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

As Enacted ( Federal Register)

Cited authorities 5

Cited in

[Federal Register: July 1, 1998 (Volume 63, Number 126)]

[Notices]

[Page 35986-36002]

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

[DOCID:fr01jy98-118]

NUCLEAR REGULATORY COMMISSION

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

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 June 8, 1998, through June 19, 1998. The last biweekly notice was published on June 17, 1998 (63 FR 33103).

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

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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 July 31, 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 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.

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

Date of amendment request: January 28, 1998 (Reference NRC-98-0027)

Description of amendment request: The proposed amendment will revise Section F and I of the Fermi, Unit 1 Technical Specifications to include requirements for control of effluents; dose limits; annual reporting in accordance with requirements of 10 CFR 50.36a; and numerical guideline criteria based on 10 CFR 50, Appendix I. Also, this amendment will correct several editorial errors.

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?

No, the proposed submittal establishes additional requirements and limits on radioactive effluent releases. No existing requirements are deleted. For these reasons, this proposed change will not significantly increase the probability or consequences of an accident at Fermi 1.

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

No, the addition of requirements for radioactive effluent releases will not cause a new kind of accident. The additional requirements involve having a functional waste system with procedures, submitting an annual report, and restricting the potential dose to the public from effluents. These changes, in themselves, do not require a different type of operation of systems. Any new system installed to enable future discharges will be evaluated at the time of design.

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

No, adding new requirements for radioactive effluents will not decrease the margin of safety. Since no existing requirements are being eliminated, this change will not reduce the margin of safety of the facility.

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.

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

Date of amendment request: January 28, 1998 (Reference NRC-98- 0025).

Description of amendment request: The proposed amendment will revise the Technical Specifications on access controls to provide flexibility while maintaining similar controls over access. Provisions will be established for cases where work is performed on the Protected Area boundary, such that the boundary temporarily will not meet the Technical Specification criteria. Redundancy between Technical Specifications will be eliminated. Figure B-1, ``Facility Plan,'' will be modified to show the buildings within the Protected Area, delete locations of the Protected Area gates and doors, and delete a building and equipment outside the Protected Area which are planned to be removed in the future. Finally, several editorial corrections will be made.

(1) Does the proposed change significantly increase the probability or consequences of an accident previously evaluated?

The proposed changes do not involve a significant increase in the probability or consequences of an accident. The proposed changes all involve access control, the Protected Area boundary, or deletion of details from a sketch, including a building and equipment planned for removal, which are outside the Protected Area. The changes still require control over the gates and doors to the Protected Area and that only authorized individuals will be issued the Fermi 1 key. Since the changes do not involve operation of any system, modifications to any required plant systems, nor eliminate the requirements for control of the Fermi 1 key and access points, the probability or consequences of an accident will be unaffected.

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

The proposed changes do not create the possibility of a new or different type of accident from any previously evaluated. The proposed changes will not lead to any different method of operating any systems, nor will they create any tests involving plant systems. The changes only affect the access control requirements, the Protected Area boundary, and deletion of details from a sketch. Changes of who issues the key, how doors are secured, provisions for temporary modifications to the boundary, requirements to observe the Protected Area boundary if degraded, wording consolidation, and more accurate building outlines cannot cause a new or different type of accident. Access points and the Fermi 1 key are still required to be controlled. The Boilerhouse and main unit output transformer are not used to support the Fermi 1 nuclear facility. Removal of the Boilerhouse and main unit output transformer from the drawing will help facilitate future removal plans, but will not cause a new or different accident from any previously evaluated, since they provide no support to the Fermi 1 nuclear facility. For these reasons, the proposed changes to the access control requirements and Figure B-1 will not create 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 changes do not involve a significant reduction in the margin of safety. The changes involve access control, the Protected Area boundary, and the sketch of the facility. Doors and gates in the Protected Area boundary will still be required to be secured when personnel are not inside. The keys will still be required to be controlled and issued only to authorized personnel. Compensatory measures will be required if the Protected Area boundary is degraded such that the requirements are not met. Therefore, there will not be a significant reduction in the margin of safety.

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, 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.

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

Date of amendment request: June 5, 1998 (NRC-98-0067).

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 2.1.2 to incorporate cycle-specific safety limit minimum critical power ratios (SLMCPRs) for the core that will be loaded during the upcoming refueling outage and update the footnote associated with the SLMCPR values to limit applicability of the SLMCPR values to Cycle 7 operation only.

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

The proposed license amendment establishes a revised SLMCPR value of 1.11 for two recirculation loop operation and 1.13 for single recirculation loop operation for use during Cycle 7 operation. The derivation of the cycle-specific SLMCPRs was performed using ``General Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13; U.S. Supplement, EDE-24011-P--A-13-US, August 1996; and the ``Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was submitted by General Electric Nuclear Energy (GENE) to the NRC on December 13, 1996.

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established, consistent with NRC approved methods, to ensure that fuel performance during normal, transient, and accident conditions is acceptable.

The probability of an evaluated accident is not increased by revising the SLMCPR

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values. The change does not require any physical plant modifications or physically affect any plant components. Therefore, no individual precursors of an accident are affected.

The proposed license amendment establishes a revised SLMCPR that ensures that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences. Specifically, the reload analysis demonstrates that a SLMCPR value of 1.11 (1.13 for single loop operation) ensures that less than 0.1 percent of the fuel rods will experience boiling transition during any plant operation if the limit is not violated.

Based on (1) the determination of the new SLMCPR values using conservative methods, and (2) the operability of plant systems designed to mitigate the consequences of accidents not having been changed;[,] the consequences of an accident previously evaluated have not been increased.

Additionally, updating of the footnote on the SLMCPR value in Technical Specification 2.1.2 to limit the applicability of the SLMCPR values to only Cycle 7 operation will not increase the probability or consequences of accidents previously evaluated. The updating of the footnote on the SLMCPR value in Technical Specification 2.1.2 is an administrative change that has no effect on the probability or consequences of accidents previously evaluated.

Therefore, the proposed TS change does not involve an increase in the probability or consequences of an accident previously evaluated.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed license amendment involves a revision of the SLMCPR from 1.09 to 1.11 for two recirculation loop operation and from 1.11 to 1.13 for single loop operation based on the results of analysis of the Cycle 7 core using the same fuel types as in previous fuel cycles, and updating of the footnote on the SLMCPR values in TS 2.1.2. Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in the allowable methods of operating the facility. This proposed license amendment does not involve any modifications of the plant configuration or changes in the allowable methods of operation. Therefore, the proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The change does not involve a significant reduction in the margin of safety.

The proposed license amendment establishes a revised SLMCPR value of 1.11 for two recirculation loop operation and 1.13 for single recirculation loop operation for use during Cycle 7 operation. The derivation of the cycle-specific SLMCPRs was performed using ``General Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13; U.S. Supplement, EDE-24011-P-A-13-US, August 1996; and the ``Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was submitted by General Electric Nuclear Energy (GENE) to the NRC on December 13, 1996. Use of these methods ensures that the resulting SLMCPR satisfies the fuel design safety criteria that less than 0.1 percent of the fuel rods experience boiling transition if the safety limit is not violated. Based on the assurance that the fuel design safety criteria will be met, the proposed license amendment does not involve a significant reduction in a margin of safety.

Additionally, updating of the footnote on the SLMCPR value in TS 2.1.2 will not decrease the margin of safety for accidents previously evaluated. The updating of the footnote on the SLMCPR value in Technical Specification 2.1.2 is an administrative change that does not reduce 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: Monroe County Library System, Ellis Reference and Information Center, 3700 South Custer Road, Monroe, Michigan 48161.

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

NRC Project Director: Cynthia A. Carpenter.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: May 27, 1998.

Description of amendment request: The request, if granted, would modify the Technical Specifications to allow the use of various controlled shift structures during a 36 to 48 hour work week. The request will allow the use of up to 12 hour shifts without routine heavy use of overtime.

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 will delete the TS 6.2.2.f. requirement ``. . . to have operating personnel work a normal 8-hour day, 40- hour week while the plant is operating.'' The proposed change will allow FPL to implement various controlled shift structures and durations during a nominal (36 to 48 hours) work week. The proposed changes will allow the use of up to 12 hour shifts without routine heavy use of overtime. The TS will continue to require the controls and guidelines for work hours to be contained in administrative procedures. The proposed amendments do not involve a change to any structure, system, or component that affects the probability or consequences of an accident previously evaluated. The proposed amendments are administrative in nature and do not involve a significant increase in the probability or consequences of any accident previously evaluated.

(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 proposed amendments will not change the physical plant or modes of plant operation and therefore, will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendments will not result in the addition or modification of equipment for any systems, structures, or components at St. Lucie.

The proposed changes modify the controls on working hours for operating personnel without significantly changing the hours worked on a weekly or annual basis, and do not alter the current guidelines on the use of overtime. The changes are administrative in nature. Consequently, operation of either unit 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 amendments would not involve a significant reduction in a margin of safety.

The proposed amendments will delete the TS 6.2.2.f. requirement ``. . . to have operating personnel work a normal 8-hour day, 40- hour week while the plant is operating.'' The proposed change will allow FPL to implement various controlled shift structures and durations during a nominal (36 to 48 hours) work week. The proposed changes will allow the use of up to 12 hour shifts without routine heavy use of overtime. The TS will continue to require the controls and guidelines for work hours to be contained in administrative procedures. This will result in fewer operating shift-to-shift turnovers per day and will allow more contiguous days off between work shifts. The net result of longer work shifts will be more rested crews with better communications between shifts.

The proposed changes do not alter the current guidelines on the use of overtime and will not alter the basis for any TS that is related to the establishment of, or maintenance of, a nuclear safety margin. Consequently, operation of St. Lucie Units 1 and 2 in accordance with the proposed amendments will 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, et al., Docket Nos. 50-335 and 50-389 St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: June 3, 1998

Description of amendment request: The request will modify the Technical Specifications to provide for the use of an interim periodic method of monitoring oxygen concentration in the service waste decay tanks in the event that continuous monitoring capability is lost.

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 license amendments are administrative in nature and will rectify an inconsistency between Surveillance Requirement 4.11.2.5.1 and the UFSAR that was inadvertently created by previous license amendments. The revisions will reinstate a previously approved conditional exception to the explicit terms of the presently stated TS requirement to continuously monitor the waste gases in the on service Waste Gas Decay Tank, and allow limited system operation using the laboratory gas partitioner to periodically analyze gas samples in the event that continuous monitoring capability becomes inoperable. Limits for potentially explosive mixtures of waste gases have not been altered, and explosive gas monitoring instrumentation does not prevent or mitigate design basis accidents or transients which assume a failure of or a challenge to a fission product barrier. The proposed revisions do not involve any change to the plant accident analyses assumptions, and do not involve accident initiators. Therefore, operation of either facility in accordance with its proposed amendment would 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 license amendments are administrative in nature and rectify an inconsistency between Technical Specification 4.11.2.5.1 and the UFSAR that was inadvertently created by previous license amendments. The revisions will not change the physical plant or the modes of plant operation defined in the Facility Licenses. The changes do not involve the addition or modification of equipment nor do they alter the design of plant systems. Therefore, operation of either facility in accordance with its 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 license amendments are administrative in nature and rectify an inconsistency between Surveillance Requirement 4.11.2.5.1 and the UFSAR that was inadvertently created by previous license amendments. The revisions will reinstate a previously approved conditional exception to the explicit terms of the presently stated TS requirement to continuously monitor the waste gases in the on service Waste Gas Decay Tank, and allow limited system operation using the laboratory gas partitioner to periodically analyze gas samples in the event that continuous monitoring capability becomes inoperable. Limits for potentially explosive mixtures of waste gases have not been altered, and explosive gas monitoring instrumentation does not prevent or mitigate design basis accidents or transients which assume a failure of or a challenge to a fission product barrier. The proposed changes do not alter the basis for any technical specification that is related to the establishment of, or the maintenance of, a nuclear safety margin. Therefore, operation of either facility in accordance with its proposed amendment would not involve a significant reduction in a margin of safety.

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

Local Public Document Room location: Indian River 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.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan.

Date of amendment requests: March 3, 1998.

Description of amendment requests: The proposed amendments would remove the word ``immediately'' from the Unit 1 hydrogen recombiner surveillance requirement 4.6.4.2.b.4 and revise the Unit 1 and Unit 2 Technical Specification 3/4.6.4 bases.

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

In accordance with 10 CFR 50.92, the proposed changes do not involve a significant hazards consideration if the changes do not:
Involve a significant increase in the probability or consequences of an accident previously evaluated;
Create the possibility of a new or different kind of accident from any accident previously evaluated; or
Involve a significant reduction in a margin of safety.

Criterion 1

This amendment request does not involve a significant increase in the probability or consequences of an accident previously evaluated. The change removes an ambiguous word from the technical specification. It does not physically alter the recombiner, nor does it adversely impact its operating characteristics.

The resistance to ground test will continue to be used to detect circuit faults. However, with the removal of the word ``immediately'', it will be possible to conduct the test near the ambient temperature, the temperature for which the 10,000 ohm criterion is applicable. The previously observed resistance value that was lower than 10,000 ohms is not indicative of a faulted heater circuit. Rather, it is the result of an elevated heater temperature and the electrical characteristics of the heater's insulating material, magnesium oxide. Magnesium oxide has a negative electrical resistance temperature coefficient, and it is not unusual or unacceptable for the measured insulation resistance to be less than 10,000 ohms when the heater temperature is elevated.

Criterion 2

This proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The hydrogen recombiner is used to mitigate the consequences of an accident, and it performs no function during normal operation. The change to the surveillance requirement removes an ambiguous word and does not affect the equipment or its installed configuration. No accident initiators that might be introduced by this change have been identified.

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Criterion 3

This proposed change does not involve a significant reduction in a margin of safety. The change removes an ambiguous word from the T/ S. The performance characteristics for the recombiner are not affected by this change, and no margin of safety is impacted.

The resistance to ground test will continue to be used to detect circuit faults. However, with the removal of the word ``immediately'', it will be possible to conduct the test near the ambient temperature, the temperature for which the 10,000 ohm criterion is applicable. The previously observed resistance values that were lower than 10,000 ohms are not indicative of a faulted heater circuit. Rather, they are the result of an elevated heater temperature and the electrical characteristics of the heater's insulating material, magnesium oxide. Magnesium oxide has a negative electrical resistance temperature coefficient, and it is not unusual or unacceptable for the measured insulation resistance to be less than 10,000 ohms when the heater temperature is elevated.

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

Local Public Document Room location: Maud Preston Palenske Memorial Library, 500 Market Street, St. Joseph, MI 49085

Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive, Buchanan, MI 49107

NRC Acting Project Director: Dr. Ronald R. Bellamy

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

Date of amendment request: May 7, 1998

Description of amendment request: The proposed revision to the Millstone Unit 3 licensing basis would address the addition of the dose from refueling water storage tank (RWST) back leakage into the design basis loss-of-coolant accident (LOCA) analysis and Chapter 15 of the Final Safety Analysis Report (FSAR).

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:

NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and concluded that the revision does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve an SHC because the revision would not:
Involve a significant increase in the probability or consequence of an accident previously evaluated.

The RWST is a standby system during normal operation, and provides the initial makeup water supply for the Emergency Core Cooling System (ECCS) when actuated in response to a Safety Injection signal. The RWST supply piping does not interface directly with the Reactor Coolant System or associated Reactor Coolant Pressure Boundary piping. All piping, up to and including the last isolation valve prior to the RWST, is rated for pressure exceeding RSS [recirculation spray system] pump discharge pressure.

The RWST is a passive tank, vented to atmosphere. Following swapover to post-LOCA recirculation cooling, the RWST is isolated and is no longer required for accident mitigation purposes. Back leakage will collect in the tank and mix with any remaining volume of water. The temperature of the mixed fluid will not significantly exceed the ambient temperature of the remaining tank volume due to the extremely low leakage rates involved. Because the tank is vented to atmosphere, pressurization of the tank [cannot] occur.

The specific condition of back leakage through the RWST isolation valves in combination with a motor operated valve failure does not contribute to the probability of a malfunction previously evaluated in the Safety Analysis Report. In lines that contain a motor operated valve and result in back leakage to the RWST, there exists another valve in series. The other valve is either another motor operated valve, a check valve, or a manually operated valve. The most limiting single failure assumed is the failure of the lowest leakage series valve to close and results in the maximum calculated leakage rate. Certain ECCS check valves are not subject to single failure consideration and are therefore credited as the barrier valve against back leakage.

The back leakage into the RWST results in sump water entering the RWST when it is at its minimum level. The RWST now becomes a radioactive source and contributes a shine dose to the surrounding areas. The increase in dose rates onsite will not prevent operators from remaining in the control room or from accessing equipment needed to mitigate the accident.

All piping and valves associated with RWST back leakage are located in harsh radiation areas. Backflow from RSS could increase dose rates in the areas where these valves are located. Since these areas are already classified as harsh radiation environments post LOCA, additional dose contributions from these pipes would not adversely impact EEQ [environmental qualification of electrical equipment] doses to vital equipment located in these rooms. Any vital equipment located within would continue to perform its safety function.

The leakage back to the RWST has no effect on the ability of the RSS pumps to perform their design function. The NPSH [net positive suction head] required by the RSS pumps is not adversely impacted by the loss of sump water back to the RWST. The RSS switchover to cold leg recirculation occurs prior to reaching a minimum level of 392,000 gallons in the RWST. Not counting the reactor coolant system volume, 774,000 gallons of water is in the sump. QSS [quench spray system] pumps shut off when the inventory in the RWST decreases to 93,000 gallons. Another 303,000 gallons will reach the sump prior to QSS shutoff. RWST back leakage displaces approximately 36,000 gallons of sump water back into the RWST at the end of 720 hours, leaving more than 1,000,000 gallons, not counting RCS [reactor coolant system] volume, in the sump. When RSS switches over to recirculation, at least 774,000 gallons of water will remain in the sump. After 720 hours, more water resides in the sump than when RSS is started. Therefore minimum NPSH requirements will not be impacted by this leakage.

Post-LOCA back leakage to the RWST has not previously been included in the radiological consequence analyses for Millstone Unit 3. Including this source in dose assessment increases the consequences of the accident. NNECO has tested the associated valves to establish bounding criteria to be used in the analysis of potential radiological consequences. The contribution of the RWST back leakage has been determined to be 2.1 Rem at the LPZ [low population zone] and 0.9 Rem at the Control Room. When combined with the present LOCA analysis radiological consequences, the results remain below the previously analyzed values reported in the FSAR. All dose estimates reflect the limiting exposure which, in this case, is Thyroid dose. All resultant doses are less than 10CFR100 and GDC [General Design Criterion] 19 limits to offsite and control room.

Back leakage to the RWST from the operation of RSS is a result of a LOCA. It cannot increase the probability of a LOCA. Therefore RWST back leakage does not increase the probability of an accident previously evaluated.

Based on the above, the proposed license amendment request does not involve a significant increase in the probability or consequence of an accident previously evaluated.
Create the possibility of a new or different kind of accident from any accident previously evaluated.

No new condition potentially impacting the ability to mitigate the accident is created by the back leakage. The low leakage rates from these valves occurs over [an] extended period of time during which other makeup water sources can be brought into service to account for lost inventory, if necessary.

Therefore, the proposed license amendment request does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Involve a significant reduction in a margin of safety.

The resultant dose from back leakage of ECCS valves to the RWST does not reduce the Margin of Safety. The offsite and control room doses, with the addition from RWST back leakage, remain below the licensing

[[Page 35992]]

base dose as listed in the SAR [safety analysis report]. Technical Specification 6.8.4 defines the basis for the leak reduction program. The basis for the program is to reduce leakage outside containment to the maximum extent possible. The Technical Specifications do not define the maximum amount of leakage or the origin of the leakage. The addition of the back leakage valves to the leak reduction program does not reduce the Margin of Safety.

Therefore, the proposed license amendment request does not involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.

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

Local Public Document Room location: Learning Resources Center, Three Rivers Community-Technical College, 574 New London Turnpike, Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, Connecticut

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

NRC Deputy Director: Phillip F. McKee.

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

Date of amendment request: June 5, 1998

Description of amendment request: The proposed revision to the Millstone Unit 3 licensing basis would address a recent steam generator tube rupture (SGTR) analysis that was determined to be an unreviewed safety question. The SGTR analyses described in the Final Safety Analysis Report (FSAR) include an offsite dose analysis and a margin to overfill analysis. Both of the analyses have been updated. The offsite dose analysis was updated to reflect a larger capacity for the steam generator atmospheric dump valve, and the margin to overfill analysis was updated to reflect a new single failure.

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:

NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve an SHC because the revision would not:
Involve a significant increase in the probability or consequence of an accident previously evaluated.

The FSAR Steam Generator Tube Rupture offsite dose analysis is being updated to reflect a larger capacity for the steam generator atmospheric dump valve. The updated analysis, as well as the current FSAR analysis, postulate the failure, in the open position, of the steam generator atmospheric dump valve associated with the steam generator with the ruptured tube. Revising the analyses does not impact the failure probability of the steam generator atmospheric dump valve. The SGTR analyses credit closure of the atmospheric dump valve block valve to isolate the failed open atmospheric dump valve. The revised SGTR analysis uses a larger flow capacity for the atmospheric dump valve. A larger flow capacity, without other changes being made, would increase the consequences associated with this failure. However, the time credited for closure of the block valve is being reduced to 20 minutes after the atmospheric dump valve fails open, instead of 30 minutes after the atmospheric dump valve fails open. A shorter isolation time, without other changes being made, would decrease the consequences associated with the atmospheric dump [valve] failing open. This faster isolation time more than compensates for the larger capacity assumed for the atmospheric dump valve. Therefore, the revised analyses does not increase the consequences of a Steam Generator Tube Rupture. The change is a revision to the analyses for a steam generator tube rupture and the description of the analyses in the FSAR. Changing the analyses and its description [cannot] cause an increase in the probability of a steam generator tube rupture.

Therefore, the proposed revision does not involve a significant increase in the probability or consequence of an accident previously evaluated.
Create the possibility of a new or different kind of accident from any accident previously evaluated.

The change is to the analyses and FSAR description of that analyses. The important changes in the analyses are the increased capacity of the atmospheric dump valve and the shorter time utilized for isolation of the failed open atmospheric dump valve. The only change in equipment credited in the analyses is the crediting of the block valve to close when there is a larger flow through the valve. The block valve can close under the postulated accident conditions.

Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Involve a significant reduction in a margin of safety.

The revised analyses reduces the time available to the Operators to isolate the failed open atmospheric dump valve from 30 minutes to 20 minutes. The actions required are unchanged. The twenty minutes allows sufficient time for the Operators to both recognize the failure of the atmospheric dump valve and to close the block valve. However, reducing the available time to the Operators from 30 minutes to 20 minutes represents a reduction in the margin for error available to the Operators and thus represents a reduction in the margin of safety. The reduction in the margin of safety is not significant since the twenty minutes allowed by the analysis is still significantly above the typical ten minute minimum assumed response time for Operator actions performed in the control room. In addition, Operator training provides assurance that the twenty minute time limit is met.

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

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.

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

Local Public Document Room location: Learning Resources Center, Three Rivers Community-Technical College, 574 New London Turnpike, Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, Connecticut.

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

NRC Deputy Director: Phillip F. McKee.

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

Date of amendment request: June 6, 1998.

Description of amendment request: The proposed revision to the Millstone Unit 3 licensing basis relates to operation of the supplementary leak collection and release system (SLCRS) after a postulated accident. Specifically, the proposed revision to the Final Safety Analysis Report (FSAR) would address (1) the manual actions required to trip the non-nuclear safety grade fans and time requirements for control room ventilation realignment, and (2) the input assumptions and results of the new loss-of-cooling accident/ control rod ejection accident analyses.

Basis for proposed no significant hazards consideration determination:

[[Page 35993]]

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:

NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve [an] SHC because the revision would not:
Involve a significant increase in the probability or consequences of an accident previously evaluated.

The potential condition of radioactive effluent bypassing the isolated boundary in the Supplemental Leak Collection and Release System after an accident cannot contribute to the probability of an accident previously evaluated. The leakage is caused by a postulated failure of the non-nuclear safety grade exhaust fans within the SLCRS boundary to trip after a safety injection signal. Operator action is needed to verify that the fans in question are tripped within a predetermined time delay after the accident in order that credit can be taken in the radiological dose analysis for the isolation of this source.

The proposed operator action will verify that the power to the fan motors is terminated, which cannot create any conditions leading to a new accident. The verification will augment the procedure to minimize the consequences of the accident itself. The trip circuits of the fan motors do not interface with safety systems.

The consequences of the limiting design basis accidents have been evaluated with the additional bypass leakage. The doses for the Exclusion Area Boundary, Low Population Zone and Unit 3 Control room remain below the previously calculated and approved licensing values. The calculated doses for the Technical Support Center are higher than previously approved, but below the radiological acceptance criteria of GDC [General Design Criterion] 19.

Therefore, the proposed license amendment does not involve a significant increase in the probability or consequence of an accident previously evaluated.
Create the possibility of a new or different kind of accident from any accident previously evaluated.

There are no conceivable conditions, created by the proposed operator action, that may lead to the possibility of a new accident. Interruption of power to the exhaust fans is, in itself, a part of accident mitigating activity. The proposed activity cannot create an adverse environment where a possibility of a new accident has to be considered.

The breakers used to de-energize the fans, control only the fan motors and no other equipment. Clear labeling ensures that no safety equipment is inadvertently de-activated. The revised ventilation system operating procedure will clearly specify the order of steps and confirmatory indicators necessary for safe shutdown of the exhaust fans. The equipment operator will be briefed before proceeding to open the breakers to the affected fan motors. To minimize the possibility of an error, this step will be done early in the sequence of procedural steps performed to re-align the control room ventilation system to the filtration/recirculation mode of operation after an accident.

Therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Involve a significant reduction in a margin of safety.

In considering the impact of the proposed revision on the margin of safety, as defined in the Technical Specifications, the impact on the design basis analysis of the fission product barriers must be evaluated.

The proposed operator action to trip the fans is done as part of personnel protective actions after a major accident, which is to stop the distribution of radioactive iodine into the vital areas through the ventilation system within a predetermined time. The maintenance of the fission product barriers is not affected by this action. This potential source of radioactivity associated with the ventilation fans discharging through the closed SLCRS boundary dampers has not been considered previously in the dose analysis. Including this source results in a small increase in the gamma and beta doses to the Technical Support Center. The GDC 19 limits for protection of personnel in the vital areas however, are not violated. The calculated doses to EAB/LPZ [exclusion area boundary and the low population zone] zones and to the control room vital area remain below the current licensing base values.

Therefore, the proposed license amendment request does not involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.

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

Local Public Document Room location: Learning Resources Center, Three Rivers Community-Technical College, 574 New London Turnpike, Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, Connecticut.

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

NRC Deputy Director: Phillip F. McKee.

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: April 23, 1998.

Description of amendment request: The amendment would update the operating licenses such that the corporate name of Pennsylvania Power and Light Company ``be changed to PP&L, Inc.''

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:
Will the change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. This request involves an administrative change only. The Operating Licenses (OLs) are being changed to reference the new corporate name of the licensee. No actual plant equipment or accident analyses will be affected by the proposed changes. Therefore, this request will have no impact on the possibility of any type of accident: new, different, or previously evaluated.
Will the change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. This request involves an administrative change only. The OLs are being changed to reference the new corporate name of the licensee. No actual plant equipment or accident analyses will be affected by the proposed change and no failure modes not bounded by previously evaluated accidents will be created. Therefore, this request will have no impact on the possibility of any type of accident: new, different, or previously evaluated.
Will the change involve a significant reduction in a margin of safety?

No. Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel and fuel cladding, Reactor Coolant System pressure boundary, and containment structure) to limit the level of radiation dose to the public. This request involves an administrative change only. The OLs are being changed to reference the new corporate name of the licensee.

No actual plant equipment or accident analyses will be affected by the proposed change. Additionally, the proposed change will not relax any criteria used to establish safety limits, will not relax any safety systems settings, or will not relax the bases for any limiting conditions of operation. Therefore, this request will not impact 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

[[Page 35994]]

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: Robert A. Capra.

Philadelphia Electric Company, Docket No. 50-171, Peach Bottom Atomic Power Station, Unit 1, York County, Pennsylvania

Date of application for amendment: March 2, 1998

Brief description of amendment: This proposed amendment will revise the Peach Bottom Atomic Power Station, Unit 1, Technical Specifications (TS) to include requirements for control of effluents and annual reporting in accordance with the requirements of 10 CFR 50.36a.

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 proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
  
  The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed changes do not impact the SAFSTOR status of Unit 1 or the design of any plant system, structure, or component (SSC). These changes are administrative in nature. They do not affect security at Unit 1 or the potential of radioactive material being released. Inspections for potential liquid and gas effluents have previously been established. These changes ensure the requirement for procedures and reporting are listed in TS. Therefore, these proposed changes do not increase the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
  
  The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because implementation of the proposed changes do not involve any physical changes to plant SSC or impact the SAFSTOR status. The changes are administrative in nature. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
  
  The proposed changes do not involve a significant reduction in a margin of safety because the proposed changes do not affect the plant SAFSTOR status. Because proposed changes are administrative in nature, they do not involve a question of safety. These changes involve reporting and adding a requirement that procedures be in place for effluent monitoring. 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration.
  
  Local Public Document Room location: Government Publications Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
  
  Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, Pennsylvania 19101.
  
  NRC Branch Chief: John W. N. Hickey.
  
  Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey
  
  Date of amendment request: May 13, 1998.
  
  Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3/4.10.8, ``Inservice Leak and Hydrostatic Testing,'' to delete the requirement for an operable High Drywell Pressure trip function. Specifically, TS 3.10.8.a is being revised to remove the reference to the Secondary Containment Isolation Actuation Instrumentation trip function 2.b.
  
  Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
  
  (1) The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
  
  The proposed TS revisions will continue to allow the performance of inservice leak and hydrostatic testing at a reactor coolant temperature of greater than 200 degrees Fahrenheit but less than or equal to 212 degrees Fahrenheit while considering the plant to remain in Operational Condition 4; however, the requirement to have an operable ``High Drywell Pressure'' Secondary Containment Isolation trip function during a leak or hydrostatic test is being deleted. This change will not have an impact on the consequences of an accident previously evaluated since the tests will continue to be performed nearly water solid and with all control rods fully inserted. The stored energy in the reactor core and coolant will continue to be very low and the potential for causing fuel failures with a subsequent increase in coolant activity will continue to be minimal. The remaining restrictions provided in Special Test Exception 3.10.8 requiring Secondary Containment Integrity and Filtration, Recirculation and Ventilation System (FRVS) operability will continue to provide assurance that potential releases into secondary containment will be restricted from direct release to the environment. With the reactor coolant continued to be limited to 212 degrees Fahrenheit, there will be little or no flashing of coolant to steam, and any release of radioactive materials will be minimized.
  
  In the event of a large primary system leak, the reactor vessel will rapidly depressurize, allowing the low pressure Emergency Core Cooling Systems (ECCS) to operate. The capability of the required ECCS in Operational Condition 4 remains adequate to maintain the core flooded under these conditions. Small system leaks will continue to be detected by leakage inspections, which are an integral part of the inservice leak and hydrostatic testing programs, before any significant inventory loss can occur. In addition, the ``High Drywell Pressure'' Secondary Containment Isolation trip function (TS Table 3.3.2-1, Trip Function 2.b) provides no additional protection against the events of concern during the inservice leak and hydrostatic tests. As a result, these changes will not increase the probability of an accident previously evaluated nor significantly increase the consequences of an accident previously evaluated.
  
  (2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
  
  The proposed changes to Special Test Exception 3.10.8 contained in this submittal will not adversely impact the operation of any safety related component or equipment. Since the proposed changes involve no hardware changes and no changes to existing structures, systems or components, there can be no impact on the potential occurrence of any accident due to new equipment failure modes. The remaining restrictions provided in proposed Special Test Exception 3.10.8 requiring Secondary Containment Integrity and Filtration, Recirculation and Ventilation System (FRVS) operability will continue to function as required, which will provide assurance that potential releases into secondary containment will be restricted from direct release to the environment. Furthermore, there is no change in plant testing proposed in this change request that could initiate an event. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
  
  (3) The proposed change does not involve a significant reduction in a margin of safety.
  
  The proposed TS revisions will still allow the performance of inservice leak and
  
  [[Page 35995]]
  
  hydrostatic testing at a reactor coolant temperature of greater than 200 degrees Fahrenheit but less than or equal to 212 degrees Fahrenheit while considering the plant to remain in Operational Condition 4; however, the requirement to have an operable ``High Drywell Pressure'' Secondary Containment Isolation trip function during a leak or hydrostatic test is being deleted. Since the reactor vessel head will remain in place, secondary containment will continue to be maintained, sufficient isolation actuation instrumentation will be maintained and all systems required in Operational Condition 4 will continue to be operable in accordance with the TS, the proposed changes will not have any significant impact on any design basis accident or safety limit. Since Hope Creek will still remain capable of meeting all applicable design basis requirements and retaining the capability to mitigate the consequences of accidents described in the UFSAR, the proposed changes contained in this submittal were determined to not result in a significant reduction in a margin of safety.
  
  The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Local Public Document Room location: Pennsville Public Library, 190 S. Broadway, Pennsville, NJ 08070.
  
  Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
  
  NRC Project Director: Robert A. Capra.
  
  Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey
  
  Date of amendment request: June 12, 1998.
  
  Description of amendment request: The proposed amendment would revise Technical Specification (TS) Limiting Condition for Operation (LCO) sections 3.7.1.1, 3.7.1.2, and 3.7.1.3. Specifically, the proposed changes implement more appropriate Ultimate Heat Sink (UHS) limits for river water temperature, which increases operational flexibility. In addition, the Station Service Water System (SSWS) and Safety Auxiliaries Cooling System (SACS) TS Action Statements are being revised to provide additional restrictions on continued plant operation. These revisions provide explicit TS guidance, which maintains SSWS/SACS operating configurations within design analysis assumptions.
  
  Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
  
  (1) The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
  
  LCO 3.7.1.3 Changes
  
  The proposed TS revisions related to UHS involve no hardware changes and no changes to existing structures, systems or components. The UHS and supported system temperature and configuration limits ensure that the UHS can remove required heat loads during design basis accidents and transients with the proposed UHS river water temperature limits. The proposed UHS TS ACTION Statements ensure that the plant is directed to enter a safe shutdown condition whenever the capability to mitigate design basis accidents and transients is lost. The existing UHS TS surveillance requirements to increase monitoring of the river water temperature at 82 deg.F adequately ensures that the actions required at elevated river water temperature conditions are taken as appropriate. Since the UHS will still remain capable of meeting all applicable design basis requirements and retaining the capability to mitigate the consequences of accidents described in the [Hope Creek] HC [Updated Final Safety Analysis Report] UFSAR, the proposed changes were determined to be justified. As a result, these changes will not increase the probability of an accident previously evaluated nor significantly increase the consequences of an accident previously evaluated.
  
  LCO 3.7.1.1 and 3.7.1.2 Changes
  
  The proposed TS revisions related to SSWS/SACS operating configuration restrictions involve no hardware changes and no changes to existing structures, systems or components. The additional restrictions requiring: 1) SACS heat exchanger operability in one SSWS/SACS pump per loop scenarios; and 2) assessments of SACS loop operability when a SSWS loop is declared inoperable; ensure that the SSWS/SACS can remove required heat loads during design basis accidents and transients with the proposed UHS river water temperature limits contained in this submittal. The proposed SSWS/SACS TS ACTION Statements ensure that the plant is directed to enter a safe shutdown condition whenever the capability to mitigate design basis accidents and transients is lost. Since SSWS/SACS will still remain capable of meeting all applicable design basis requirements and retaining the capability to mitigate the consequences of accidents described in the HC UFSAR, the proposed changes were determined to be justified. As a result, these changes will not increase the probability of an accident previously evaluated nor significantly increase the 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.
  
  LCO 3.7.1.3 Changes
  
  The proposed changes to the UHS TS contained in this submittal will not adversely impact the operation of any safety related component or equipment. Since the proposed changes involve no hardware changes and no changes to existing structures, systems or components, there can be no impact on the potential occurrence of any accident due to new equipment failure modes. The system configuration limits imposed by the UHS LCO ensure that supported systems can remove required heat loads during design basis accidents and transients with the proposed UHS river water temperature limits. Furthermore, there is no change in plant testing proposed in this change request that could initiate an event. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
  
  LCO 3.7.1.1 and 3.7.1.2 Changes
  
  The proposed changes to the SSWS/SACS TS contained in this submittal will not adversely impact the operation of any safety related component or equipment. Since the proposed changes involve no hardware changes and no changes to existing structures, systems or components, there can be no impact on the potential occurrence of any accident due to new equipment failure modes. The system configuration limits imposed by the SSWS/SACS LCOs ensure that systems can remove required heat loads during design basis accidents and transients with the proposed UHS river water temperature limits. Furthermore, there is no change in plant testing proposed in this change request that could initiate an event. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
  
  (3) The proposed change does not involve a significant reduction in a margin of safety.
  
  LCO 3.7.1.3 Changes
  
  The proposed changes for the TS related to the UHS ensure continued capability of the UHS to mitigate the consequences of design basis accidents and transients. The UHS supported systems' configuration limits and changes to the operating limits of the UHS ensure that the UHS can remove required heat loads during design basis accidents and transients with the proposed river water temperature limits. The proposed UHS TS ACTION Statements ensure that the plant is directed to: 1) enter a safe shutdown condition whenever the capability to mitigate design basis accidents and transients is lost; or 2) enter a conservatively short period of continued operation when supported system redundancy is reduced. Since the UHS will still remain capable of meeting all applicable design basis requirements and retaining the capability to mitigate the consequences of accidents described in the HC UFSAR, the proposed changes contained were determined to not result in a significant reduction in a margin of safety.
  
  LCO 3.7.1.1 and 3.7.1.2 Changes
  
  The proposed changes for the TS related to the SSWS/SACS ensure continued capability of these systems to mitigate the consequences
  
  [[Page 35996]]
  
  of design basis accidents and transients. The proposed configuration limits ensure that the safety-related heat removal systems can perform their safety functions during design basis accidents and transients with the proposed river water temperature limits. The SSWS/SACS TS ACTION Statements ensure that the plant is directed to: 1) enter a safe shutdown condition whenever the capability to mitigate design basis accidents and transients is lost; or 2) enter a conservatively short period of continued operation when supported system redundancy is reduced. Since the SSWS/SACS will still remain capable of meeting all applicable design basis requirements and retaining the capability to mitigate the consequences of accidents described in the HC UFSAR, the proposed changes contained were determined to not result in a significant reduction in a margin of safety.
  
  The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
  
  Local Public Document Room location: Pennsville Public Library, 190 S. Broadway, Pennsville, NJ 08070.
  
  Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
  
  NRC Project Director: Robert A. Capra.
  
  STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas
  
  Date of amendment request: May 7, 1998.
  
  Description of amendment request: The proposed amendment would change the Technical Specifications (TSs) to reflect reactor coolant system flow differences between the existing Model E and the replacement Delta 94 steam generators (SGs). Specifically, it would (1) add a new reactor core safety limit figure in TS 2.1.1, Reactor Core Safety Limits, that shows curves that are a function of core temperature, power and operating pressure, applicable to the Delta 94 SGs, (2) add a footnote in TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, to specify a new design loop flow rate applicable to the Delta 94 SGs, and (3) add a new flow rate requirement to TS 3.2.5, Departure from Nucleate Boiling (DNB) Parameters, applicable to the Delta 94 SGs. Related changes to the TS Bases were also proposed for Bases 2.1.1, Reactor Core Safety Limits, and Bases 3/ 4.2.5, DNB Parameters.
  
  Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed Technical Specification changes are necessary to reflect new conditions associated with replacement of the steam generators. The differences in the replacement steam generators only require small changes to parameters modeled in existing accident analyses. Accident analyses affected by the replacement steam generator parameter changes have each been evaluated to establish that there is no significant change in the documented results. In cases where an evaluation was not adequate, new analyses have been performed to verify that there is no significant change in the consequences of the affected accidents.

The Technical Specification changes specify new requirements (i.e., changed RCS [reactor coolant system] flow) which support the new and existing accident analyses. The accident analysis performed for these new requirements determined that neither the probability, nor the consequences, of accidents previously evaluated in the UFSAR

[Updated Final Safety Analysis Report] would be increased.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed Technical Specification changes are necessary to reflect new conditions associated with replacement of the steam generators. The differences in the replacement steam generators only require small changes to parameters modeled in existing accident analyses. The replacement of the original steam generators with new Model Delta 94 steam generators improves the structural integrity of the steam generator tubes. The improved structural integrity of the new steam generators does not increase the possibility of a new or different kind of accident from any accident previously evaluated such as a multiple steam generator tube rupture event.
The proposed change does not involve a significant reduction in a margin of safety.

The proposed change does not alter the manner in which Safety Limits, Limiting Safety System Setpoints, or Limiting Conditions for Operations are determined. Changes in parameters assumed in safety analyses associated with replacement of the steam generators have been analyzed and new Technical Specification limits are proposed. The new limits proposed for SL [Safety Limit] 2.1.1, ``Reactor Core''; Table 2.2-1, ``Reactor Trip System Instrumentation Trip Setpoints''; and LCO [Limiting Condition for Operation] 3.2.5, ``DNB

[Departure from Nucleate Boiling] Parameters'' maintain or improve the margin of safety.

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

Local Public Document Room location: Wharton County Junior College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.

NRC Project Director: John N. Hannon.

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

Date of amendment request: June 1, 1998.

Description of amendment request: The proposed amendment would revise the minimum steam generator (SG) tube roll expansion distances for the F* and elevated F* (EF*) repair criteria that were approved in Amendment 129.

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

The proposed change was reviewed in accordance with the provisions of 10 CFR 50.92 to show no significant hazards exist. The proposed change will not:

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

The changes to the minimum engagement lengths for F* and EF* do not change any of the conclusions of the original F* and EF* analyses. The technical justification for the repair criteria has not changed due to changes in the engagement lengths. The calculated engagement lengths continue to preclude tube pullout and rupture during all postulated conditions. Based on the geometry of the Model 51 SG, tube rupture type release rates are not expected for a postulated failure at an F* or EF* repair location. Engagement lengths were calculated such that structural integrity of the repaired tube meets the RG [Regulatory Guide] 1.121 requirements. Therefore, application of the new F* and EF* distances will not increase the probability of an accident previously evaluated.

The new calculated engagement lengths continue to preclude primary to secondary leakage during all conditions. Leakage for both F* and EF* remains negligible at normal operating conditions. The amount of leakage expected at faulted conditions from F* and EF* repaired tubes remains a small percentage of the maximum allowable leak rate during a[n] SLB [steamline break] and is considered negligible. Therefore, it can be concluded that leakage will be restricted such that off-site doses will not exceed a small fraction of 10 CFR part 100 and control

[[Page 35997]]

room doses will not exceed GDC [General Design Criterion] 19 criteria. Therefore, the proposed change to the F* and EF* distances will not increase the consequences of an accident previously evaluated.

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

Implementation of the proposed changes in F* and EF* distances does not introduce any significant changes to the plant design basis. As with the original acceptance of the amendment for using the original F* and EF* criteria, use of the proposed F* and EF* engagement lengths will not introduce a mechanism that will result in an accident initiated outside of the tubesheet crevice region. As previously discussed, the structural integrity of F* and EF* tubes will be maintained during all plant conditions. Any hypothetical accident as a result of tube degradation in the tubesheet crevice region of the tube will be bounded by the existing tube rupture analysis. Therefore, implementation of the proposed engagement lengths for F* and EF* will not create the possibility of a new or different kind of accident.

(3) Involve a significant reduction in the margin of safety.

The calculation for the new F* and EF* minimum engagement lengths used the same methodology as the original F* and EF* analysis. The only change was the assumed normal operating primary to secondary differential pressure. The new assumed differential pressure is the design differential pressure for the KNPP [Kewaunee Nuclear Power Plant] SGs. The calculation for the engagement lengths continues to use the appropriate safety factors from RG 1.121. The revised F* and EF* engagement lengths continue to preclude tube pullout at all plant conditions and to maintain the structural integrity of the tube. Additionally, primary to secondary leakage during all plant conditions is precluded as described in the preceding sections. Since the structural and leakage integrity is not changed by the proposed changes in engagement length, the margin of safety is not significantly reduced.

Additionally, use of the F* and EF* repair criteria will decrease the number of tubes removed from service by plugging or repaired by sleeving. Since both plugging and sleeving reduce reactor coolant flow margin, implementation of the F* and EF* repair criteria helps to maintain that flow margin.

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 Acting Project Director: Ronald R. Bellamy.

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

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

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

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

Date of amendment request: February 2, 1998, as supplemented February 18, 1998.

Description of amendment request: The proposed amendments would revise the wording to specify refueling outage surveillances. The changes clarify that these surveillances are to be performed on an 18- month frequency and need not be constrained to refueling outage conditions.

Date of publication of individual notice in Federal Register: February 10, 1998 (63 FR 6784).

Expiration date of individual notice: For comments February 24, 1998; For hearing March 12, 1998.

Local Public Document Room location: Oconee County Library, 501 West South Broad Street, Walhalla, South Carolina.

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

Date of amendment request: June 4, 1998.

Description of amendment request: The proposed amendments would revise Technical Specification 4.17.2 to allow continued operation with certain steam generator tubes that exceed their repair limit as a result of tube end anomalies. This action temporarily exempts these tubes from the requirement for sleeving, rerolling, or removal from service until they are repaired during or before the next scheduled refueling outages for the respective unit.

Date of publication of individual notice in Federal Register: June 17, 1998 (63 FR 33097).

Expiration date of individual notice: For comments July 1, 1998; For hearing July 17, 1998.

Local Public Document Room location: Oconee County Library, 501 West South Broad Street, Walhalla, South Carolina.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

Date of application for amendments: June 6 and December 11, 1996, April 11, May 1, August 14, October 15, November 5 and 14, December 3, 4, 15, 22, 23, 29 and 30, 1997, January 23, March 12 and 13, April 16, 20 and 28, May 7, 14 and 19, and June 2, 1998.

Brief description of amendments: Conversion to Standard Improved Technical Specifications (TSs). Supplements requested less restrictive changes to the planned conversion. These changes involve (1) plant- specific application of generically approved methodology supporting extended instrument surveillance intervals and allowed outage times, (2) operating practice to treat secondary containment as a single zone, (3) TS changes to support installation of a Power Range Neutron Monitoring System, Average Power Range Monitor and Rod Block Monitor TS improvements, and the Maximum Extended Load Line Limit analysis, (4) TSs to specify reactor vessel water level should be greater than the top of the irradiated fuel, (5) reflect plant-specific design condition that excludes average U-235 enrichment, (6) all spiral off-load procedures and adopt revision to Surveillance Requirement (SR). Also, changes to (1) SR relating to core reactivity difference between actual and expected critical rod configuration, (2) calibration frequency for local power range monitors and (3) an alternate SR for Unit 3 for position verification of the low pressure core injection cross tie valves.

Date of publication of individual notices in the Federal Register: June 1, 1998 (63 FR 29763), and June 12, 1998 (63 FR 32252).

Expiration dates of individual notices: July 1, 1998 (63 FR 29763) and July 13, 1998 (63 FR 32252).

Local Public Document Room location: Athens Public Library, South Street, Athens, Alabama 35611.

[[Page 35998]]

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama

Date of application for amendments: October 1, 1997, as supplemented October 14, 1997, March 16, April 1 and 28, May 1 and 20, 1998.

Brief description of amendments: Change Technical Specifications to allow operation at the uprated power level of 3458 MWt which represents a power level increase of 5 percent.

Date of publication of individual notice in the Federal Register: June 9, 1998 (63 FR 31533).

Expiration date of individual notice: July 9, 1998.

Local Public Document Room location: Athens Public Library, South Street, Athens, Alabama 35611.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at the local public document rooms for the particular facilities involved.

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

Date of application for amendment: March 25, 1998, as supplemented on April 8, and May 5, 1998.

Brief description of amendment: The amendment modifies the Pilgrim Nuclear Power Station Technical Specification Section 3.6.A.1 with respect to the monitoring requirements for the vessel flange and adjacent shell differential temperature during heatup and cooldown and removes the 145 deg.Fahrenheit differential temperature limit.

Date of issuance: June 19, 1998.

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

Amendment No.: 175.

Facility Operating License No. DPR-35: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 28, 1998 (63 FR 23304). The May 5, 1998, 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 June 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11 North Street, Plymouth, Massachusetts 02360

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: July 15, 1996, as supplemented on June 19, 1997, and February 2, 1998.

Brief description of amendments: The amendments relocate requirements related to fire protection from the Technical Specifications (TS) to the Updated Final Safety Analysis Report. The TS sections to be relocated are: 3/4.3.7.9, Fire Detection Instrumentation; 3/4.7.5, Fire Suppression Systems; 3/4.7.6, Fire Rated Assemblies; and 6.1.C.4, Fire Brigade Staffing. The amendments also replace License Condition 2.C.(25) for Unit 1 and License Condition 2.C.(15) for Unit 2.

Date of issuance: June 10, 1998.

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 127 and 112.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the operating licenses and the Technical Specifications.

Date of initial notice in Federal Register: September 25, 1996 (61 FR 50340). The June 19, 1997, and February 2, 1998, supplements clarified the license conditions by providing specific approval dates for previous fire protection safety evaluations. This information was within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 10, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library, Illinois Valley Community College, Oglesby, Illinois 61348

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

Date of application for amendment: March 13, 1998, as supplemented March 30, 1998.

Brief description of amendment: The amendment revises the auxiliary feedwater system technical specification to allow two auxiliary feedwater flow control valves in one train to be inoperable for up to 72 hours.

Date of issuance: June 10, 1998.

Effective date: June 10, 1998.

Amendment No.: 183.

Facility Operating License No. DPR-20: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR 19967) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 10, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope College, Holland, Michigan 49423-3698

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

Date of amendment request: May 20, 1998 (NRC-98-0099)

Description of amendment request: The amendment revises the action specified in Technical Specification 3.1.3.1, ``Control Rod Operability,'' by changing the action statements associated with the scram discharge volume vent and drain valves to align with those in the NUREG-1433, Revision 1, ``Standard Technical

[[Page 35999]]

Specifications General Electric Plants, BWR/4.''

Date of issuance: June 12, 1998.

Effective date: June 12, 1998.

Amendment No: 120.

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

Public comments requested as to proposed no significant hazards considerations (NSHC): Yes (63 FR 29254 dated May 28, 1998). The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided for an opportunity to request a hearing by June 29, 1998, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of exigent circumstances, and final determination of no significant hazards consideration are contained in a Safety Evaluation dated June 12, 1998.

Local Public Document Room location: Monroe County Library System, Ellis Reference and Information Center, 3700 South Custer Road, Monroe, Michigan 48161.

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

NRC Project Director: Cynthia A. Carpenter.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: May 22, 1998.

Brief description of amendments: The amendments revise Surveillance Requirement Section 4.4.3.3 of each unit's Technical Specification to be consistent with the plant design; specifically, deleting the reference to manual transfer of power supply from normal to emergency.

Date of issuance: June 17, 1998.

Effective date: As of the date of issuance.

Amendment Nos.: Unit 1--166; Unit 2--158.

Facility Operating License Nos. NPF-35 and NPF-52: The amendments revised the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration: Yes. (63 FR 29759 dated June 1, 1998). That notice provided an opportunity to submit comments on the Commission's proposed no significant hazards consideration determination. No. comments have been received. The notice also provided for an opportunity to request a hearing by July 1, 1998, but indicated that if the Commission makes a final no significant hazards consideration determination, any such hearing would take place after issuance of the amendments.

The Commission's related evaluation of the amendments, finding of exigent circumstances, and final no significant hazards consideration determination are contained in a Safety Evaluation dated June 17, 1998.

Local Public Document Room location: York County Library, 138 East Black Street, Rock Hill, South Carolina 29730.

Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: March 20, 1998.

Brief description of amendment: The amendment revised the Improved Technical Specification 5.6.2.8 to reflect the current schedule for performing the required reactor coolant pump flywheel inspection.

Date of issuance: June 8, 1998.

Effective date: June 8, 1998.

Amendment No.: 167.

Facility Operating License No. DPR-31: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR 25110).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 8, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619 W. Crystal Street, Crystal River, Florida 34428.

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

Date of application for amendments: November 22, 1996, as revised and replaced February 2, 1998.

Brief description of amendments: The amendments revise the Technical Specifications (TS) to allow for the installation of a temporary fuel oil storage and transfer system in order to maintain the operability of one Unit 3 emergency diesel generator during the performance of a required surveillance to clean the permanent fuel oil storage tank.

Date of issuance: June 9, 1998.

Effective date: June 9, 1998.

Amendment Nos.: 197 and 191.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments revised the TS.

Date of initial notice in Federal Register: February 25, 1998 (63 FR 9604).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 9, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International University, University Park, Miami, Florida 33199.

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

Date of amendment request: March 23, 1998.

Description of amendment request: The proposed change would revise the Seabrook Station Technical Specifications (TSs) to add a new TS 3.0.5 that would provide an exception to TSs 3.0.1 and 3.0.2 to allow the performance of required testing to demonstrate the operability of the equipment being returned to service or the operability of other equipment.

Date of issuance: June 16, 1998.

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

Amendment No.: 57.

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

Date of initial notice in Federal Register: April 22, 1998 (63 FR 19972)

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

No significant hazards consideration comments received: No.

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

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

Date of application for amendment: December 8, 1997.

Brief description of amendment: The changes modify the Technical Specifications to resolve several compliance issues by rewording of the text, changing terminology, correcting a

[[Page 36000]]

mode applicability, correcting a formula, updating the Design Features section, and updating the Bases section to reflect the changes.

Date of issuance: June 16, 1998.

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

Amendment No.: 216.

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

Date of initial notice in Federal Register: January 28, 1998 (63 FR 4319).

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

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center, Three Rivers Community-Technical College, 574 New London Turnpike, Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, Connecticut

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 14, 1997, as supplemented by letters dated October 9, 1997, March 31, 1998, and April 15, 1998.

Brief description of amendments: The amendments revised the combined Technical Specifications (TS) for the Diablo Canyon Power Plant, Unit Nos. 1 and 2 to change the surveillance frequencies from at least once every 18 months to at least once per refueling interval (nominally 24 months) for (1) eight slave relays, (2) 20 electrical system tests, (3) one electrical Bases change, and (4) five miscellaneous tests.

Date of issuance: June 5, 1998.

Effective date: June 5, 1998, to be implemented within 90 days from the date of issuance.

Amendment Nos.: Unit 1--126; Unit 2--124.

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

Date of initial notice in Federal Register: March 26, 1997 (62 FR 14466). The October 9, 1997, March 31, 1998, and April 15, 1998, supplemental letters provided additional 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 June 5, 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

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendments: January 27, 1998.

Brief description of amendments: These amendments revise Table 3.6.3-1 of the Technical Specifications by removing the isolation time for the high pressure coolant injection turbine exhaust valves and adding a notation that the isolation is not required.

Date of issuance: June 16, 1998.

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

Amendment Nos.: 129 and 90.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 11, 1998 (63 FR 11921).

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

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500 High Street, Pottstown, PA 19464.

Power Authority of the State of New York, Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: June 25, 1997, as supplemented by letter dated June 2, 1998.

Brief description of amendment: The amendment changes the Technical Specifications (TSs) to allow for up to +17\1/2\ steps of control rod misalignment when power is greater than 85%.

Date of issuance: June 17, 1998.

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

Amendment No.: 180.

Facility Operating License No. DPR-64: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 27, 1997 (62 FR 45461)

The June 2, 1998, supplement provided a clarification to the wording of the TSs and did not change the staff's proposed finding of no significant hazards consideration. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 17, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10610.

Power Authority of the State of New York, Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: September 3, 1997.

Brief description of amendment: The amendment changes the Technical Specifications (TSs) by revising the number of hours operating personnel can work in a normal shift. The proposed amendment also contains some administrative changes to the TS.

Date of issuance: June 17, 1998.

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

Amendment No.: 181

Facility Operating License No. DPR-64: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 22, 1997 (62 FR 54875).

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

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York 10610.

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 97-03).

Brief description of amendments: The amendments change the Technical Specifications by adding a new Limiting Condition for Operation 3.7.1.6 that addresses the requirements for the main feedwater isolation valve functions required by the Sequoyah Nuclear Plant accident analysis.

Date of issuance: June 8, 1998.

Effective date: As of the date of issuance to be implemented no later than 45 days after issuance.

Amendment Nos.: Unit 1-232; Unit 2-222.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revise the technical specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR 19979).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 8, 1998.

No significant hazards consideration comments received: No.

[[Page 36001]]

Local Public Document Room location: Chattanooga-Hamilton County Library, 101 Broad Street, Chattanooga, Tennessee 37402.

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

Date of application for amendment: April 29, 1998.

Brief description of amendment: The requested changes would allow, temporarily, both trains of hydrogen igniters to be declared inoperable for up to 72 hours.

Date of issuance: June 9, 1998.

Effective date: June 9, 1998.

Amendment No.: 10.

Facility Operating License No. NPF-90: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: May 7, 1998 (63 FR 25243).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 9, 1998.

No significant hazards consideration comments received: None.

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

Toledo Edison Company, Centerior Service Company, and The Cleveland Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: August 26, 1997.

Brief description of amendment: This amendment changed Technical Specification (TS) Section 3/4.2, ``Power Distribution Limits.'' The departure from nucleate boiling parameters limiting condition for operation was modified due to an industry notification.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment No.: 222.

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

Date of initial notice in Federal Register: October 8, 1997 (62 FR 52590)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 11, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William Carlson Library, Government Documents Collection, 2801 West Bancroft Avenue, Toledo, OH 43606.

Toledo Edison Company, Centerior Service Company, and The Cleveland Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: August 26, 1997.

Brief description of amendment: This amendment revises Technical Specification (TS) Section 3/4.6.1.3, ``Containment Systems-- Containment Air Locks,'' and the associated bases. The limiting condition for operation and the surveillance requirements were modified. The application also proposed a change to TS Bases 3/4.9.4, ``Refueling Operations--Containment Penetrations.'' That bases change was approved by letter dated March 19, 1998.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment No.: 223.

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

Date of initial notice in Federal Register: October 22, 1997 (62 FR 54876)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 11, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William Carlson Library, Government Documents Collection, 2801 West Bancroft Avenue, Toledo, OH 43606

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50- 281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: December 18, 1997.

Brief Description of amendments: These amendments revise the Technical Specifications (TS) to clarify the terminology used for describing equipment surveillances performed on a refueling interval frequency, and to use consistent wording.

In two cases the proposed changes are denied. These two exceptions, TS 4.6.A.1.b and 4.6.C.1.e, do not include required specific Mode restrictions and could not be approved at this time. If appropriate revisions are submitted, these two exceptions could be found to be acceptable at a later time.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment Nos.: 213 and 213.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR 25118). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 11, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of William and Mary, Williamsburg, Virginia 23185.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50- 281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: November 5, 1997, as supplemented January 28, 1998 and May 12, 1998.

Brief Description of amendments: These amendments permit an increase in the maximum allowable fuel enrichment for core reloads from 4.1 to 4.3 weight percent U\235\.

Date of issuance: June 19, 1998.

Effective date: June 19, 1998.

Amendment Nos.: 214 and 214.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: December 31, 1997 (62 FR 68320)

The January 28 and May 12, 1998 submittals provided clarifying information that did not affect the initial no significant hazards determination.

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

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of William and Mary, Williamsburg, Virginia 23185

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50- 281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: March 25, 1998.

Brief Description of amendments: These amendments revise the Technical Specifications to change certain management titles. There is no change in duties or responsibilities proposed. Specifically, the Station Manager's title is changed to Site Vice President. The title of Assistant Station Manager Operations and Maintenance is changed to Manager-Operations and Maintenance. The title of Assistant Station Manager Nuclear Safety and Licensing is changed to Manager-Station Safety and Licensing.

Date of issuance: June 19, 1998.

Effective date: June 19, 1998.

Amendment Nos.: 215 and 215.

[[Page 36002]]

Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR 25119) The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of William and Mary, Williamsburg, Virginia 23185.

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

Date of application for amendment: October 13, 1997, supplemented on February 10, 1998.

Brief description of amendment: The amendment involves miscellaneous changes to the TS to (1) relocate information to the Updated Safety Analysis Report (USAR), (2) delete redundant information, (3) incorporate new references, (4) delete incorrect references, (5) correct errors, and (6) augment existing requirements.

Date of issuance: June 9, 1998.

Effective date: June 9, 1998.

Amendment No.: 137.

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

Date of initial notice in Federal Register: April 8, 1998 (63 FR 11926).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 9, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin, Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power Station, Franklin County, Massachusetts

Date of application for amendment: September 5, 1997 and March 30, 1998.

Brief description of amendment: Revises Technical Specifications and bases in order to allow loads of up to 80-tons to travel over the spent fuel pool.

Date of issuance: June 17, 1998.

Effective date: June 17, 1998.

Amendment No.: 149.

Facility Opertating (Possession Only) License No. DPR-3: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 22, 1997 (62 FR 54879) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 17, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Greenfield Community College, 1 College Drive, Greenfield, Massachusetts 01301

Dated at Rockville, Maryland, this 24th day of June 1998.

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

[FR Doc. 98-17352Filed6-30-98; 8:45 am]

BILLING CODE 7590-01-P

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

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