[Federal Register Volume 64, Number 66 (Wednesday, April 7, 1999)] [Notices] [Pages 17021-17040] From the Federal Register Online via the Government Publishing Office [www.gpo.gov] [FR Doc No: 99-8503] ----------------------------------------------------------------------- NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. Background Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Pub. L. 97- [[Page 17022]] 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 March 13, 1999, through March 26, 1999. The last biweekly notice was published on March 24, 1999 (64 FR 14278). Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently. Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of 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 April 23, 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 [[Page 17023]] 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 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. Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50- 529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 1, 2, and 3, Maricopa County, Arizona Date of amendments request: February 26, 1999. Description of amendments request: The proposed amendment would revise Technical Specification (TS) 3.5.3, ``Emergency Core Cooling System--Operating,'' to extend the completion time for one inoperable low pressure safety injection subsystem from 72 hours to 7 days. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment will extend the Completion Time for one inoperable low pressure safety injection (LPSI) subsystem in Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems (ECCE)[S]--Operating, from 72 hours to 7 days. The LPSI subsystem is part of the ECCS train and part of the shutdown cooling subsystem. The LPSI components are not accident initiators in any accident previously evaluated. Therefore, this change does not involve a significant increase in the probability of an accident previously evaluated. The LPSI system is primarily designed to mitigate the consequences of a large break loss of coolant accident (LOCA). These proposed changes do not affect any of the assumptions used in the deterministic LOCA analysis. In order to evaluate the LPSI Completion Time extension with respect to the ECCS, probabilistic safety analysis (PSA) methods were utilized. The results of these analyses show no significant increase in the core damage frequency. As a result, there would be no significant increase in the consequences of an accident previously evaluated. These analyses are detailed in CE NPSD-995, Combustion Engineering Owners Group ``Joint Applications Report for Low Pressure Safety Injection System AOT Extension,'' May 1995, as supplemented by updated PVNGS data provided in the attachment to this enclosure. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment will extend the Completion Time for one inoperable low pressure safety injection (LPSI) subsystem in Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems (ECCE)[S]--Operating, from 72 hours to 7 days. The proposed change does not change the design, configuration, or method of operation of the plant. Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. The proposed change does not involve a significant reduction in a margin of safety. The proposed change does not involve a significant reduction in a margin of safety. The proposed amendment will extend the Completion Time for one inoperable low pressure safety injection (LPSI) subsystem in Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems (ECCE)[S]--Operating, from 72 hours to 7 days. The proposed change does not affect the limiting conditions for operation or their bases used in the deterministic analyses to establish the margin of safety. PSA evaluations were used to evaluate these changes. These evaluations demonstrate that the changes will be risk neutral or risk beneficial for PVNGS. These evaluations are detailed in CE NPSD-995, as supplemented by updated data provided in the attachment to this enclosure. The NRC staff has reviewed the licensee's analysis and, based on that review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involve no significant hazards consideration. Local Public Document Room location: Phoenix Public Library, 1221 N. Central Avenue, Phoenix, Arizona 85004. Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999. NRC Project Director: William H. Bateman. Consolidated Edison Company of New York, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York Date of amendment request: January 22, 1999. Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) Sections 3.7.D.1.g, 6.2.2.h and 6.3.1. Specifically, (1) Section 3.7.D.1.g would be revised to correct an editorial error; (2) Section 6.2.2.h would be revised to change the senior reactor operator license requirement for the Operations Manager; and (3) Section 6.3.1 would modify the qualification requirement for the Operations Manager. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change [to Section 3.7.D.1.g] is administrative in nature. It involves making an editorial change to provide the correct functional description of the breakers. This change does not affect possible initiating events for accidents previously evaluated or alter the configurations or operation of the facility. The Limiting Safety Systems Settings and Safety Limits specified in the current Technical Specifications [[Page 17024]] remain unchanged. Therefore, the proposed change to the subject Technical Specification would not increase the probability or consequences of an accident previously evaluated. The proposed change [to Section 6.2.2.h] is administrative in nature. The individual who provides the day to day direction of the activities of the operating shift will still possess an SRO [Senior Reactor Operator] license and this proposed change is consistent with the statement in NUREG-1431, Section 5.2.2.f. This change does not affect possible initiating events for accidents previously evaluated or alter the configuration or operation of the facility. The Limiting Safety Systems Settings and Safety Limits specified in the current Technical Specifications remain unchanged. Therefore, the proposed change to the subject Technical Specification would not increase the probability or consequences of an accident previously evaluated. The proposed change [to Section 6.3.1] is administrative in nature. The individual who provides the day to day direction of the activities of the operating shift will still possess an SRO license and this proposed change is consistent with the statement in NUREG- 1431, Section 5.2.2.f. This change does not affect possible initiating events for accidents previously evaluated or alter the configuration or operation of the facility. The Limiting Safety Systems Settings and Safety Limits specified in the current Technical Specifications remain unchanged. Therefore, the proposed change to the subject Technical Specification would not increase the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. As stated above, the proposed change [to Section 3.7.D.1.g] is administrative in nature. The safety analysis of the facility remains complete and accurate. There are no physical changes to the facility and the plant conditions for which the design basis accidents have been evaluated are still valid. The operating procedures and emergency procedures are unaffected. Consequently, no new failure modes are introduced as a result of the proposed change. Therefore, the proposed change will not initiate any new or different kind of accident. The proposed change [to Section 6.2.2.h] is administrative in nature. The safety analysis of the facility remains complete and accurate. There are no physical changes to the facility and the plant conditions for which the design basis accidents have been evaluated are still valid. The operating procedures and emergency procedures are unaffected. Consequently, no new failure modes are introduced as a result of the proposed changes. Therefore, the proposed change will not initiate any new or different kind of accident. The proposed change [to Section 6.3.1] is administrative in nature. The safety analysis of the facility remains complete and accurate. There are no physical changes to the facility and the plant conditions for which the design basis accidents have been evaluated are still valid. The operating procedures and emergency procedures are unaffected. Consequently, no new failure modes are introduced as a result of the proposed changes. Therefore, the proposed change will not initiate any new or different kind of accident. 3. The proposed change does not involve a significant reduction in a margin of safety. The proposed change [to Section 3.7.D.1.g] is administrative in nature. Since there are no changes to the operation of the facility or physical design the Updated Final Safety Analysis Report (UFSAR) design basis, accident assumptions, or Technical Specification Bases are not affected. Therefore, the proposed changes will not result in a reduction in the margin of safety. The proposed change [to Section 6.2.2.h] is administrative in nature. Since there are no changes to the operation of the facility or physical design the Updated Final Safety Analysis Report (UFSAR) design basis, accident assumptions, or Technical Specification Bases are not affected. Therefore, the proposed changes will not result in a reduction in the margin of safety. The proposed change [to Section 6.3.1] is administrative in nature. Since there are no changes to the operation of the facility or physical design the Updated Final Safety Analysis Report (UFSAR) design basis, accident assumptions, or Technical Specification Bases are not affected. Therefore, the proposed changes will not result in a 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 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: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10610. Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, New York, New York 10003. NRC Project Director: S. Singh Bajwa, Director. Consolidated Edison Company of New York, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York Date of amendment request: January 22, 1999. Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) Section 4.3. Specifically, the revision would permit the reactor coolant system (RCS) leak test to be performed at normal operating pressure after it has been closed following normal opening in lieu of a hydrostatic test being performed at 2335 psig. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The change proposes a system leakage test for the RCS that is comparable to the hydrostatic test that it replaces, as acknowledged by the NRC approval of ASME Code Case N-498, ``Alternative Rules for 10-Year Hydrostatic Pressure Testing for Class 1 and 2 Systems Section XI, Division 1,'' and the ASME [American Society for Mechanical Engineers] Boiler and Pressure Vessel Code, Section XI. [. . .] The proposed change to substitute a system leak test at normal operating pressure in lieu of the hydrostatic test at 2335 psig will minimize challenge to plant safety and demonstrate leak tightness of the RCS. Therefore, the proposed change would not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not involve the addition of any new or different type of equipment, nor do they involve the operation of equipment required for safe operation of the facility in a manner different from those addressed in the Updated Final Safety Analysis Report. [. . .] Based on industry experience, it is expected that any leaks would be discovered by the leak test at normal operating pressure. 3. The proposed change does not involve a significant reduction in a margin of safety. The proposed license amendment does not involve a significant reduction in a margin of safety. The proposed changes do not adversely affect performance of any safety related system or component, instrument operation, or safety system setpoints and do not result in increased severity of any of the accidents considered in the safety analysis. Although the current basis states that if the system does not leak at 2335 psig (operating pressure + 100 psig) it will be leak tight during normal operation, industry experience demonstrates that leaks are not discovered as a result of hydrostatic test pressure propagating a preexisting flaw through wall. In most cases, leaks are discovered when the system is at normal operating pressure. Also, testing will continue to be performed as required by the ASME Boiler and Pressure Vessel Code Section XI. 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. [[Page 17025]] Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10610. Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, New York, New York 10003. NRC Project Director: S. Singh Bajwa, Director. Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Plant Units 3 and 4, Dade County, Florida Date of amendment request: March 8, 1999. Description of amendment request: The proposed amendments would delete certain requirements from Technical Specification (TS) Section 6.0 ``Administrative Controls'' that are adequately controlled by existing regulations, other than 10 CFR 50.36 and the TS. The amendments also relocate selected requirements from TS Section 6.0 to the licensee's controlled documents such as the Turkey Point Units 3 and 4 Updated Final Safety Analysis Report (UFSAR). The amendments also clarify certain provisions of TS Section 6.0. The proposed changes are to relocate, revise, delete, or clarify the following provisions of the TS: ------------------------------------------------------------------------ Existing TS section Subject Proposed change ------------------------------------------------------------------------ 6.2.2.f................ Administrative Controls Partly delete, partly on Working Hours of relocate within TS. Plant Staff. Table 6.2-1............ Minimum Shift Crew Clarify. Composition. 6.2.3.................. Shift Technical Advisor Clarify. 6.4.................... Training............... Delete. 6.5.................... Review and Audit....... Relocate to UFSAR. 6.6.................... Reportable Event Action Partly delete, partly relocate to UFSAR. 6.8.2.................. Review and Approval of Relocate to UFSAR. Procedures. 6.8.3.................. Temporary Changes to Relocate to UFSAR. Procedures. 6.8.4.b................ In-Plant Radiation Relocate to UFSAR. Monitoring. 6.8.4.g................ Radiological Relocate to UFSAR. Environmental Monitoring Program. 6.10................... Record Retention....... Relocate to UFSAR. 6.11................... Radiation Protection Relocate to UFSAR. Program. 6.12................... High Radiation Area.... Clarify. 6.13................... Process Control Program Relocate to UFSAR. (PCP). 6.14................... Offsite Dose Revise to reflect Calculation Manual changes to 6.5 & (ODCM). 6.10. ------------------------------------------------------------------------ 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 plant in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed changes are administrative in nature. These proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated because they do not affect assumptions contained in plant safety analyses, the physical design and/or operation of the plant, nor do they affect Technical Specifications that preserve safety analysis assumptions. None of the proposed changes involve a physical modification to the plant, a new mode of operation or a change to the UFSAR transient analyses. No Limiting Condition for Operation, ACTION statement or Surveillance Requirement is affected by any of the proposed changes. Also, these proposed changes, in themselves, do not reduce the level of qualification or training such that personnel requirements would be decreased. Further, the Proposed changes do not alter the design, function, or operation of any plant component. Therefore, the proposed changes do not affect the probability or consequences of accidents previously evaluated. 2. Operation of the plant in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The changes being proposed are administrative in nature and do not affect assumptions contained in plant safety analyses, the physical design and/or modes of plant operation defined in the plant operating license, or Technical Specifications that preserve safety analysis assumptions. The proposed changes do not introduce a new mode of plant operation or surveillance requirement, nor involve a physical modification to the plant. The proposed changes are administrative in nature. The changes propose to revise, delete, or relocate the stated administrative control provisions from the TS to the UFSAR whereby adequate control of information is maintained. Furthermore, the proposed changes do not alter the design, function, or operation of any plant components. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Operation of the plant in accordance with the proposed amendments would not involve a significant reduction in a margin of safety. The proposed changes do not involve a significant reduction in a margin of safety because they are administrative in nature. The operating limits and functional capabilities of the affected systems, structures, and components are unchanged by the proposed amendments. None of the proposed changes involve a physical modification to the plant, a new mode of operation or a change to the UFSAR transient analyses. No Limiting Condition for Operation, ACTION statement, or Surveillance Requirement is affected. Additionally, the proposed changes do not alter the scope of equipment currently required to be OPERABLE or subject to surveillance testing, nor does the proposed change affect any instrument setpoints or equipment safety functions. Therefore, the 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: Florida International University, University Park, Miami, Florida 33199. Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420. NRC Project Director: Cecil O. Thomas. [[Page 17026]] GPU Nuclear, Inc. etal., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey Date of amendment request: December 23, 1998. Description of amendment request: The proposed Technical Specification (TS) change request will change the surveillance frequency for verifying the operability of motor-operated isolation valves and condensate makeup valves in the Isolation Condenser TS 4.8.A.1 and Bases page from once per month to once per 3 months. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed surveillance interval change does not alter the actual surveillance requirements, nor does it alter the limits and restrictions on plant operations. The reliability of systems and components relied upon to prevent or mitigate the consequences of accidents previously evaluated is not degraded by the proposed change to the surveillance interval. Assurance of system and equipment availability is maintained. The proposed change does not alter any system or equipment configuration. Based on the above, the proposed change does not significantly increase the probability or consequences of a[n] accident previously evaluated. 2. The proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed surveillance interval change does not alter the actual surveillance requirements, nor does it alter the limits and restrictions on plant operations. Assurance of system and equipment availability is maintained. The proposed change does not alter any system or equipment configuration nor does it introduce any new mechanisms which could contribute to the creation of a new or different kind of accident than previously evaluated. 3. The proposed TS changes do not involve a significant reduction in a margin of safety. The proposed change extends the surveillance interval for verifying the operability of Isolation Condenser motor-operated isolation valves and condensate makeup valves from once per month to once per three months. The proposed change does not alter the actual surveillance requirements, the limits and restrictions on plant operations nor the design, function or manner of operation of any structures, systems or components. System availability and reliability are maintained. Accordingly, the proposed TS change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Ocean County Library, Reference Department, 101 Washington Street, Toms River, NJ 08753. Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Project Director: Elinor G. Adensam. GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey Date of amendment request: February 12, 1999. Description of amendment request: The proposed Technical Specification (TS) change will delete the organizational chart and the related organizational references from the Appendix B Environmental TS and revise the appearance and format of the Environmental TS. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because deletion of the organization charts and other organizational references in the [Environmental Technical Specifications] ETS does not affect plant operation. GPU Nuclear will continue to inform the NRC of organizational changes through other required controls. 2. The proposed change does not create the possibility of a new or different type of accident than previously evaluated because the proposed change is administrative in nature, and no physical alteration of plant configuration, changes to setpoints or operating parameters are proposed. 3. The proposed change does not involve a significant reduction in the margin of safety because it does not alter the design, function or manner of operation of any structures, systems or components. Organizational structure or its representation does not directly impact 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: Ocean County Library, Reference Department, 101 Washington Street, Toms River, NJ 08753. Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Project Director: Elinor G. Adensam. IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa. Date of amendment request: February 18, 1999. Description of amendment request: The proposed amendment would revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) Table 3.3.6.1-1, ``Primary Containment Isolation Instrumentation,'' by deleting the manual initiation function of the high pressure coolant injection (HPCI) system and reactor core isolation cooling (RCIC) system isolation. A related condition as well as corresponding surveillance requirements and bases would also be deleted. Thus, the change would (1) revise Table 3.3.6.1-1 by removing items 3j. and 4.j.; (2) revise Note 2 to Surveillances to Licensing Condition for Operation (LCO) 3.3.6.1 by deleting information regarding items 3 j. and 4.j.; and (3) revise LCO 3.3.6.1 by removing Condition G and Surveillance Requirement 3.3.6.1.10. 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: After reviewing this proposed amendment, we [the licensee] have concluded: 1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The Manual Initiation Function for HPCI and RCIC Isolation is not considered to be an initiator for any accident previously evaluated in the UFSAR. Therefore, this change does not involve a significant increase in the probability of any previously evaluated accidents. The Manual Initiation push button channels introduce signals into HPCI and RCIC System isolation logics that are redundant to the automatic protective instrumentation and provide manual isolation capability only if a system initiation signal is present. Technical Specification Section 3.3.6.1 Condition G requires isolation of the System flowpath, which renders the System inoperable and reduces the availability of the System due to the failure of a manually initiated isolation, an isolation which is not assumed in any transient or accident analysis in the UFSAR. Removal of the Manual Initiation Function [[Page 17027]] for HPCI and RCIC from the Primary Containment Isolation Instrumentation Section of Technical Specifications does not affect the automatic protective instrumentation and the automatic isolation capability. Therefore, this change does not significantly increase the consequences of a previously analyzed accident. 2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change introduces no new mode of plant operation and does not involve physical modification to the plant. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed amendment will not involve a significant reduction in a margin of safety. The proposed change deletes the Manual Initiation Function from Technical Specifications, but no significant reduction in a margin of safety is involved. Technical Specification Section 3.3.6.1 Condition G requires isolation of the System flowpath, which renders the System inoperable and reduces the availability of the System due to the failure of a manually initiated isolation, an isolation that is not assumed in any transient or accident analysis in the UFSAR. Removal of the Manual Initiation Function for HPCI and RCIC from the Primary Containment Isolation Instrumentation Section of Technical Specifications does not affect the automatic protective instrumentation and the isolation capability. This change is acceptable based on the fact that the Manual Initiation Function is not assumed in any accident or transient analysis in the UFSAR. Based upon the above, we [licensee] have determined that the proposed amendment will not involve a significant hazards consideration. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Cedar Rapids Public Library, 500 First Street, SE., Cedar Rapids, IA 52401. Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869. NRC Project Director: T.J. Kim, Acting. Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska Date of amendment request: March 1, 1999. Description of amendment request: The proposed amendment would change the Cooper Nuclear Station (CNS) Technical Specifications (TSs) to revise the calibration frequency of the reactor recirculation flow transmitters from once every 184 days to once every 18 months. This calibration is required as part of TS Surveillance Requirement (SR) 3.3.1.1.10. 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 proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. Changing the calibration frequency of the recirculation loop flow transmitters from 184 days to 18 months may increase the amount of drift experienced by the transmitters. However, CNS calculation (NEDC 98-024 [forwarded by letter dated March 10, 1999]) takes into account the 18 month calibration intervals. This calculation, performed in accordance with the General Electric (GE) setpoint methodology for CNS, demonstrates that the expected drift is not significant, and is consistent with past operating experience. Changing the calibration frequency of the flow transmitters does not change any of the precursors assumed in the accident analysis. Therefore, changing the calibration frequency for flow transmitters from 184 days to 18 months does not involve a significant increase in the probability of an accident previously evaluated in the USAR [Updated Safety Analysis Report]. The proposed change will not create the possibility of a new or different kind of accident than evaluated in the USAR. The proposed change does not result in any physical change to plant structures, systems, or components. The proposed change does not alter the form, fit, or function of any equipment or components credited in the accident analyses described in the USAR. Therefore, changing the test frequency does not create the possibility of a new or different kind of accident. The proposed change will not involve a significant reduction in a margin of safety. This conclusion is based on the fact that the proposed change is consistent with the drift assumptions used in CNS approved calculation (NEDC 98-024). The calibration frequency of 18 months is consistent with the operating practices prior to conversion to Improved Technical Specifications, and is consistent with past operating practice at CNS. Therefore, changing the calibration frequency from 184 days to 18 months does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Auburn Memorial Library, 1810 Courthouse Avenue, Auburn, Nebraska 68305. Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power District, Post Office Box 499, Columbus, Nebraska 68602-0499. NRC Project Director: George Dick, Acting. Northeast Nuclear Energy Company (NNECO), et al., Docket Nos. 50-336 and 50-423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London County, Connecticut Date of amendment request: March 5, 1999. Description of amendment request: The proposed amendment would relocate certain Technical Specification (TS) Section 6.0 administrative controls to the NRC-approved Northeast Utilities Quality Assurance Program (NUQAP) Topical Report. Specifically, Sections 6.2.3 (Unit 3 only), 6.5, 6.6 (partial), 6.7 (partial), and 6.10. The proposed amendment would also delete parts of Section 6.6 and 6.7 because their requirements are duplicated in existing regulations or elsewhere in the TS. In addition, the proposed amendment would modify the table of contents and other TS sections to incorporate the aforementioned changes (e.g., correct references). 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: In accordance with 10 CFR 50.92, NNECO has reviewed the attached proposed changes and has concluded that they do not involve a Significant Hazards Consideration (SHC). The basis for this conclusion is that the three criteria of 10 CFR 50.92 are not compromised. The proposed changes are not a SHC because the proposed change will not: 1. Involve a significant increase in the probability or consequence of an accident previously evaluated. No design basis accidents are affected by these proposed changes. The proposed changes relocate portions of the Technical Specifications to the NUQAP Topical Report or remove duplicate sections and are being proposed to eliminate the need for a T.S. change each time there is a related change in the administrative controls for the site. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated. There are no changes in the way the plant is operated due to these revisions. The potential for an unanalyzed accident is not created. There is no impact on plant response, and no new failure modes are introduced. The proposed deletions and [[Page 17028]] editorial changes have no impact on safety limits or design basis accidents, and have no potential to create a new or unanalyzed event. 3. Involve a significant reduction in a margin of safety. These changes do not directly affect any protective boundaries nor do they impact the safety limits for the protective boundaries. These proposed changes relocate portions of the administrative controls to the NUQAP Topical Report or are editorial in nature. Therefore, there is no reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. 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. Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, Connecticut. NRC Project Director: Elinor G. Adensam. PECO Energy Company, Docket No. 50-353, Limerick Generating Station, Unit 2, Montgomery County, Pennsylvania Date of amendment request: March 11, 1999. Description of amendment request: The proposed revision to the Technical Specifications (TSs) involves a change to TS Section 2.1 and its associated TS Bases to revise the minimum critical power ratio (MCPR) Safety Limits for Cycle 6. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The revised MCPR Safety Limits for LGS Unit 2 Technical Specifications, and their use to determine cycle-specific thermal limits, have been calculated using NRC-approved methods (i. e, GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific inputs. The use of these methods assures that the SLMCPR [safety limit minimum critical power ratio] value is within the existing design and licensing basis, and cannot increase the probability or severity of an accident. The basis for the MCPR Safety Limit calculation is to ensure that greater than 99.9 percent of all fuel rods in the core avoid transition boiling if the limit is not violated. The MCPR Safety Limit preserves the existing margin to transition boiling and fuel damage in the event of a postulated accident. The probability of fuel damage is not increased. Therefore, the proposed TS changes do not involve an increase in the probability or consequences of an accident previously evaluated. 2. The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The MCPR Safety Limit is a Technical Specification numerical value designed to ensure that fuel damage from transition boiling does not occur as a result of the limiting postulated accident. The MCPR Safety Limit is not an accident initiator; therefore, it cannot create the possibility of any new type of accident. The new MCPR Safety Limits are calculated using NRC-approved methods (i.e., GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific inputs. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed TS changes do not involve a significant reduction in the margin of safety. The margin of safety as defined in the TS Bases will remain the same. The new MCPR Safety Limits are calculated using NRC-approved methods (i.e., GESTAR-II, Rev. 13), which are in accordance with the current fuel design and licensing criteria, and are based on LGS, Unit 2, Cycle 6 specific inputs. The MCPR Safety Limit remains high enough to ensure that greater than 99.9 percent of all fuel rods in the core will avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. Therefore, the proposed TS changes do not involve a 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: Pottstown Public Library, 500 High Street, Pottstown, PA 19464. Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, PA 19101. NRC Project Director: Elinor G. Adensam. PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit 1, Luzerne County, Pennsylvania Date of amendment request: March 12, 1999. Description of amendment request: The amendment would modify the Susquehanna Steam Electric Station, Unit 1, Technical Specifications Table 3.3.5.1-1 ``Emergency Core Cooling System Instrumentation.'' The change updates the allowable values for both the Core Spray (CS) and Low Pressure Coolant Injection System (LPCI) ``Reactor Steam Dome Pressure--Low'' functions for initiation and injection permissive. Specifically, the allowable values are being changed from a specified minimum pressure to a specified allowable pressure band. This more restrictive allowable value range will prevent CS and LPCI system overpressurization while still permitting injection to prevent fuel clad temperature limits from being exceeded. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. This proposal does not involve an increase in the probability or consequences of an accident previously evaluated. The proposed amendment changes the ``Reactor Steam Dome Pressure-Low'' Allowable Values so to provide further assurance that the Core Spray and RHR systems will perform their LOCA [Loss-of-coolant accident] design basis function. The functional design basis of the Core Spray and LPCI is to inject water into the reactor vessel to cool the core during a LOCA by opening the Core Spray and LPCI injection valves when reactor pressure drops below the reactor vessel low pressure permissive. The upper analytical limit for the permissive is the Core Spray and LPCI systems' maximum design pressure, and the lower analytical limit is the lowest pressure which allows injection to prevent exceeding the fuel cladding temperature limit. The new allowable values were selected to lie within the upper and lower limits to ensure there will be no change in the required logic or functions of the Core Spray and LPCI systems. These new values do not affect the LOCA or its ``limiting fault'' frequency of occurrence and do not introduce any new accidents or malfunctions of equipment important to safety. Since they do not affect the LOCA, they do not change the probability of occurrence of the LOCA. The new allowable values do not change the logic or function of the reactor vessel low pressure permissive. These new values simply provide the basis for which the associated pressure [[Page 17029]] instruments are to be set to ensure proper operation of Core Spray and LPCI within the design pressures as described above. Therefore, the change in allowable values does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. Based upon the analysis presented above, PP&L [PP&L, Inc.] concludes that the proposed action does not involve an increase in the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. This proposal does not create the probability of a new or different type of accident from any accident previously evaluated. The new allowable values do not change any plant systems, structures, or components, nor do they change any existing or create any new Core Spray and LPCI logic or functions. The new allowable values were selected to ensure the required operation of the Core Spray and LPCI systems within the design pressures described above. 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 the margin of safety. The change does not involve a reduction in the margin of safety. Technical Specification Bases Section B3.3.5.1 9 (ECCS Instrumentation) identifies that the low reactor steam dome pressure signals are used as permissives for operation of the low pressure ECCS subsystems. The new allowable values were selected so to not impact the logic, redundancy, operability or surveillance requirements for these subsystems. The new allowable values maintain the margin requirements that the Core Spray and LPCI system pressures such that they do not exceed their system maximum design pressures and that system pressures are high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10CFR50.46. The margin of safety is unaffected by the proposed changes. 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 proposed to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Osterhout Free Library, Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701. Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PP&L, Inc., 2 North Ninth St., GENTW3, Allentown, PA 18101- 1179. NRC Project Director: Elinor G. Adensam. PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit 1, Luzerne County, Pennsylvania Date of amendment request: March 12, 1999. Description of amendment request: This proposed amendment would revise the minimum critical power ratio safety limit in Technical Specification (TS) Section 2.1.1.2. Also, the proposed amendment would modify the references in TS Section 5.6.5 in order to include only those references that directly support the generation of the Core Operating Limit and to remove the reference for the Lead Use Assemblies, which will be discharged during the next Unit 1 refueling outage. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The applicable sections of the FSAR are Chapters 4.4 and 15. FSAR Chapter 4.4 describes the MCPR Safety Limit, and Chapter 15 describes the transient and accident analyses. The reference to be added to Section 5.6.5 of the Unit 1 Technical Specifications describes a NRC approved critical power correlation for ATRIUMTM-10 fuel. This correlation is appropriate for use in conservative methodologies for generating MCPR Safety Limits and MCPR Operating Limits to assure safe operation of Unit 1 with ATRIUMTM-10 fuel. A discussion of the impact of the proposed Technical Specification change is provided below. The proposed change in critical power correlation does not physically affect the plant or its systems. Thus, it does not increase the probability of an accident previously evaluated. A Unit 1 Cycle 12 MCPR Safety Limit analysis was performed for PP&L by SPC. This analysis used NRC approved methods described in ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2. These methods will be used each cycle to calculate the Unit 1 Safety Limits. For Unit 1 Cycle 12, the critical power performance of the 9 x 9-2 and ATRIUMTM-10 fuel was determined using the NRC approved ANFB and ANFB-10 correlations, respectively. The SAFETY LIMIT MCPR calculations statistically combine uncertainties on feedwater flow, feedwater temperature, core flow, core pressure, core power distribution, and uncertainties in the Critical Power Correlation. The SPC analysis used cycle specific power distributions and calculated MCPR values such that at least 99.9% of the fuel rods are expected to avoid boiling transition during normal operation or anticipated operational occurrences. The resulting two-loop and single-loop MCPR Safety Limits are included in the proposed Technical Specification change. Thus, the cladding integrity and its ability to contain fission products are not adversely affected. Analyses of the Single Loop Pump Seizure accident with the NRC approved ANFB-10 correlation for ATRIUMTM-10 fuel (Reference 1) will be performed to demonstrate that the NRC acceptance criterion (i.e., small fraction of 10CFR100 dose limits) is met. Analyses will also be performed to validate the conclusion that two-loop transients are more severe than those events analyzed in single-loop operation. Changes to Section 2.1.1.2 reflect the change from a flow dependent MCPR Safety Limit to a single value MCPR Safety Limit for two-loop operation and single-loop operation. Changes to Reference 5.6.5 delete the methodology used for critical power analyses for ATRIUMTM-10 fuel and add the NRC approved ANFB-10 methodology to the list of approved methodologies. Other changes in Reference 5.6.5 are administrative in nature because they delete references not directly related to the generation of Core Operating Limits. No new analysis approaches are used due to these changes. Changes to BASES Sections 2.1.1 and 3.2.2 reflect the inclusion of the ANFB-10 critical power correlation. The range of the applicability of the ANFB-10 is valid for pressures > 571 psia and bundle mass fluxes > 0.115 x 10\6\ lb/hr-ft \2\. These values assure that a valid CPR calculation will result at or above 25% of rated core thermal power, that is, reactor steam dome pressure785 psig and core flow 10 Mlbm/hr. Changes to BASES Sections 3.2.1, 3.2.2, 3.2.3, and 3.2.4 reflect the removal of Reference 7 for the ABB LUAs, since the four LUAs will be discharged from Unit 1 during the Unit 1 11th Refueling and Inspection Outage. The consequences of transients and accidents will remain within the criteria approved by the NRC. The methodology used to perform the analyses has been previously approved by the NRC. Thus, analysis results using the new methodology will continue to provide assurance that the reactor will perform its design safety function during normal operation and design basis events. Therefore, the proposed action does not involve an increase in the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes to the Unit 1 Technical Specifications (MCPR Safety Limits, removal of methodology references not directly supporting the generation of Core Operating Limits, removal of the two references describing previously approved methodology for applying ANFB to ATRIUMTM-10 fuel, removal of the ABB LUA reference, and inclusion of the ANFB-10 correlation reference) do not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Removal of the Unit 1 Cycle 11 footnote allows Unit 1 Cycle 12 and future cycle operation with NRC [[Page 17030]] approved methodology. Thus, the proposed change does not create the possibility of a previously unevaluated operator error or a new single failure. The consequences of transients and accidents will remain within the criteria approved by the NRC. 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 applicable Technical Specification Sections include 2.1.1.2 and 5.6.5. The changes to the Unit 1 Technical Specifications discussed in Item 1 above do not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. The consequences of transients and accidents will remain within the criteria approved by the NRC. The proposed MCPR Safety Limits and use of the ANFB-10 critical power correlation described in the reference added to Section 5.6.5 do not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections. Therefore, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Osterhout Free Library, Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701. Attorney for licensee: Bryan A. Snapp, Esquire, PP&L, Inc., 2 North Ninth St., Allentown, PA 18101. NRC Project Director: Elinor G. Adensam. 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: March 2, 1999. Description of amendment request: The proposed amendment would clarify the use of a ``check valve with flow through the valve secured'' as a means to isolate an affected containment penetration (i.e., a penetration with an inoperable penetration barrier) in Technical Specification 3.6.3 Action b. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed change does not involve an increase in the probability or consequences of an accident previously evaluated. The proposed change does not involve any hardware changes. The proposed change will clarify Technical Specification 3.6.3 Action b to allow the use of a check valve with the flow through the valve secured as a means to isolate an inoperable containment penetration. This change is consistent with the changes identified in NUREG-1431, ``Improved Standard Technical Specifications for Westinghouse Plants'', Specification 3.6.3 (Containment Isolation Valves), which identifies check valves with flow through the valve secured as a type of deactivated automatic valve, and with 10 CFR 50 Appendix A General Design Criteria 55 and 56, which include the use of check valves as ``automatic isolation valves''. The proposed change will not affect the containment isolation valve OPERABILITY requirements or associated isolation time limits established in the Specifications. Therefore the proposed change will not affect any safety margin or safety limit applicable to the facility. Therefore no increase in the probability or consequences of any accident previously evaluated will occur. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change will clarify Technical Specification 3.6.3 Action b to allow the use of a check valve with the flow through the valve secured as a means to isolate an inoperable containment penetration. The proposed change will not involve any physical change to plant systems, structures, or components (SSC). This change is consistent with the changes identified in NUREG-1431, ``Improved Standard Technical Specifications for Westinghouse Plants'', Specification 3.6.3 (Containment Isolation Valves), which identifies check valves with flow through the valve secured as a type of deactivated automatic valve, and with 10 CFR 50 Appendix A General Design Criteria 55 and 56, which include the use of check valves as ``automatic isolation valves''. The proposed change only provides clarification to the existing Specification 3.6.3, and will not affect the established containment isolation valve OPERABILITY requirements or associated isolation time limits. Since the proposed change does not impact operation of the facility as presently approved, no possibility exists for a new or different kind of accident from those previously evaluated. 3. Does this change involve a significant reduction in a margin of safety? The proposed change will clarify Technical Specification 3.6.3 Action b to allow the use of a check valve with the flow through the valve secured as a means to isolate an inoperable containment penetration. This change is consistent with the changes identified in NUREG-1431, ``Improved Standard Technical Specifications for Westinghouse Plants'', Specification 3.6.3 (Containment Isolation Valves), which identifies check valves with flow through the valve secured as a type of deactivated automatic valve, and with 10 CFR 50 Appendix A General Design Criteria 55 and 56, which include the use of check valves as ``automatic isolation valves''. The proposed change only provides clarification to the existing Specification 3.6.3, and will not affect the established containment isolation valve OPERABILITY requirements or associated isolation time limits. The proposed change does not involve a significant reduction in a margin of safety because the ability to isolate containment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment will be maintained. The margin of safety is defined by the established containment isolation valve OPERABILITY requirements and associated isolation time limits. The proposed change does not alter these operating restrictions and the margin of safety which assures the ability to isolate containment is not affected. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration. Local Public Document Room location: Wharton County Junior College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488. Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869. NRC Project Director: George Dick, Acting. STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas Date of amendment request: March 9, 1999. Description of amendment request: The amendment request proposes that reference to the Independent Safety Engineering Group be removed from Technical Specification requirements, with supporting changes to the Operations Quality Assurance Plan. 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 proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment is a programmatic and administrative change that [[Page 17031]] does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. The functions assigned to the Independent Safety Engineering Group are addressed by other organizations. Because the design of the facility and system operating parameters are not being changed, the proposed amendment does not involve an increase in the probability or consequences of any accident previously evaluated. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment is a programmatic and administrative change that does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. The functions assigned to the Independent Safety Engineering Group are addressed by other organizations. Because the design of the facility and system operating parameters are not being changed, the proposed amendment does not create the possibility of a new or different kind of accident previously evaluated. The proposed change does not involve a significant reduction in a margin of safety. The proposed amendment is a programmatic and administrative change that provides assurance that plant operations continue to be conducted in a safe manner. The functions assigned to the Independent Safety Engineering Group are addressed by other organizations. As stated above the proposed amendment does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. Because the design of the facility and system operating parameters are not being changed, the proposed amendment 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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration. Local Public Document Room location: Wharton County Junior College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488. Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869. NRC Project Director: George F. Dick, Acting. STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas Date of amendment request: March 15, 1999 (Supplement to October 29, 1998). Description of amendment request: The proposed amendments were submitted by application dated October 29, 1998, to relocate Technical Specification (TS) 3/4.7.9 requirements for snubbers to the Technical Requirements Manual. The Commission issued a Notice of Consideration of Issuance of Amendments regarding its proposed no significant hazards consideration determination that was published in the Federal Register on December 16, 1998 (63 FR 69346). Subsequently, by letter dated March 15, 1999, supplemental information was submitted to include TS 6.10.3.l to be relocated to the Technical Requirements Manual. This information is being noticed to provide for public comment on the issue of no significant hazards consideration. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The supplement to the amendment request relocates the record keeping requirements of Technical Specification 6.10.3.l to the Technical Requirements Manual. The change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The change will not impose different requirements, and adequate control of information will be maintained. This change will not alter assumptions made in the safety analysis and licensing basis. The Technical Requirements Manual is incorporated in the South Texas Project Updated Final Safety Analysis Report and will be maintained pursuant to 10 CFR 50.59. In addition, snubber operability is addressed in existing surveillance procedures that are also controlled by 10 CFR 50.59 and subject to the change control provisions imposed by plant administrative procedures, which endorse applicable regulations and standards. Therefore, the supplement to the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The supplement to the amendment request relocates the record keeping requirements of Technical Specification 6.10.3.l to the Technical Requirements Manual. The change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The change will not impose different requirements, and adequate control of information will be maintained. This change will not alter assumptions made in the safety analysis and licensing basis. The Technical Requirements Manual is incorporated in the South Texas Project Updated Final Safety Analysis Report and will be maintained pursuant to 10 CFR 50.59. In addition, snubber operability is addressed in existing surveillance procedures that are also controlled by 10CFR50.59 and subject to the change control provisions imposed by plant administrative procedures, which endorse applicable regulations and standards. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the change involve a significant reduction in a margin of safety? The supplement to the amendment request relocates the record keeping requirements of Technical Specification 6.10.3.l to the Technical Requirements Manual. The relocated requirements remain the same as the existing Technical Specifications. The change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Future changes to the relocated requirements will be evaluated per the requirements of 10CFR50.59. Therefore, the supplement will not result in a reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration. Local Public Document Room location: Wharton County Junior College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488. Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869. NRC Project Director: George F. Dick, Acting. Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin Date of amendment request: February 15, 1999. Description of amendment request: The proposed amendment would revise requirements of Technical Specifications Section 6, ``Administrative Controls,'' related to (1) plant manager's responsibilities, (2) plant staff titles and organization, (3) offsite and onsite review committee (4) reportable events, and (5) actions required in event of a safety limit violation. 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 [[Page 17032]] consideration, which is presented below: The proposed amendment will not change the intent of the TS or decrease WPSC's management support or involvement in activities at the Kewaunee Plant. Furthermore, it will not result in a decrease in the engineering or technical support supplied by the plant staff or the corporate support staff. The proposed changes are administrative in nature. They primarily involve the relocation of existing requirements to owner controlled documents; therefore, there are no significant hazards associated with this change. As an administrative change this will not result in a significant increase in the probability of occurrence or consequences of an accident. As an administrative change this will not create the possibility of a new or different kind of accident from any previously analyzed. This administrative change relocates existing requirements, and therefore, will not involve a significant decrease in the margin of safety. In addition, the staff analyzed the proposed changes in accordance with the provisions of 10 CFR 50.92. The proposed change will not: 1. Involve a significant increase in the probability or consequence of an accident previously evaluated. The analyses for the previously evaluated accidents are presented in Chapter 14 of the Updated Safety Analysis Report. There are 19 postulated accidents addressed therein. The proposed amendment would not affect the safety analysis assumptions or analytical models used for any of these analyses. Also, the calculated dose consequences for analyzed accidents would be unaffected. Therefore the proposed changes do not involve a significant increase in the probability or consequence of an accident previously evaluated. 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed accident does not involve any physical change to the design of the physicality, or operation of the facility outside the bounds of the existing analyses. Thus, there is no possibility of creating a new or different kind of accident. 3. Involve a significant reduction in the margin of safety. The proposed changes do not involve any physical changes to any of the fission product barriers or to the design or operation of any safety systems. Also, no safety limits, limiting safety systems settings, limiting conditions for operation or testing requirements would be affected. Therefore, the proposed changes do not involve a significant reduction in the 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: University of Wisconsin, Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001. Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, P.O. Box 1497, Madison, WI 53701-1497. NRC Project Director: Cynthia A. Carpenter. Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power Station, Franklin County, Massachusetts Date of amendment request: March 17, 1999. Description of amendment request: Licensee submitted a License Amendment request to delete administrative Technical Specification (TS) requirements related to overtime restrictions. The licensee stated it will provide appropriate constraints on excessive overtime in its Administrative Procedures. 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 proposed changes are administrative in nature and simply eliminate outdated requirements from the YNPS Technical Specifications. As such the changes will not: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The administrative nature of the changes will not affect safety-related systems or components or their mode of operation and therefore, will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Create the possibility of a new or different accident from any previously evaluated. The proposed changes do not modify any plant systems or components and, therefore, do not create the possibility of a new or different accident from any previously evaluated. 3. Involve a significant reduction in the margin of safety. The changes are administrative in nature involving the deletion of outdated requirements in the technical specifications; therefore, there will be no reduction in the margin of safety. Based on the considerations noted above, it is concluded that the proposed changes will not endanger the public health and 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: Greenfield Community College, 1 College Drive, Greenfield, Massachusetts 01301. Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One International Place, Boston, Massachusetts 02110-2624. NRC Project Director: Seymour H. Weiss. Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power Station, Franklin County, Massachusetts Date of amendment request: March 17, 1999. Description of amendment request: Licensee submitted a License Amendment request to transfer Technical Specification Sections 6.7-- Procedures and Programs and 6.9--Record Retention to the Yankee Decommissioning Quality Assurance Program (YDQAP). 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 proposed changes are administrative in nature. Administrative requirements in Sections 6.7 and 6.9 of the YNPS Technical Specifications are to be transferred to the YDQAP which is the current location of related administrative requirements. As such the changes will not: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The administrative nature of the changes will not affect safety-related systems or components or their mode of operation and therefore, will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Create the possibility of a new or different accident from any previously evaluated. The proposed changes do not modify any plant systems or components and, therefore, will not create the possibility of a new or different accident from any previously evaluated. 3. Involve a significant reduction in the margin of safety. The changes are administrative in nature involving the relocation of administrative requirements from one licensing document to another licensing document currently containing related requirements; therefore, there will be no significant reduction in the margin of safety. Based on the considerations noted above, it is concluded that the proposed changes will not endanger the public health and safety. The NRC staff has reviewed the licensee's analysis and, based on this [[Page 17033]] 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: Greenfield Community College, 1 College Drive, Greenfield, Massachusetts 01301. Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One International Place, Boston, Massachusetts 02110-2624. NRC Project Director: Seymour H. Weiss. Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power Station, Franklin County, Massachusetts Date of amendment request: March 17, 1999. Description of amendment request: Licensee submitted a License Amendment request to consolidate management positions and to transfer Technical Specification review and audit functions to the Yankee Decommissioning Quality Assurance Program (YDQAP). 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 proposed changes are administrative in nature and reflect a streamlining of the YAEC/YNPS management structure and procedures consistent with the on-going requirement to complete the remaining scope of YNPS decommissioning safely and efficiently. As such the changes will not: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The administrative nature of the changes will not affect safety-related systems or components or their mode of operation and therefore, will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Create the possibility of a new or different accident from any previously evaluated. The proposed changes do not modify any plant systems or components and, therefore, will not create the possibility of a new or different accident from any previously evaluated. 3. Involve a significant reduction in the margin of safety. Elimination of the Manager of Operations position and the Plant Superintendent position will not eliminate any of the responsibilities or functions currently assigned to these positions. These responsibilities or functions will be reassigned to an appropriately qualified YAEC/YNPS manager, i.e., the Decommissioning Manager. This change and replacement of the PORC and the NSARC review and audit functions with an independent safety review and an IRAC are consistent with the significant reduction in the scope and the complexity of activities at YNPS as the facility moves into the later stages of the decommissioning effort; therefore, there will be no significant reduction in the margin of safety. Based on the considerations noted above, it is concluded that the proposed changes will not endanger the public health and 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: Greenfield Community College, 1 College Drive, Greenfield, Massachusetts 01301. Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One International Place, Boston, Massachusetts 02110-2624. NRC Project Director: Seymour H. Weiss. Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration. For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice. Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas Date of application for amendment: February 24, 1999. Brief description of amendment: The amendment would revise Technical Specification Table 3.3-1, ``Reactor Protective Instrumentation,'' Action 2, for Arkansas Nuclear One, Unit No. 2. The proposed change would add a footnote to Action 2 that would allow startup and operation with the functional units associated with the Channel ``D'' ex-core nuclear instrumentation to be maintained in the bypassed or tripped condition following the restart from Refueling Outage 2R13. This footnote is intended to support normal plant operations until such time that the Channel ``D'' ex-core detector assembly can be restored to an operable status. Date of publication of individual notice in Federal Register: March 8, 1999 (64 FR 11067). Expiration date of individual notice: April 7, 1999. Local Public Document Room location: Tomlinson Library, Arkansas Tech University, Russellville, AR 72801. 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 17034]] CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz Mill, Pennsylvania Date of application for amendment: September 28, 1998 supplemented on November 17, 1998. Brief description of amendment: This amendment changes the license to reflect the new legal name of the licensee for the Westinghouse Test Reactor to CBS Corporation. Date of issuance: March 25, 1999. Effective Date: March 25, 1999. Amendment No: 9. Facility License No. TR-2: This amendment changes the license. Date of initial notice in Federal Register: December 16, 1998, (63 FR 69334). The Commission has issued a Safety Evaluation for this amendment dated March 25, 1999. No significant hazards consideration comments received: No. Local Public Document: N/A. Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Date of application for amendments: December 17, 1998. Brief description of amendments: The amendments revised the respective facility Technical Specifications (TS) by adding a new Limiting Condition for Operations that provided an administrative enhancement by allowing testing required to return equipment to service to be conducted under administrative controls. Date of issuance: March 16, 1999. Effective date: Immediately, to be implemented within 60 days. Amendment Nos.: 172, 167; 184, 181; 132, 117. Facility Operating License Nos. DPR-19, DPR-25, DPR-29, DPR-30, NPF-11 and NPF-18. The amendments revised the Technical Specifications. Date of initial notice in Federal Register: January 27, 1999 (64 FR 4153) The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 16, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: for Dresden, Morris Area Public Library District, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021; for LaSalle, the Jacobs Memorial Library, 815 North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, Illinois 61348-9692. Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Date of application for amendments: August 14, 1998, as supplemented on October 13 and December 23, 1998. Brief description of amendments: The amendments revised the Technical Specifications (TSs) to reflect the use of Siemens Power Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments incorporate the following into the TSs: (1) new methodologies that will enhance operational flexibility and reduce the likelihood of future plant derates; (2) administrative changes that adopt Improved Standard Technical Specification (iSTS) language where appropriate; and (3) changes to the Minimum Critical Power Ratio. Date of issuance: March 16, 1999. Effective date: Immediately, to be implemented prior to startup of Cycle 9 for Unit 1 and prior to startup of Cycle 8 for Unit 2. Amendment Nos.: 131, 116. Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the TSs. Date of initial notice in Federal Register: November 4, 1998 (63 FR 59588). The December 23, 1998, submittal provided additional clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 16, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Jacobs Memorial Library, 815 North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, Illinois 61348-9692. Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of application for amendments: August 14, 1998, as supplemented by letters dated October 13, 1998, and December 23, 1998. Brief description of amendments: The amendments changed the Quad Cities Technical Specifications (TS) to reflect the use of Siemens Power Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments incorporate the following into the TS: (a) new methodologies that will enhance operational flexibility and reduce the likelihood of future plant derates; (b) administrative changes that eliminate the cycle- specific implementation of ATRIUM-9B fuel and adopt Improved Standard Technical Specification language where appropriate; and (c) changes to the Minimum Critical Power Ratio (MCPR). The amendment for Unit 1 also reflects the removal of Unit 1 specific pages incorporated into Unit 1 TS by Amendment No. 182 and are no longer applicable. The August 14, 1998, application superseded an August 29, 1997, application in its entirety (63 FR 2274). Date of issuance: March 17, 1999. Effective date: Immediately, to be implemented within 60 days. Amendment Nos.: 185 & 182. Facility Operating License Nos. DPR-29, DPR-30: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: September 9, 1998 (63 FR 48258) and November 4, 1998 (63 FR 59588). The October 13, 1998, submittal changed a reference to a recently NRC-approved additive constant uncertainty (ACU) generic methodology for ATRIUM-9B fuel (ANF- 1125 (P)(A), supplement 1, Appendix E) from Appendix D which provided an interim value for ACU. This change was noticed on November 4, 1998 (63 FR 48258). The December 23, 1998, submittal provided additional clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 17, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021. Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of application for amendments: November 30, 1998. Brief description of amendments: The amendments changed the technical specifications (TSs) by decreasing the Allowed Outage Time (AOT) from 67 days to 14 days for the Safe Shutdown Makeup Pump (SSMP). Date of issuance: March 26, 1999. Effective date: Immediately, to be implemented within 60 days. Amendment Nos.: 186 & 183. [[Page 17035]] Facility Operating License Nos. DPR-29 and DPR-30: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: January 13, 1999 (64 FR 2246). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021. Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina. Date of application of amendments: September 30, 1998. Brief description of amendments: The amendments increase the maximum fuel rod internal pressure in the spent fuel pool from 1200 pounds per square inch gauge (psig) to 1300 psig by changing the Updated Final Analysis Report (UFSAR) reference to the computer code used to determine the fuel rod internal pressure (TACO3 computer code would be added) in UFSAR Chapter 15. In addition, the amendments justify not increasing the overall effective decontamination factor for iodine as a consequence of a fuel handling accident and change the terminology used in the UFSAR from ``fuel assembly gap gas pressure'' to ``fuel rod internal pressure.'' Date of Issuance: March 26, 1999. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment Nos.: Unit 1-301; Unit 2-301; Unit 3-301. Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments authorized change(s) to the FSAR. Date of initial notice in Federal Register: November 4, 1998 (63 FR 59590). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Oconee County Library, 501 West South Broad Street, Walhalla, South Carolina. Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas. Date of application for amendment: February 25, 1999. Brief description of amendment: This amendment revises Technical Specification (TS) Table 3.3-1, ``Reactor Protective Instrumentation,'' Action 2, for Arkansas Nuclear One, Unit No. 2 (ANO-2). This change adds a footnote to Action 2 that allows startup and operation with the functional units associated with the Channel ``D'' ex-core nuclear instrumentation to be maintained in the bypassed or tripped condition following the restart from Refueling Outage 2R13. This footnote is intended to support normal plant operations until such time that the Channel ``D'' ex-core detector assembly can be restored to an operable status. This footnote will be in effect for a time period not to extend beyond Mid-Cycle Outage 2P99, which is the next planned entry into cold shutdown conditions for ANO-2. A Notice of Enforcement Discretion (NOED) related to TS Table 3.3-1, Action 2, was issued verbally on February 23, 1999. The NOED is documented in a letter dated February 25, 1999. Date of issuance: March 23, 1999. Effective date: As of the date of issuance. Amendment No.: 202. Facility Operating License No. NPF-6: Amendment revised the Technical Specifications. Public comments requested as to proposed no significant hazards consideration (NSHC): Yes (64 FR 11067 dated March 8, 1999). The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided for an opportunity to request a hearing by April 7, 1999, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment. The Commission's related evaluation of the amendment, finding of exigent circumstances, and final NSHC determination are contained in a Safety Evaluation dated March 23, 1999. Attorney for Licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., Washington DC 20005-3502. 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: November 2, 1995, and as supplemented by submittal dated January 7, 1999. Brief description of amendment: This amendment revises technical specification requirements for handling irradiated fuel in the Primary Containment and the Fuel Handling Building, and selected specifications associated with performing core alterations. Date of issuance: March 11, 1999. Effective date: March 11, 1999. Amendment No.: 102. Facility Operating License No. NPF-58: This amendment revised the Technical Specifications. Date of initial notice in Federal Register: December 6, 1995 (60 FR 62497). The supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not change the scope of the original application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 11, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Perry Public Library, 3753 Main Street, Perry, OH 44081. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio. Date of application for amendment: August 27, 1996, as supplemented by submittals dated April 9, 1997, July 22, 1998, December 3, 1998, and January 18, 1999. Brief description of amendment: This amendment revised Technical Specification 3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' and 3.6.1.9, ``Main Steam Isolation Valve (MSIV) Leakage Control System (LCS).'' The amendment reflects implementation of the revised accident source term in NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power Plants'' and permits the licensee to eliminate the MSIV LCS and increase the allowable leak rates of the MSIVs. Date of issuance: March 26, 1999: Effective date: March 26, 1999. Amendment No.: 103. Facility Operating License No. NPF-58: This amendment revised the Technical Specifications. Date of initial notice in Federal Register: October 7, 1998 (63 FR 53958). The supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. [[Page 17036]] Local Public Document Room location: Perry Public Library, 3753 Main Street, Perry, OH 44081. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio. Date of application for amendment: October 27, 1998. Brief description of amendment: This amendment revised the minimum critical power ratio (MCPR) safety limit contained in TS 2.1.1.2. In addition, the amendment removes a note to TS 2.1.1.2 and a footnote to TS 5.6.5.b that references MCPR safety limit values as cycle specific. Date of issuance: March 26, 1999: Effective date: March 26, 1999. Amendment No.: 104. Facility Operating License No. NPF-58: This amendment revised the Technical Specifications. Date of initial notice in Federal Register: December 2, 1998 (63 FR 66603). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Perry Public Library, 3753 Main Street, Perry, OH 44081. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio. Date of application for amendment: September 9, 1998, as supplemented by submittals dated January 6, March 4, and March 18, 1999. Brief description of amendment: This amendment revises the design and licensing basis of containment isolation valves in the feedwater system. The amendment revises (1) Surveillance Requirement 3.6.1.3.11 of Technical Specification (TS) 3.6.1.3, ``Primary Containment Isolation Valves (PCIVs)'' to exclude the feedwater check valves from the hydrostatic test program, (2) TS 5.5.2, ``Primary Coolant Sources Outside Containment,'' to stipulate that water leakage past the feedwater motor-operated containment isolation valves and the reactor water cleanup system return to feedwater line is added to the program, and (3) TS 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' to state that the feedwater check valves will be tested in accordance with the Inservice Testing Program (TS 5.5.6). Date of issuance: March 26, 1999. Effective date: March 26, 1999. Amendment No.: 105. Facility Operating License No. NPF-58: This amendment revised the Technical Specifications. Date of initial notice in Federal Register: October 21, 1998 (63 FR 56262). The supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Perry Public Library, 3753 Main Street, Perry, OH 44081. Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida. Date of application for amendments: August 24, 1998. Brief description of amendments: These amendments change the St. Lucie Technical Specifications (TSs) by both removing obsolete license conditions and revising the TSs. The amendments change the TSs to modify the St. Lucie Unit 1 TSs to add components, not previously described in the TSs, to the list of components that comprise an operable control room emergency ventilation system, to modify the Unit 1 and Unit 2 TSs surveillance requirements to clarify component operations, not previously described, that must be verified in response to a containment sump recirculation actuation signal, to delete from the facility operating license No. NPF-16 for Unit 2, license condition 2.C.19 to reflect the completion of the Unit 1 spent fuel pool re-rack and delete license condition 2.I to reflect the resolution of litigation and to modify license condition 2.B.5 to restore the original syntax of the license condition and license condition 2.F to update the references to current license conditions. Date of Issuance: March 17, 1999. Effective Date: These amendments shall be implemented within 30 days of receipt. Amendment Nos.: 160 and 99. Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the TSs. Date of initial notice in Federal Register: September 23, 1998 (63 FR 50937). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 17, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003. GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey. Date of application for amendment: September 3, 1998. Brief description of amendment: The amendment revises Technical Specifications 3.4.A.10.e and 3.5.a.2.e to incorporate a Condensate Storage Tank water level of greater than 35 feet. Date of Issuance: March 17, 1999. Effective date: March 17, 1999, to be implemented within 30 days Amendment No.: 204. Facility Operating License No. DPR-16. Amendment revised the Technical Specifications. Date of initial notice in Federal Register: February 10, 1999 (64 FR 6698). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated March 17, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Ocean County Library, Reference Department, 101 Washington Street, Toms River, NJ 08753. Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois. Date of application for amendment: January 20, 1999, as supplemented February 4, 8, and 25, and March 5, 1999. Brief description of amendment: The amendment changes the undervoltage relay setpoints. Date of issuance: March 26, 1999. Effective date: March 26, 1999. Amendment No.: 122. Facility Operating License No. NPF-62: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: January 28, 1999 (64 FR 4474). The four supplemental submittals provided additional information and did not change the requested amendment or affect the proposed no significant hazards consideration. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: The Vespasian Warner Public Library, 120 West Johnson Street, Clinton, IL 61727. [[Page 17037]] Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee Atomic Power Station, Lincoln County, Maine Date of application for amendment: April 13, 1998, as supplemented November 5, 1998. Brief description of amendment: The proposed amendment would revise the Appendix A Technical Specifications to base the Limiting Condition for Operation for the fuel storage pool water level on a revised analysis of the fuel handling accident and a new analysis for radiological shielding during movement of irradiated fuel. Date of issuance: March 16, 1999. Effective date: March 16, 1999 (and shall be implemented no later than 30 days). Amendment No.: 162. Facility Operating License No. DPR-36: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: May 20, 1998 (63 FR 27763). The November 5, 1998, submittal provided additional clarifying information and did not change the initial proposed no significant hazards determination and did not expand the scope of the original application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Wiscasset Public Library, High Street, P.O. Box 367, Wiscasset, ME 04578. Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point Nuclear Station Unit No. 1, Oswego County, New York Date of application for amendment: December 30, 1998. Brief description of amendment: The amendment changes Technical Specification (TS) Tables 3.6.14-2 and 4.6.14-2 regarding the noble gas activity monitor channel operability requirement and daily sensor check surveillance requirement to be consistent with the conditions specified in TS 3.1.3.a for operability of the emergency cooling system. Also, this amendment corrects a clerical error in TS 4.6.15.d. Date of issuance: March 16, 1999. Effective date: As of the date of issuance to be implemented within 30 days. Amendment No.: 165. Facility Operating License No. DPR-63: Amendment revises the Technical Specifications. Date of initial notice in Federal Register: February 10, 1999 (64 FR 6699). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Reference and Documents Department, Penfield Library, State University of New York, Oswego, New York 13126. Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York Date of application for amendment: November 19, 1998. Brief description of amendment: This amendment changes surveillance frequencies in Technical Specifications 4.8.4.4a and 4.8.4.5a to require testing of the Electrical Protection Assemblies once every 6 months with the plant on-line rather than shut down. Date of issuance: March 18, 1999. Effective date: As of the date of issuance to be implemented within 30 days. Amendment No.: 86. Facility Operating License No. NPF-69: Amendment revises the Technical Specifications Date of initial notice in Federal Register: December 30, 1998 (63 FR 71970). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 18, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Reference and Documents Department, Penfield Library, State University of New York, Oswego, New York 13126. North Atlantic Energy Service Corporation, et al., Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire Date of amendment request: May 20, 1998, as supplemented by letter dated January 28, 1999. Description of amendment request: Revise Technical Specifications Table 3.3-4 and associated bases to depict a change to the refueling water storage tank low-low level setpoint Date of issuance: March 12, 1999. Effective date: As of its date of issuance, to be implemented within 60 days. Amendment No.: 60. Facility Operating License No. NPF-86. Amendment revised the Technical Specifications Date of initial notice in Federal Register: August 12, 1998 (63 FR 43205). The supplemental letter provided clarifying information and did not change the staff's proposed no significant hazards determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 12, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Exeter Public Library, Founders Park, Exeter, NH 03833. Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear Power Station, Unit No. 2, New London County, Connecticut Date of application for amendment: September 9, 1998, as supplemented February 19 and 26, 1999. Brief description of amendment: The amendment resolves several previously identified technical specifications (TSs) compliance issues. Specifically, the amendment: (1) changed TS definitions 1.24, ``Core Operating Limits Report,'' 1.27, ``Engineering Safety Feature Response Time,'' and 1.31, ``Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM)''; (2) changed TS 3.0.2, ``Limiting Condition for Operation,'' by adding a new TS 3.0.6 to the Limiting Condition for Operation TS section; (3) changed TS 4.0.5, ``Surveillance Requirements''; (4) changed the mode applicability of TS 3.2.3, ``Total Unrodded Integrated Radial Peaking--F r T''; (5) changed TS 3.3.2.1, ``Engineered Safety Features Actuation System Instrumentation,'' by modifying TS Table 4.3-2 Table Notation (1) which it references; and (6) changed TS 3.4.1.1, ``Reactor Coolant System-- Reactor Coolant System Vents.'' The associated TS Bases sections were also changed. Date of issuance: March 11, 1999. Effective date: As of the date of issuance to be implemented within 60 days from the date of issuance. Amendment No.: 230. Facility Operating License No. DPR-65: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: October 21, 1998 (63 FR 56251). The supplemental letters provided clarifying information that did not change the original 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 March 11, 1999. [[Page 17038]] 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. Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear Power Station, Unit No. 2, New London County, Connecticut Date of application for amendment: July 17, 1998, as supplemented November 10, 1998, and February 11, 1999. Brief description of amendment: The amendment revises certain diesel generator (DG) action statements and surveillance requirements to improve overall DG reliability and availability. Date of issuance: March 12, 1999. Effective date: As of the date of issuance to be implemented within 60 days from the date of issuance. Amendment No.: 231. Facility Operating License No. DPR-65: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: August 12, 1998 (63 FR 43207). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 12, 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 No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of application for amendment: November 25, 1997, as supplemented September 25 and November 11, 1998, and January 28, 1999. Brief description of amendment: The amendment revises the Technical Specifications for the condensate storage tank (CST) low level suction transfer setpoint for the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems to allow removing one CST from service for maintenance. Date of issuance: March 19, 1999. Effective date: March 19, 1999, with full implementation within 30 days. Amendment No.: 105. Facility Operating License No. DPR-22. Amendment revised the Technical Specifications. Date of initial notice in Federal Register: December 18, 1998 (63 FR 69344) The November 25, 1997, letter and September 25 and November 11, 1998, supplements were referenced in the original Federal Register notice. The January 28, 1999, supplement provided an updated Technical Specification page following the incorporation of Amendment 103, issued December 23, 1998. This information was within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards considerations determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 19, 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. 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: November 25, 1998. Brief description of amendments: The amendments revise Technical Specifications 3.2 and Table 3.5-2B to allow limited inoperability of boric acid storage tank level channels and transfer logic channels to provide for required testing and maintenance of the associated components. Date of issuance: March 17, 1999. Effective date: March 17, 1999, with full implementation within 30 days Amendment Nos.: 143 and 134. Facility Operating License Nos. DPR-42 and DPR-60. Amendments revised the Technical Specifications. Date of initial notice in Federal Register: December 16, 1998 (63 FR 69345). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 19, 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. Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California Date of application for amendments: July 30, 1997, as supplemented by letter dated December 23, 1998. Brief description of amendments: The amendments revise the combined Technical Specifications (TS) for the Diablo Canyon Power Plant (DCPP) Unit Nos. 1 and 2 by adding a Limiting Condition for Operation, trip setpoints, and surveillance requirements for a residual heat removal pump trip on refueling water storage tank level-low. Date of issuance: March 26, 1999. Effective date: March 26, 1999, to be implemented within 30 days from the date of issuance. Amendment Nos.: Unit 1--130; Unit 2--128. Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Technical Specifications. Date of initial notice in Federal Register: December 31, 1997 (62 FR 68312). The December 31, 1997 supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noted, and did not change the staff's proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: California Polytechnic State University, Robert E. Kennedy Library, Government Documents and Maps Department, San Luis Obispo, California 93407. Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California Date of application for amendments: August 26, 1997, as supplemented by letters dated October 14 and November 13, 1997, and January 29, 1998. Brief description of amendments: The amendments approve a modification to the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2 auxiliary saltwater (ASW) system to bypass approximately 800 feet of Unit 1 and 200 feet of Unit 2 Class 1 ASW pipe, a portion of which is buried below sea level in the tidal zone outside the intake structure. Date of issuance: March 26, 1999. Effective date: March 26, 1999, and shall be implemented in the next [[Page 17039]] periodic update to the FSAR Update in accordance with 10 CFR 50.71(e). Amendment Nos.: Unit 1--131; Unit 2--129. Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Final Safety Analysis Report Update. Date of initial notice in Federal Register: September 16, 1997 (62 FR 48677). The October 14 and November 13, 1997, and January 29, 1998, supplemental letters provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: California Polytechnic State University, Robert E. Kennedy Library, Government Documents and Maps Department, San Luis Obispo, California 93407. 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 application for amendments: September 25, 1996, as supplemented on October 29, 1997, March 16, 1998, and February 8, 1999. Brief description of amendments: The amendments revise the Technical Specifications by revising the voltage and frequency acceptance criteria and the start-timing methodology for the emergency diesel generator surveillance testing. Date of issuance: March 23, 1999. Effective date: As of date of issuance, to be implemented within 60 days. Amendment Nos: 218 and 200. Facility Operating License Nos. DPR-70 and DPR-75. The amendments revised the Technical Specifications. Date of initial notice in Federal Register: October 23, 1996 (61 FR 5039). The October 29, 1997, March 16, 1998, and February 9, 1999, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 23, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Salem Free Public Library, 112 West Broadway, Salem, NJ 08079. 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 application for amendments: September 17, 1998. Brief description of amendments: The amendments revise Technical Specification 3/4.8.2, ``Electrical Power Sources--Shutdown,'' for the AC distribution system and the 125-volt and 28-volt DC distribution systems. Specifically, the amendments change the Applicability and Action Statements, if less than the complement of equipment and buses are operable, to eliminate the need to establish containment integrity and to add the action to suspend core alterations, positive reactivity additions, and movement of irradiated fuel assemblies. Date of issuance: March 24, 1999. Effective date: As of the date of issuance, to be implemented within 60 days. Amendment Nos.: 219 and 201. Facility Operating License Nos. DPR-70 and DPR-75. The amendments revised the Technical Specifications. Date of initial notice in Federal Register: October 21, 1998 (63 FR 56257). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 24, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Salem Free Public Library, 112 West Broadway, Salem, NJ 08079. 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: April 6, 1995, as supplemented on August 21, 1995. (TS 95-19). Brief description of amendments: The amendments change the licenses for Sequoyah Nuclear Plant, Units 1 and 2 by removing the license conditions that reference the post-accident sampling system (PASS). The PASS information has been placed in the Sequoyah Final Safety Analysis Report (FSAR). This Change is consistent with NUREG-1431, ``Standard Technical Specifications--Westinghouse Plants.'' Date of issuance: March 16, 1999. Effective date: March 16, 1999. Amendment Nos.: 243 and 233. Facility Operating License Nos. DPR-77 and DPR-79: The amendments revise the licenses. Date of initial notice in Federal Register: April 26, 1995 (60 FR 20527). The August 21, 1995, letter provided clarifying information that did not change the original no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 1999. No significant hazards consideration comments received: None. Local Public Document Room location: Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402. Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont Date of application for amendment: April 23, 1998, as supplemented on January 25, 1999. Brief description of amendment: The amendment changes the existing requirements for the Residual Heat Removal Service Water (RHRSW), Station Service Water (SSW) and Alternate Cooling Tower Systems (ACS) as identified in Technical Specifications (TSs) 4.5.C and 3/4.5.D. Specifically, the changes are as follows: (1) Specifications 3.5.D.3 and 4.5.D.3: This requirement is revised to delete the existing allowance for 7 days of operation after both SSW subsystems are made or found to be inoperable. (2) Specification 4.5.C.1 and Specification 4.5.D.1: These requirements have been revised to relocate testing information related to pump flow and pressure testing characteristics for the RHRSW and SSW Systems, respectively, to the Technical Requirements Manual. (3) Specifications 3.5.D.1, 3.5.D.2, 3.5.D.3, 4.5.D.2, 4.5.D.3, and associated Bases: All references to SSW ``subsystem'' have been replaced by ``essential equipment cooling loop'' to more accurately reflect the Vermont Yankee design and operation. In addition, certain operability clarifications have been made to the Bases relative to affected Specifications. (4) Bases for Specification 3.5.D: The Bases have been revised to omit statements that imply that the ACS could provide adequate heat removal following a postulated accident. Other Bases additions have been made that include certain operability clarifications relative to affected Specifications. Date of Issuance: March 11, 1999. Effective date: As of the date of issuance, to be implemented within 30 days. Amendment No.: 169. Facility Operating License No. DPR-28: Amendment revised the Technical Specifications. [[Page 17040]] Date of initial notice in Federal Register: February 10, 1999 (64 FR 6713). The January 25, 1999, supplement did not affect the original proposed no significant hazards consideration. The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated March 11, 1999. No significant hazards consideration comments received: No. Local Public Document Room location: Brooks Memorial Library, 224 Main Street, Brattleboro, Vermont 05301. Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas Date of amendment request: March 20, 1998, as supplemented by letters dated May 28, June 30, August 28, September 4, November 20, and December 8, 1998. Brief description of amendment: The amendment revised the technical specifications (TS) to support a modification to the plant to increase the storage capacity of the spent fuel pool and increase the nominal fuel enrichment to 5% weight percent of U-235. The amendment also revised the TS to allow the storage of an additional 279 assemblies in the cask loading pit. Date of issuance: March 22, 1999. Effective date: March 22, 1999, to be fully implemented no later than December 31, 1999, except that the racks in the cask loading pit may be installed at a future time after the completion of the next refueling outage. Amendment No.: 120. Facility Operating License No. NPF-42: The amendment revised the Technical Specifications. Date of initial notice in Federal Register: July 13, 1998 (63 FR 37601). The June 30, August 28, September 4, November 20, and December 8, 1998, supplemental letters provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the staff's proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 22, 1999. No significant hazards consideration comments received: No. Local Public Document Room locations: Emporia State University, William Allen White Library, 1200 Commercial Street, Emporia, Kansas 66801 and Washburn University School of Law Library, Topeka, Kansas 66621. Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas Date of amendment request: February 4, 1998, as supplemented by letter dated October 20, 1998. Brief description of amendment: The amendment revises the requirements in Technical Specification Tables 3.3-3, 3.3-4 and 4.3-2 regarding the engineered safety features actuation system (ESFAS) Functional Unit 6.f, and adds a note to Table 4.3-2 to clarify the verification of time delays associated with ESFAS Functional Units 8.a and 8.b. Date of issuance: March 23, 1999. Effective date: March 23, 1999, to be implemented within 30 days from the date of issuance. Amendment No.: 121. Facility Operating License No. NPF-42. The amendment revised the Technical Specifications. Date of initial notice in Federal Register: March 25, 1998 (63 FR 14491). The October 20, 1998, supplemental letter provided additional clarifying information, did not expand the scope of the application as originally noticed and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 23, 1999. No significant hazards consideration comments received: No. Local Public Document Room locations: Emporia State University, William Allen White Library, 1200 Commercial Street, Emporia, Kansas 66801 and Washburn University School of Law Library, Topeka, Kansas 66621. Dated at Rockville, Maryland, this 31st day of March 1999. For the Nuclear Regulatory Commission. Suzanne C. Black, Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation. [FR Doc. 99-8503 Filed 4-6-99; 8:45 am] BILLING CODE 7590-01-P
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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
[Federal Register Volume 64, Number 66 (Wednesday, April 7, 1999)] [Notices] [Pages 17021-17040] From the Federal Register Online via the Government Publishing Office [ www.gpo....
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