [of the April 15, 2002, license amendment request]) included the review and acceptance of the DCPP Control of Heavy Loads Program.

The rupture of MPC dewatering, vacuum, forced helium dehydration or related closure system lines or the malfunction of equipment during cask handling operations resulting in radiological consequences are bounded by the DCPP Final Safety Analysis Report (FSAR) Update fuel-handling accident analysis.

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

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

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

With the Holtec HI-STORM 100 System, and the associated design and handling procedures, most cask drops and other events have been completely precluded through redundant load-handling systems, providing defense-in-depth as described in NUREG-0612, and meeting PG&E's commitments to the guidance of NUREG-0612. In those remaining cases where a cask drop is still credible, impact limiter design ensures that the deceleration of the contained spent fuel remains below fuel design limits, preventing damage to the contained fuel assemblies (and associated structures), and meeting the analysis guidelines of NUREG-0612. As a result of this design approach, the margin of safety has been maintained through the elimination of certain drops and the associated structural challenges.

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

Revision of DCPP Control of Heavy Loads Program to incorporate the additional restrictions on heavy loads movement will not affect the procedures or methodology used and will, therefore, not affect margins.

The addition of restraint structures and use of impact limiters preclude adverse effects from seismic events and/or cask drops or tipovers, assuring that the fuel, MPC, transfer cask, and other potentially affected 10 CFR 50 structures remain within their design bases. Since design-basis criteria are fully satisfied, there is no impact on the margin of safety.

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

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

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

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

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

Date of amendment request: April 8, 2002.

Brief description of amendments: The proposed change would revise Technical Specification (TS) 3.4.16, “RCS [Reactor Coolant System] Specific Activity,” to lower the Limiting Condition For Operation and associated Surveillance Requirements for Dose Equivalent Iodine-131 in the Reactor Coolant System from a specific activity of 1.0 μCi/gm to 0.45 μCi/gm. The change also includes approval of proposed changes to Technical Specification Bases for Main Steam Line Break post-accident radiological dose consequences analysis that was previously approved for implementing the Comanche Peak Steam Electric Station Steam Generator Alternate Repair Criteria.

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

Response: No.

The proposed change to revise Technical Specification (TS) 3.4.16 “Reactor Coolant System Specific Activity” to reduce the Limiting Condition For Operation (LCO) for Dose Equivalent I-131 in the reactor coolant from a specific activity of 1.0 μCi/gm to 0.45 μCi/gm and the revised main steam line break (MSLB) radiological consequence analysis are used to determine post-accident dose. They are not related to any accident initiator. Therefore, this change cannot increase the probability of an accident.

The revised MSLB offsite and control room radiological consequences analysis dose results are within 10 CFR Part 100 and 10 CFR Part 50, Appendix A Criterion 19 limits and the NUREG-0800 SRP [Standard Review Plan] section 15.1.5 and section 6.4 guideline values.

Therefore, the proposed change does 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 change to revise TS 3.4.16 “Reactor Coolant System Specific Activity” to reduce the LCO for Dose Equivalent I-131 in the reactor coolant from a specific activity of 1.0 μCi/gm to 0.45 μCi/gm and the revised MSLB radiological consequence analysis do not involve any physical plant changes. The change does not involve changes in operation of the plant that could introduce a new failure mode for creating an accident or affect the mitigation of an accident.

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

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

Response: No.

The proposed change to revise TS 3.4.16 “Reactor Coolant System Specific Activity” to reduce the LCO for Dose Equivalent I-131 in the reactor coolant from a specific activity of 1.0 μCi/gm to 0.45 μCi/gm is a conservative change in that this reduced TS limit, when used in applicable plant radiological dose consequence analysis models with all other input parameters held constant, calculates decreased dose consequences to the thyroid. The change, with all other analysis input parameters held constant, increases the margin to acceptance limits. Therefore, this change does not result in a significant reduction in the margin provided by TS 3.4.16.

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

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, 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.

Notice of Issuance of Amendments to Facility Operating Licenses

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of application for amendments: January 31, 2002.

Brief description of amendments: The amendments revised Surveillance Requirement (SR) 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours, or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours, or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”

Date of issuance: May 22, 2002.

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

Amendment Nos.: 253 and 229.

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

Date of initial notice in Federal Register: March 19, 2002 (67 FR 12600). The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated May 22, 2002.

No significant hazards consideration comments received: No.

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 amendment request: August 1, 2001, as supplemented by letters dated November 28, 2001, December 17, 2001, January 24, 2002, February 4, 2002 (two letters), April 25, 2002, May 10, 2002 and May 28, 2002.

Description of amendment request: The amendments changed the Technical Specifications (TS) to replace the current accident source term used in design basis radiological analyses with an alternative source term pursuant to 10 CFR 50.67, “Accident Source Term.” License Conditions were added to the Unit 2 Operating License.

Date of issuance: May 30, 2002.

Effective date: Unit 1, upon issuance. Unit 2, upon completion of Refueling Outage 15.

Amendment Nos: 221 and 246.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments changed the Technical Specifications and added License Conditions to DPR-62 only.

Date of initial notice in Federal Register: September 5, 2001 (66 FR 46477). The supplements contained clarifying information only, and did not change the initial no significant hazards consideration determination or expand the scope of the initial Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 30, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: August 24, 2002, as supplemented March 26, 2002.

Brief description of amendment: The amendment revises the Technical Specifications to delete Required Action 3.3.1.1.J.2, which specifies that the oscillation power range monitor upscale trip function be restored to operable status within 120 days when it is determined to be inoperable.

Date of issuance: May 24, 2002.

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

Amendment No.: 146.

Facility Operating License No. NPF-43: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: November 28, 2001 (66 FR 59503). The March 26, 2002, supplemental letter provided additional clarifying information that was within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 24, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 20, 2001, as supplemented on January 25 and April 29, 2002.

Brief description of amendment: The amendment revises Technical Specification (TS) 3.8, “Refueling, Fuel Storage and Operations with the Reactor Vessel Head Bolts Less Than Fully Tensioned,” TS Table 4.1-2, “Frequencies for Sampling Tests,” and TS 5.4, “Fuel Storage,” to allow credit for soluble boron in the criticality analysis for the spent fuel pit (SFP). The amendment also incorporates changes to the SFP rack layout by dividing it into sub-regions and specifying requirements for fuel assembly burnup and soluble boron concentration for various loading configurations in these sub-regions.

Date of issuance: May 29, 2002.

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

Amendment No.: 227.

Facility Operating License No. DPR-26: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR 55012). The January 25 and April 29, 2002, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 29, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 30, 2001.

Brief description of amendments: These amendments revised Surveillance Requirement (SR) 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”

Date of issuance: May 23, 2002.

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

Amendments Nos.: 243, 247.

Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 19, 2002 (67 FR 7417). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 23, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 22, 2001.

Brief description of amendments: The amendments allowed the relocation of the Technical Specification (TS) sections associated with the curie content limit for liquid and gaseous waste storage and the TS sections associated with the explosive gas concentration limits to licensee controlled documents. In addition, the amendments allow for revisions to the reporting requirements of TS 6.9.3, “Annual Radioactive Release Report.”

Date of issuance: May 21, 2002.

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

Amendment Nos.: 250, 130.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 3, 2001 (66 FR 50467). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 21, 2002.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

Date of application for amendment: November 8, 2000, as supplemented February 6, May 7, and November 21, 2001.

Brief description of amendment: The amendment changed the technical specifications associated with the deletion of TS 3/4.4.1.6, “Reactor Coolant Pump—Startup.”

Date of issuance: May 30, 2002.

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

Amendment No: 131.

Facility Operating License No. NPF-73: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 27, 2000 (65 FR 81917). The February 6, May 7, and November 21, 2001, letters provided additional information that clarified the application but did not expand the scope of the application as originally noticed or change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 30, 2002.

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: April 9, 2002, as supplemented April 25, 2002.

Brief description of amendments: The amendments consist of changes to the Technical Specifications (TSs) in response to the application dated April 9, 2002, as supplemented April 25, 2002. In the April 25, 2002, supplemental letter, the licensee requested that the portion of the original application dealing with the Unit 2 AB and CD train batteries for Unit 2 only be processed on an emergency basis. By letter dated April 26, 2002, the Nuclear Regulatory Commission issued Amendment No. 249 for Unit 2. The amendments revise the Surveillance Requirement (SR) for the Train AB, and CD batteries in TS 4.8.2.3.2.c.1 for Unit 1 and SR TS 4.8.2.5.2.c.1 for the N train batteries in both Units 1 and 2. The amendments modify the requirements to verify that battery cells, cell plates and racks show no visual indication of physical damage or abnormal deterioration. The amendments would allow the operability of batteries exhibiting damage or deterioration to be determined by an evaluation. The amendments are consistent with an NRC-approved change to the Standard Technical Specifications for Westinghouse plants (NUREG 1431, Revision 1) as documented in Technical Specification Task Force Standard Technical Specification.

Date of issuance: May 30, 2002.

Effective date: As of the date of issuance and shall be implemented within 7 days.

Amendment Nos.: 269 and 250.

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

Date of initial notice in Federal Register: April 25, 2002 ( 67 FR 20552). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 30, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 28, 2002.

Brief description of amendment: Amendment changes Technical Specification 3.0.3 to allow a longer time before entering a limiting condition for operation in the event of a missed surveillance and adds requirements to (1) perform a risk evaluation for any surveillance delayed greater than 24 hours and (2) manage the risk impact.

Date of issuance: May 30, 2002.

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

Amendment No.: 246.

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

Date of initial notice in Federal Register: April 30, 2002 (67 FR 21290). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 30, 2002.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 14, 2001, as supplemented by letters dated January 15 and April 15, 2002.

Brief description of amendment: The amendment revises Technical Specification (TS) 2.10.4(5)(a)(iii), “DNBR [departure from nucleate boiling ratio] Margin During Power Operation Above 15% Rated Power,” to decrease the minimum required reactor coolant system flow rate from 206,000 gallons per minute (gpm) to 202,500 gpm. In addition, the Bases section for TS 2.10.4 has been revised to be consistent with the approved change to the TS.

Date of issuance: May 24, 2002.

Effective date: May 24, 2002 , and to be implemented within 30 days from the date of issuance.

Amendment No.: 209.

Facility Operating License No. DPR-40. Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 22, 2002 (67 FR 2927). The January 15 and April 15, 2002, supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 24, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 10, 2001.

Description of amendment request: The proposed amendments revised the Technical Specifications (TSs) to incorporate the Nuclear Regulatory Commission (NRC)-approved generic change Technical Specification Task Force-287, Revision 5, to the “Standard Technical Specifications for General Electric Plants (BWR/4),” NUREG-1433, Revision 1. Specifically, the changes: (a) Inserted a note in the Limiting Condition for Operation (LCO) in TS 3.7.3 to state that the control room habitability envelope boundary may be opened intermittently under administrative control; (b) inserted a new LCO Action B in TS 3.7.3 to allow 24 hours to restore the control room habitability envelope boundary to operable status if two control room emergency outside air supply (CREOAS) subsystems should become inoperable due to an inoperable control room habitability envelope boundary in Modes 1, 2 and 3; (c) re-labeled the existing LCO Actions B, C, D, and E to C, D, E, and F respectively; and (d) revised the existing LCO Action D to require immediate entry into LCO 3.0.3 when two CREOAS subsystems are inoperable for situations other than when the inoperability is due to an inoperable control room habitability envelope boundary. Minor formatting and editorial changes were also made.

Date of issuance: May 17, 2002.

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

Amendment Nos.: 203, 177.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 5, 2002 (67 FR 10014). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 17, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 11, 2002.

Brief description of amendments: The amendments revise TS Section 1.1, Definitions, to change the definition of response time testing as it is applied to the Engineered Safety Features, and the Reactor Protective System, based on approved Technical Specification Task Force (TSTF) Traveler TSTF-368, Revision 0, “Incorporate Combustion Engineering Owners Group (CEOG) Topical Report to Eliminate Pressure Sensor Response Time Testing.”

Date of issuance: 1 May 22, 2002.

Effective date: May 22, 2002, to be implemented within 60 days of issuance.

Amendment Nos.: Unit 2-188; Unit 3-179.

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

Date of initial notice in Federal Register: April 16, 2002 (67 FR 18648). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 22, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 31, 2001, as supplemented by letter dated November 15, 2001, February 20 (two letters), dated February 21, and March 14, 2002.

Brief description of amendments: The amendments revised the Technical Specifications to extend the completion times for the required actions associated with restoring an inoperable emergency diesel generator.

Date of issuance: May 17, 2002.

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

Amendment Nos.: 231 and 172.

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

Date of initial notice in Federal Register: November 17, 2001 (66 FR 52803). The supplements dated November 15, 2001, February 20 (two letters), February 21, and March 14, 2002, provided clarifying information that did not change the scope of the August 31, 2001, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 17, 2002.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County, Georgia

Date of application for amendment: May 21, 2001.

Brief description of amendment: The amendment revised the Technical Specifications to eliminate the response time testing requirements for the reactor protection system signals of reactor high steam dome pressure and reactor vessel water level low.

Date of issuance: May 17, 2002.

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

Amendment No.: 173.

Renewed Facility Operating License No. NPF-5: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 12, 2001 (66 FR 31713). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 17, 2002.

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.

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.

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, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Assess and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.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, 304-415-4737 or by email to pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By July 11, 2002, 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 which is available at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852, and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). 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.

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 20555-001, Attention: Rulemakings and Adjudications Staff, or may be delivered to the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-001, and 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, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of application for amendment: May 29, 2002.

Brief description of amendment: This amendment revises the Technical Specification (TS) 3/4.3.3.6 “Accident Monitoring Instrumentation” and associated Bases for Reactor Vessel Level and In Core Temperature monitoring to be consistent with NUREG-1431, Revision 2, “Standard Technical Specifications Westinghouse Plants.”

Date of issuance: May 30, 2002.

Effective date: May 30, 2002.

Amendment No. 110.

Facility Operating License No. NPF-63. Amendment revises the TS.

Public comments requested as to proposed no significant hazards consideration (NSHC):

No. The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a Safety Evaluation dated May 30, 2002.

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: Thomas Koshy, Acting.

Dated at Rockville, Maryland, this 3rd day of June, 2002.

For the Nuclear Regulatory Commission.

Stuart A. Richards,

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

[FR Doc. 02-14339 Filed 6-10-02; 8:45 am]

BILLING CODE 7590-01-P