Meetings: Operating licenses, amendments; no significant hazards considerations; biweekly notice,

[Federal Register: April 8, 1998 (Volume 63, Number 67)]

[Notices]

[Page 17219-17242]

From the Federal Register Online via GPO Access [wais.access.gpo.gov]

[DOCID:fr08ap98-142]

NUCLEAR REGULATORY COMMISSION

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

1. Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from March 16, 1998, through March 27, 1998. The last biweekly notice was published on March 25, 1998.

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 May 8, 1998, 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 filedin 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 filedby 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

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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 fileda petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross- examine witnesses.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.

If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filedwith 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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: February 11, 1998.

Description of amendment request: The proposed amendment would modify the Pilgrim Nuclear Power Station (PNPS) Updated Final Safety Analysis Report (UFSAR) Section 10.7, Salt Service Water System, by identifying that certain single active failures do exist that could leave the Salt Service Water (SSW) system in a configuration with one SSW pump serving both SSW trains through open crossover (division) valves for the first 10 minutes of an accident.

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

   Operation with one (1) SSW pump supplying two (2) SSW trains is not an accident or transient precursor and does not prevent the [Reactor Building Closed Cooling Water] RBCCW system from providing adequate cooling during an accident. Core cooling requires no SSW for the first ten minutes, and no containment cooling is assumed for the first ten minutes. Pump testing has proved no SSW pump damage will result from this configuration so there will be no effect on the containment cooling function. The current licensing basis includes operator action after ten minutes to align the SSW system to achieve containment cooling. This amendment does not affect operator action after ten minutes since pump and valve manipulations are already required to align containment cooling. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The SSW system operating modes are not accident precursors. They cannot influence the types of accidents that can occur. The SSW pumps can withstand operation under the full range of conditions and for the time periods considered under a one pump, two train system configuration with no adverse effects. The SSW system is properly designed as a common header arrangement with five (5) pumps in which any combination of one to five pumps may operate without damaging effects.

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

   Operation with one (1) SSW pump supplying two (2) SSW trains does not impact the ability to provide adequate core or containment cooling during an accident. Although SSW system flow will be diminished during the first ten minutes of the accident, no system flow at all is needed at that time. The current licensing basis credits operator action after ten minutes to align the [Residual Heat Removal] RHR, RBCCW, and SSW systems for containment cooling.

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   Operators are expected to isolate the SSW loops or start additional SSW pumps as necessary given the existing specific conditions.

   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: Plymouth Public Library, 11 North Street, Plymouth, Massachusetts 02360.

   Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

   NRC Project Director: Cecil O. Thomas.

   Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

   Date of amendment request: February 20, 1998.

   Description of amendment request: The proposed amendment would change the Pilgrim Nuclear Power Station Technical Specification (TS) 3/4.5.B and its Bases to incorporate the ultimate heat sink (UHS) temperature of 75 deg.F, as required by Amendment No. 173. The introduction of a UHS temperature restriction requires new specifications, actions, and surveillances for the salt service water system.

   The amendment would also replace existing Specification 3.5.B ``Containment Cooling System'' with new Specification 3/4.5.B.1 ``Residual Heat Removal (RHR) Suppression Pool Cooling,'' 3/4.5.B.2 ``Residual Heat Removal (RHR) Containment Spray,'' 3/4.5.B.3 ``Reactor Building Closed Cooling Water (RBCCW) System,'' and 3/4.5.B.4 ``Salt Service Water (SSW) System and Ultimate Heat Sink (UHS).'' The proposed new subsections will more clearly define the various subsystems that comprise the containment cooling system and the operating states in which they are applicable. The proposed changes also provide clarity with respect to the application of limiting conditions of operation (LCOs), actions, completion times, and surveillances for the containment cooling subsystems.

   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?

   Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: Administrative Changes

   These proposed changes (editorial rewording, reformatting, repagination, and renumbering) are made to restructure the section, accounting for the new specifications replacing Specification 3/4.5.B. These proposed administrative changes do not alter any existing requirements. Technical Changes--More Restrictive

   The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. The more stringent requirements provide greater assurance that the affected systems will remain capable of providing the safety functions assumed in design basis accidents and transients. If anything, the new requirements may decrease the probability or consequences of an analyzed event. The change will not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements will not alter the operation of process variables, structures, systems, or components as described in the safety analyses. Technical Changes--Relocations

   This proposed change relocates requirements from the Technical Specifications to the Inservice Testing (IST) Program. The (IST) Program documents containing the relocated requirements must be maintained using the provisions of 10 CFR 50.55a and 10 CFR 50.59. Since any changes to the (IST) Program documents will be evaluated per 10 CFR 50.55a and 10 CFR 50.59, no increase in the probability or consequences of an accident previously evaluated will be allowed without NRC review. Technical Changes--Less Restrictive

   This change relaxes the current requirements to declare the affected RBCCW subsystem inoperable when one of the required RBCCW pumps is inoperable. Since the RBCCW system is not assumed as an initiator of any analyzed event, the proposed change will not affect the probability of an accident occurring. The safety function of the RBCCW system is to support the operability of the RHR suppression pool cooling and spray functions, and component cooling for the RHR and core spray pumps, and area coolers. With one required RBCCW pump inoperable, the remaining pump in the affected subsystem is capable of supporting the component cooling requirements for the RHR and core spray pumps, and area coolers, and the remaining OPERABLE subsystem is capable of supporting the suppression pool cooling and spray functions.

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

   Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: Administrative Changes

   The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing plant operation. The proposed changes will not impose any new or different requirements or eliminate any existing requirements. Technical Changes--More Restrictive

   The proposed more restrictive requirements will not alter the plant configuration (no new or different type of equipment will be installed) or change methods governing plant operation. The change does impose different requirements. However, the changes are consistent with assumptions made in the safety analyses. Technical Changes--Relocations

   This change relocates requirements to the (IST) Program. This change will not alter the plant configuration (no new or different type of equipment will be installed) or changes in methods governing plant operation. This 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. Technical Changes--Less Restrictive

   The proposed change will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these systems are operated, maintained, modified, tested, or inspected.

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

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   Administrative Changes

   Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: safety analysis margin of safety.

   The changes are administrative in nature and do not involve any technical changes. Since no technical changes (either actual or interpretational) were made, there is no impact on any safety analysis margin of safety. Technical Changes--More Restrictive

   The proposed more restrictive requirements will not alter assumptions relative to mitigation of an accident or transient event or alter the operation of process variables, structures, systems, or components as described in the safety analyses. Technical Changes--Relocations

   This change relocates requirements from the Technical Specifications to the Inservice Testing (IST) Program. The requirements to be transposed to the IST program are the same as the existing Technical Specifications. Since any changes to the (IST) Program documents will be evaluated per 10 CFR 50.55a and 10 CFR 50.59, no reduction in margin of safety previously approved will be allowed without NRC review. Technical Changes--Less Restrictive

   The 7 day completion time is consistent with the completion times for one inoperable loop of suppression pool cooling system or containment spray system, and the remaining pump in the affected subsystem is capable of supporting the component cooling requirements for the RHR and core spray pumps, and area coolers.

   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: Plymouth Public Library, 11 North Street, Plymouth, Massachusetts 02360.

   Attorney for licensee: W.S. Stowe, Esquire, Boston Edison Company, 800 Boylston Street, Massachusetts 02199.

   NRC Project Director: Cecil O. Thomas.

   Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

   Date of amendment request: March 12, 1998.

   Description of amendment request: The proposed amendment revises Technical Specification (TS) 3/4.9.12, ``Fuel Handling Building Emergency Exhaust System.'' Specifically, Harris Nuclear Plant (HNP) proposes to delete Surveillance Requirement 4.9.12.d.4, which requires verifying that the filter cooling bypass valve for the Fuel Handling Building Emergency Exhaust System is locked in the balanced position at least once per 18 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 amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   Fuel Handling Building Emergency Exhaust System (FHBEES) is not an accident initiating system as described in the Final Safety Analysis Report. The proposed change allows the elimination of the filter cooling bypass flowpath for FHBEES units. Engineering calculations were performed which demonstrate this filter cooling path is not required to mitigate the consequences of a fuel handling accident.

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

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

   FHBEES is a ventilation system designed to limit off-site dose releases in the event of a fuel handling accident. FHBEES is not an accident initiating system as described in the Final Safety Analysis Report [FSAR]. The proposed change ensures the seismic and safety classification is maintained while not affecting another Structure, System, or Component.

   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 amendment does not involve a significant reduction in the margin of safety.

   The proposed change to FHBEES does not affect any of the parameters that relate to the margin of safety as described in the Bases of the TS or the FSAR. Accordingly, NRC Acceptance Limits are not affected by this change.

   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: Cameron Village Regional Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

   NRC Project Director: Pao Tsin Kuo, Acting Director.

   Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

   Date of amendment request: September 3, 1997, as supplemented March 13, 1998.

   Description of amendment request: The proposed amendment would revise the technical specifications (TS) to delete snubber operability requirements (Change A), action requirements for inoperable snubbers (Change B), and snubber testing requirements (Change E). The snubber testing requirements would be relocated to the Palisades Operating Requirements Manual (ORM). Each proposed change has been classified by the licensee as either Administrative, More Restrictive, or Less Restrictive.

   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:

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

1. Administrative Change (Change A):

   ``Administrative'' changes make wording changes which clarify existing TS requirements, without affecting their technical content. Since ``Administrative'' changes do not alter the technical content of any requirements, they cannot involve a significant increase in the probability or consequences of an accident previously evaluated.

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2. More Restrictive Change (Change B):

   ``More Restrictive'' changes only add new requirements, or revise existing requirements to result in additional operational restrictions. The TS, with all ``More Restrictive'' changes incorporated, will still contain all of the requirements which existed prior to the changes. Therefore, ``More Restrictive'' changes cannot involve a significant increase in the probability or consequences of an accident previously evaluated.

3. Less Restrictive Change (Change E):

   Change E deletes the TS requirements for snubber testing, but adds identical requirements to a document (the ORM) controlled under 10 CFR 50.59.

   10 CFR 50.59 specifically prohibits changes to the facility as described in the safety analysis report, and to procedures described in the safety analysis report (without prior NRC approval) ``if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased''. Since the conditions which limit changes performed under 50.59 are more restrictive than the conditions which define changes considered to involve a significant hazards consideration, moving of a requirement from the TS to a document which is controlled under 50.59 cannot involve a significant increase in the probability or consequences of an accident previously evaluated.

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

1. Administrative Change (Change A):

   ``Administrative'' changes make wording changes which clarify existing TS requirements, without affecting their technical content. Since ``Administrative'' changes do not alter the technical content of any requirements, they cannot create the possibility of a new or different kind of accident from any previously evaluated.

2. More Restrictive Change (Change B):

   ``More Restrictive'' changes only add new requirements, or revise existing requirements to result in additional operational restrictions. The TS, with all ``More Restrictive'' changes incorporated, will still contain all of the requirements which existed prior to the changes. Therefore, ``More Restrictive'' changes cannot create the possibility of a new or different kind of accident from any previously evaluated.

3. Less Restrictive Change (Change E):

   Change E deletes the TS requirements for snubber testing, but adds identical requirements to a document (the ORM) controlled under 10 CFR 50.59.

   10 CFR 50.59 specifically prohibits changes to the facility as described in the safety analysis report, and to procedures described in the safety analysis report (without prior NRC approval) ``if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created''. Since the conditions which limit changes performed under 50.59 are more restrictive than the conditions which define changes considered to involve a significant hazards consideration, relocation of a requirement from the TS to a document which is controlled under 50.59 cannot create the possibility of a new or different kind of accident from any previously evaluated.

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

1. Administrative Change (Changes A):

   ``Administrative'' changes make wording changes which clarify existing TS requirements, without affecting their technical content. Since ``Administrative'' changes do not alter the technical content of any requirements, they cannot involve a significant reduction in a margin of safety.

2. More Restrictive Change (Change B):

   ``More Restrictive'' changes only add new requirements, or revise existing requirements to result in additional operational restrictions. The TS, with all ``More Restrictive'' changes incorporated, will still contain all of the requirements which existed prior to the changes. Therefore, ``More Restrictive'' changes cannot involve a significant reduction in a margin of safety.

3. Less Restrictive Change (Change E):

   Change E deletes the TS requirements for snubber testing, but adds identical requirements to a document (the ORM) controlled under 10 CFR 50.59.

   10 CFR 50.59 specifically prohibits changes to the facility as described in the safety analysis report, and to procedures described in the safety analysis report (without prior NRC approval) ``if the margin of safety as defined in the basis for any technical specification is reduced''. Since the conditions which limit changes performed under 50.59 are more restrictive than the conditions which define changes considered to involve a significant hazards consideration, relocation of a requirement from the TS to a document which is controlled under 50.59 cannot 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: Van Wylen Library, Hope College, Holland, Michigan 49423.

   Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

   NRC Project Director: Cynthia A. Carpenter.

   Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power Plant, Unit 1, Monroe County, Michigan

   Date of amendment request: December 15, 1997 (Reference NRC-98- 0023).

   Description of amendment request: The proposed amendment will add a subpart 3 to Part 2.B of the Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1), that would allow the licensee to receive, acquire, possess, use and transfer byproduct material without restriction to chemical or physical form for sample analysis, instrument calibration, or associated with radioactive apparatus, hardware, tools, and equipment, provided the cumulative radioactive material quantity of the byproduct material does not exceed the criteria contained in Section 30.72, Schedule C, ``Quantities of Radioactive Material Requiring Consideration of the Need for an Emergency Plan for Responding to a Release.''

   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 using the standards in 10 CFR 50.92(c). The licensee's analysis is presented below:

   (1) Does the proposed change significantly increase 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. Using slightly contaminated apparatus or a small non-exempt radioactive source cannot affect the probability of the analyzed sodium or liquid waste accidents. The ability to possess such equipment does not in itself change any methods of handling liquid waste or sodium. Use of

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   contaminated equipment could potentially increase the consequences of an accident if it was in use or in the vicinity if an accident occurs. However, the increase in consequences would not be significant due to the limitations on radioactivity content of such equipment. The limit was selected to be that in 10 CFR Part 30.72, Schedule C, as the threshold beyond which offsite emergency plans are required. Since the quantity is below that requiring an offsite emergency plan, even if all the byproduct material allowed to be possessed by the proposed amendment were released during a postulated accident, the consequences would be significantly increased. The quantity contained in any specific piece of contaminated apparatus or a source would be expected to be even less. Therefore, this amendment does not involve a significant increase in the probability or consequences of an accident.

   (2) Will the proposed amendment create the possibility of a new or different kind of accident from any accident previously analyzed?

   The proposed amendment does not create the possibility of a new or different type of accident from any previously evaluated. Allowing possession of contaminated apparatus, tools, or equipment does not change methods of monitoring the facility or operation or surveillance of any system at Fermi 1. While possession of a different source will permit other instruments to be calibrated, source checked, or tested at Fermi 1, testing of instrumentation is routine, ordinary activity. It is not an activity which creates the possibility of a new or different type of accident.

   (3) Will the proposed change significantly reduce the margin of safety at the facility?

   The proposed amendment does not involve a significant reduction in the margin of safety at Fermi 1. No change to any system or the status of any systems or structures, are created by this amendment. Being able to have limited amounts of additional radioactive material at Fermi 1 in the form of contaminated apparatus, tools, equipment or hardware or non-exempt radioactive sources will not significantly reduce the margin of safety because a 10 CFR Part 20 program is already in place and the amount of radioactive material is being limited below the amount in 10 CFR Part 30.72, Schedule C. For these reasons, this amendment will not significantly reduce the margin of safety at Fermi 1.

   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, NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

   Local Public Document Room location: Monroe County Library System, 3700 South Custer Road, Monroe, Michigan 48161.

   Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 2000 Second Avenue, Detroit, Michigan 48226.

   NRC Branch Chief: John W. N. Hickey.

   Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No 2, St. Lucie County, Florida

   Date of amendment request: March 3, 1998.

   Description of amendment request: The amendment request proposes to revise the applicability of the St. Lucie Unit 2 technical specifications (TSs) to be consistent with St. Lucie Unit 1 TSs for reactor coolant system (RCS) chemistry. In addition, the amendment request proposes to modify the St. Lucie Unit 2 TSs by making administrative changes to the TS discussion of the criticality design features for fuel storage, and administrative changes to the technical review responsibilities under the cognizance of the Company Nuclear Review Board.

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

   (1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

   The proposed change to TS 3.4,7 will replace the existing applicability statement of ``At all times'' with ``All MODES.'' This revision will obviate the burden and personnel radiation exposures associated with sampling the RCS for chloride and fluoride concentrations during low temperature, defueled conditions. The existing limits, corrective actions for above limit conditions, and sampling requirements will be applicable for all operational MODES defined in the TS. The proposed applicability will continue to assure consistency with the bases for the RCS chemistry specification, and the potential for occurrence, initial conditions, or consequences of events considered in the safety analyses are not changed. The revisions proposed for TS 5.6.1.a.1 and 6.5.2.9.d are administrative in nature, and assure consistency with the bases for previously approved license amendments. Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

   (2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

   The proposed amendment will not change the physical plant or the operational MODES defined in the facility license. The changes do not involve the addition of new equipment or the modification of existing equipment, nor do they alter the design of St. Lucie plant systems. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

   (3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

   The proposed revision to TS 3.4.7 will not change the existing RCS chemistry requirements that are applicable to the operational MODES defined in the technical specifications. However, the change will allow the chloride and fluoride concentrations to go unmonitored during certain refueling operations when there is no fuel in the reactor vessel. For the limited time intervals associated with this defueled condition, the RCS is depressurized, coolant temperature is near ambient, it is unlikely that the chloride and fluoride concentrations could be significantly increased above the concentrations that existed during MODE 6 prior to the core off-load, and susceptibility to corrosive attack from these halides is, therefore, significantly reduced. The existing bases for the RCS chemistry limiting conditions for operation are not changed, and both the bases and the proposed specification are consistent with the corresponding TS at St. Lucie Unit 1. The proposed revisions to TS 5.6.1.a.1 and TS 6.5.2.9.d are administrative in nature and ensure that descriptions contained therein are consistent with the bases for previously approved license amendments. Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

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

   Local Public Document Room location: Indian River Community College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

   Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

   NRC Project Director: Frederick J. Hebdon.

   Florida Power and Light Company, Dockets Nos. 50-250 and 50-251, Turkey Point Plant Units 3 and 4, Dade County, Florida

   Date of amendment request: March 12, 1998

   Description of amendment request: The licensee proposed to amend Turkey Point Unit 3 Facility Operating License DPR-31 to delete license conditions 3.I, ``Steam Generator Repair Program,'' 3.K, ``Integrated Schedule,'' and Section 4 of the Operating License Conditions and renumber Section 5 to Section 4; and to amend Turkey Point Unit 4 Facility Operating License DRP-41 to delete license conditions 3.H, ``Steam Generator Repair Program,'' and 3.K, ``Integrated Schedule''. In addition, the proposed amendments would modify Appendix A of Facility Operating Licenses DPR-31 and DPR-41 of the Turkey Point Units 3 and 4 Technical Specifications (TS) to delete outdated references from TS Figure 5.1-2, ``Plant Area Map'' and to incorporate a recent organization change in TS 6.5.1.2, and 6.5.3.1.a.

   The proposed changes are administrative in nature because they would remove fulfilled license conditions and outdated TS references, and incorporate an organizational change.

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

   (1) Operation of the facility in accordance with the proposed 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 removing fulfilled license conditions, outdated Technical Specification referenced material, and reflecting an organizational change. These amendments 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. Therefore, the proposed changes do not affect the probability or consequences of accidents previously analyzed.

   (2) Operation of the facility 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 use of the modified specifications cannot create the possibility of a new or different kind of accident from any previously evaluated since the proposed amendments will not change the physical plant or the modes of plant operation defined in the facility operating license. No new failure mode is introduced due to the administrative changes since the proposed changes do not involve the addition or modification of equipment nor do they alter the design or operation of affected plant systems, structures, or components.

   (3) Operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

   The operating limits and functional capabilities of the affected systems, structures, and components are unchanged by the proposed amendments. The organizational change from Services Manager to Protection Services Manager maintained the associated level of management controls and the required qualifications. The proposed changes to the Facility Operating License Conditions and to the Technical Specifications are administrative and do not significantly reduce any of the margins 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: Library, Florida International University, University Park Campus, 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: Frederick J. Hebdon.

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

   Date of amendment request: March 2, 1998.

   Description of amendment request: The proposed change would revise Technical Specification (TS) 4.5.2.b.1 to delete the requirement to vent the operating chemical volume and control system (CVCS) centrifugal charging pump casing.

   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 does not affect accident initiators or precursors and does not alter the design assumptions affecting the ability of the ECCS [emergency core cooling system] pumps to mitigate the consequences of an accident.

   The proposed change will align the surveillance requirements with the installed system design and normal operating conditions. The intent of the surveillance requirement ensures operability of the CVCS centrifugal charging pumps by verifying that the ECCS pumps and piping is full of water and not subjected to gas binding or hydraulic transients.

   Excluding the venting of the operating CVCS centrifugal charging pump will not effect pump operation nor subject the high head safety injection portion of the ECCS to potential hydraulic transients. Venting the operating pump under a dynamic condition at high system pressure is ineffective.

   The design and installation of the CVCS centrifugal charging pumps is such that significant non-condensable gasses do not collect in the pumps, whether they are running or not. Therefore, it is unnecessary to require periodic pump casing venting to ensure the pumps will remain operable. Venting of the non-operating centrifugal charging pump will continue to be performed, as required by TS 4.5.2b.1.

   Therefore, the proposed change does not involve a significant increase in the

   [[Page 17226]]

   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 previously analyzed.

   The proposed change will not result in new failure modes because no new components or physical changes are involved with this change nor are the components operated in a new or different manner. The proposed change does not alter the ability of the CVCS centrifugal charging pumps to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR). The proposed change has no impact on component or system interactions, or the plant design basis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

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

   There is no impact on equipment design or operation and there are no changes being made to the Technical Specification required safety limits or safety system settings that would adversely affect plant safety. The CVCS centrifugal charging pumps are designed and installed to be self-venting, such that, accumulation, if any, of non-condensable gasses would have no significant impact on pump operation. Since the proposed change will not result in new failure modes, then, the designed margins of safety to minimize/preclude the consequences of a radiological event resulting from a design basis accident remain unchanged. Therefore, the proposed change to eliminate the requirement to vent the operating CVCS centrifugal charging pump casing does not involve a significant reduction in any margin of safety.

   The NRC staff has reviewed the licensee's analysis, and based on this review, it appears that the three standards of 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: Exeter Public Library, Founders Park, Exeter, NH 03833.

   Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270.

   NRC Project Director: Cecil O. Thomas.

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

   Date of amendment request: January 30, 1998.

   Description of amendment request: The proposed amendment would revise the technical specifications (TS) by relocating pressure- temperature (P-T) curves, predicted radiation induced NDTT shift curves, and the low temperature overpressure protection (LTOP) limits and values from the TS to an OPPD controlled document.

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

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

   The proposed changes relocate the reactor coolant system (RCS) pressure-temperature (P-T) curves, the predicted radiation induced NDTT shift curve and the low temperature overpressure protection (LTOP) limits to the Fort Calhoun Station Unit No. 1 RCS Pressure-Temperature Limits Report (PTLR).

   Compliance with these curves and limits continues to be required by the Technical Specifications. Changes to the curves and limits will be controlled by TS 5.9.6, and must be in accordance with the NRC and ASME approved methodologies listed there and with 10 CFR 50.59.

   The FCS PTLR in combination with the limitations imposed by the TS, will ensure the integrity of the reactor vessel pressure boundary. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   There will be no physical alterations to the plant configuration (no new or different equipment is being installed). No changes in operating modes or limits are proposed. The TS retain requirements to maintain the RCS within acceptable operational limits established in accordance with NRC and ASME approved methodologies and assure operability of the LTOP system. As such, the TS will continue to require compliance with the limitations being relocated to the FCS PTLR. Therefore, these proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

   This proposed change to the FCS TS is administrative in nature relocating the P-T curves, NDTT curve, LTOP limits and associated TS requirements to the FCS PTLR in accordance with GL 96-03. Future updates of the FCS PTLR will be conducted under the 10 CFR 50.59 process utilizing NRC and ASME approved methodologies (as described in FCS Unit No. 1 PTLR, Rev. 0 and CEOG Task 942, Report CE NPSD-683, Rev. 02). Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

   Local Public Document Room location: W. Dale Clark Library, 215 South 15th Street, Omaha, Nebraska 68102.

   Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

   NRC Project Director: William H. Bateman.

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

   Date of amendment request: January 30, 1998.

   Description of amendment request: The proposed amendment would revise Facility Operating License No. DPR-40 to delete the License Term based on a reevaluation of the end of license fluence.

   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 previously evaluated accidents affected by this change are limited to the pressurized thermal shock (PTS) events. Vessel embrittlement due to fast neutron associated damage to the limiting beltline region reactor vessel material, which for Fort Calhoun Station is the lower course axial welds, is a

   [[Page 17227]]

   component in the PTS analysis. The fast neutron, thermal neutron and dpa values of the FCS reactor vessel were recalculated using actual power history values for Cycles 1 through 14 rather than conservative estimates, with the revised BUGLE-93 cross sections from the ENDF/B-VI cross section library to appropriately account for the iron atoms in the thermal shield and a methodology that the NRC has previously approved for neutron fluence calculations performed by Westinghouse. The evaluation included data from the three surveillance capsules (W- 225, W-265, and W-275) previously removed and analyzed. The evaluation results indicate that the FCS reactor vessel is able to reach current licensed life without exceeding the 10 CFR 50.61 screening criteria for RTPTS of 270 deg.F for limiting axial welds.

   In accordance with 10 CFR 50.61, this assessment must be updated whenever there is a significant change in projected values of RT‹INF›PTS‹/INF› or upon request for a change in the expiration date of the facility. Since these requirements are contained in 10 CFR 50.61, Section 3.E can be deleted from Operating License No. DPR-40 without resulting in a significant increase in the probability or consequences of any accident previously evaluated.

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

   The proposed change does not physically alter the configuration of the plant and no new or different mode of operation is proposed. Increasing the long term load factor from 0.77 to 0.85 more accurately projects RT‹INF›PTS‹/INF› by accounting for improvement in FCS operating cycle efficiency. Requirements for assessing and reporting RT‹INF›PTS‹/INF› are contained in 10 CFR 50.61 and therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

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

   The margin of safety is defined by the draft regulatory guide DG- 1053 for neutron fluence calculations which requires the methodology to be capable of providing best estimate fluence evaluations within plus or minus 20 percent (1‹greek-s›). The analysis shows that the applicable regulatory criteria are met and therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

   Local Public Document Room location: W. Dale Clark Library, 215 South 15th Street, Omaha, Nebraska 68102.

   Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

   NRC Project Director: William H. Bateman.

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

   Date of amendment request: March 18, 1998.

   Description of amendment request: The proposed amendment would revise the technical specifications by changing the title of the Shift Supervisor to Shift 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.

   OPPD proposes to change the title of the Shift Supervisor to Shift Manager. The qualifications required of these individuals and the duties they perform are unchanged. The title of Shift Manager better conveys the appropriate level of responsibility and authority required of the position. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   There will be no physical alterations to the plant configuration (no new or different equipment is being installed). No changes in operating modes or limits are proposed. The qualifications required of these individuals and the duties they perform are unchanged. Therefore, these proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

   The proposed change in the title of the Shift Supervisor to Shift Manager is strictly administrative. The qualifications required of these individuals and the duties they perform are unchanged. The title of Shift Manager better conveys the appropriate level of responsibility and authority required of the position. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   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: W. Dale Clark Library, 215 South 15th Street, Omaha, Nebraska 68102.

   Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

   NRC Project Director: William H. Bateman.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne Count, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test intervals for: (1) the standby liquid control (SLC) system that ensures that there is a functioning flow path from the boron injection tank to the reactor pressure vessel, and (2) the scram discharge volume (SDV) that verifies system performance of the vent and drain valves. Specifically, the interval for SLC testing is being increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period; and, from once every 36 months to once every 48 months for those surveillances on a staggered test basis. The frequency for testing the SDV vent and drain valves would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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

   [[Page 17228]]

   consequences of an accident previously evaluated?

   The proposed changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the SAR. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal. Since the impact on the systems is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test intervals for performance of channel calibrations on: (1) the reactor protection system (RPS) instrumentation, (2) the source range monitor (SRM) instrumentation, (3) the feedwater-main turbine high-water-level trip instrumentation, (4) the post accident monitoring (PAM) instrumentation, (5) the remote shutdown system instrumentation, (6) the end-of-cycle recirculation pump trip (EOC-RPT) instrumentation, (7) the anticipated transient without scram recirculation pump trip (ATWS-RPT) instrumentation, (8) the emergency core cooling system (ECCS) instrumentation, (9) the rector core isolation cooling (RCIC) system instrumentation, (10) the primary containment isolation instrumentation, (11) secondary containment isolation instrumentation, (12) the control room emergency outside air supply (CREOAS) system instrumentation, (13) the loss of power (LOP) instrumentation, and (14) the RPS electric power monitoring instrumentation. Specifically, the intervals for the associated channel calibration would be increased from either once every 18 months or refueling cycle to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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 changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the SAR. Furthermore, the instrument drift has been evaluated and found to be acceptable for the extended operating cycle[.] Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal and instrument drift over the extended operating cycle has been evaluated and found to be acceptable. Since the impact on the systems and from instrument drift is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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,

   [[Page 17229]]

   Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

   Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test intervals for: (1) the integrated leak test of each system listed as a primary coolant source outside containment, and (2) the engineered safety feature filter ventilation systems in the ventillation filter testing program. Specifically, the interval for these tests would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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 changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the SAR [safety analysis report]. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal. Because the impact on the systems is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test intervals for the AC and DC electrical power system sources. Specifically, the intervals for various functional tests would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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 changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the [safety analysis report] SAR. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal. Because the impact on the systems is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three

   [[Page 17230]]

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would lower the minimum allowable low power setpoint for the control rod block instrumentation rod worth minimzer (RWM) from less than or equal to 20 percent rated thermal power (RTP) to less than or equal to 10 percent RTP.

   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?

   This change establishes the minimum allowable low power setpoint of the RWM as less than or equal to 10% RTP. This change will not result in a significant increase in the probability of an accident previously evaluated because the Operability of the RWM not considered an initiator for any accidents previously analyzed. This change will not result in a significant increase in the consequences of an accident previously evaluated because, as documented in Amendment 17 to NEDE- 24011-P-A (GESTAR-II) and the associated NRC SER [safety evaluation report], if core power level exceeds 10% RTP, no control rod pattern can generate rod worths such that the fuel enthalpy would exceed the 280 cal/gm fuel enthalpy limit during the worst RDA [rod drop accident].

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SSC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The proposed change does not involve a significant reduction in a margin of safety because, as documented in Amendment 17 to NEDE-24011- P-A (GESTAR-II) and the associated NRC SER, if core power level exceeds 10% RTP, no control rod pattern can generate rod worths such that the fuel enthalpy would exceed the 280 cal/gm fuel enthalpy limit during the worst RDA.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test intervals for the: (1) drywell-to-suppression chamber vacuum breaker leakage test, (2) the primary containment isolation valves functional tests, (3) each reactor instrumentation line excess flow check valve (EFCV) functional tests, (4) the suppression chamber- to-drywell vacuum breaker opening setpoint test, (5) the system functional test, visual examination, and heater phase resistance to ground tests for the drywell and suppression chamber hydrogen recombiners, (6) the secondary containment vacuum tests of the standby gas treatment (SGT) subsystem, (7) the seconday containment isolation valves (SCIVs) functional tests, and (8) the SGT subsytem functional tests. Specifically, the intervals for these tests would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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 changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the [Safety Analysis Report] SAR. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal. Since the impact on the systems is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the

   [[Page 17231]]

   subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test intervals for: (1) the system functional test of the core spray and low pressure coolant injection system, and (2) the high pressure coolant injection (HPCI) and the low pressure HPCI flow test. Specifically, the intervals for system functional tests and response time tests would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period. Additionally, the surveillance test intervals for: (1) the system functional test of the automatic depressurization system (ADS), and (2) the system functional test and low pressure flow test of the reactor core isolation cooling (RCIC) system.

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

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

   The proposed changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the SAR. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal. Since the impact on the systems is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test interval for the channel calibration of the reactor coolant system leakage detection instrumentation. The surveillance test interval would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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 changes involve a change in the [S]urveillance Frequency from 18 months to 24 months. The change in [S]urveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the SAR [safety analysis report]. Furthermore, the instrument drift has been evaluated and found to be acceptable for the extended operating cycle. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal and instrument drift over the extended operating cycle has been evaluated and found to be acceptable. Since the impact on the systems and from instrument drift is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant

   [[Page 17232]]

   systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would remove the operability requirement for the 480 volt engineered safeguards systems bus 0565 undervoltage relay (degraded voltage 65 percent and 92 percent) in the loss of power instrumentation.

   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 changes remove from the SSES CTS [Susquehanna Steam Electric Station current technical specifications] items that are informational or implementing details that are adequately and more appropriately controlled by the licensee. Additionally, the proposed changes remove from the SSES CTS items that are contained in the Code of Federal Regulations or other regulatory documents and, therefore, do not need to be repeated in the SSES ITS [improved technical specifications]. These requirements being moved to another controlled document or removed from Technical Specifications are not deleted or changed. Therefore, these changes will not result in any changes to the requirements specified in the SSES CTS, but will reduce the level of regulatory control on the identified requirements. The level of regulatory control has no impact on the probability or the consequences of an accident previously evaluated, therefore, these changes have no impact on 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 proposed changes will not involve any physical changes to plant systems, structures, or components (SSC), or the manner in which these SSC are operated, maintained, modified, tested, or inspected. The proposed changes will not impose or eliminate any requirements. Therefore, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety as defined in the bases of any Technical Specification is not reduced. The requirements being moved to another controlled document or removed from Technical Specifications remain the same as stated in the SSES CTS. Therefore, no reduction in a margin of safety will be permitted.

   Removal of these items from SSES CTS eliminates the requirement for NRC review and approval of revisions in accordance with 10 CFR 50.92. Elimination of this administrative process does not have a margin of safety that can be evaluated. However, the proposed changes are consistent with the BWR [Boiling-Water Reactor] Standard Technical Specification, NUREG-1433, Rev. 1, which was approved by the NRC. Revising the Technical Specifications to reflect the approved level of detail ensures no 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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996.

   Description of amendment request: The change would increase the surveillance test intervals for: (1) the reactor protection system (RPS) instrumentation, (2) the feedwater-main turbine high-water-level trip instrumentation, (3) the end of cycle recirculation pump trip (EOC-RPT) instrumentation, (4) the anticipated transient without scram recirculation pump trip (ATWS-RPT) instrumentation, (5) the emergency core cooling system (ECCS) instrumentation, (6) the reactor core isolation cooling (RCIC) system instrumentation, (7) RPS electric power monitoring system instrumentation, (8) primary containment isolation instrumentation, (9) secondary containment isolation instrumentation, (10) the control room emergency outside air supply (CREOAS) system instrumentation, and (11) the loss of power (LOP) instrumentation. Specifically, the intervals for various logic system functional tests and response time tests would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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?

   [[Page 17233]]

   The proposed changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the SAR. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal. Since the impact on the systems is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996, and March 2, 1998.

   Description of amendment request: The change would increase the surveillance test interval for the: (1) emergency service water (ESW) system functional test, (2) the control room emergency outside air supply (CREOAS) system functional test and control room pressurization test, and (3) the main turbine bypass system functional and response time tests. Specifically, the interval for these tests would be increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25 percent grace period.

   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 changes involve a change in the surveillance Frequency from 18 months to 24 months. The change in surveillance Frequency is not assumed to be an accident initiator for any accidents previously evaluated in the [safety analysis report] SAR. Therefore, this change will have no impact on the probability of an accident previously evaluated. By changing the Surveillance Frequency from 18 months plus grace to a maximum of 30 months, the consequences of an accident previously evaluated in the SAR are not significantly increased. This is based on the fact that the evaluation of the subject changes demonstrated that the overall impact, if any, on the systems availability is minimal. Since the impact on the systems is minimal, it can be concluded that the overall impact on the plant accident analysis is negligible. Furthermore, it is shown that the performance history for the subject systems does not indicate any failures which would invalidate the conclusions reached in this evaluation.

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

   This proposed change will not involve any physical changes to plant systems, structures, or components (SCC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety has not been significantly reduced. Although, there will be an increase in the interval between the subject surveillance tests, the evaluation of the changes demonstrates that there is no evidence of any failures which would impact the subject systems availability. Based on the fact that the increased testing interval has a minimal impact on the subject systems, it can be concluded that the assumptions in the licensing basis are not impacted by the changes in the subject requirements and commitments.

   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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996, and March 2, 1998.

   Description of amendment request: The change would add a surveillance requirement and acceptance criteria to verify the source range monitor (SRM) count rate versus the signal to noise ratio of the SRMs. This change also incorporates a new SRM count rate to signal to noise ratio curve which is based on General Electric Service Information Letter (SIL) 478.

   Basis for proposed no significant hazards consideration determination:

   [[Page 17234]]

   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 changes provide requirements determined to be more conservative than the existing requirements for operation of the facility.

   Therefore, these changes establish or maintain adequate assurance that components are operable when necessary for the prevention or mitigation of accidents or transients and that plant variables are maintained within limits necessary to satisfy the assumptions for initial conditions in the safety analysis. Therefore, these changes do not involve any 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 proposed changes will not involve any physical changes to plant systems, structures, or components (SSC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The imposition of more restrictive requirements either has no impact on or increases the margin of plant safety. As provided in the discussion of each of the changes, each change in this category provides additional requirements designed to enhance plant safety. Each of the changes maintains requirements within the safety analyses and licensing basis. Therefore, these changes do not involve 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 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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

   Date of amendment request: August 1, 1996, as supplemented March 2, 1998.

   Description of amendment request: The change would reduce the allowable values for the reactor protection system instrumentation scram discharge volume water level--high scram setpoints: (1) for the level transmitter from less than or equal to 88 gallons to less than or equal to 66 gallons, and (2) for the float switch from less than or equal to 88 gallons to less than or equal to 62 gallons in order to be consistent with the design setpoint calculations.

   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 changes provide requirements determined to be more conservative than the existing requirements for operation of the facility. Therefore, these changes establish or maintain adequate assurance that components are operable when necessary for the prevention or mitigation of accidents or transients and that plant variables are maintained within limits necessary to satisfy the assumptions for initial conditions in the safety analysis. Therefore, these changes do not involve any 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 proposed changes will not involve any physical changes to plant systems, structures, or components (SSC). The changes in normal plant operation are consistent with the current safety analysis assumptions. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The imposition of more restrictive requirements either has no impact on or increases the margin of plant safety. As provided in the discussion of each of the changes, each change in this category provides additional requirements designed to enhance plant safety. Each of the changes maintains requirements within the safety analyses and licensing basis. Therefore, these changes do not involve 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 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: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.

   NRC Project Director: John F. Stolz.

   Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

   Date of amendments request: December 30, 1997.

   Description of amendments request: The proposed amendments would revise the Technical Specification surveillance requirements for the Auxiliary Building and Service Water Building batteries to remove the existing 1.75 volt minimum individual cell voltage associated with the ``service test'' acceptance criterion and replace it with a reference to the battery load profile specified in the Final Safety Analysis Report, Section 8.3.2.

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

1. The proposed changes to remove and replace specific acceptance criterion in the Technical Specifications with a reference to more detailed and bounding criteria in the FSAR [Final Safety Analysis Report] for service tests on the batteries do not involve a significant increase in the probability or consequences of an accident previously evaluated in the Farley FSAR. The AB [Auxiliary Building] and SWB

   [Service Water Building] batteries do not initiate any accident. Clarification of testing acceptance criteria does not adversely affect the batteries ability to mitigate the consequences of any accident in the

   [[Page 17235]]

   Farley FSAR. No new accident initiators are identified as a result of this proposed revision. No new performance requirements for any system that is used to mitigate dose consequences have been imposed by this proposed change. No input assumptions to any dose consequence calculations are affected by this proposed change. All previously reported dose consequences remain bounding. Therefore, the radiological consequences resulting from any accident previously evaluated in the FSAR are not increased.

2. The proposed changes to remove and replace specific acceptance criterion in the Technical Specifications with a reference to more detailed and bounding criteria in the FSAR for service tests on the batteries do not create the possibility of a new or different kind of accident from any previously evaluated in the Farley FSAR. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the clarifications to the battery service test acceptance criteria. No new challenges to the safety-related AB or SWB 125VDC Distribution Systems have been identified. The 125VDC Systems including the batteries have not been modified. Farley will continue to perform service discharge surveillance tests in accordance with the frequency requirements of the Technical Specifications to demonstrate battery operability. Previously identified accident scenarios remain bounding because the performance requirements of the batteries have not been changed. Therefore, the possibility of a new or different kind of accident is not created.

3. The proposed changes to remove and replace specific acceptance criterion in the Technical Specifications with a reference to more detailed and bounding criteria in the FSAR for service tests on the batteries do not involve a significant reduction in the margin of safety. All previously established acceptance limits continue to be met for all events since the battery function is to provide power during the time between LOSP [loss of offsite power] & D/G [diesel generator] start and in the event of battery charger failure to mitigate the consequences of any accident scenario. Relocating and clarifying service test acceptance criteria will not invalidate the battery function. There are no physical modifications required to the AB or SWB 125VDC Distribution Systems or the batteries. This change will not affect the operation of the batteries or any other safety-related equipment. Applicable values, reflected in the governing electrical design calculations, will be incorporated into the FSAR and will remain or be included in the surveillance test procedures. Since current battery performance acceptance limits will continue to be met, 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: Houston-Love Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

   Attorney for licensee: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama.

   NRC Project Director: Herbert N. Berkow.

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

   Date of application for amendments: February 13, 1998 (TS 97-04).

   Brief description of amendments: The amendments change the Sequoyah (SQN) Technical Specifications (TS) by relocating the mechanical snubber requirements from Section 3.7.9 of the TS to the SQN Technical Requirements Manual.

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

   TVA has concluded that operation of SQN Units 1 and 2, in accordance with the proposed change to the TS, does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).

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

      The proposed revision to the TS relocates the requirements for SQN snubbers without changing the current requirements and deletes an obsolete License Condition. TVA does not consider the snubbers to be the source of any accident; therefore, this administrative relocation of the requirements and License Condition deletion will not increase the possibility of an accident. The capability of the snubbers will continue to provide the same function in support of accident mitigation. Changes to the relocated requirements will be processed, in accordance with 10 CFR 50.59, to ensure the snubber functions will be properly maintain[ed]. Therefore, the proposed relocation of the snubber requirements and License Condition deletion will not increase the consequences of an accident.

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

      The SQN safety-related snubbers provide support for mitigation functions associated with previously evaluated accidents and are not the initiator of any accident. The proposed change does not alter the current functions of the snubbers; therefore, it will not create the possibility of a new or different kind of accident.

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

      The requirements for SQN safety-related snubbers are unchanged by the proposed relocation of the requirements to the SQN TRM [Technical Requirements Manual] and the License Condition deletion. The function of the snubbers and surveillances to ensure operability will remain the same as currently required by the TS. Changes to these requirements will be evaluated, in accordance with 10 CFR 50.59, to ensure acceptability and NRC review as required. Therefore, the proposed change will not result in a reduction in a margin of safety.

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

      Local Public Document Room location: Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

      NRC Project Director: Frederick J. Hebdon.

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

      Date of application for amendments: February 25, 1998 (TS 97-06).

      Brief description of amendments: The amendments change the Sequoyah

      [[Page 17236]]

      (SQN) Technical Specifications (TSs) for the emergency diesel generators (D/Gs) by 1) incorporating vendor-recommended changes to the D/G inspection program, 2) revising the D/G surveillance program, and 3) changing the allowable D/G steady-state voltage range.

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

      TVA has concluded that operation of SQN Units 1 and 2, in accordance with the proposed change to the TSs (or operating license[s]), does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).

      Part 1--Vendor Recommended Inspections:

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

      The proposed revision to the TS deletes the requirements for 18- month inspections from the TS. TVA does not consider the inspections to be the source of any accident; therefore, this deletion will not increase the possibility of an accident. The D/Gs come within the purview of 10 CFR 50.65, which monitors the effectiveness of maintenance at nuclear power plants. The capability of the D/Gs to provide the required safety function in support of accident mitigation will be unaffected. Therefore, the proposed deletion of the inspection requirements will not increase the consequences of an accident.

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

      The emergency D/Gs provide support for mitigation functions associated with previously evaluated accidents and are not the initiator of any accident. The proposed change does not alter the current functions of the D/Gs; therefore, it will not create the possibility of a new or different kind of accident.

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

      The requirements for emergency D/Gs are unchanged by the proposed deletion of the requirements from TSs. The function of the emergency D/ Gs and surveillances to ensure operability will remain the same as currently required by the TS. NRC will continue to monitor the effectiveness of D/G maintenance as required by 10 CFR 50.65. Therefore, the proposed change will not result in a reduction in a margin of safety.

      Part 2--D/G Online Testing:

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

      The proposed amendment to allow the load rejection tests and the 24-hour D/G endurance run to be conducted during any mode of operation does not significantly increase the probability or consequences of an accident previously evaluated in Chapter 15 of the Final Safety Analysis Report (FSAR) since the capability to safely shutdown the plant following a LOOP [loss of offsite power], LOCA [loss of coolant accident] or LOCA/LOOP coincident with a single failure is maintained throughout the surveillance test. Other aspects of D/G parallel testing (protective devices, risks interactions with offsite power capabilities, and operation) are unaffected by the proposed TS change. Required Class-lE onsite power operability during normal operation, shutdown cooling, LOOP, and accident conditions will be the same.

      Performance of the new SR [Surveillance Requirement] 4.8.1.1.2.g.4 requires the D/Gs to be at the same system conditions prior to the test (stabilized operating temperature) as previously required. The LOOP start will continue to be performed as required by SR 4.8.1.1.2.d.4.b.

      In addition, the performance of proposed SRs 4.8.1.1.2.g.1, 4.8.1.1.2.g.2, 4.8.1.1.2.g.3, or 4.8.1.1.2.g.4 during Modes 1, 2 or 3 will not significantly increase the consequences of perturbations to any of the electrical distribution systems that could result in a challenge to steady state operation or to plant safety systems.

      Performance of proposed SR 4.8.1.1.2.g.1, 4.8.1.1.2.g.2, or 4.8.1.1.2.g.3 during Modes 1, 2 or 3 or failure of the surveillance, will not cause, or result in, an anticipated operational occurrence with attendant challenges to plant safety systems that has not been previously analyzed for the existing monthly surveillances.

      Therefore, TVA concludes that the above change does not involve a significant increase in the probability or consequences of an 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 requested changes do not result in a new or different kind of accident from that previously analyzed in SQN's FSAR. The changes propose to eliminate restrictions of the plant operating modes in which standby D/G system testing may be performed, but does not change the type of testing performed and are not due to modification of the system design. NRC's assessment of the testing of the D/Gs in the configuration proposed is documented in Section 8.3.1, Supplement 1 of the SER (NUREG-0011).

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

      As previously stated, performance of proposed SRs 4.8.1.1.2.g.1, 4.8.1.1.2.g.2, 4.8.1.1.2.g.3, or 4.8.1.1.2.g.4 during Modes 1, 2 or 3 will not cause, or result in, an anticipated operational occurrence with attendant challenges to plant safety systems that has not been previously analyzed for the existing monthly surveillances. It also does not change any setpoints or limits established for accident mitigation. Therefore, implementation of the proposed amendment will not reduce the margin of safety for this system.

      Part 3--D/G Steady State Allowable Voltage Range:

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

      The proposed revisions to the SRs conservatively restrict the allowable range of the D/G steady state voltage. The capability of the D/Gs to provide the required safety function, in support of accident mitigation, will be unaffected or enhanced. Therefore, the proposed revision of the SRs will not increase the consequences of an accident.

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

      The proposed changes do not alter the current functions of the D/ Gs; therefore, they will not create the possibility of a new or different kind of accident.

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

      The requirements for emergency D/Gs are unchanged by the conservative revision of the allowable range of the D/G steady state voltage or clarification of the required voltage and frequency after 10 seconds. The function of the emergency D/Gs and surveillances to ensure operability will remain the same as currently required by the TS. Therefore, the proposed changes will not result in a reduction in a margin of safety.

      [[Page 17237]]

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

      Local Public Document Room location: Chattanooga-Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

      NRC Project Director: Frederick J. Hebdon.

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

      Date of amendment request: February 25, 1998.

      Description of amendment request: Requests Technical Specifications changes to permit use of Option B of 10 CFR 50, Appendix J, for containment leakage testing.

      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 KNPP in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   The proposed TS changes do not involve any physical or operational changes to structures, systems or components. The current safety analysis and design basis for the accident mitigation functions of the containment, the airlocks, and the containment isolation valves are maintained. On-site and off-site dose consequences remain unaffected. Containment leakage rate testing is not an accident initiator.

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

   The accidents considered are found in the Safety Analysis, Section 14 of the USAR. The proposed change does not involve a change to the plant design (structures, systems or components) or operation. No new failure mechanisms beyond those already considered in the current plant Safety Analysis are introduced. No new accident is introduced and no safety-related equipment or safety functions are altered. The proposed change does not affect any of the parameters or conditions that contribute to initiation of any accidents.

3. The proposed license amendment does not involve a significant reduction in the margin of safety.

   The implementation of Option B potentially affects the frequency of Type A, B, and C containment testing. Except for the determination of test frequency, the methods for performing the actual tests are not changed. NUREG-1493, ``Performance-Based Containment Leak-Test Program'', dated September, 1995, which forms the basis for the Appendix J revision, concludes that adoption of performance-based testing will not significantly reduce the margin of safety. Therefore, the proposed TS amendment will not involve a significant reduction in a margin of safety and will continue to support the design and licensing basis of ensuring an essentially leak-tight containment boundary.

   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: 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: Richard P. Savio.

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

   Date of amendment request: March 4, 1998.

   Description of amendment request: Requests Technical Specifications changes to provide a one hour Limiting Condition for Operation (LCO) that will permit a safety injection pump to be used for addition of make-up fluid to safety injection accumulators during power operation.

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

1. Operation of the KNPP in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   While filling a safety injection (SI) accumulator, the large break loss of coolant accident (LOCA) would be the bounding accident for pump runout concerns. The proposed LCO would allow relaxation of a single failure being assumed during the short duration of the accumulator fill. The SI pump filling the SI accumulator will be considered to be operable while filling the accumulator.

   Using current KNPP PRA methods, this configuration results in a core damage frequency (CDF) of 5x10‹SUP›-5‹/SUP›/year during the five minutes it exists. The increased core damage probability (CDP) due to an accumulator fill is 8x10‹SUP›-11‹/SUP›. Conservatively assuming that the accumulator fill occurs every three weeks, the total CDP increase is 1.3x10‹SUP›-9‹/SUP› in a year. The configuration specific DF and CDP increase are well below the limits of 1.0x10‹SUP›-3‹/SUP›/year and 1.0x10‹SUP›-6‹/SUP›, respectively, in the Electric Power Research Institute's PRA Applications Guide. The increase in probability is extremely low and well within industry PRA limits.

   With entry into a one hour action statement, the single failure criterion is relaxed (i.e., a postulated failure of an SI pump is not required) and both SI pumps will provide the required flow to ensure accident mitigation and prevent pump run out. By assuming both SI pumps are available, there is no impact on the accident analysis.

   By remaining within the bounds of the accident analysis and the extremely low increase in the probability of a LOCA concurrent with an accumulator fill, WPSC concludes that this change does not significantly increase the probability or consequences of an accident previously evaluated.

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

   The change allows relaxation of single failure criteria during the short time an SI accumulator would be filled. The SI pump filling the accumulator will be available during the short filling period.

   With entry into a one hour action statement, the single failure criterion is relaxed (i.e., a postulated failure of an SI pump is not required) and both SI pumps will provide the required flow to ensure accident mitigation and prevent pump runout.

   The proposed change is not a result of a hardware change, and with one SI pump considered to be available during an accumulator fill, all the accident analysis requirements are satisfied. Therefore, WPSC concludes that this

   [[Page 17238]]

   proposed change does not create the possibility of a new or different kind of accident.

3. The proposed license amendment does not involve a significant reduction in the margin of safety.

   With both SI pumps available during an accumulator fill, there is not an SI pump runout concern and all the requirements of the accident analysis are met. Due to the infrequent occurrence, short duration and extremely low probability of LOCA occurring during an accumulator fill, WPSC concludes there is not 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: 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: Richard P. Savio.

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

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

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

   Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power Station, Unit 2, Grundy County, Illinois

   Date of amendment request: March 19, 1998.

   Description of amendment request: The proposed amendment would reflect a change in the Dresden, Unit 2, minimum critical power ratio (MCPR) Safety Limit and revise footnotes in Technical Specifications (TS) Section 5.3, to allow the use of Siemens Power Corporation (SPC) ATRIUM-9B fuel.

   Date of publication of individual notice in Federal Register: March 26, 1998 (63 FR 14735).

   Expiration date of individual notice: April 27, 1998.

   Local Public Document Room location: Morris Area Public Library District, 604 Liberty Street, Morris, Illinois 60450.

   Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

   Date of amendment request: March 16, 1998.

   Brief description of amendment request: These amendments add a new Limiting Condition for Operation (LCO) 3.0.6 to TS Section 3/4.0, ``APPLICABILITY.'' The new LCO 3.0.6 provides specific guidance for returning equipment to service under administrative control to perform testing required to demonstrate OPERABILITY.

   Date of publication of individual notice in Federal Register: March 24, 1998 (63 FR 14142).

   Expiration date of individual notice: Comment period April 7, 1998, and hearing period April 23, 1998.

   Local Public Document Room location: B. F. Jones Memorial Library, 663 Franklin Avenue, Aliquippa, PA 15001.

   Northern States Power Company, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

   Date of amendment request: March 13, 1998.

   Description of amendment request: The proposed amendment would revise Section 2.1.A of the Technical Specifications (TS) to change the safety limit minimum critical power ratio (SLMCPR) values from 1.08 to 1.10 for two recirculation pump operation, and from 1.09 to 1.11 for single loop operation. The amendment would also revise pages 6 and 249b of the TS to indicate that the revised SLMCPR values are applicable only to operating cycle 19.

   Date of individual notice in the Federal Register: March 20, 1998 (63 FR 13704).

   Expiration date of individual notice: April 20, 1998.

   Local Public Document Room location: Minneapolis Public Library, Technology and Science Department, 300 Nicollet Mall, Minneapolis, Minnesota 55401.

   Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW, Washington, DC 20037.

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

   Date of amendment request: March 12, 1998, TXX-98076.

   Description of amendment request: The proposed amendment would provide a temporary Technical Specification change for SRs 4.8.1.1.2f.4)b) and 4.8.1.1.2f.6)b) to allow the verification of the auto connected shut-down loads through the load sequencer to be performed at power for fuel cycle 6 on Unit 1 and fuel cycle 4 on Unit 2.

   Date of individual notice in the Federal Register: March 27, 1998 (63 FR 14974).

   Expiration date of individual notice: April 13, 1998.

   Local Public Document Room location: University of Texas at Arlington Library, Government Publications/Maps, 702 College, P.O. Box 19497, Arlington, TX 76019.

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

   Date of amendment request: March 20, 1997.

   Description of amendment request: The licensee requested to modify their licensing basis by limiting the time the large (18'') purge and vent valves may be open to containment.

   Date of publication of individual notice in Federal Register: March 27, 1998. (63 FR 14976).

   Expiration date of individual notice: April 27, 1998.

   Local Public Document Room location: Brooks Memorial Library, 224 Main Street, Brattleboro, VT 05301.

   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

   [[Page 17239]]

   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.

   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 application for amendment: December 17, 1997.

   Brief description of amendment: These amendments modify the technical specifications (TS) to remove the reference to Exide batteries with a generic reference to low specific gravity cell batteries.

   Date of issuance: March 16, 1998.

   Effective date: March 16, 1998.

   Amendment No.: Unit 1--116; Unit 2--109; Unit 3--88.

   Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: January 14, 1998 (63 FR 2272).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Phoenix Public Library, 1221 N. Central Avenue, Phoenix, Arizona 85004.

   Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

   Date of application for amendments: October 22, 1997.

   Brief description of amendments: The amendments change the Technical Specifications (TSs) to incorporate both steady state and transient degraded voltage setpoints as opposed to the current single degraded voltage setpoints. Additionally, the TS decreases the 4 kV voltage range of the emergency diesel generators to assure that the new steady state degraded voltage relays are not actuated during testing.

   Date of issuance: March 17, 1998.

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

   Amendment Nos.: 226 and 200.

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

   Date of initial notice in Federal Register: November 19, 1997 (62 FR 61838).

   The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated March 17, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Calvert County Library, Prince Frederick, Maryland 20678.

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

   Date of application for amendments: November 6, 1997, as supplemented by letters dated January 27, March 3, March 6, March 13, and March 18, 1998.

   Brief Description of amendments: The amendments change the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP) Units 1 and 2 to allow three 18-month diesel generator (DG) surveillance requirements (SR) to be performed during both plant operation (Operational Conditions 1 and 2) and shutdown (Operational Conditions 3, 4, and 5) rather than, as currently required, only during shutdown. The first SR is an inspection of the DG involving a partial disassembly. The second ensures that non-critical DG protective functions are bypassed on an Emergency Core Cooling system actuation signal. The third verifies that the DG operates for greater than or equal to 60 minutes while loaded to at least 3500 kw, which bounds the maximum expected post-accident DG loading. The proposed amendments additionally remove an expired footnote from the BSEP Unit 2 DG TS.

   Date of issuance: March 26, 1998.

   Effective date: March 26, 1998

   Amendment Nos.: 192 and 223.

   Facility Operating License Nos. DPR-71 and DPR-62: Amendments authorize changes to the facility's Technical Specifications.

   Date of initial notice in Federal Register: December 3, 1997 (62 FR 63971). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 26, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: University of North Carolina at Wilmington, William Madison Randall Library, 601 S. College Road, Wilmington, North Carolina 28403-3297.

   Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

   Date of application for amendment: April 23, 1997.

   Brief description of amendment: This amendment changes the Technical Specifications Surveillance Requirements for TS 4.3.2.1.1.a, 4.3.2.1.4.b, 4.3.2.1.10.a, 4.3.2.1.10.b, and 4.7.3.b.3. to provide more specific information about the tests performed and the components tested.

   Date of issuance: March 18, 1998.

   Effective date: March 18, 1998.

   Amendment No.: 76.

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

   Date of initial notice in Federal Register: June 18, 1997 (62 FR 33119).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 18, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Cameron Village Regional Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

   Date of application for amendment: September 29, 1997 (NRC-97- 0089), as supplemented on March 10, 1998 (NRC-98-0036).

   Brief description of amendment: The amendment revises the technical specifications by relocating the requirements for selected

   [[Page 17240]]

   instrumentation and the associated Bases from the technical specifications (TS) to the updated final safety analysis report. The affected instrumentation is seismic monitoring (TS 3.7.2), meteorological monitoring (TS 3.7.3), the traversing in-core probe system (TS 3.7.7), the chlorine detection system (TS 3.7.8), and the loose-parts detection system (TS 3.7.10). The TS index and list of tables are also revised to reflect the relocation of these TS and associated Bases. NRC Generic Letter 95-10, ``Relocation of Selected Technical Specification Requirements Related to Instrumentation,'' dated December 15, 1995, provided information concerning relocation of the requirements for these instruments.

   Date of issuance: March 17, 1998.

   Effective date: March 17, 1998, with full implementation within 90 days.

   Amendment No.: 115.

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

   Date of initial notice in Federal Register: October 22, 1997 (62 FR 54870). The March 10, 1998, supplement requested a change in the implementation period and was not outside the scope of the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 17, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Monroe County Library System, 3700 South Custer Road, Monroe, Michigan 48161.

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

   Date of amendment request: May 24, 1997.

   Brief description of amendment: The amendment modifies Technical Specification (TS) 3/4.7.4, Ultimate Heat Sink, Table 3.7-3, by incorporating more restrictive dry cooling tower fan requirements, and changes the wet cooling tower water consumption in the TS Bases.

   This amendment modifies the TS to be consistent with revised design-basis calculations.

   Date of issuance: March 23, 1998.

   Effective date: March 23, 1998, to be implemented within 60 days.

   Amendment No.: 139.

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

   Date of initial notice in Federal Register: June 18, 1997 (62 FR 33123).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 23, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: University of New Orleans Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

   Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point Nuclear Station Unit No. 2, Oswego County, New York

   Date application for amendment: July 31, 1997.

   Brief description of amendment: This amendment changes Action Statement 36 to TS Table 3.3.3-1, ``Emergency Core Cooling System Actuation Instrumentation,'' to include actions to be taken if more than one channel per trip function should be inoperable in the high- pressure core spray drywell pressure and reactor water level instrumentation.

   Date of issuance: March 16, 1998.

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

   Amendment No.: 79.

   Facility Operating License No. DPR-63: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 27, 1997 (62 FR 45460).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 1998.

   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, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

   Date of amendment request: October 16, 1996.

   Description of amendment request: The amendment revises the Technical Specifications (TSs) relating to the requirements for AC power sources. The amendment changes certain requirements stated in TS 3/4.8.1, ``AC Sources.'' The requirements are related to the emergency diesel generators.

   Date of issuance: March 17, 1998.

   Effective date: As of the date of issuance, with full implementation within 60 days.

   Amendment No.: 54.

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

   Date of initial notice in Federal Register: December 18, 1996 (61 FR 66711).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 17, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Exeter Public Library, Founders Park, Exeter, NH 03833.

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

   Date of amendment request: February 12, 1997.

   Description of amendment request: The amendment modifies Technical Specification (TS) Section 6.0 ``Administrative Controls,'' to reflect recent organizational changes and changes to the approval title for the Station Qualified Reviewer Program and corrects an incorrect reference in TS 6.4.3.9.b.

   Date of issuance: March 26, 1998.

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

   Amendment No.: 55.

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

   Date of initial notice in Federal Register: May 21, 1997 (62 FR 27797).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 26, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Exeter Public Library, Founders Park, Exeter, NH 03833.

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

   Date of amendment request: July 25, 1997, as supplemented by letters dated November 21, 1997, and March 3, 1998.

   Brief description of amendment: The amendment revises Technical Specifications (TS) 3.5(2), 3.5(3) through 3.5(7), 5.19 and associated Basis to implement Option B of 10 CFR 50 Appendix J.

   Date of issuance: March 23, 1998.

   Effective date: March 23, 1998, to be implemented within 30 days from the date of issuance.

   Amendment No.: 185.

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

   Date of initial notice in Federal Register: November 5, 1997 (62 FR 59919).

   [[Page 17241]]

   The November 21, 1997, and March 3, 1998, supplemental letters provided additional clarifying information that did not change the original no significant hazards determination consideration.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 23, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: W. Dale Clark Library, 215 South 15th Street, Omaha, Nebraska 68102

   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: February 26, 1997, as supplemented by letters dated December 23, 1997, January 30, 1998, and February 9, 1998.

   Brief description of amendments: The amendments revised the combined Technical Specifications (TS) for the Diablo Canyon Power Plant (DCPP) Unit Nos. 1 and 2 to change TS 3/4.4.5 and 3.4.6.2, including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the implementation of steam generator (SG) tube voltage based repair criteria for outside diameter stress corrosion cracking (ODSCC) indications at tube-to-tube support plant (TSP) intersections. The allowed primary-to-secondary operational leakage from any one SG would be reduced from 500 gpd to 150 gpd.

   Date of issuance: March 12, 1998.

   Effective date: March 12, 1998, to be implemented within 30 days from the date of issuance.

   Amendment Nos.: Unit 1-124; Unit 2-122.

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

   Date of initial notice in Federal Register: April 4, 1997 (62 FR 17239).

   The December 23, 1997, January 30, 1998, and February 9, 1998, supplemental letters provided additional clarifying information and did not change the staff's initial no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 12, 1998.

   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: March 4, 1997.

   Brief description of amendments: These amendments revise the emergency core cooling system surveillance test acceptance criteria in Technical Specification 3/4.5.2 for the centrifugal charging and safety injection pumps. Specifically, the change would reduce the maximum specified flow rate values for system alignments that affect the suction pressure to the pumps. In the recirculation mode, increased system flow occurs when the charging and safety injection pumps take suction from the discharge of the residual heat removal pumps.

   Date of issuance: March 12, 1998.

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

   Amendment Nos: 208 and 189.

   Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: April 23, 1997 (62 FR 19834).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 12, 1998.

   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 No. 50-311, Salem Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

   Date of application for amendment: October 29, 1997, as supplemented on January 27, 1998.

   Brief description of amendment: The amendment provides a one-time change to Technical Specification 3/4.4.6, ``Steam Generators,'' to require that the next inspection be performed within 24 months from initial criticality for fuel cycle 10, or during the next refueling outage, whichever is first for fuel cycle 10. In addition, the amendment eliminates a description of an alternate steam generator tube sampling plan that was applicable only during the fourth refueling outage.

   Date of issuance: March 19, 1998.

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

   Amendment No: 190.

   Facility Operating License No. DPR-75: This amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: December 17, 1997 (62 FR 66142).

   The January 27, 1998, supplemental letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 19, 1998.

   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: November 4, 1997.

   Brief description of amendments: These amendments revise the containment systems surveillance test acceptance criteria in Technical Specification 3/4.6.2 for the containment spray pumps. Specifically, the change would replace the Salem Unit 2 minimum specified discharge pressure requirement with an acceptance criterion based on pump differential pressure, and add this surveillance as a new requirement on Salem Unit 1.

   Date of issuance: March 24, 1998.

   Effective date: As of the date of issuance, to be implemented within 60 days Amendment Nos.: 209 and 191.

   Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: December 17, 1997 (62 FR 66141).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 24, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Salem Free Public Library, 112 West Broadway, Salem, NJ 08079.

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

   Date of application for amendment: February 9, 1998.

   Brief description of amendment: The amendment revises the Virgil C. Summer Nuclear Station Technical Specifications (TS) to remove

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   emergency diesel generator (1) accelerated testing requirements (TS 3/ 4.8.1, Table 4.8-1), and (2) special reporting requirements (TS Surveillance Requirement 4.8.1.1.3) in accordance with NRC Generic Letter (GL) 94-01, ``Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators.''

   Date of issuance: March 30, 1998.

   Effective date: March 30, 1998.

   Amendment No.: 139.

   Facility Operating License No. NPF-12: Amendment revises the Technical Specifications.

   Date of initial notice in Federal Register: February 25, 1998 (63 FR 9614) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 30, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Fairfield County Library, 300 Washington Street, Winnsboro, SC 29180.

   Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

   Date of application for amendment: April 24, 1997, as supplemented by letters dated June 6, 1997, and June 27, 1997.

   Brief description of amendment: The amendment revises Section 6.0 of the Callaway Plant, Unit 1 Technical Specifications to change the title ``Senior Vice President Nuclear'' to ``Vice President and Chief Nuclear Officer.''

   Date of issuance: March 23, 1998.

   Effective date: March 23, 1998.

   Amendment No.: 122.

   Facility Operating License No. NPF-30: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: July 30, 1997 (62 FR 40859).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 23, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: University of Missouri- Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

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

   Date of application for amendment: October 10, 1997, as supplemented on October 31, 1997.

   Brief description of amendment: The amendment revises and clarifies the offsite power requirements.

   Date of Issuance: March 24, 1998.

   Effective date: March 24, 1998, to be implemented within 60 days.

   Amendment No.: 155.

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

   Date of initial notice in Federal Register: December 31, 1997 (62 FR 68319).

   The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated March 24, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Brooks Memorial Library, 224 Main Street, Brattleboro, VT 05301.

   Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

   Date of application for amendments: January 21, 1997, as supplemented on December 15, 1997.

   Brief description of amendments: These amendments revise TS Section 15.6.11, ``Radiation Protection Program,'' references to Title 10, Code of Federal Regulations, Part 20.

   Date of issuance: March 17, 1998.

   Effective date: March 17, 1998, with full implementation within 45 days.

   Amendment Nos.: 182 and 186.

   Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: April 23, 1997 (62 FR 19837) The December 15, 1997, supplement provided clarifying information and modified proposed language within the scope of the original application and did not change the staff's initial proposed no significant hazards considerations determination.

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 17, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: The Lester Public Library, 1001 Adams Street, Two Rivers, Wisconsin 54241.

   Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

   Date of application for amendments: November 17, 1995 (TSCR 182), as supplemented on July 29, 1996, and December 15, 1997.

   Brief description of amendments: These amendments revise Technical Specifications 15.6.3.2, 15.6.3.3, and 15.6.5 designation of health physics manager to health physicist.

   Date of issuance: March 24, 1998.

   Effective date: March 24, 1998, with full implementation within 45 days.

   Amendment Nos.: 183 and 187.

   Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: September 11, 1996 (61 FR 47983).

   The December 15, 1997, letter provided additional clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards considerations determination.

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 24, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: The Lester Public Library, 1001 Adams Street, Two Rivers, Wisconsin 54241.

   Dated at Rockville, Maryland, this 1st day of April 1998.

   For the Nuclear Regulatory Commission. Elinor G. Adensam, Acting Director, Division of Reactor Projects--III/IV, Office of Nuclear Reactor Regulation.

   [FR Doc. 98-9040Filed4-7-98; 8:45 am]

   BILLING CODE 7590-01-P