[Federal Register Volume 64, Number 183 (Wednesday, September 22, 1999)] [Notices] [Pages 51343-51356] From the Federal Register Online via the Government Publishing Office [www.gpo.gov] [FR Doc No: 99-24573] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION 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 August 28, 1999, through September 10, 1999. The last biweekly notice was published on September 8, 1999 (64 FR 48858). Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) [[Page 51344]] involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently. Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administration 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 October 22, 1999, 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 at the local public document room for the particular facility involved. 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: 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 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 [[Page 51345]] amendment which is available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at the local public document room for the particular facility involved. Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland Date of amendments request: August 27, 1999. Description of amendments request: The proposed amendment revises Technical Specification (TS) 3.7.13, ``Spent Fuel Pool (SFP) Water Level'' to allow placement of one or more fuel assemblies on SFP rack spacers to support fuel reconstitution activities while irradiated fuel assembly movement continues in the SFP. Although the plant TSs do not prohibit fuel reconstitution, the effect of the current wording of TS 3.7.13, in conjunction with the specific design of the SFP and storage racks, limits reconstituting only one fuel assembly at a time and only when no irradiated fuel assembly movement occurs in the SFP. Specifically, the proposed change adds a new statement to the limiting condition for operation that would require the water level over fuel assemblies placed on rack spacers to be 19.8 feet while irradiated fuel assemblies are being moved in the SFP. The proposed administrative controls will ensure that the current design basis fuel handling accident described in the Updated Final Safety Analysis Report (UFSAR) bounds a fuel handling accident associated with reconstitution activities. 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. Would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change will require a minimum water level of 19.8 feet over fuel assemblies that are placed on rack spacers for fuel reconstitution activities while fuel movement continues in the SFP. This proposed change does not cause any spent fuel handling equipment to be operated in a new or different manner. No structural changes or modifications are being made to the spent fuel handling machine (SFHM) or to the spent fuel storage racks. Administrative controls will be put in place to ensure that the SFHM or an assembly being carried by the SFHM will not strike assemblies placed on rack spacers. This proposed change does not make any changes to equipment, procedures, or processes that increase the likelihood of dropping the fuel assembly from the SFHM. Administrative controls will be put in place to limit the movement of heavy loads such that only a single-failure-proof crane will be used in the area of the affected fuel assembly and the adjacent storage rack cells when the assemblies are seated on rack spacers with their upper end fittings removed. Therefore, this proposed change does not involve a significant increase in the probability of an accident previously evaluated. A Fuel Handling Incident (FHI) during reconstitution activities is bounded by those previously analyzed and described in the Updated Final Safety Analysis Report (UFSAR) for the limiting FHI. The number of fuel pins that could be ruptured in a raised fuel assembly does not exceed that previously analyzed. Also, by requiring that reconstitution activities do not occur until 10 days after shutdown ensures that a[n] FHI during these activities will be bounded by the most limiting FHI described in the UFSAR. Therefore, the proposed change does not significantly increase the consequences of an accident previously evaluated. Based on the above, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Would not create the possibility of a new or different type of accident from any accident previously evaluated. The proposed change will not make any physical changes to the plant. Specifically, no modifications will be made to the SFHM, the spent fuel storage racks, or the spent fuel assemblies. No changes are made to the operation of the SFHM. The only change made by this activity is that multiple fuel assemblies may be placed on rack spacers in the SFP for reconstitution activities. Administrative controls will be put in place to ensure that this proposed change does not create the potential of a[n] FHI during reconstitution activities that is not bounded by our current accident analysis. This proposed change does not have any impact on the cooling or safe geometry functions of the SFP storage racks. This proposed change does not create any new interactions between any plant components. Therefore, the possibility of a new or different type of accident is not created by this proposed change. 3. Would not involve a significant reduction in a margin of safety. The Technical Specification requires a minimum water level to be maintained above the fuel assemblies stored in the SFP storage racks to ensure that sufficient water depth is available to remove the assembled iodine gap activity released from the rupture of an irradiated fuel assembly. The proposed change will allow multiple fuel assemblies to be placed on rack spacers for fuel reconstitution activities while fuel movement continues in the spent fuel pool. These activities will reduce the amount of water maintained above the fuel assemblies that are placed on rack spacers. However, the proposed change does not involve a significant reduction in a margin of safety based on the administrative controls that require an increase in the decay time before these activities can be started. Additional administrative controls will be put in place that include, in part, restricting load movements over the affected fuel assembly and the adjacent storage rack cells, as well as controlling the SFHM. The administrative controls will ensure that the FHI associated with reconstitution activities is bounded by the current design basis FHI described in the UFSAR. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Calvert County Library, Prince Frederick, Maryland 20678. Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW, Washington, DC 20037. NRC Section Chief: S. Singh Bajwa. Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida Date of amendment request: August 18, 1999. Description of amendment request: The proposed amendment will change the required surveillance interval for cycling the steam valves in the turbine overspeed protection system from monthly to quarterly. The license requirement is documented in the St. Lucie, Unit 2 Updated Final Safety Analysis Report (UFSAR) Section 13.7.1.6.2, and the proposed change does not satisfy the 10 CFR 50.59 standards for a change that can be made by the licensee without prior Commission approval. 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 of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The small increase in turbine missile ejection frequency resulting from extending the test interval for turbine valves is acceptable with respect to the NRC probabilistic acceptance criterion and supports quarterly testing. In addition, there are no physical changes to plant equipment or changes in plant operation that could initiate or adversely affect the mitigation or [[Page 51346]] consequences of an accident previously evaluated. Turbine disk integrity remains unchanged since the turbine rotor inspection cycle is not affected by the change in valve testing frequency. Further, there are no changes to protective barriers or changes in separation of equipment important to safety. Therefore, safety related structures, systems, and components remain adequately protected against potential turbine missiles and the potential for turbine missile generation has not significantly increased. The change to extend the turbine valve test interval maintains the intent and design basis function being verified by the surveillance requirement. Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no physical changes to plant equipment or changes in plant operation that could create a new or different kind of accident. This proposed change does not result in any plant configuration changes or create new failure modes. The small increase in turbine missile ejection frequency resulting from extending the test interval for turbine valves is acceptable with respect to the NRC probabilistic acceptance criterion and supports quarterly testing. New types of turbine missiles or strike probabilities are not created by extending the turbine valve test interval. No new or different kind of accident is created. In addition, turbine disk integrity remains unchanged since the turbine rotor inspection cycle is not affected by the change in valve testing frequency. Further, there are no changes to protective barriers or changes in the separation of equipment important to safety. Safety related structures, systems, and components remain adequately protected against potential turbine missiles, the potential for turbine missile generation has not significantly increased, and new or different kinds of accidents are not created. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. This proposed surveillance change extends the turbine overspeed protection system turbine valve test frequency from monthly to quarterly. The results of turbine missile ejection frequency remain within NRC acceptance criterion and therefore supports quarterly testing. There are no physical changes to plant equipment or changes in plant operation that involve a significant reduction in the margin of safety. Turbine disk integrity remains unchanged since the turbine rotor inspection cycle is not affected by the change in valve testing frequency. There are no changes to protective barriers or changes in separation of equipment important to safety. Therefore, safety related structures, systems, and components remain adequately protected against potential turbine missiles and the potential for turbine missile generation has not significantly increased. The change in turbine valve test interval maintains the intent and design basis function being verified by the surveillance requirement. As such, the assumptions and conclusions of the accident analyses in the UFSAR remain valid and associated safety limits will continue to be met. Therefore, operation of the facility in accordance with the proposed amendment would 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003. Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420. NRC Section Chief: Sheri R. Peterson. GPU Nuclear Inc., Docket No. 50-320, Three Mile Island--Unit 2 (TMI-2), Dauphin County, Pennsylvania Date of amendment request: June 29, 1999, as supplemented August 27, 1999 (LAR No. 77). Description of amendment request: The proposed amendment would grant authority for the licensee to possess limited amounts and types of radioactive materials without unit distinction so that after the sale and transfer of the Three Mile Island--Unit 1 (TMI-1) license to AmerGen, radioactive materials may continue to be moved between the TMI-1 and TMI-2 units. After the license transfer, GPU Nuclear will need to access the waste handling and processing facilities at TMI-1 (currently common facilities) for its normal post-defueling monitored storage (PDMS) activities. Similarly, AmerGen as the TMI-1 licensee and PDMS contractor, will need to move radioactive apparatus and materials between units, principally during TMI-1 outages. The amendment would not authorize receipt or possession of radioactive material or waste from other sites. 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 changes would not involve a significant increase in the probability of an accident previously evaluated because no accident initiators or assumptions are affected. The proposed changes have no effect on any plant systems. All Limiting Conditions for PDMS and Safety Limits specified in the Technical Specifications will remain unchanged. [The proposed changes would] not involve a significant increase in the consequences of an accident previously evaluated because no accident conditions or assumptions are affected. The proposed changes do not alter the source term, containment isolation, or allowable radiological consequences. The staging of radioactive materials such as the contaminated reactor coolant pump and motor components will not result in a source term, that if released, would exceed that previously analyzed in the PDMS SAR [safety analysis report] in terms of off-site dose consequences. The proposed changes have no adverse effect on any plant system. 2. [The proposed changes would] not create the possibility of a new or different kind of accident from any previously evaluated because no new accident initiators or assumptions are introduced by the proposed changes. The proposed changes have no direct effect on any plant system. The changes do not affect any system functional requirements, plant maintenance, or operability requirements. 3. [The proposed changes would] not involve a significant reduction in the margin of safety because the proposed changes do not involve significant changes to the initial conditions contributing to accident severity or consequences. The proposed changes have no direct effect on any plant systems. 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. Local Public Document Room location: Law/Government Publications Section, State Library of Pennsylvania, (Regional Depository) Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105. Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Section Chief: Michael T. Masnik. Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois Date of amendment request: August 23, 1999. Description of amendment request: The proposed amendment would delete certain license conditions that are obsolete and no longer apply. Basis for proposed no significant hazards consideration determination: [[Page 51347]] 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 activity does not involve a significant increase in the probability or consequences of any accident previously evaluated. The proposed changes delete various license conditions each of which has been fulfilled and no longer warrants a license condition. As such, the changes are purely administrative in nature, and involve no physical or operational changes to the facility. The initial conditions and methodologies used in the accident analyses consequently remain unchanged. Further, the proposed changes do not change or alter the design assumptions for the systems or components used to mitigate the consequences of an accident. Therefore, accident analyses results are not impacted. On this basis, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. (2) The proposed activity does not create the possibility of a new or different kind of accident from any accident previously evaluated. As noted above, the proposed changes are purely administrative and involve no physical or operational changes to the facility. As such, the proposed changes do not affect the design or operation of any system, structure, or component in the plant. The safety functions of the related structures, systems, or components are not changed in any manner, nor is the reliability or[f] any structures, systems or components reduced. No new or different type of equipment will be installed, and consequently, no new failure modes are introduced. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated. (3) The proposed activity does not involve a significant reduction in the margin of safety. The proposed changes are administrative in nature and have no impact on the margin of safety of any Technical Specification. There is no impact on safety limits or limiting safety system settings. The changes do not affect any plant safety parameters or setpoints. All active/applicable license conditions set forth in the CPS Operating License will remain in effect, and no physical or operational changes to the facility will result from these changes. Therefore, the proposed changes do not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Vespasian Warner Public Library, 120 West Johnson Street, Clinton, IL 61727. Attorney for licensee: Leah Manning Stetzner, Vice President, General Counsel, and Corporate Secretary, 500 South 27th Street, Decatur, IL 62525. NRC Section Chief: Anthony J. Mendiola. Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point Nuclear Station Unit No. 1, Oswego County, New York Date of amendment request: August 26, 1999. Description of amendment request: The proposed amendment would revise the Technical Specifications (TS) to reflect the proposed implementation of Noble Metal Chemical Addition (NMCA) so as to enhance the effectiveness of Hydrogen Water Chemistry (HWC) in mitigating Intergranular Stress Corrosion Cracking (IGSCC) in reactor vessel internal components. Specifically, the proposed amendment would raise the reactor water conductivity limit in TS 3.2.3.a from 1.0 micromho/cm to 20 micromho/cm and in TS 3.2.3.c.1 from 5.0 micromho/cm to 20.0 micromho/cm during NMCA application. The proposed amendment will also raise the limit in TS 3.2.3.a and 3.2.3.b from 1 micromho/cm to 2 micromho/cm for up to a 5-month period at power operation following NMCA application. The reactor water conductivity would be restored to within the limit currently specified in TS 3.2.3 after the NMCA process is complete. The Bases for TS 3.2.3 and 4.2.3, ``Coolant Chemistry,'' would be supplemented to explain the changes resulting from NMCA. 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: The operation of Nine Mile Point Unit 1, in accordance with the proposed amendment, will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment to TS 3.2.3 will raise the reactor water conductivity limit during and following NMCA application. This change will allow the application of a layer of noble metals to the reactor vessel internals to enhance the effectiveness of HWC in mitigating IGSCC. An increased conductivity is expected both during and following NMCA. However, during NMCA, this increase is caused principally by residual ionic species which do not contribute to IGSCC. Following NMCA application, the increased conductivity is expected to be due to soluble iron and increased pH which has no adverse affect on crack growth. Accordingly, the proposed change will not adversely affect reactor vessel internals or reactor fuel such that the probability of an accident is increased. The proposed change will not alter the current TS requirements concerning equipment needed to mitigate the consequences of an accident nor affect the performance of this equipment. Therefore, operation in accordance with the proposed amendment will not create an increase in the probability or consequences of an accident previously evaluated. The operation of Nine Mile Point Unit 1, in accordance with the proposed amendment, will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment to TS 3.2.3 will raise the reactor water conductivity limit during and following NMCA application. This change will allow the application of a layer of noble metals to the reactor vessel internals to enhance the effectiveness of HWC in mitigating IGSCC. Except for these temporary exceptions to the existing reactor coolant chemistry specification, no new plant or system operating modes are being introduced and plant equipment will continue to perform their intended function. An increased conductivity is expected both during and following NMCA. However, during NMCA, this increase is caused by ionic species which do not contribute to IGSCC. Following NMCA application, the increased conductivity is due to soluble iron and increased pH which has no adverse affect on crack growth. Accordingly, the proposed changes will not affect plant equipment in a way to create a new or different kind of accident. Therefore, operation in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The operation of Nine Mile Point Unit 1, in accordance with the proposed amendment, will not involve a significant reduction in a margin of safety. The proposed amendment to TS 3.2.3 will raise the reactor water conductivity limit during and following the application of NMCA. During NMCA, the proposed change will raise the reactor water conductivity limit in TS 3.2.3a and 3.2.3c.1 to 20 [micromho/cm]. However, the expected increase in coolant conductivity is caused principally by ionic species which do not contribute to IGSCC and, therefore, will not adversely affect reactor vessel internals or reactor fuel. Following NMCA application, industry experience indicates that there may be an elevated conductivity approaching the 1 [micromho/ cm] conductivity limit delineated in TS 3.2.3a and 3.2.3b. To provide operating margin, NMPC proposes to raise this limit to 2 [micromho/cm] for up to 5 months of power operation following application. The expected increase in the conductivity is attributed to an increase in soluble iron and pH in the reactor coolant which results from the application of the noble metals and its affect on the deposits on the fuel. Soluble iron nor increased pH contribute to IGSCC crack growth. The existing 1 [micromho/cm] limit is based on EPRI [Electric Power Research Institute] guidelines action Level 2 for power operation, which assumes normal [[Page 51348]] conductivity below .3 [micromho/cm]. Increasing the limit to 2 [micromho/cm] during the period when soluble iron levels are high provides an equivalent operating margin consistent with the chloride and sulfate limits. Accordingly, this temporary ([less than] 5 months) elevated conductivity is expected, acceptable, and not considered ``abnormal'' as discussed in TS 4.2.3 and associated Bases. Daily samples of coolant for conductivity, chlorides and sulfates will continue to be performed to assure water quality. Therefore, operation in accordance with the proposed amendment 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Reference and Documents Department, Penfield Library, State University of New York, Oswego, New York 13126. Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502. NRC Section Chief: S. Singh Bajwa. Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: August 26, 1999 Description of amendment request: The proposed amendment would raise the condensate storage tank (CST) low level setpoint and the corresponding allowable value in Technical Specification (TS) Tables 3.3.3-2 and 3.3.5-2. The subject setpoint is associated with the automatic transfer of the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) pump suctions from the CST to the suppression pool in the event of low CST level. These changes are being made to address concerns regarding potential vortexing in the HPCI and RCIC suction flowpaths. 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 changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The systems affected by the proposed change provide accident mitigation functions. Neither the proposed increase in level setpoint nor the reliance on operator action to maintain the required 135,000 gallon reserve volume in the condensate storage tank (CST) can affect initiation of a design basis accident. Raising the CST low level setpoint to account for potential vortexing in the HPCI and RCIC suction flowpaths provides assurance that the functions of these systems can be properly carried out. There will no longer be a possibility of air entrainment into the RCIC and HPCI pumps suction at low levels in the CST. Initiation of RCIC or HPCI flow is unaffected by this modification. Execution of the suction line transfer to the suppression pool remains an entirely automatic function, utilizing the same safety related instrument signals as previously. Reliance on level alarms and operator action to maintain the 135,000-gallon minimum reserve water volume in the CST, in lieu of internal standpipes, cannot increase the consequences of an accident. This is an operational condition that establishes initial conditions prior to an accident occurring. Operators would have sufficient time to respond to a CST level decrease under non- accident conditions. Manually transferring HPCI and RCIC suction to the safety related suppression pool should CST level decline below 203,000 gallons (the 135,000 gallons required inventory, plus 68,000 gallons unusable) ensures HPCI and RCIC remain fully capable of performing their design basis functions. All parameters pertaining to the accident analysis, including pump initiation time, flowrate, volume and duration of flow delivered to the reactor vessel remain satisfied following implementation of this proposed change. Therefore, no accident scenario evaluated in the SAR [Safety Analysis Report] will be affected, and the radiological consequences of accidents previously evaluated in the SAR are not increased. These changes, therefore, do not modify or add any initiating parameters that would significantly increase the probability or consequences of any previously analyzed accident. (2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Implementation of these proposed changes cannot create the possibility of a different type of accident from any previously considered. First, the affected systems only perform mitigation functions, so postulated failures of any of these systems would not initiate a design basis accident. The function credited in the safety analysis is automatic transfer of the HPCI and RCIC suction lines from the CST to the suppression pool. This automatic transfer will still occur as required, with the only difference being execution earlier at a higher CST water level. Any considerations associated with maintaining the required minimum CST water level, including reliance on an alarm and operator action in lieu of a passive design feature, cannot lead to an accident of a different type since the CST itself is explicitly excluded from consideration in the accident analysis. Although the preference is to provide shutdown cooling with the reactor grade water of the CST, failure to do so will neither impact the ability to achieve shutdown cooling nor create a new type of accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. (3) The proposed change does not involve a significant reduction in a margin of safety. The margin of safety of the affected TS is maintained. RCIC is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and concurrent loss of feedwater flow to the reactor vessel without requiring actuation of ECCS [Emergency Core Cooling System] equipment. This function will be accomplished. HPCI provides a backup to RCIC for safe shutdown and the ECCS function of ensuring the reactor core is adequately cooled to limit fuel clad temperature during a small break loss of coolant accident. The safety analysis does not credit CST water. Since the automatic transfer to the suppression pool is assured with the same high quality and reliability as before, the ECCS function is not affected. Should CST level decline below the required minimum volume, operators would align HPCI and RCIC suction to the suppression pool. System design functions, including containment isolation, continue to be maintained in this alignment. The CST also provides a source of water for shutdown during station blackout (SBO) scenarios. The proposed changes do not affect the ability to recover from a SBO scenario. Core spray is provided to assure that the core is adequately cooled following a LOCA [Loss of Coolant Accident] and provides core cooling capacity for all break sizes. Core spray is a primary cooling source after the reactor vessel is depressurized and a source for flooding in case of accidental draining. In Operational Conditions 4 or 5, the CST is relied upon as the cooling water source if the suppression pool is drained below its minimum level. Operator actions in response to a CST alarm ensure sufficient condensate inventory is available to accomplish this function. ECCS instrumentation (HPCI) is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. RCIC instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. The HPCI and RCIC level instruments continue to provide their automatic function thereby preserving the design requirements of these systems. Remote shutdown instrumentation and controls ensure that sufficient capability is available to permit shutdown and maintenance of Hot Shutdown of the unit from locations outside the control room in the event control room habitability is lost. RCIC continues to satisfy this function. All design basis requirements of HPCI, RCIC, core spray and the CST continue to be satisfied to ensure safe shutdown and [[Page 51349]] mitigate a LOCA. Required water volumes remain available for core cooling, as is the automatic transfer to the safety related suppression pool source. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the 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. Local Public Document Room location: Pennsville Public Library, 190 S. Broadway, Pennsville, NJ 08070. 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. 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: July 29, 1999. Description of amendment request: The proposed amendment would revise Technical Specification (TS) Surveillance Requirement 4.6.1.1 to clarify when verification of primary containment integrity may be performed by administrative means and to change the surveillance interval for verification of manual valves and blind flanges inside of containment. 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 operation of Salem Nuclear Generating Station, Unit Nos. 1 and 2, in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The licensee has determined that the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change revises means for verification of containment integrity in certain cases by allowing the verification to be conducted by administrative means such as tagging requests, other TS surveillance procedures and previously performed valve alignments. Although the current Salem TSs allow the use of administrative means to verify valve position, its application is limited to valves that are open under administrative controls. The proposed amendment does not change the position of containment isolation valves or otherwise modify the containment integrity. Thus, the assumptions made in evaluating the occurrence and radiological consequences of accidents described in the Safety Analysis Report (SAR) have not been changed. The proposed change to use administrative means continues to ensure that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analysis. Allowing the use of administrative means to verify compliance with the surveillance requirement for these valves is acceptable based on the limited access to these areas in Modes 1 through 4 (power operation through hot shutdown). The probability of misalignment of these containment isolation valves, once they have been verified in the proper position is small. The probability of occurrence of any previously evaluated accident is independent of valve position verification. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated in the SAR. 2. The operation of Salem Nuclear Generating Station, Unit Nos. 1 and 2, in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The licensee has determined that the proposed amendment does not physically alter the facility or change the operation of the facility. The proposed change does not affect the current operation and response of any systems, structures or components assumed to function in the accident analysis. Additionally, the proposed change does not increase the consequences of a malfunction of equipment important to safety. The proposed change to use administrative means in lieu of field verification continues to ensure that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analysis. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The operation of Salem Nuclear Generating Station, Unit Nos. 1 and 2, in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. The licensee has determined that the proposed amendment does not involve a significant reduction in a margin of safety. The proposed change involves a revision of certain TSs surveillance requirements and frequency of performance. The proposed change does not modify hardware or plant operation, and the accident analyses are unchanged. The proposed amendment will continue to ensure that the proper valves are identified and tested in accordance with the TS requirements. 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Salem Free Public Library, 112 West Broadway, Salem, NJ 08079 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 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: August 25, 1999. Description of amendment request: The proposed amendment would revise Technical Specification (TS) Appendix C, ``Additional Conditions,'' to authorize the performance of single cell charging of operable safety-related batteries by using non-Class 1E single cell battery chargers, with proper electrical isolation. The single cell chargers would be used to restore individual cell float voltage to the normal TS limit. 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 changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change permits the use of an industry accepted method to restore a battery cell to its design basis from an OPERABLE but degraded condition or to prevent a cell from becoming degraded. IEEE Std [Institute of Electrical and Electronics Engineers Standard] 450-1995, ``IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead Storage Batteries for Stationary Applications,'' states that single cell charging is an acceptable method of correcting low cell voltage or low specific gravity conditions for a single cell or for a small number of cells. At least two class 1E fuses in series will be used on both the positive and negative leads between the battery and the charger to protect the battery if a fault should develop in the charger. The battery charger design includes diodes, a power transformer and control circuitry to prevent draining the connected cells in the event of a short circuit in the 120 Volt ac source or a loss of charger input or output voltage. Charger output is controlled automatically to prevent overcharging the connected cells. In the event of a controller failure resulting in charger overvoltage, procedural controls [[Page 51350]] governing the use of the charger ensure the condition is detected and corrected before failure of a connected cell occurs. While the single cell charger is connected, procedures will require periodic checks to verify proper charger operation and to measure electrolyte level, temperature and specific gravity for the cells being charged. Monitoring will be performed at least once every eight hours, a frequency sufficient to ensure compliance with the requirements of the Technical Specifications. An insulating material will be used to minimize the possibility of shorting leads or clips at the battery. Administrative controls governing the use and storage of transient loads are sufficient to ensure the use of single cell battery chargers does not create a potential missile hazard to safety related systems, structures and components. The Class 1E DC system is not an accident initiator. The Class 1E DC system supports the operation of safety related equipment required for the safe shutdown of the plant and for the mitigation of accident conditions. Therefore, the proposed change does not increase the probability of an accident previously evaluated. The station's dc systems will be operable to mitigate the consequences of an accident previously evaluated. Single cell charging would be limited to one OPERABLE class 1E battery bank at a time for either the 28 VDC or 125 VDC systems. Therefore, failure of a class 1E battery as a result of single cell charging would be limited to a single channel and would not reduce the number of OPERABLE dc sources below that required to safely shutdown the plant. Administrative controls would also prohibit the use of single cell charging for an OPERABLE class 1E battery if less than the minimum number of class 1E batteries required by Technical Specifications are OPERABLE. The proposed change does not cause the capability of the class 1E DC system to be degraded below the level assumed for any accident described in the SAR [Safety Analysis Report]. It would enhance the availability of safety related equipment required for the safe shutdown of the plant and for the mitigation of accident conditions. Therefore the radiological consequences of an accident will remain inside the design basis while single cell charging is performed on an OPERABLE battery. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The potential to adversely affect the Class 1E batteries is minimized by the use of Class 1E fuses and by appropriate administrative controls. Failure modes associated with the proposed change are bounded by the loss of a Class 1E battery bank which was previously evaluated. Therefore, the proposed change does 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 proposed change permits the use of non-Class 1E single cell battery chargers, with proper electrical isolation, for charging connected cells in OPERABLE class 1E batteries. This would allow parameters for an individual cell or for a small number of cells to be restored to the normal values specified in Technical Specifications without affecting the remainder of the cells in the battery. Increased cell monitoring after single cell charging, together with PSE&G's corrective action program which requires degraded and non-conforming conditions to be documented and evaluated, provides assurance that the use of single cell charging will not cause long-term cell degradation to go undetected. Since all battery cells are required to be maintained within the allowable values specified in Technical Specifications, and since the use of the single cell charger will not adversely affect battery capacity or capability, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Salem Free Public Library, 112 West Broadway, Salem, NJ 08079. 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. Previously Published Notice 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. Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois Date of amendment request: July 30, 1999. Description of amendment request: The proposed amendments would temporarily change the Technical Specifications (TS) to increase the upper temperature limit for the Ultimate Heat Sink (UHS) from 98 degrees Fahrenheit to 100 degrees Fahrenheit. The proposed temporary change would be in effect until September 30, 1999. Date of publication of individual notice in Federal Register: August 18, 1999 (64 FR 44962). Expiration date of individual notice: September 17, 1999. Local Public Document Room location: Wilmington Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481. 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 at the local public document rooms for the particular facilities involved. [[Page 51351]] Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland Date of application for amendments: November 30, 1998, as supplemented May 25, 1999. Brief description of amendments: The amendments revise the appropriate Technical Specifications to permit the use of leak-limiting Alloy 800 repair sleeves developed by AAB--Combustion Engineering (ABB- CE) to be used at Calvert Cliffs. Date of issuance: September 1, 1999. Effective date: As of the date of issuance to be implemented during the spring 2000. Amendment Nos.: 231 and 207. Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: January 13, 1999 (64 FR 2244). The May 25, 1999, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated September 1, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Calvert County Library, Prince Frederick, Maryland 20678. Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina Date of application for amendment: January 28, 1999. Brief description of amendment: The amendment revises Technical Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to add two references to the list of approved topical reports. Date of issuance: September 1, 1999. Effective date: September 1, 1999. Amendment No.: 185. Facility Operating License No. DPR-23. Amendment revised the Technical Specifications. Date of initial notice in Federal Register: February 24, 1999 (64 FR 9184). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 1, 1999. Local Public Document Room location: Hartsville Memorial Library, 147 West College Avenue, Hartsville, South Carolina 29550. 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: January 22, 1999. Brief description of amendment: The amendment revises Technical Specifications 4.3.a and 4.3.b and Basis Section 4.3 to permit reactor coolant system leak test to be performed at normal operating pressure following each refueling outage according to the requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, and implemented in accordance with 10 CFR 50.55a(g). Date of issuance: September 2, 1999. Effective date: As of the date of issuance to be implemented within 60 days. Amendment No.: 203. Facility Operating License No. DPR-26: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: April 7, 1999 (64 FR 17023). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 2, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10610. Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren County, Michigan Date of application for amendment: June 17, 1998, as supplemented June 23 and December 2, 1998, and March 18, 1999. Brief description of amendment: The amendment revises the Technical Specifications to reduce the minimum reactor vessel flow rate requirement and revise the units of measurement for consistency with the flow measurement procedure. Date of issuance: September 3, 1999. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment No.: 187. Facility Operating License No. DPR-20. Amendment revised the Technical Specifications. Date of initial notice in Federal Register: July 2, 1998 (63 FR 36271). The December 2, 1998, letter provided additional clarifying information and the March 18, 1999, letter requested a 60-day allowance for implementation of the amendment. The additional information and proposed change to the implementation period were within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 3, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Van Wylen Library, Hope College, Holland, Michigan 49423-3698. Duquesne Light 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: December 24, 1998, as supplemented June 15, June 17, and July 7, 1999. Brief description of amendments: The amendments revise the Technical Specification (TS) requirements for the axial flux difference (AFD) monitor, quadrant power tilt ratio (QPTR) monitor, rod position deviation monitor, and rod insertion limit (RIL) monitor. Specifically, the changes (1) relocate requirements for the AFD monitor and the QPTR monitor to the Licensing Requirements Manual; (2) delete requirements for the rod position deviation monitor and RIL monitor from the TSs; (3) modify Unit 1 surveillance requirements (SR) 4.1.3.5 and 4.1.3.6 by incorporating the Unit 2 wording to provide surveillances more consistent with the Limiting Condition for Operation; (4) change Unit 1 SR 4.1.3.2.2, SR 4.1.3.5, SR 4.1.3.6 and Unit 2 SR 4.1.3.5 from 24-hour surveillance frequencies to 12-hour frequencies; and (5) delete Unit 1 SR 4.1.3.2.3. Date of issuance: August 30, 1999. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 225 and 102. Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: January 27, 1999 (64 FR 4155) The June 15, June 17, and July 7, 1999, letters provided additional information but 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 amendments is contained in a Safety Evaluation dated August 30, 1999. No significant hazards consideration comments received: No [[Page 51352]] Local Public Document Room location: B. F. Jones Memorial Library, 663 Franklin Avenue, Aliquippa, PA 15001. Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas Date of amendment request: April 9, 1999, as supplemented by letter dated July 14, 1999. Brief description of amendment: Revises requirements affecting the surveillance methods for the containment tendons, the conduct of containment visual inspections, and the reporting methods employed in disseminating the results of these inspections to the NRC. Date of issuance: September 9, 1999. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment No.: 199. Facility Operating License No. DPR-51: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: May 19, 1999 (64 FR 27320). The July 14, 1999, letter provided clarifying information that did not change the scope of the April 9, 1999, application and the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 9, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Tomlinson Library, Arkansas Tech University, Russellville, Arkansas 72801. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio Date of application for amendment: March 17, 1999. Brief description of amendment: This amendment approves a proposed modification that changes the Perry facility as described in the Updated Safety Analysis Report. The change incorporates a leak-off line in the residual heat removal system. The leak-off line is designed to eliminate an operator work around, which will significantly reduce the collective dose to operations personnel. Date of issuance: August 31, 1999. Effective date: August 31, 1999. Amendment No.: 106. Facility Operating License No. NPF-58: This amendment authorizes the revision of the Updated Safety Analysis Report. Date of initial notice in Federal Register: May 19, 1999 (64 FR 27322) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 31, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Perry Public Library, 3753 Main Street, Perry, OH 44081 Florida Power Corporation, et al., Docket No. 50-302, Crystal River Nuclear Generating Plant, Unit 3, Citrus County, Florida Date of application for amendment: May 10, 1999. Brief description of amendment: The amendment corrects an invalid reference in Section 5.8, ``High Radiation Area,'' of the Crystal River Unit 3 Improved Technical Specifications (ITS). Date of issuance: September 3, 1999. Effective date: September 3, 1999. Amendment No.: 186. Facility Operating License No. DPR-72: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: July 14, 1999 (64 FR 38026) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 3, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Coastal Region Library, 8619 W. Crystal Street, Crystal River, Florida 34428. 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: May 17, 1999. Brief description of amendment: The amendment revises Technical Specification (TS) section 4.4.6.2.2.e to replace the reference to American Society of Mechanical Engineers (ASME) Code paragraph IWV- 3472(b) which pertains to the frequency of leakage rate testing for 6- inch, nominal pipe size valves and larger with the requirement that the surveillance interval and frequency of surveillance leakage rate testing for these valves be performed pursuant to the requirements of TS 4.0.5, ``Operations and Surveillance Requirements.'' Date of issuance: September 10, 1999. Effective date: As of the date of issuance. Amendment No.: 174. Facility Operating License No. NPF-49: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: July 14, 1999 (64 FR 38033). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 10, 1999. No significant hazards consideration comments received: No Local Public Document Room location: Learning Resources Center, Three Rivers Community-Technical College, 574 New London Turnpike, Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, Connecticut. Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota Date of application for amendments: May 13, 1999. Brief description of amendments: The amendments revise Technical Specifications 6.2.A.2, ``Onsite and Offsite Organizations,'' to reflect a change in the plant organizational structure that was implemented on March 1, 1999. Date of issuance: August 26, 1999. Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment Nos.: 146 and 137. Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Technical Specifications. Date of initial notice in Federal Register: July 14, 1999 (64 FR 38034). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Minneapolis Public Library, Technology and Science Department, 300 Nicollet Mall, Minneapolis, Minnesota 55401. Power Authority of The State of New York, Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York Date of application for amendment: April 12, 1999. Brief description of amendment: The amendment removes from the Technical Specifications a footnote regarding departure from nucleate boiling analysis. Date of issuance: September 2, 1999. Effective date: September 2, 1999. [[Page 51353]] Amendment No.: 191. Facility Operating License No. DPR-64: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: May 19, 1999 (64 FR 27324). No significant hazards consideration comments received: No. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 2, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10610. Power Authority of the State of New York, Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York Date of application for amendment: January 28, 1999, as supplemented May 4, 1999 Brief description of amendment: The amendment changes the reactor trip on turbine trip from at or above 10 percent rated power to at or above the P-8 setpoint. Date of issuance: September 8, 1999. Effective date: As of the date of issuance to be implemented within 30 days. Amendment No.: 192. Facility Operating License No. DPR-64: Amendment revises the Technical Specifications. Date of initial notice in Federal Register: April 21, 1999 (64 FR 19563). The May 4, 1999, letter provided additional information that did not change the staff's proposed finding of no significant hazards consideration. No significant hazards consideration comments received: No. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 8, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10610. Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey Date of application for amendment: March 29, 1999, as supplemented June 21, 1999. Brief description of amendment: This amendment revises the Technical Specifications (TSs) by relocating the procedural details of the Radiological Effluent Technical Specifications (RETS) to the Offsite Dose Calculation Manual. The TSs were also revised to relocate procedural details associated with solid radioactive wastes to the Process Control Program. In addition, the Administrative Controls section of the TSs was revised to incorporate programmatic controls for radioactive effluents and environmental monitoring. These changes are consistent with the guidance provided in Generic Letter 89-01, ``Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program.'' Date of issuance: September 8, 1999. Effective date: As of the date of issuance, and shall be implemented within 60 days. Amendment No.: 121. Facility Operating License No. NPF-57: This amendment revised the Technical Specifications. Date of initial notice in Federal Register: May 19, 1999 (64 FR 27324). The June 21, 1999, supplement 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 the amendment is contained in a Safety Evaluation dated September 8, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Pennsville Public Library, 190 S. Broadway, Pennsville, NJ 08070. STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas Date of amendment request: June 7, 1999, as supplemented by letters dated June 24 and August 24, 1999. Brief description of amendments: The amendments revised Technical Specification (TS) 2.0, ``Safety Limits and Limiting Safety System Settings,'' TS 3.2.5, ``DNB [Departure from Nucleate Boiling] Parameters,'' and the associated Bases, and Administrative Controls Section 6.9.1.6, ``Core Operating Limits Report [(COLR)],'' by relocating cycle-specific reactor coolant system-related parameter limits from the TSs to the COLR. Date of issuance: September 2, 1999. Effective date: September 2, 1999, to be implemented within 30 days. Amendment Nos.: Unit 1--115; Unit 2--103. Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: July 14, 1999 (64 FR 38036). The August 24, 1999, supplement provided revised TS pages and clarifying information that was within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 2, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Wharton County Junior College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488. Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama Date of application for amendments: September 4, 1998, as supplemented by letter dated November 25, 1998. Brief description of amendments: Revises the licensing basis to credit containment pressure in excess of atmospheric pressure in the analysis for Emergency Core Cooling Systems pump. Date of issuance: September 3, 1999. Effective date: As of date of issuance, to be incorporated into the Final Safety Analysis Report (FSAR) with the next update. Amendment Nos.: 261 and 220. Facility Operating License Nos. DPR-52 and DPR-68: Amendments approves changes to the FSAR. Date of initial notice in Federal Register: September 23, 1998 (63 FR 5093). The November 25, 1998 supplemental letter did not change the original proposed no significant hazards determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 3, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Athens Public Library, 405 E. South Street, Athens, Alabama 35611. 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: January 15, 1999 (TS 98-09). [[Page 51354]] Brief description of amendments: The amendments relocate seismic instrumentation requirements from the Technical Specifications to the Technical Requirements Manual. Date of issuance: September 7, 1999. Effective date: As of the date of issuance to be implemented no later than 45 days after issuance. Amendment Nos.: Unit 1--245; Unit 2--236. Facility Operating License Nos. DPR-77 and DPR-79: Amendments revise the technical specifications. Date of initial notice in Federal Register: February 10, 1999 (64 FR 6712). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 7, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402. TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: May 24, 1999, as supplemented by letter dated July 9, 1999. Brief description of amendments: The amendments remove several cycle-specific parameter limits from the Technical Specifications (TSs). These parameter limits are added to the Core Operating Limits Report (COLR). Appropriate references to the COLR are inserted in the affected TSs. In addition, the core safety limit curves are replaced with safety limits more directly applicable to the fuel and fuel cladding fission product barriers. The affected TSs are: (1) TS 2.0, ``Safety Limits (Sls),'' (2) TS 3.3.1, ``Reactor Trip System Instrumentation Setpoints,'' (3) TS 3.4.1, ``RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,'' and (4) TS 5.6.5, ``Core Operating Limits Report.'' Date of issuance: August 30, 1999. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment Nos.: 67 and 67. Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: June 30, 1999 (64 FR 35213) and July 28, 1999, (64 FR 40908). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 30, 1999. No significant hazards consideration comments received: No Local Public Document Room location: University of Texas at Arlington Library, Government Publications/Maps, 702 College, P.O. Box 19497, Arlington, Texas 76019. TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station (CPSES), Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: May 14, 1999. Brief description of amendments: The amendments change the licenses to accurately reflect the new corporate name of the current licensee, ``TXU Electric Company'' in Facility Operating Licenses NPF-87 and NPF- 89 for CPSES, Units 1 and 2, respectively. Date of issuance: August 31, 1999. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment Nos.: Unit 1--Amendment No. 68; Unit 2--Amendment No. 68. Facility Operating License Nos. NPF-87 and NPF-89: The amendments change the Operating Licenses. Date of initial notice in Federal Register: June 30, 1999 (64 FR 35213). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 31, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: University of Texas at Arlington Library, Government Publications/Maps, 702 College, P.O. Box 19497, Arlington, Texas 76019. Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont Date of application for amendment: May 26, 1999. Brief description of amendment: The amendment revises the suppression pool water temperature surveillance requirements to specify monitoring the temperature every 5 minutes when performing testing that adds heat to the suppression pool. In addition, the amendment revises the requirement to check the suppression chamber water level and temperature from ``once per shift'' to ``daily'' and specifies that it is the average temperature that is checked. Date of Issuance: August 30, 1999. Effective date: As of the date of issuance, and shall be implemented within 30 days. Amendment No.: 174. Facility Operating License No. DPR-28.: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: July 28, 1999 (64 FR 40909). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated August 30, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Brooks Memorial Library, 224 Main Street, Brattleboro, VT 05301. Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power Station (YNPS) Franklin County, Massachusetts Date of application for amendment: March 17, 1999. Brief description of amendment: Revises the Possession Only License by deleting License Condition 2.C.(10) related to the Fitness-For-Duty program. Date of issuance: August 27, 1999. Effective date: August 27, 1999. Amendment No.: 152. Facility Operating License No. DPR-3. Amendment revises the license. Date of initial notice in Federal Register: June 2, 1999 (64 FR 29717). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 27, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Greenfield Community College, 1 College Drive, Greenfield, Massachusetts 01301. 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 [[Page 51355]] 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 at the local public document room for the particular facility involved. The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By October 22, 1999, 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 at the local public document room for the particular facility involved. 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 [[Page 51356]] 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). For the Nuclear Regulatory Commission. Dated at Rockville, Maryland, this 15th day of September, 1999. Elinor G. Adensam, Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation. [FR Doc. 99-24573 Filed 9-21-99; 8:45 am] BILLING CODE 7590-01-P