00-12302. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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 April 22, 2000, through May 5, 2000. The last biweekly notice was published on May 3, 2000 (65 FR 25761).

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; (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 Review and Directives Branch, Division of Freedom of Information and Publications 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 Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By June 16, 2000, 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, the Gelman Building, 2120 L Street, NW., Washington, DC, 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.

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.

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: Docketing and Services Branch, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 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 Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Maricopa County, Arizona

Date of amendments request: May 26, 1999, as supplemented March 31, 2000.

Description of amendments request: The licensee proposes to change the allowable values in Technical Specification Section 3.3.1, Table 3.3.1-1, Item 12, “Reactor Coolant Flow, Steam Generator No. 1-Low” and Item 13, “Reactor Coolant Flow, Steam Generator No. 2-Low,” to reduce the demonstrated spurious trip hazard associated with this setpoint. This application was originally noticed in the Federal Register on June 30, 1999 (64 FR 35201).

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:

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

No. The proposed change will change the Reactor Protection System (RPS) reactor coolant flow trip setpoints. The RPS functions to mitigate the consequences of an accident. The changes to the low reactor coolant flow trip setpoints will reduce or eliminate unnecessary challenges to the RPS. Therefore, the proposed change will not involve a significant increase in the probability of an accident previously evaluated.

These changes will result in an increased time delay for the RPS low reactor coolant flow trip. The reanalysis of the affected Updated Final Safety Analysis Report (UFSAR) Chapter 15 event (UFSAR 15.3.4, Reactor Coolant Pump Shaft Break with Loss of Offsite Power), with the increased time delay, shows that the dose consequences for this event remains bounded by the UFSAR analysis. Therefore, this change does not involve a significant increase in the consequences of an accident previously evaluated.

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

No. The proposed change will change the RPS reactor coolant flow trip setpoints. The RPS functions to mitigate the consequences of an accident. The changes to the low reactor coolant flow trip setpoints will reduce or eliminate unnecessary challenges to the RPS. The proposed change only changes the mitigating actions of the RPS, without changing the required function of the RPS. Therefore, the change to the low reactor coolant flow trip setpoints does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

No. The proposed change will change the RPS reactor coolant flow trip setpoints. The reanalysis of the affected UFSAR Chapter 15 event (UFSAR 15.3.4, Reactor Coolant Pump Shaft Break with Loss of Offsite Power), with the revised reactor coolant flow trip setpoints, shows that the minimum departure from nucleate boiling ratio (DNBR) and specified acceptable fuel design limits (SAFDLs) for this event remains bounded by the UFSAR analysis. 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 that review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999 NRC Section Chief: Stephen Dembek

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

Date of amendments request: April 26, 2000.

Description of amendments request: The proposed amendments would increase the maximum average ultimate heat sink (UHS) temperature allowed by Technical Specification (TS) 3.7.2, “Service Water System and Ultimate Heat Sink.”

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. Operation with the maximum 24 hour average UHS water temperature as high as 90.5°F does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The BSEP SW [Service Water] system is designed to provide cooling water for the removal of heat from equipment required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. This equipment includes the Diesel Generators (DGs), Residual Heat Removal (RHR) pump seal coolers, room cooling units for Emergency Core Cooling System (ECCS) equipment, and Residual Heat Removal Service Water (RHRSW) heat exchangers. The SW system also provides cooling to other components, as required, during normal operation. The SW system is not an initiator of any previously evaluated accident. The safety related components associated with SW cooling have been analyzed for a maximum UHS temperature of 92°F. The proposed change maintains this maximum UHS temperature. As such, the qualification of safety related components is not affected. Therefore, the probability of occurrence of a previously evaluated accident is not increased.

The new maximum 24 hour average UHS water temperature limit of 90.5° F has been evaluated and it was determined that the SW system will maintain sufficient heat removal capability. Existing TS operability requirements for the UHS ensure that conservatively bounding assumptions used in the analysis of the SW system's heat removal capability will be met, or the UHS will be declared inoperable. As such, the consequences of previously analyzed accidents are not affected[.]

2. Operation with the maximum 24 hour average UHS water temperature as high as 90.5°F will not create the possibility of a new or different kind of accident from any accident previously evaluated.

Increasing the maximum 24 hour average UHS water temperature does not create the possibility of an accident of a different type than any evaluated previously in the safety analysis report. UHS water temperature does not represent an accident initiator. There is no physical change to any plant structure, system, or components. Therefore, there is no possibility of an accident of a different type.

Increasing the maximum 24 hour average UHS water temperature does not create the possibility of a malfunction of a different type than any evaluated previously. The safety related components associated with SW cooling have been analyzed for a maximum UHS temperature of 92°F. This maximum UHS temperature is maintained by the proposed change. As such, this condition does not introduce the possibility of a malfunction of a different type than any evaluated.

3. Operation with the maximum 24 hour average UHS water temperature as high as 90.5°F does not involve a significant reduction in a margin of safety.

UHS temperature limits are established to ensure that the SW system is able to provide sufficient cooling water for the removal of heat from equipment, such as the DGs, RHR pump seal coolers, ECCS room cooling units, and RHRSW heat exchangers, required for a safe reactor shutdown following a DBA or transient. CP&L has performed an analysis which demonstrates that this capability is not reduced with the increased maximum 24 hour average UHS water temperature limit. Existing TS operability requirements for the UHS ensure that conservatively bounding assumptions used in the analysis of the SW system's heat removal capability will be met, or the UHS will be declared inoperable. As such, the ability of the SW system to perform its intended safety function is not affected and the margin of safety is not reduced.

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

NRC Section Chief: Richard P. Correia.

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

Date of amendment request: April 13, 2000

Description of amendment request: The proposed amendments would revise the Technical Specifications to accommodate the use of Framatome Cogema Fuels Mark-B11 fuel with M5 cladding.

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

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

No. The proposed change to the technical specifications and bases incorporate the use of Mark-B11 fuel assemblies with M5 cladding. The analyzed events are initiated by the failure of specific plant structures, systems, or components. The change in fuel assembly design or cladding material does not impact the condition or performance of those structures, system, or components. Therefore, the proposed changes will not increase the probability of an accident previously evaluated.

The accident analyses have been evaluated to address the changes in the fuel design and cladding material. The results of this evaluation demonstrate that the applicable acceptance criteria are met. Thus, the proposed changes will not increase the consequences of an accident previously evaluated.

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

No. The proposed changes to the technical specifications are to support implementation of Mark-B11 fuel assemblies with M5 cladding. The changes in fuel design and cladding material do not alter the operating characteristics of the plant. In addition, the fuel handling equipment is compatible with the Mark-B11 fuel assembly design. Therefore, the proposed changes do 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?

No. The margin of safety is established through the design of the plant systems, structures, components, and the parameters within which the plant is operated. The proposed change does not involve any significant physical change to the plant. The primary design changes, which enhance nuclear, thermal-hydraulic and mechanical performance, include the following:

1. Reduced diameter fuel rod,

2. Flow mixing vanes on five of the six intermediate spacer grids,

3. Improved grid restraint system, and

4. M5 fuel rod cladding.

The changes in fuel design and cladding material have been evaluated which demonstrates that all of the applicable acceptance criteria are met. Based on this, the proposed changes do not involve a significant reduction in a margin of safety.

Duke has concluded based on the above that there are no significant hazards considerations involved in this request.

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

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 17th Street, NW., Washington, DC 20005.

NRC Section Chief: Richard L. Emch, Jr.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: December 1, 1999.

Description of amendment request: The proposed amendment would revise the standard by which GPU Nuclear tests charcoal used in engineered safeguards features (ESF) systems to American Society for Testing and Materials D3803-1989. These proposed changes are made in accordance with Generic Letter (GL) 99-02.

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

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

The proposed change is in accordance with NRC guidance in GL 99-02 which states that new testing protocol is more accurate and demanding than older tests. The acceptance criteria for charcoal efficiency has been made more stringent and there is no change to an operating parameter of any system, component or structure. Therefore, the probability of occurrence of the consequences of an accident previously evaluated in the SAR [Safety Analysis Report] will not increase as a result of this change.

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

The proposed change revises the testing standard for activated charcoal efficiency to a more conservative methodology while increasing the acceptance criteria through the application of a safety factor. There is no change to an operating parameter of any system, component, or structure. Therefore, the proposed activity does not create the possibility for an accident or malfunction of a different type than any previously identified in the SAR.

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

The proposed change does not involve a reduction in the margin of safety. The change is primarily administrative, adheres to NRC guidance, and is more conservative than the previously employed standard. The change does not modify an operating parameter of any system, or component structure. Therefore, there is no reduction in the margin of safety.

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

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Acting Section Chief: M. Gamberoni.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee Atomic Power Station, Lincoln County, Maine

Date of amendment request: January 13, 2000.

Description of amendment request: The amendment would add a license condition that requires Maine Yankee Atomic Power Company (MYAPC) to implement and maintain in effect all provisions of the License Termination Plan (LTP). MYAPC submitted the LTP in accordance with 10 CFR 50.82(a)(9) to demonstrate that the remainder of decommissioning activities will be performed in accordance with Title 10 of the Code of Federal Regulations, will not be inimical to the common defense or security or to the health and safety of the public, and will not have a significant effect on the quality of the environment.

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

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

The requested license amendment does not authorize any plant activities beyond that allowed by 10 CFR Chapter I or beyond that considered in the DSAR [Defueled Safety Analysis Report]. The bounding accident described in the DSAR for potential airborne activity is the postulated resin cask drop accident in the Low Level Radioactive Waste Storage Building. This accident is expected to contain more potential airborne activity than can be released from other decommissioning events. The radionuclide distribution assumed for the spent resin cask has more transuranics (the major dose contributor) than the distribution in the components involved in other decommissioning accidents. The accidents considered in the DSAR include: 1) Explosion of Liquid Petroleum Gas (LPG) Leaked from a Front End Loader or Forklift, 2) Explosion of Oxyacetylene During Segmenting of the Reactor Vessel Shelf, 3) Release of Radioactivity from the RCS Decontamination Ion Exchange Resins, 4) Gross Leak During In-Situ Decontamination, 5) Segmentation of RCS Piping with Unremoved Contamination, 6) Fire Involving Contaminated Clothing or Combustible Waste, 7) Loss of Local Airborne Contamination Control During Blasting or Jackhammer Operations, 8) Temporary Loss of Services, 9) Dropping of Contaminated Concrete Rubble, 10) Natural Phenomena and 11) Transportation Accidents. The probabilities and consequences for these accidents are estimated in the basis documentation for DSAR Section 7. No systems, structures, or components that could initiate or be required to mitigate the consequences of an accident are affected by the proposed change in any way not previously evaluated in the DSAR. Since Maine Yankee does not exceed the salient parameters associated with the plant referenced in the basis documentation in any material respects, it is concluded that these probabilities and consequences are not increased. Therefore, the proposed change to the Maine Yankee License does not involve any increase in the probability or consequences of any 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.

The requested license amendment does not authorize any plant activities which could precipitate or result in any accidents beyond that considered in the DSAR. The accidents previously evaluated in the DSAR are described above. These accidents are described in the basis documentation for DSAR Section 7. The proposed change does not affect plant systems, structures, or components in any way not previously evaluated in the DSAR. Since Maine Yankee does not exceed the salient parameters associated with the plant referenced in the basis documentation in any material respects, it is concluded that these accidents appropriately bound the kinds of accidents possible during decommissioning. Therefore, the proposed change to the Maine Yankee License would not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The margin of safety defined in Maine Yankee's license basis for the consequences of decommissioning accidents has been established as the margin between the bounding decommissioning accident and the dose limits associated with the need for emergency plan offsite protection, namely the Environmental Protection Agency Protective Action Guidelines EPA-PAGs. As described above, the bounding decommissioning accident is the postulated resin cask drop accident in the Low Level Radioactive Waste Storage Building. Since the bounding decommissioning accident is expected to contain more potential airborne activity than can be released from other decommissioning events and since the radionuclide distribution assumed for the spent resin cask has more transuranics (the major dose contributor) than the distribution in the components involved in other decommissioning accidents, the margin of safety associated with the consequences of decommissioning accidents cannot be reduced. The margin of safety defined in the statements of consideration for the final rule on the Radiological Criteria for License Termination is described as the margin between the 100 mrem/yr public dose limit established in 10 CFR 20.1301 for licensed operation and the 25 mrem/yr dose limit to the average member of the critical group at a site considered acceptable for unrestricted use. This margin of safety accounts for the potential effect of multiple sources of radiation exposure to the critical group. Since the license termination plan was designed to comply with the radiological criteria for license termination for unrestricted use, the margin of safety cannot be reduced. Therefore, the proposed changes to the Maine Yankee License would not involve a significant reduction in any margin of safety.

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

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic Power Company, 321 Old Ferry Road, Wiscasset, Maine 04578.

NRC Section Chief: Michael T. Masnik.

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

Date of amendment request: November 30, 1999, as supplemented on April 28, 2000.

Description of amendment request: The licensee proposes to change the technical specifications (TSs) relating to the emergency diesel generator fuel sampling/testing surveillance requirements (SRs). The changes would provide a new administrative control to establish, implement, and maintain a diesel fuel oil testing program, relocate fuel oil sampling/testing surveillance requirements and fuel oil storage tank cleaning frequency requirement to a new diesel fuel oil testing program which will reside in the Seabrook Station Technical Requirements (SSTR) Manual. The change will also add references to the A.C. Sources—Shutdown surveillance requirement to perform additional activities while in modes 5 and 6.

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 NRC staff's review is presented below:

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

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (SSCs) 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).

The proposed changes do not affect the source term, containment isolation or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated in the Seabrook Station UFSAR. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed change to SR 4.8.1.2 provides additional requirements for operation of the facility. These additional requirements are not initiators of analyzed events and will not alter assumptions relative to mitigation of accident or transient events. The proposed change does not adversely affect previously evaluated accidents.

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

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

The proposed changes do not involve physical alteration of plant SSCs or changes in parameters governing the manner in which the plant is operated and maintained in a state of readiness. The changes do not introduce a new mode of plant operation.

As discussed in the above narrative, the proposed change to SR 4.8.1.2 provides additional requirements for operation of the facility. These additional requirements are not initiators of analyzed events and will not alter assumptions relative to mitigation of accident or transient events. The proposed change does not adversely affect previously evaluated accidents.

Therefore, the proposed changes do 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 do not involve a reduction in a margin of safety because they do not adversely affect assumptions used in transient or safety analyses. The details associated with the involved specifications are not required to be in the TS to provide adequate protection of the public health and safety, since the TS still retains the requirement for compliance with the applicable standards. The level of safety of facility operation is unaffected by the changes since there is no change in the intent of the TS requirements of ensuring fuel oil is of the appropriate quality for diesel generator use.

The proposed change to the A.C. Sources—Shutdown SR imposes an additional level of requirements that are more restrictive than the current TS requirements for operation of the facility in Modes 5 and 6. The additional requirements being proposed enhance assurance that the same fuel oil quality requirements are met, and visual inspection activities conducted, whenever a diesel generator is required to be OPERABLE.

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

Based on this review, it appears that the three standards of 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, Esq., Senior Nuclear Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270.

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 14, 2000.

Description of amendment request: The licensee proposes to relocate Technical Specification (TS) Sections TS 3/4.9.5, “Communications”, TS 3/4.9.6, “Refueling Machine”, and TS 3/4.9.6, “Crane Travel—Spent Fuel Storage Area” to the Seabrook Station Technical Requirements Manual.

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

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

The proposed changes to relocate Technical Specifications 3/4.9.5, 3/4.9.6 and 3/4.9.7 to the Technical Requirements Manual (TRM) are administrative in nature and do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, configuration of the facility or the manner in which it is operated. The proposed changes do not alter or prevent the ability o[f] 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].

The subject specifications relocated to the Technical Requirements Manual will continue to be administratively controlled. The TRM is a licensee-controlled document, which contains certain technical requirements and is the implementing manual for the Technical Specification Improvement Program. Changes to these requirements are reviewed and approved in accordance with Seabrook Station Technical Specification, Section 6.7.1.i, and as outlined in the TRM.

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

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

The proposed change does not alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated. There are no changes to the source term or radiological release assumptions used in evaluating the radiological consequences in the Seabrook Station UFSAR. The proposed change has no adverse impact on component or system interactions. The proposed change will not adversely degrade the ability of systems, structures and components important to safety to perform their safety function nor change the response of any system, structure or component important to safety as described in the Seabrook Station Updated Final Safety Analysis Report (UFSAR). The proposed changes are administrative in nature and do not change the level of programmatic and procedural details of assuring operation of the facility in a safe manner. Since there are no changes to the design assumptions, conditions, configuration of the facility, or the manner in which the plant is operated and surveilled, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

3. 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 Technical Specification required safety limits or safety system settings that would adversely affect plant safety. The proposed change is administrative in nature and does not reduce the level of programmatic or procedural controls associated with the activities presently performed via Technical Specifications 3/4.9.5, 3/4.9.6 and 3/4.9.7.

Future changes to the subject technical requirements will be reviewed and approved in accordance with Seabrook Station Technical Specification, Section 6.7, and as outlined in the Technical Requirements Manual. Specifically, all changes to the Technical Requirements Manual require a 10 CFR 50.59 safety evaluation and will be reviewed and approved by the Station Operations Review Committee (SORC) prior to implementation.

Therefore, relocation of the requirements contained in Technical Specifications 3/4.9.5, 3/4.9.6 and 3/4.9.7 to the Technical Requirements Manual does not involve a significant reduction in the margin of safety provided in the existing specifications.

The NRC staff has reviewed the licensee's analysis, and based on this review, it appears that the three standards of 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, Esq., Senior Nuclear Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270.

NRC Section Chief: James W. Clifford.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: November 24, 1999, as supplemented by letter dated February 10, 2000.

Description of amendment request: The amendment will establish charcoal filter testing requirements in the technical specifications (TSs) for the Auxiliary Building Ventilation (ABV) System, the Control Room Envelope Air Conditioning System (CREACS), and the Fuel Handling Building Ventilation (FHV) System that are consistent with Generic Letter 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal,” dated June 3, 1999.

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 NRC staff's review is presented below:

The operation of the Salem units in accordance with the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed TS changes do not involve any physical changes to plant structures, systems, or components (SSC). The FHV, CREACS, and ABV systems will continue to function as designed. The FHV, CREACS, and ABV systems are designed to mitigate the consequences of an accident. The proposed changes also will not affect the sequence of any accidents previously analyzed. The proposed TS surveillance requirement changes implement testing methods that demonstrate charcoal filter capability and establish acceptance criteria, which ensure that Salem's design basis assumptions continue to be met. The proposed surveillance requirement acceptance criteria ensure that the FHV, CREACS, and ABV safety functions will be accomplished. Therefore, the proposed TS changes would not result in a significant increase of the consequences of an accident previously evaluated, nor do they involve an increase in the probability of an accident previously evaluated.

The operation of the Salem units in accordance with the proposed changes does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed TS changes do not involve any physical changes to the design of any plant SSC. The design and operation of the FHV, CREACS, and ABV systems are not changed from those currently described in Salem's licensing basis. The FHV, CREACS, and ABV systems will continue to function as designed to mitigate the consequences of an accident. Implementing the proposed charcoal filter testing methods and acceptance criteria does not change the operation of the FHV, CREACS, and ABV systems that would create a different type of accident previously evaluated. In addition, the proposed TS changes do not alter the conclusions described in Salem's licensing basis regarding the safety-related functions of these systems. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated.

The operation of the Salem units in accordance with the proposed license amendment will not be changed nor result in a significant reduction to margins of safety. The licensee is not proposing any modifications to FHV, CREACS, and ABV systems design or operation, and there are no changes being made to the TS-required safety limits or safety system settings that would adversely affect plant safety. The proposed changes modify the TSs to reference appropriate test parameters for performing laboratory testing of nuclear-grade charcoal in engineered safety feature filtration systems in accordance with ASTM D3803-1989. The imposition of the more conservative charcoal filter testing requirements associated with ASTM D3803-1989 will not involve a significant 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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

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 amendment requests: April 20, 2000 (PCN-503).

Description of amendment requests: The amendment application proposes to revise the San Onofre Nuclear Generating Station, Units 2 and 3, Technical Specification (TS) 5.5.2.5, “Reactor Coolant Pump Flywheel Inspection Program.” The proposed change would revise the required volumetric examination frequency of the upper flywheel on each of the primary reactor coolant pump motors from a 3-year to a 10-year cycle.

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

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

Response: No.

Missile generation from a Reactor Coolant Pump (RCP) flywheel could damage the reactor coolant system, the containment, or other equipment or systems important to safety. The fracture mechanics analysis performed to support the change shows that a preexisting flaw of an initial size at the detection threshold level will not grow to a flaw size necessary to create flywheel missiles within the life of the plant. The fracture mechanics analysis conservatively assumes minimum material toughness properties, maximum flywheel speed, location of flaw in the highest stress region of the flywheel, and a number of start/stop cycles eight times greater than the design basis. Therefore, an existing flaw in the flywheel will not grow to a size that exceeds the allowable flaw size for either normal operating or accident conditions over the plant life. On this basis, the extension of the 3-year interval inspection to a 10-year interval will not involve a significant increase in the probability of an accident previously considered. The proposed changes do not increase the amount of radioactive material available for release or modify any systems used for preventing or mitigating such releases during accident conditions. Therefore, these changes do not involve a significant increase in the consequences of any accident previously evaluated.

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

Response: No.

The proposed changes will not change the design configuration, or method of operation of the plant. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any previously evaluated.

3. Involve a significant reduction in a margin of safety?

Response: No.

Significant conservatisms have been used in the calculation of allowable flaw size (critical flaw size) and flaw growth for each RCP flywheel design. These include minimum fracture toughness properties, code reference crack growth rate curves, maximum flywheel accident speed, postulated flaw location at the highest stress region of the flywheel, and a number of start/stop cycles that is eight times the number expected in a plant life. The final flaw size has been determined to remain smaller than the allowable flaw size for the flywheel under the relevant design conditions, including postulated accident conditions. Therefore, the extension of the 3-year interval inspection to a 10-year interval will not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

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: February 4, 2000 (TS 99-14).

Brief description of amendments: The proposed amendments would change the Sequoyah Nuclear Plant (SQN) Technical Specification Limiting Conditions for Operations for the reactor coolant system cold leg accumulators (CLAs). The upper CLA water limit and required pressure range would both be decreased to more appropriately account for instrument uncertainties and instrument line tap locations.

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

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

The capability of the accumulators to perform their safety function is not affected by this change. All components and system functional requirements remain the same. There are no new sequences of events which would increase the probability of an accident analyzed in the Final Safety Analysis Report (FSAR). Therefore, the proposed activity does not increase the probability of an accident previously evaluated in the FSAR. The fuel cladding peak temperature established by the ECCS [Emergency Core Cooling System] evaluation model remains below 2200 degrees Fahrenheit for a loss-of-coolant accident (LOCA). As such, the assumptions on fuel failure and isotope release post-LOCA do not change from the information presented in the FSAR.

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

The primary function of the CLAs is in the event of a large break LOCA to support accident mitigation. CLAs are not a contributor to events that could generate accidents. The CLA system volume capability bounds this change in operational limits and the system is not physically changing. Therefore, the proposed activity does not create a possibility for an accident of a different type than any evaluated previously.

C. The proposed amendment does not involve a significant reduction in a margin of safety.

The safety function provided by the CLAs is to inject core cooling water into the reactor coolant system when system pressure decreases below a predetermined value during a LOCA. The timing (function of pressure) and amount (function of volume) of cooling water is modeled in the ECCS evaluation model. The proposed changes to the accumulator operational limits have been evaluated using the Sequoyah plant specific ECCS model. The evaluation shows an increase in the peak fuel cladding temperature from 2162 degrees Fahrenheit to 2185 degrees Fahrenheit. The results confirm that existing LOCA safety analysis acceptance criteria (established by 10 CFR 50.46) continue to be met for the revised accumulator limits. The safety analysis acceptance criteria continues to be met with the revised limits. The 23 degree increase in the peak fuel cladding temperature associated with accumulator operation is not a significant reduction in the margin of safety.

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Richard P. Correia.

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

Date of amendment request: March 6, 2000 (TS 99-09).

Description of amendment request: The proposed amendment would revise the Watts Bar Nuclear Plant (WBN) Unit 1 Technical Specifications (TS) and associated TS Bases for Limiting Condition for Operation (LCO) 3.9.4 Containment Penetrations. The revision would permit both doors of the containment personnel airlocks to be open during refueling operations to facilitate personnel and equipment access to containment. It would also allow containment penetration flow paths to be open under administrative controls to facilitate maintenance activities during refueling operations.

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

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

The proposed change to WBN Technical Specification LCO 3.9.4, Refueling Operations—Containment Penetrations, would allow both doors of the containment personnel airlocks and certain containment penetration flow paths to be open during core alterations and movement of irradiated fuel within containment under specific administrative controls. The proposed change is consistent with NRC approved TS travelers TSTF-68, R2 and TSTF-312, R1, and proposes controls similar to the administrative controls currently allowed by WBN TS (LCO 3.6.3) for containment penetrations during more restrictive, higher operational modes. The administrative controls will ensure appropriate personnel are aware of the open personnel airlocks and penetration flow paths and ensure designated individual(s) are assigned to promptly close the airlock doors and penetration flow paths in the event of a fuel handling accident (FHA) inside containment. Timely closure of penetration flow paths and closure of the airlock doors following containment evacuation will ensure that the unlikely transmission of radioactive material from the reactor building to the auxiliary building is minimized.

In order to minimize the consequences of any leakage of radionuclides past these open penetrations during the period of time before their closure, additional procedural controls will be provided to ensure the integrity of the WBN auxiliary building secondary containment enclosure (ABSCE) boundary and proper auxiliary building gas treatment system (ABGTS) operation. These controls will ensure that in the event of a fuel handling accident (FHA) inside containment, the following will be promptly accomplished: shutdown and isolation of the reactor building purge air ventilation system, auxiliary building isolation, and initiation of ABGTS. Therefore, through the use of these controls for the proposed license amendment, the offsite dose consequences of a FHA inside containment with open airlock doors and/or open penetration flow paths remain well within the 10 CFR 100 limits and within the limits of 10 CFR 50, Appendix A, General Design Criteria 19 for control room operator dose.

[The licensee's application also states that “The results for the fuel handling analysis inside containment with open airlock doors and/or open penetration flow paths are bounded by the current analysis.”]

The containment personnel airlock doors and containment penetration flow paths are not initiators to any previously evaluated accident for WBN. In addition, the position of the airlock doors and penetration flow paths during refueling operations has no affect on the probability of the occurrence of any accident previously evaluated. The proposed revision does not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. Since the probability of a accident is not affected by the positions of the containment personnel airlock doors, and because the doses remain within acceptable limits, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The open containment personnel airlock doors and containment penetration flow paths are not accident initiators and do not represent a significant change in the configuration of the plant. The proposed allowance to open the containment personnel airlock doors and penetrations during refueling operations will not adversely affect plant safety functions or equipment operating practices such that a new or different accident could be created. Therefore, since plant safety functions are not adversely affected and the isolation status of containment personnel airlock doors and penetration flow paths do not contribute to the initiation of postulated accidents, the proposed revision will not create a new or different kind of accident from any accident previously evaluated.

C. The proposed amendment does not involve a significant reduction in a margin of safety.

WBN Technical Specification LCO 3.9.4 closure requirements for containment penetrations ensure that the consequences of a postulated FHA inside containment during core alterations or fuel handling activities remain within acceptable limits. The LCO establishes containment closure requirements, which limit the potential escape paths for fission products by ensuring that there is at least one integral barrier to the release of radioactive material. The proposed change to allow the containment personnel airlock doors and containment penetration flow paths to be open during refueling operations under administrative controls does not significantly affect the expected dose consequences of a FHA because of the absence of containment pressurization during refueling. Without this motive force, the potential for additional offsite dose consequence is unlikely. The proposed administrative controls provide assurance that prompt closure of the airlock doors and penetration flow paths will be accomplished in the event of a FHA inside containment thus minimizing the transmission of radioactive material from the reactor building to the auxiliary building. Under the proposed TS change, the provisions to ensure shutdown and isolation of the reactor building purge air ventilation system, auxiliary building isolation, and initiation of ABGTS and to promptly isolate open penetration flow paths and close the airlock doors following containment evacuation, provide assurance that the offsite dose consequences of a FHA inside containment will remain well within the 10 CFR 100 limits and within the limits of 10 CFR 50, Appendix A, General Design Criteria 19 for control room operator dose. Therefore, the proposed change to the WBN Technical Specifications does not involve a significant reduction in the margin of safety.

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Richard P. Correia.

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

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

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

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: September 16, 1999.

Brief description of amendment: The proposed amendment would increase the licensed capacity for spent fuel assembly storage in the Spent Fuel Pool and revise the configuration for storage of fresh fuel.

Date of publication of individual notice in the Federal Register: December 8, 1999 (64 FR 68702).

Expiration date of individual notice: January 7, 2000.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva Power and Light Company, and Atlantic City Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of amendment request: February 29, 2000, as supplemented on March 31, 2000.

Brief description of amendment request: The amendments would add a note to the completion time of Condition A for Technical Specification 3.7.2, “Emergency Service Water (ESW) System and Normal Heat Sink.” This note would provide a one-time extension to the completion time for one ESW subsystem inoperable from 7 to 14 days. This note would allow the replacement of one ESW pump currently scheduled to occur in May 2000 and will expire on May 31, 2000.

Date of publication of individual notice in Federal Register: March 9, 2000 (65 FR 12589).

Expiration date of individual notice: April 10, 2000.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of amendment request: March 14, 2000, as supplemented March 27, 2000.

Brief description of amendment request: The proposed amendment would amend the licenses to change the required implementation date for previously issued license Amendment No. 184 to Facility Operating License NPF-14 and Amendment No. 158 to Facility Operating License NPF-22. The proposed amendment would not alter any of the requirements of the SSES Unit 1 and 2 Technical Specifications (TSs).

Date of publication of individual notice in Federal Register: April 27, 2000 (65 FR 24718).

Expiration date of individual notice: May 30, 2000.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of amendment request: April 10, 2000.

Brief description of amendment request: Permits deferral of testing of primary containment penetration flange o-rings on spectacle flanges 2S299A and 2S299B until the Unit 2 10th refueling outage, scheduled for spring 2001 or a prior Unit 2 outage requiring entry into Mode 4.

Date of publication of individual notice in Federal Register: April 21, 2000 (65 FR 21487).

Expiration date of individual notice: May 22, 2000.

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, the Gelman Building, 2120 L Street, NW., Washington, DC, and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Date of application for amendment: October 25, 1999 (U-603281).

Brief description of amendment: The amendment revised the Technical Specification definitions for channel calibrations, channel functional tests, and logic system functional tests.

Date of issuance: April 25, 2000.

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

Amendment No.: 128.

Facility Operating License No. NPF-62: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 12, 2000 (65 FR 1920).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 25, 2000.

No significant hazards consideration comments received: No.

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: November 19, 1999, as supplemented on March 16, 2000.

Brief description of amendment: This amendment revises the Technical Specifications (TS) to incorporate the American Society for Testing and Materials (ASTM) D3803-1989, Standard Test Method for Nuclear-Grade Activated Carbon,” in accordance with NRC Generic Letter (GL) 99-02, “Laboratory Testing Of Nuclear-Grade Activated Charcoal,” dated June 3, 1999. Specifically, TS 4.7.6 has been revised for the Control Room Emergency Filtration System, TS 4.7.7 has been revised for the Reactor Auxiliary Building Emergency Exhaust System, and TS 4.9.12 has been revised for the Fuel Handling Building Emergency Exhaust System.

Date of issuance: May 2, 2000.

Effective date: May 2, 2000.

Amendment No. 98.

Facility Operating License No. NPF-63. Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: December 15, 1999 (64 FR 70081).

The March 16, 2000, submittal contained clarifying information only, and did not change the initial no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 2, 2000. No significant hazards consideration comments received: Yes. One comment was received, and is addressed in the above-referenced Safety Evaluation.

Consolidated Edison Company of New York, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: March 17, 2000.

Brief description of amendment: The amendment revises Technical Specifications (TSs) associated with probes used in steam generator tube inspections, specifically TS Section 4.13.A.3.f. The proposed change would provide more flexibility in the type of probe used and would reflect current technological advances in inspection equipment, while still maintaining the current 610-mil diameter probe restriction. ]

Date of issuance: April 28, 2000.

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

Amendment No.: 209.

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

Date of initial notice in Federal Register: March 27, 2000 (65 FR 16230).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 28, 2000.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas Nuclear One, Units 1 and 2, Pope County, Arkansas

Date of amendment request: November 16, 1999.

Brief description of amendments: The proposed changes to the Arkansas Nuclear One, Units 1 and 2, Technical Specifications (TSs) and associated Bases provided a 30-day allowed outage time (AOT) for startup transformer No. 2, which is an offsite power source shared by both units. This 30-day AOT will be used infrequently for the purpose of performing preventative maintenance to increase the reliability of the transformer. In addition, changes have been made to the requirements associated with demonstrating the operability of the emergency diesel generators (EDGs), in the event a required power source is inoperable, to increase the reliability of the EDGs.

Date of issuance: April 28, 2000.

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

Amendment Nos.: 206 and 215.

Facility Operating License Nos. DPR-51 and NPF-6: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 26, 2000 (65 FR 4271).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 28, 2000.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas

Date of application for amendment: February 24, 2000.

Brief description of amendment: The amendment revised Technical Specification (TS) 4.4.11 on reactor coolant system vent flow verification, TS 4.6.1.1.a on containment penetration closure verification (non-automatic), and TS 4.6.3.1.2 on containment isolation valve actuation verification. The changes eliminated unnecessary mode restrictions on these surveillance requirements.

Date of issuance: April 26, 2000.

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

Amendment No.: 214.

Facility Operating License No. NPF-6: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 22, 2000 (65 FR 15379).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 26, 2000.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 6, 1998, as supplemented by letter dated March 3, 2000, Moderator Temperature Coefficient test near the end of each cycle.

Brief description of amendment: The proposed change modifies the requirement to perform a Moderator Temperature Coefficient test near the end of each cycle.

Date of issuance: April 21, 2000.

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

Amendment No.: 159.

Facility Operating License No. NPF-38: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 25, 1999 (64 FR 46435). The March 3, 2000, letter did not change the scope of the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 21, 2000.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 15, 1999, as supplemented by letter dated January 6, 2000.

Brief description of amendment: The proposed change modifies plant technical specifications to extend the Reactor Coolant System Pressure Temperature Curve Limit to 16 Effective Full Power Years.

Date of issuance: April 24, 2000.

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

Amendment No.: 160.

Facility Operating License No. NPF-38: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 26, 2000 (65 FR 4276).

The January 6, 2000, letter reduced EFPY from 20 years, requested in the July 15, 1999, letter, to 16 years. This change is bounded by, and did not change the scope of, the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 24, 2000.

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, Shippingport, Pennsylvania

Date of application for amendments: November 23, 1999, as supplemented February 22, 2000.

Brief description of amendments: The amendments make the following changes to the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2) Technical Specifications (TSs): (1) For BVPS-1, surveillance requirement (SR) 4.8.1.1.2.b.3.b is revised to reflect a narrower required diesel generator (DG) frequency band; an associated footnote is deleted; associated Bases are revised to reflect these TS changes. (2) For BVPS-2, SR 4.8.1.1.2.f is revised to clarify that the DGs are only required to achieve a minimum frequency and voltage within the first 10 seconds of the related test, and that the stated voltage and frequency bands are requirements for steady state operation of the DGs; a footnote is also added to this SR. (3) Page formats are revised as needed to permit the addition or deletion of text.

Date of issuance: April 25, 2000.

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

Amendment Nos.: 230 and 109.

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

Date of initial notice in Federal Register: March 8, 2000 (65 FR 12292). The February 22, 2000, letter provided supplemental information and did not change the initial proposed no significant hazards consideration determination or expand the amendments beyond the scope of the initial notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 25, 2000.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 22, 1999, as supplemented April 27, 2000.

Brief description of amendment: The amendment allowed a one-time only extension to the surveillance interval of the Technical Specification Surveillance 4.7.12.d for functional testing of snubbers. The extension is limited to the first re-entry into MODE 6 following the defueled condition during the 8th refueling outage or November 30, 2000, whichever occurs sooner.

Date of issuance: May 3, 2000.

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

Amendment No.: 110.

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

Date of initial notice in Federal Register: November 17, 1999, (64 FR 62711).

The April 27, 2000, letter did not change the initial proposed no significant hazards consideration determination or expand the amendment beyond the scope of the initial notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 3, 2000.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 2, 1999.

Brief description of amendment: This amendment revised the Technical Specifications (TSs) to modify 1) TS Table 3.3-4, “Safety Features Actuation System Instrumentation Trip Setpoints,” to remove the “Trip Setpoint” values for Instrument String Functional Unit “f”, Borated Water Storage Tank (BWST) Level, 2) the “Allowable Values” entry for this same Functional Unit, consistent with updated calculations using current setpoint methodology, 3) TS 3/4.3.2.1, “Safety Features Actuation System Instrumentation,” and Bases to reflect the removal of “Trip Setpoints” described above, and 4) TS 3/4.5.4, “Emergency Core Cooling Systems—Borated Water Storage Tank,” and Bases to increase the minimum volume of water in the BWST.

Date of issuance: May 4, 2000.

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

Amendment No.: 241.

Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 15, 1999 (64 FR 70087).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 4, 2000.

No significant hazards consideration comments received: No.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Date of application for amendment: October 23, 1998, as supplemented February 22 and June 24, 1999, and March 31, 2000.

Brief description of amendment: The amendment would allow implementation of a feedwater leakage control system to address leakage through the primary containment feedwater penetration valve.

Date of issuance: April 25, 2000.

Effective date: April 25, 2000.

Amendment No.: 127.

Facility Operating License No. NPF-62: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63 FR 64118).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 25, 2000.

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: February 18, 2000, as supplemented March 31, 2000.

Brief description of amendments: The proposed license amendments would approve a change to the facility involving an unreviewed safety question discovered by the licensee during a 10 CFR 50.59 evaluation of modifications to the auxiliary feedwater (AFW) pump rooms to protect the equipment in the rooms from the environmental effects of a postulated high-energy line break. This will be accomplished by sealing the AFW pump rooms to ensure that the rooms do not communicate with the turbine buildings or each other.

Date of issuance: April 25, 2000.

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

Amendment Nos.: 244.

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

Date of initial notice in Federal Register: February 25, 2000 (65 FR 10116).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 25, 2000.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 13, 1999.

Description of amendment request: The amendment changes the license to delete expired license conditions and to make editorial and administrative changes to correct or clarify the license.

Date of issuance: April 27, 2000.

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

Amendment No.: 68.

Facility Operating License No. NPF-86: Amendment revised the License.

Date of initial notice in Federal Register: February 9, 2000 (65 FR 6408).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 27, 2000.

No significant hazards consideration comment received: No.

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

Date of amendment request: February 18, 2000.

Description of amendment request: The amendment revises Technical Specifications (TSs) Surveillance Requirements 4.0.5.a, 4.0.5.b, 4.0.5.e, and 4.4.6.2.2.e. These changes are required to ensure consistency between the TSs and the second 10-year inservice test program by approval to use the 1995 Edition and 1996 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). The revision to TSs Surveillance Requirement 4.0.5.a also incorporates semi-quarterly and biennial intervals to the list of required frequencies for performing inservice test and inspection activities.

Date of issuance: May 8, 2000.

Effective date: As of its date of issuance, and shall be implemented by August 18, 2000.

Amendment No.: 69.

Facility Operating License No. NPF-86: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 5, 2000 (65 FR 17917).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 8, 2000.

No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: December 14, 1999, as supplemented February 11, March 30, and April 26, 2000.

Brief description of amendment: This amendment will revise Technical Specifications (TSs) Sections: 3.3.2.1, “Instrumentation—Engineered Safety Feature Actuation System Instrumentation;” 3.3.3.1, “Instrumentation—Monitoring Instrumentation—Radiation Monitoring;” 3.7.6.1, “Plant Systems—Control Room Emergency Ventilation System;” 3.9.3.1, “Refueling Operations—Decay Time;” 3.9.4, “Refueling Operations—Containment Penetrations;” 3.9.9, “Refueling Operations—Containment Radiation Monitoring;” 3.9.10 “Refueling Operations—Containment Purge Valve Isolation System;” 3.9.13, “Refueling Operations—Storage Pool Radiation Monitoring;” 3.9.14, “Refueling Operations—Storage Pool Area Ventilation System—Fuel Movement;” 3.9.15, “Refueling Operations—Storage Pool Area Ventilation System—Fuel Storage;” 3.9.16.1, “Refueling Operations—Shielded Cask;” 3.9.16.2, “Refueling Operations—Shielded Cask;” 3.9.17, “Refueling Operations—Movement of Fuel in Spent Fuel Pool;” and 3.9.19.2, “Refueling Operations—Spent Fuel Pool—Storage Pattern”; and add new TS 3.3.4, “Containment Purge Valve Isolation Signal.” The requested changes would make the TSs and the Final Safety Analysis Report (FSAR) consistent with the new analyses of the fuel handling and cask drop accidents. The Index Pages and the Bases for these TSs will be modified to reflect these proposed changes.

Date of issuance: April 28, 2000.

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

Amendment No.: 245.

Facility Operating License No. DPR-65: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 17, 2000 (65 FR 14632).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 28, 2000.

No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: June 15, 1999, as supplemented July 20, September 3, and November 29, 1999, and January 18, 2000.

Brief description of amendment: The amendment modifies the license to change the number of owners from 14 to 13 and to remove Montaup Electric Company as an owner as a result of the transfer of its interest in Millstone Nuclear Power Station, Unit No. 3 to New England Power Company, an existing owner.

Date of issuance: May 1, 2000.

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

Amendment No.: 180.

Facility Operating License No. NPF-49: Amendment revised the License.

Date of initial notice in Federal Register: January 19, 2000 (65 FR 2990).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 24, 2000, issued with the February 24, 2000, Order approving the transfer as noticed in the Federal Register on March 1, 2000 (65 FR 11091).

No significant hazards consideration comments received: No.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva Power and Light Company, and Atlantic City Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: February 29, 2000, as supplemented on March 31, 2000.

Brief description of amendments: The amendments will add a note to the completion time of Condition A for Technical Specification 3.7.2, “Emergency Service Water (ESW) System and Normal Heat Sink.” This note will provide a one-time extension to the completion time for one ESW subsystem inoperable from 7 to 14 days. This note will allow the replacement of one ESW pump currently scheduled to occur in May 2000 and will expire on May 31, 2000.

Date of issuance: April 25, 2000.

Effective date: Both units, as of the date of issuance and shall be implemented no later than May 31, 2000.

Amendments Nos.: 231 and 236.

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

Date of initial notice in Federal Register: March 9, 2000 (65 FR 12589).

The March 31, 2000, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 25, 2000.

No significant hazards consideration comments received: No.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of application for amendment: April 10, 2000.

Brief description of amendment: The amendment adds a note to Technical Specification Surveillance Requirement 3.6.1.1.1 to defer performance of this test on a one-time basis for spectacle flanges 2S299A and 2S299B o-rings until the Unit 2 10th Refueling Outage (Spring 2001) or a prior Unit 2 outage requiring entry into Mode 4. The change allowed Unit 2 operation to continue until an outage occurs where leak rate surveillance testing on spectacle flanges 2S299A and 2S299B can be performed.

Date of issuance: May 8, 2000.

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

Amendment No.: 160.

Facility Operating License No. NPF-22. This amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 21, 2000 (65 FR 21487).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 8, 2000.

No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of application for amendment: March 15, 2000.

Brief description of amendment: This amendment changes Technical Specification (TS) Definition 1.7, CORE ALTERATION. The definition has been revised to be similar to the definition of CORE ALTERATION that is documented in NUREG-1433, Revision 1, “Standard Technical Specifications, General Electric Plants, BWR/4.”

Date of issuance: April 25, 2000.

Effective date: As of the date of issuance, and shall be implemented within 3 days.

Amendment No.: 125.

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

Date of initial notice in Federal Register: March 23, 2000 (65 FR 15657).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 25, 2000.

No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of application for amendment: September 30, 1999, as supplemented March 27, 2000.

Brief description of amendment: This amendment revised the Technical Specifications (TSs) associated with the Safety Limit Minimum Critical Power Ratios in order to support the operation of Hope Creek Generating Station in the upcoming Cycle 10 with a mixed core of General Electric (GE) and Asea Brown Bovieri/Combustion Engineering (ABB/CE) fuel. In addition, administrative changes have been made to the TSs to reflect the change in fuel vendor from GE to ABB/CE.

Date of issuance: May 1, 2000.

Effective date: As of the date of issuance, and shall be implemented within 60 days after completion of Cycle 9.

Amendment No.: 126.

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

Date of initial notice in Federal Register: November 3, 1999 (64 FR 59805).

The March 27, 2000 letter 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 1, 2000.

No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: July 20, 1999, as supplemented on October 25, 1999.

Brief description of amendment: The amendment revises Technical Specifications to reflect the implementation of increased core flow.

Date of Issuance: April 25, 2000.

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

Amendment No.: 187.

Facility Operating License No. DPR-28: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 25, 1999 (64 FR 46450). The October 25, 1999, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the original Federal Register notice.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 25, 2000.

No significant hazards consideration comments received: No.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: March 2, 2000.

Brief description of amendment: The amendment increases the minimum refueling boron concentration to 2200 parts per million (ppm) from 2100 ppm as specified in the Technical Specification 3.8.a.5.

Date of issuance: May 1, 2000.

Effective date: Immediately upon its date of issuance and is to be implemented within 30 days of the date of issuance.

Amendment No.: 147.

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

Date of initial notice in Federal Register: March 30, 2000 (65 FR 16969).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 1, 2000.

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.

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, the Gelman Building, 2120 L Street, NW., Washington, DC, and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By June 16, 2000, 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, the Gelman Building, 2120 L Street, NW., Washington, DC 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.

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 Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 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 the factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear Station, Unit 2, York County, South Carolina

Date of application for amendments: May 3, 2000.

Brief description of amendments: The amendment revised the Technical Specifications (TS) and associated Bases Section 3.6.9 for the Hydrogen Ignition System. Specifically, the proposed amendment modifies Surveillance Requirements (SRs) 3.6.9.1, 3.6.9.2, and 3.6.9.3 to exclude the two hydrogen ignitors located beneath the reactor vessel missile shield from the applicability of the SRs. These two ignitors are presently considered to be inoperable at Unit 2 and cannot be accessed for replacement with the unit in its current operating mode (Mode 1). This change is effective for Unit 2 Cycle 11 only, or until such time that the unit enters Mode 5 (cold shutdown) such that the inoperable ignitors can be accessed for replacement.

Date of issuance: May 5, 2000.

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

Amendment No.: 178.

Facility Operating License No. NPF-52: Amendment revised the Technical Specifications and associated Bases.

Public Comments requested as to proposed no significant hazards consideration: No.

The Commission's related evaluation of the amendment, finding of emergency circumstances, consultation with the State of South Carolina, and final no significant hazards consideration determination are contained in a Safety Evaluation dated May 5, 2000.

Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

NRC Section Chief: Richard L. Emch, Jr.

Dated at Rockville, Maryland, this 10th day of May 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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