Agency information collection activities: Operating licenses, amendments; no significant hazards considerations; biweekly notices,

[Federal Register: January 27, 1999 (Volume 64, Number 17)]

[Notices]

[Page 4152-4165]

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

[DOCID:fr27ja99-99]

NUCLEAR REGULATORY COMMISSION

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

1. Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Pub. L. 97-415 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 January 4, 1999, through January 14, 1999. The last biweekly notice was published on January 13, 1999.

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administration Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By February 26, 1999, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be 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

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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.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois.

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

Date of application for amendment request: December 17, 1998.

Description of amendment request: The amendments would revise the respective facility Technical Specifications (TS) by adding a new Limiting Conditions for Operations which provides an administrative enhancement by allowing testing required to return equipment to service to be conducted under administrative controls.

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

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

   The proposed change has no impact on the design basis of the plant. The change has no impact on the response of the plant during normal or transient conditions. Incorporation of ISTS [improved Standard Technical Specification] 3.0.5 provides the necessary administrative controls that allow the return of equipment to service to complete testing required to demonstrate operability. Without this allowance, certain components could not be restored to operable status and a plant shutdown would ensue. It is not the intent of the TS to preclude the return to service of a component in order to confirm its operability or the operability of other equipment. This allowance is deemed to be a safer operation than requiring a plant shutdown to complete necessary testing. This allowance is considered acceptable because it: (1) is temporary; (2) accompanied by appropriate administrative controls, and; (3) provides a safety enhancement by restoring the plant status to, or confirming the existing plant status is in, a condition that is expected to provide for safe operation.

   ISTS 3.0.5 was adopted to address the ambiguity that ACTION requirements do not strictly allow the restoration of equipment to its normal configuration to perform functional testing required to demonstrate operability. The components involved will have completed maintenance and or testing that will demonstrate, with reasonable assurance, that the component can perform its intended safety function.

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

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

   The proposed changes do not introduce new features or modify plant structures, systems or components that may impact station operations under normal or abnormal conditions. The proposed changes will allow the necessary testing to ensure safety related equipment will perform its design basis safety function.

   Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in the margin of safety for the following reasons:

   The proposed changes have no impact on any of the Safety Limits provided in the Technical Specifications, nor does the change impact the operation of structures, systems and components import to plant safety. The purpose of the proposed change is to return equipment to service, under administrative controls, to complete operability testing. Therefore, allowing the return of equipment to service will promote timely restoration of, or confirmation of, equipment operability thereby increasing the margin of safety from that existing with this equipment remaining out of service. Temporarily returning inoperable equipment to service for the purpose of confirming operability places the plant in a condition which has been previously evaluated and determined to be acceptable for short periods. Therefore, the proposed change does not involve a significant reduction in safety.

   Therefore, the proposed changes do not involve a significant reduction in the 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 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments requested involve no significant hazards consideration.

   Local Public Document Room location: for Dresden, Morris Area Public Library District, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021; for LaSalle, Jacobs Memorial Library, 815 North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, Illinois 61348-9692.

   Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice President and General Counsel, Commonwealth Edison Company, P.O. Box 767, Chicago, Illinois 60690-0767.

   NRC Project Director: Stuart A. Richards.

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

   Date of amendment request: December 24, 1998.

   Description of amendment request: These amendment requests change the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2) Technical Specifications (TSs) to ensure that Emergency Diesel Generator (EDG) requirements contained in Technical Specification 3/ 4.8.1 for both units are consistent with assumptions contained in design analyses and requirements of plant procedures. Revisions to TS 3/4.8.1 ``A.C. Sources,'' contained in this amendment provide more conservative limiting conditions for operation (LCO) and surveillance requirements that affect EDG fuel oil storage volume, EDG load rejection and overspeed testing, and EDG operating frequency requirements. The applicable bases for each unit are also refined, as necessary, to strengthen the explanations regarding EDG fuel oil storage systems and provide the EDG overspeed in terms of frequency (Hertz) and speed (Revolutions Per Minute).

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

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

   The addition of the term ``usable'' to LCO 3.8.1.1 and 3.8.1.2 for both Units will assure that the required quantity of fuel oil will be available to operate the diesel during emergency conditions. This revision including the discussion contained in the Technical Specification Bases has no physical impact on the diesels or their setpoints. These revisions also do not delete any function previously provided by the diesels. There are no design bases accidents for which failure of the diesel is considered an initiating event. Therefore, the probability of an accident previously evaluated in the safety analysis is not increased by this change. The proposed changes do not involve an increase in the consequences of an accident previously analyzed, as they make the limiting condition for operation and associated bases more conservative and involve no physical changes to the diesels.

   The revised EDG single largest load rejection and overspeed criteria do not involve an increase in the probability or the consequences of accidents previously analyzed. The surveillance tests impacted by the proposed revision are performed only during shutdown when the opposite train EDG and its connected AC power system are relied upon as the emergency AC power source. Further, there are no design basis accidents for which changes to EDG load rejection test acceptance criteria can be an initiating event. The proposed changes affect the diesel testing requirements but do not affect the operating or design parameters. The changes also do not affect the diesels' ability to mitigate the consequences of an accident. They serve to ensure the ability of the diesel to reject the largest load. The overspeed criteria ensures that diesel frequency does not exceed a certain value subsequent to a load rejection. This criteria also ensures compliance with the guidance of Safety Guide 9 for Unit 1 and Regulatory Guide 1.9 for Unit 2. It does not involve an increase in the consequences of an accident previously analyzed. The revision does not impact accidents previously analyzed and would not, therefore, affect the consequences of accidents previously analyzed.

   Revising the EDG operating frequency as discussed in the proposed amendment protects [engineered safety feature] ESF pumps from runout conditions and motors from operating in an unanalyzed condition. The narrower frequency limits are more restrictive and have no adverse effect on the diesel generator operability. The proposed revision to decrease the EDG operating frequency limit does not involve an increase in the probability of an accident as described in the [Updated Final Safety Analysis Report] UFSAR. There are no design basis accidents for which failure of the diesel is considered an initiating event. A narrower operating frequency does not increase the probability of a design basis accident; it ensures that equipment performs their intended function. This change is intended to prevent the diesel from being loaded beyond analyzed loading limits and protect ESF equipment. The more conservative surveillance requirements being applied to operating limits will provide greater assurance that the diesels will be operable and that greater performance requirements are not imposed on ESF equipment. This change, therefore, will not result in an increase in the consequences of an accident previously described.

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

   The proposed revisions do not create the possibility of a new or different kind of accident from any accident previously evaluated. They also will have no adverse impact on the design basis accidents previously evaluated in the UFSAR. The revisions contained in the proposed amendment are more restrictive to assure that diesel and ESF equipment are available and fully operable to perform their intended safety function following a design basis accident and a loss of offsite power. The proposed changes do not involve physical changes to plant equipment or the AC power system configuration. New failure modes are not introduced as a result of the proposed revisions. A revision of the diesel frequency will prevent motors and pumps from being subjected to over-frequency conditions which could reduce the life of the equipment. Increasing the load rejection criteria for Unit 1 and including overspeed criteria for both units revises surveillance test criteria for verifying load rejection capability. This does not affect the probability of malfunction of a diesel or its connected emergency AC power system. Further, it does not create a new failure mode. Revising diesel fuel oil storage requirements to include the term ``usable'' reduces the potential for misinterpretation of this specification; it does not create a new kind of accident from any accident previously evaluated.

   The revisions contained in this license amendment have the effect of making the BVPS Technical Specifications more conservative than previously. This license amendment request will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The margin of safety is not reduced as a result of the proposed revisions. The margin of safety depends on the maintenance of specific operating parameters within design limits. The margin of safety derived from limiting condition for operation 3.8.1.1 and 3.8.1.2 for both Units is enhanced by adding ``usable'' in these requirements. This revision reduces the possibility of misinterpreting Technical Specification requirements. The addition of diesel overspeed criteria (both units) and increasing load rejection criteria for Unit 1 does not reduce the margin of safety. Diesel reliability and performance during a loss of offsite power and a design basis accident are enhanced by this more conservative surveillance test requirement. Revision of diesel operating frequency limits protects engineered safety features equipment from overfrequency conditions; this would not be a significant reduction in the margin of safety. Though the temporary Unit 1 EDG loading limit of 2791.51 exceeds the Safety Guide 9 value of 2745, it still is below the EDG 2000 hour rating limit of 2850 kW contained in Surveillance Requirement 4.8.1.1.2.b.6. Further, the loading value of 2791.51 kW does not exceed the design

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   loading capability of the EDG. Based on engineering analyses, the revisions contained in the proposed amendment will not significantly reduce the margin of safety. Engineered safety features equipment will continue to function, as assumed in the safety analysis, to ensure that fuel, reactor coolant system and containment design limits are not exceeded.

   Therefore, this change will not involve a significant reduction in a margin of safety due to the continued availability and reliability of the A.C. electrical power sources.

   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: B.F. Jones Memorial Library, 663 Franklin Avenue, Aliquippa, PA 15001.

   Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

   NRC Project Director: S. Singh Bajwa.

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

   Date of amendment request: December 24, 1998.

   Description of amendment request: The proposed amendments would revise the Technical Specification (TS) requirements for the axial flux difference [AFD] monitor, quadrant power tilt ratio [QPTR] monitor, rod position deviation monitor, and rod insertion limit (RIL) monitor. The changes would (1) relocate requirements for the AFD monitor and the QPTR monitor to the Licensing Requirements Manual (LRM); (2) delete requirements for the rod position deviation monitor and RIL monitor from the TSs; (3) modify Unit 1 surveillance requirements (SR) 4.1.3.5 and 4.1.3.6 by incorporating the Unit 2 wording to provide surveillances more consistent with the Limiting Condition for Operation (LCO); (4) change Unit 1 SR 4.1.3.2.2, SR 4.1.3.5, SR 4.1.3.6 and Unit 2 SR 4.1.3.5 from 24 hour surveillance frequencies to 12 hour frequencies; and (5) delete Unit 1 SR 4.1.3.2.3.

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

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

   The proposed amendment would modify applicable Technical Specifications (TS) by deleting requirements associated with the rod position deviation monitor and rod insertion limit (RIL) monitor and relocating the requirements associated with the axial flux difference (AFD) monitor and quadrant power tilt ratio (QPTR) monitor from the following specifications and Bases:

   Unit 1: 4.1.3.1.2, 3.1.3.2, 4.1.3.2.2, 4.1.3.2.3, 4.1.3.6, 4.2.1.1, 4.2.4; Unit 2: 4.1.3.1.2, 4.1.3.2, 4.1.3.6, 4.2.1.1, 4.2.4.

   The TS contains requirements where a reduced surveillance interval is required in the event the monitors referenced in the above specifications, surveillance requirements (SR) and associated Bases are inoperable. Removing the requirements associated with these monitors from the TS will not affect the ability of any system to perform its design function.

   Nuclear Electric Institute (NEI) Technical Specification Task Force (TSTF) 110 Revision 2 provides the basis for these changes and recommends relocating the requirements for these monitors to ``plant administrative practices.'' The AFD monitor and the QPTR monitor requirements will be relocated to the LRM and changes to these requirements will be controlled in accordance with the 10 CFR 50.59 process which will require NRC approval if the change constitutes an unreviewed safety question. However, based on the smaller change in surveillance intervals, deletion and not relocation of the rod position deviation monitor and the RIL monitor requirements can be justified and is proposed.

   Although these monitors are being removed from the TSs, they will continue to be maintained as described in the [Updated Final Safety Analysis Report] UFSAR (subject to revisions via the 10 CFR 50.59 process). Removing the rod deviation monitor requirements from Unit 1 SR 4.1.3.2.3 makes the remaining portion of SR 4.1.3.2.3 redundant to SR 4.1.3.2.2.a; therefore, SR 4.1.3.2.3 has been deleted. In addition, the 24-hour surveillance frequency in Unit 1 SR 4.1.3.2.2, 4.1.3.5 and 4.1.3.6 as well as in Unit 2 SR 4.1.3.5 is being changed to 12 hours to assure the required parameters are adequately monitored and to provide consistency between the units and related requirements as well as the Improved Standard Technical Specifications (ISTS).

   Removing these monitors from the TS is consistent with the NRC approved changes to the ISTS identified in TSTF-110, Revision 2. Verification that plant conditions are within specified limits at the frequency specified in the normal SR provides sufficient information that allows the operator to detect a parameter that is beginning to deviate from its expected limits. The specified frequency takes into account other information (i.e., rod position indication system, rod bottom alarm and excore neutron detectors) that is continuously available to the operator in the control room, so that during changes in plant conditions, deviation from the limits can be readily detected.

   The proposed changes do not affect the operation of the system or the accident analyses and are consistent with the NRC approved changes to the surveillances identified for the ISTS of NUREG-1431 identified in TSTF-110, Revision 2. These changes do not involve a change to plant equipment and do not affect the performance of plant equipment used to mitigate an accident. Although the deletion of these monitor requirements from the TS results in elimination of the reduced surveillance interval when the alarm is inoperable (for those requirements not being relocated to the LRM) the change in frequency is not significant considering the indications available to the operator and the relatively slow changes in the parameters being monitored during steady state operation. Therefore, based on the above, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

   Unit 1 SRs 4.1.3.5 and 4.1.3.6 have been additionally modified by incorporating the Unit 2 wording which more closely provides a surveillance appropriate for the LCO. The LCO requires the shutdown rods/control banks to be within the insertion limits and the revised SR requires a determination that each shutdown rod/control bank is within the insertion limits on a 12-hour frequency. Therefore, the revised SRs are consistent with the LCO requirements and more clearly provide verification that the LCO is met. This change does not affect the operation of the rod position indication system or any other system and is consistent with the Unit 2 and ISTS wording. This change will not affect the ability of any system to perform its design function; therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   Changing the surveillance frequency from 24 to 12 hours is more conservative and assures the affected parameters are adequately monitored. In addition, the change removes monitors from the TSs and provides consistency between the SRs, the units and the ISTS. Changing the surveillance frequency, correcting the Unit 1 SRs and removing reference to the identified monitors from the TS will not cause a significant reduction in system reliability nor affect the ability of any system to perform its design function. There are no hardware changes associated with this license amendment nor are there any changes in the method by which any safety-related plant system performs its safety function. No new accident scenarios, transient precursors, failure mechanisms or limiting single failures are introduced as a result of these changes. These changes do not introduce any adverse effects or challenges to any safety-related systems. No change is required to any system configurations, plant equipment or analyses. Therefore, these changes will not create the possibility of any new or different kind of accident from any accident previously evaluated.

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

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   The proposed changes do not affect the acceptance criteria for any analyzed event nor impact any plant safety analyses since the assumptions used will remain unchanged. The safety limits assumed in the accident analyses and the design function of the equipment required to mitigate the consequences of any postulated accidents will not be changed since the proposed changes do not affect the accident analyses assumptions or equipment required to mitigate design basis accidents described in the UFSAR. Although the deletion of these monitor requirements from the TSs results in elimination of the reduced surveillance interval when the alarm is inoperable (for those requirements not being relocated to the LRM) the effect is not significant considering the indications available to the operator and the relatively slow changes in the parameters being monitored during steady state operation. The TSs continue to assure the applicable operating parameters are maintained within the required limits. Based on engineering judgement, incorporating these changes will not involve a significant reduction in the margin of safety.

   The margin of safety depends upon maintenance of specific operating parameters within design limits. The TSs continue to require that these limits be maintained and provide appropriate remedial actions if a limit is exceeded. The maintenance of these limits continues to be assured through performance of the normal surveillance at the proposed frequency and the requirements for increased monitoring that are relocated to the LRM. Additional assurance that the required parameters are adequately monitored is provided through other information readily available (i.e., rod position indication system, rod bottom alarm and excore neutron detectors) that allows the operator to detect a parameter that is beginning to deviate from its expected limits and through the proposed changes which reduce the normal surveillance interval from 24 hours to 12 hours to assure the affected parameters are adequately monitored. Although these monitors are being removed from the TSs, they will continue to be maintained as described in the UFSAR (subject to revisions via the 10 CFR 50.59 process). Therefore, the plant will be maintained within the analyzed limits and the proposed changes will not involve a significant reduction in a margin of safety.

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

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

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

   NRC Project Director: S. Singh Bajwa.

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

   Date of amendment request: November 24, 1998.

   Description of amendment request: The proposed changes implement the consolidated Entergy Operations Quality Assurance Plan Manual approved by the NRC on November 6, 1998. The proposed changes also clarify the responsibilities of the shift technical advisor position on shift, simplify the contents of the monthly operating report description in accordance with Generic Letter (GL) 97-02, complete the relocation of fire protection requirements from the TS to the fire protection program in accordance with GL 88-12, and replace position titles with descriptions of functional responsibility in accordance with GL 88-06.

   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:

   Criterion 1--Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

   The proposed changes only affect the administrative controls contained in Section 6.0 of the Arkansas Nuclear One--Unit 1 (ANO-1) and Unit 2 (ANO-2) Technical Specifications (TSs). The proposed changes either add additional administrative controls, reduce regulatory duplication of requirements consistent with NUREG-1430 ``Standard Technical Specifications--Babcock and Wilcox Plants'' dated April 1995, and NUREG-1432 ``Standard Technical Specifications--Combustion Engineering Plants'' dated April 1995, or revise or relocate administrative controls in accordance with NRC guidance. The proposed changes do not affect the operation of any structure, system, or component or the assumptions of any accident analysis. The details relocated from the ANO-1 and ANO-2 TSs, and changes to these details, are controlled under the ANO 10 CFR 50.59 or 10 CFR 50.54 processes as appropriate.

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

   Criterion 2--Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

   The proposed changes to the ANO-1 and ANO-2 Section 6.0 administrative controls do not involve a change in the plant design or affect the configuration or operation of any structure, system, or component.

   Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

   Criterion 3--Does Not Involve a Significant Reduction in the Margin of Safety.

   The proposed changes to the ANO-1 and ANO-2 TSs affect only administrative requirements and do not involve changes to safety limits, limiting conditions for operation, or surveillance requirements on equipment required to operate the station.

   Therefore, this change does not involve a significant reduction in the margin of safety.

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

   Local Public Document Room location: Tomlinson Library, Arkansas Tech University, Russellville, AR 72801.

   Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW, Washington, DC 20005-3502.

   NRC Project Director: John N. Hannon.

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

   Date of amendment request: November 30, 1998.

   Description of amendment request: The proposed amendment would change the CR-3 Improved Technical Specifications (ITS) Section 3.9.3, Containment Penetrations. The proposed changes recognize the use of an outage equipment hatch (OEH) during refueling operations. The proposed changes would also allow both doors in the personnel air locks, and the single door in the OEH, to be open during core alterations or movement of irradiated fuel assemblies within containment provided certain specified conditions are met.

   The licensee stated that the ability to open these doors under administrative controls would assist in the maintenance of cleanliness and housekeeping, and would provide a safer work environment inside containment. In addition, the licensee stated that evacuation of personnel could be quickly achieved in the unlikely event of a fuel handling accident or other radiological event inside containment, reducing the potential for exposures.

   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.

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

   The proposed change would allow both doors in the personnel air locks and the door in the outage equipment hatch (OEH) to remain open during core alterations or the movement of irradiated fuel inside containment. These doors are normally closed during this period in order to prevent the escape of radioactive materials in case of a fuel handling accident.

   Operations involving the personnel air locks during refueling operations cannot be an initiator of a fuel handling accident or other radiological event inside containment. Similarly, operations involving the OEH during refueling operations cannot be an initiator of a fuel handling accident or other radiological event inside containment. The personnel air locks and the OEH are remotely located to the fuel handling equipment and cannot affect the function of this equipment. The personnel air locks and the OEH are not in the immediate vicinity of the reactor vessel and the contained irradiated fuel, or any of the paths used for movement of irradiated fuel. Additionally, allowing both doors in the personnel air locks and the door in the OEH to be open during core alterations or the movement of irradiated fuel inside containment cannot create the possibility of a fuel handling accident or other radiological event inside containment. Therefore, the probability of occurrence of any accident previously evaluated is unaffected.

   The approved fuel handling accident analysis does not take credit for containment closure. This analysis results in a maximum calculated offsite dose well within the limits of 10 CFR 100, and the existing analysis as presented in the CR-3 Final Safety Analysis Report does not require revision as a result of this proposed change. By providing a designated individual readily available to close at least one door in the personnel air locks and the door in the OEH, containment closure is assured following any required evacuation of containment terminating any release of radioactive materials outside of the containment. Therefore, the consequences of accidents will not be greater than that previously evaluated.

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

   The operations involving the personnel air locks and the OEH cannot be an initiator of any type of accident during refueling operations. The personnel air locks and the OEH are passive structural features designed to retain structural integrity under the expected environmental conditions when installed. Operation of the personnel air lock doors and the door in the OEH does not affect any safety-related component or structure. Additionally, allowing both doors in the personnel air locks and the door in the OEH to be open during core alterations or the movement of irradiated fuel inside containment cannot initiate any type of accident. Therefore, the possibility of a new or different kind of accident occurring as a result of this change is not created.

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

   The margin of safety as defined by 10 CFR 100 has not been reduced. The existing approved fuel handling accident analysis does not credit containment closure, and remains bounding with both doors in the personnel air locks and the door in the OEH open. Closing at least one door in the personnel air locks and the door in the OEH after evacuation of containment further reduces the offsite doses in case of a fuel handling accident, and provides additional margin to the calculated offsite doses. Therefore, the existing margin of safety will not be reduced.

   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: Coastal Region Library, 8619 W. Crystal Street, Crystal River, Florida 34428.

   Attorney for licensee: R. Alexander Glenn, General Counsel, Florida Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida 33733-4042.

   NRC Project Director: Cecil O. Thomas

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

   Date of amendment request: November 4, 1998.

   Description of amendment request: The proposed change would revise Technical Specifications Surveillance Requirement 4.5.2b.1 to delete the prescribed method of venting the Emergency Core Cooling System (ECCS) which would allow alternate methods to verify that the ECCS piping is full of water. In addition, the associated Bases would be expanded to reflect the intent of the surveillance requirement.

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

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

   The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, configuration of the facility or the manner in which the plant is operated. The proposed change does not alter or prevent the ability of structures, systems and components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR).

   Removal of the prescriptive requirements will not subject the ECCS system to conditions adverse to nuclear safety. The proposed change does not affect the source term, containment isolation or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated in the Seabrook Station UFSAR. The use of proven alternative techniques to verify that the ECCS piping is full of water will continue to ensure that the ECCS system is capable of performing its intended designed safety function. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The proposed change does not alter the design assumptions, conditions, configuration of the facility or the manner in which the plant is operated and maintained in a state of readiness. Existing system and component redundancy is not being changed by the proposed change. The proposed change has no adverse affect on component or system interactions. The use of proven alternative techniques to verify that the ECCS piping is full of water will continue to ensure that the ECCS system is capable of performing its intended designed safety function. Therefore, since there are no changes to the design assumptions, conditions, configuration of the facility, or the manner in which the plant is operated and maintained in a state of readiness, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

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

   The proposed change does not adversely affect equipment design or operation and there are no changes being made to the Technical Specification required safety limits or safety system settings that would adversely affect plant safety. The proposed change does not change the intent of the surveillance requirement of ensuring that the system will perform properly, injecting its full capacity into the RCS upon demand without subjecting the system to hydraulic transients, pump cavitation, and pumping of non-condensable gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following a safety injection (SI) signal or during shutdown cooling.

   Thus, it is concluded that the ECCS will continue to be available upon demand to mitigate the consequences of an accident and, therefore, there is no 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.

   [[Page 4158]]

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

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

   NRC Project Director: William M. Dean.

   Northeast Nuclear Energy Company (NNECO), et al., Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New London County, Connecticut

   Date of amendment request: December 22, 1998.

   Description of amendment request: The proposed amendment would replace specific titles in Section 6.0 of the Technical Specifications of all three Millstone units with generic titles.

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

   In accordance with 10 CFR 50.92, NNECO has reviewed the attached proposed changes and ha[s] concluded that they do not involve a Significant Hazard Consideration (SHC). The basis for this conclusion is that the three criterion of 10 CFR 50.92 are not compromised. The proposed change is not a[n] SHC because the proposed change will not:

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

   No design basis accidents are affected by these proposed changes. The proposed changes are administrative in nature and are being proposed to eliminate the need for a Technical Specification change each time there is a change in the organization.

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

   There are no changes in the way the plant is operated due to these administrative changes. The potential for an unanalyzed accident is not created. There is no impact on plant response, and no new failure modes are introduced. The proposed administrative and editorial changes have no impact on safety limits or design basis accidents, and have no potential to create a new or unanalyzed event.

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

   These changes do not directly affect any protective boundaries nor do they impact the safety limits for the protective boundaries. These proposed changes are administrative and editorial in nature. Therefore there is no reduction in the margin of safety.

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

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

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

   NRC Project Director: William M. Dean.

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

   Date of amendment request: November 13, 1998.

   Description of amendment request: NNECO has determined that the increase in radiological consequences, due to changes in the assumptions used in the updated dose consequence analysis of the Steam Generator Tube Rupture (SGTR) event in the Millstone Unit No. 2 Final Safety Analysis Report (FSAR), involves an unreviewed safety question (USQ). The changes include a change in High Pressure Safety Injection (HPSI) pump runout flowrate, a change in Auxiliary Feedwater Pump (AFW) flowrate, a change in the iodine partition factor for the air ejector, inclusion of the potential of flashing of the primary-to-secondary leakage, and a change in the atmospheric release point assumed following actuation of the Enclosure Building Filtration Actuation Signal (EBFAS). Therefore, per 10CFR50.59(c), NNECO requested that the NRC review and approve the changes to the FSAR through an amendment to Operating License DPR-65, pursuant to 10CFR50.90.

   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 10CFR50.92, NNECO has reviewed the proposed changes and has concluded that they do not involve a Significant Hazards Consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve an SHC because the changes would not:

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

   The FSAR changes reflect changes in the updated SGTR analysis. The analysis was updated because of changes in the assumptions used in the dose consequence analysis of the SGTR event in Millstone Unit No. 2 FSAR. These changes include a change in the iodine partition factor for the air ejector, inclusion of the potential of flashing of the primary-to-secondary leakage, and a change in the atmospheric release point assumed following actuation of the EBFAS. In addition, the operator actions associated with Reactor Coolant System (RCS) cooldown that are specified in the Emergency Operating Procedures have been incorporated, mass releases assuming an RCS cooldown to Shutdown Cooling Entry conditions have been used in the dose consequence analysis, thyroid doses were calculated using ICRP-30 dose conversion factors, Iodine releases account for potential flashing of the primary-to-secondary leakage, and the Reactor Coolant pumps are assumed to be tripped following actuation of a safety injection actuation signal. The revised HPSI flowrate is higher than that used in the previous analysis. Higher HPSI flowrates would increase the primary-to-secondary break flow and, thereby, increase the dose consequences. A more conservative iodine partition factor for the air ejector has been used along with more limiting atmospheric dispersion coefficients as a result of manual realignment of the air ejector discharge path to the atmosphere. These changes in radiological assumptions are the major reason for the increase in calculated dose. The revised AFW flowrate is lower than that used in the previous analysis. Lower AFW flowrate would tend to increase the steaming required and, thereby, increase the dose consequences. The probability that an accident could occur due to these changes is not increased since changing the analysis and its description can not cause a steam generator tube rupture. Therefore, these changes will not significantly increase the probability of an accident previously evaluated.

   The dose consequences for the updated SGTR analysis are higher than the dose consequences for the previous analysis. However, the dose consequences are within the acceptance criteria of SRP

   [Standard Review Plan] 15.6.3 and GDC [General Design Criterion] 19. Therefore, these changes will not significantly increase the consequences of an accident previously evaluated.

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

   The FSAR changes reflect changes in the updated SGTR analysis. The updated analysis does not introduce any new or unanalyzed failure modes of equipment or systems, and does not change the configuration of the plant. While the updated analysis incorporates operator actions that are in accordance with the Emergency Operating Procedures, it does not alter the way any structure, system, or component functions, and does not alter the manner in which the plant is operated. Therefore, there are no new or different types of failures of systems or equipment important to safety

   [[Page 4159]]

   which could cause a new or different type of accident from any accident previously evaluated.

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

   The FSAR changes reflect changes in the updated SGTR analysis. The updated analysis shows that the dose consequence acceptance criteria are met. The updated analysis incorporates operator actions that are in accordance with the Emergency Operating Procedures, and credits equipment consistent with its capabilities. Therefore, the updated analysis does not reduce the margin of safety. The FSAR changes do not alter the acceptance limits of the safety parameters of the accident analyses stated in the FSAR. Therefore, these changes do not significantly reduce the margin of safety.

   The NRC has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (March 6, 1986, 51 FR 7751) of amendments that are considered not likely to involve an SHC. The changes proposed herein are covered by example (vi) in that the consequences for the updated SGTR analysis are higher than dose consequences for the previous analysis. However, the dose consequences are within the acceptance criteria of SRP 15.6.3 and GDC 19.

   As described above, this License Amendment Request does not involve a significant increase in the probability of an accident previously evaluated, does not involve a significant increase in the consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident from any accident previously evaluated, and does not result in a significant reduction in a margin of safety. Therefore, NNECO has concluded that the proposed changes do 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 Project Director: William M. Dean.

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

   Date of amendment request: November 20, 1998.

   Description of amendment request: This amendment request updates the Emergency Diesel Generator (EDG) day tank volume Surveillance Requirement (SR) 3.8.1.4.

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

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

   This proposal does not involve an increase in the probability or consequences of an accident previously evaluated. The proposed amendment changes EDG day tank volume requirements to reflect the

   [Susquehanna Steam Electric Station] SSES design.

   The safety function of the EDG day tanks is to supply the EDG's with enough fuel to ensure the availability of necessary power to

   [engineered safety feature] ESF systems so that fuel, reactor coolant and containment system design limits are not exceeded. The proposed change increases the minimum diesel fuel oil day tank volume for Unit 1 and Unit 2 SR 3.8.1.4 from 325 gallons to 420 gallons for EDG A-D and 425 gallons for EDG E.

   This volume corresponds to the tank volume at which automatic refill occurs. This volume provides for 55 minutes of EDG A-D and 62 minutes for EDG E operation at continuous rated load conditions.

   Currently, the bases for SR 3.8.1.4 identifies that ``administrative controls ensure a useable volume of the fuel oil in the day tank adequate for approximately 60 minutes of DG operation plus 10% at the continuous rated load.'' These administrative controls ensure compliance with the Regulatory Guide 1.137 requirements. Regulatory Guide 1.137 revision 1 endorses American National Standards Institute (ANSI) N195-1976. The ANSI N195-1976 requires each diesel to be equipped with a day tank whose capacity is sufficient to maintain at least 60 minutes of operation. This capacity is to be based on the fuel consumption at a load of 100% of the continuous rating of the diesel plus a minimum margin of 10%.

   These administrative controls on day tank level ensure that the required initial fuel oil supply is available to meet the intent of the Standard as it applies to the Technical Specification surveillance. This Technical Specification change eliminates these unnecessary controls needed to conform to the ANSI standard.

   An assessment of the proposed change based on the guidance provided in Regulatory Guide 1.174, July 1998, ``An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis'' concludes that the increase in risk is insignificant. It is therefore concluded that the proposed changes to SSES Unit 1 and Unit 2 Technical Specification SR 3.8.1.4 day tank volume requirements ensures the volume is adequate to support the EDG's post accident design basis safety function to ensure the availability of necessary power to ESF systems so that fuel, reactor coolant system, and containment design limits are not exceeded.

   Based upon the above, PP&L concludes that the proposed action does not involve an increase in the probability or consequences of an accident previously evaluated.

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

   This proposal does not create the probability of a new or different type of accident from any accident previously evaluated. The change to the day tank required minimum volume does not change any plant systems, structures, or components, nor does the change affect any existing or create any new or different kind of accident.

   An assessment of the proposed change based on the guidance provided in Regulatory Guide 1.174, July 1998, ``An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis'' concludes that the increase in risk is insignificant. Based on this, it is concluded that the proposed changes to SSES Unit 1 and Unit 2 Technical Specification SR 3.8.1.4 day tank volume requirements ensures the volume is adequate to support the EDG's post accident design basis safety function to ensure the availability of necessary power to ESF systems so that fuel, reactor coolant system, and containment design limits are not exceeded.

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

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

   An assessment of the proposed change based on the guidance provided in Regulatory Guide 1.174, July 1998, ``An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis'' concludes that the increase in risk is insignificant.

   It is concluded that the proposed changes to SSES Unit 1 and Unit 2 Technical Specification SR 3.8.1.4 day tank volume requirements ensures the volume is adequate to support the EDG's post accident design basis safety function to ensure the availability of necessary power to ESF systems so that fuel, reactor coolant system, and containment design limits are not exceeded.

   Based on this, the proposed changes do not involve a reduction in the margin of safety.

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

   [[Page 4160]]

   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: S. Singh Bajwa.

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

   Date of amendment request: November 23, 1998.

   Description of amendment request: These amendments would modify the Susquehanna Steam Electric Station, Units 1 and 2, Technical Specifications (TS) limiting condition for operation (LCO) 3.8.3 and surveillance requirement (SR) 3.8.3.1 to increase the minimum fuel oil storage tank (FOST) volume ranges. The Bases would be modified to reflect that the proposed volumes equal the 7-day fuel oil consumption at the continuous emergency diesel generator (EDG) ratings, which are greater than design basis analysis (DBA) loads, plus the unusable volume in the storage tanks.

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

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

   This proposal does not involve an increase in the probability or consequences of an accident previously evaluated. The proposed amendment increases FOST volume requirements so to increase the margin of safety thus providing further assurance that the EDG FOST volume is adequate to support the EDG's post accident design basis safety function.

   The safety function of the EDG FOST is to supply the emergency diesel generators with enough fuel to ensure the availability of necessary power to ESF systems so that fuel, reactor coolant and containment system design limits are not exceeded. The current Technical specification FOST specified volume is based on the EDG post DBA load profile. The proposed FOST volume is based on EDG continuos [sic] [continuous] rated load rating which is greater than the post DBA load profile providing margin and further assurance that the EDG FOST will support the EDG safety function. The proposed required FOST volumes are calculated in accordance with ANSI N195- 1976.

   Based upon the above, PP&L concludes that the proposed action does not involve an increase in the probability or consequences of an accident previously evaluated.

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

   This proposal does not create the probability of a new or different type of accident from any accident previously evaluated. The FOST required minimum values do not change any plant systems, structures, or components, nor do they change any existing or create any new or different kind of accident. The proposed amendment changes FOST volume requirements so to increase the margin of safety thus providing further assurance that the EDG FOST volume is adequate to support the EDG's post accident design basis safety function. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The proposed change increases the margin of safety since the proposed FOST values are based on the EDG continuos [sic]

   [continuous] rated load ratings which bound the post DBA load profile.

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

   Local Public Document Room location: Osterhout Free Library, Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

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

   NRC Project Director: S. Singh Bajwa.

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

   Date of amendment request: November 6, 1998.

   Description of amendment request: The proposed amendments would revise the Technical Specifications for the Nuclear Instrumentation System [NIS] Power Range daily surveillance requirement.

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

   The proposed surveillance change does not significantly increase the probability or consequences of an accident previously evaluated in the FSAR [Final Safety Analysis Report]. This modification does not directly initiate an accident. The consequences of accidents previously evaluated in the FSAR are not adversely affected by this proposed change because the change to the NIS Power Range channel adjustment requirement ensures the conservative response of the channel even at part power levels.

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

   The proposed surveillance change does not create the possibility of a new or different kind of accident than any accident already evaluated in the FSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed Technical Specifications change does not challenge the performance or integrity of any safety- related systems. Therefore, the possibility of a new or different kind of accident is not created.

3. Does the proposed surveillance change involve a significant reduction in a margin of safety?

   The proposed surveillance change does not involve a significant reduction in a margin of safety. The proposed change does require a revision to the criterion for implementation of Power Range channel adjustment based on secondary power calorimetric calculation; however, the change does not eliminate any RTS [Reactor Trip Setpoint] surveillances or alter the frequency of surveillances required by the Technical Specifications. The revision to the criterion for implementation of the daily surveillance will have a conservative effect on the performance of the NIS Power Range channel, particularly at part power after normalization at 100% RTP

   [Rated Thermal Power] conditions. The nominal trip setpoints specified by the Technical Specifications and the safety analysis limits assumed in the transient and accident analysis are unchanged. The margin of safety associated with the acceptance criteria for any accident is unchanged. Therefore, the proposed change will not significantly reduce the margin of safety as defined in the Technical Specifications.

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

   Local Public Document Room location: Houston-Love Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

   [[Page 4161]]

   NRC Project Director: Herbert N. Berkow.

   Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia.

   Date of amendment request: December 4, 1998.

   Description of amendment request: The proposed amendments would make two changes to the Technical Specifications (TSs). Change 1 would delete the footnote in Hatch Unit 1 TS Section 2.1.1.2 that ties the Safety Limit Minimum Critical Power Ratio to Cycle 18. Change 2 would delete TS Section 5.6.5.b.2 for Units 1 and 2, and incorporate TS Section 5.6.5.b.2 into TS Section 5.6.5.b.1 for both units.

   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:

   Basis for Proposed Change 1

   The change does not involve a significant hazards consideration for the following reasons:

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

   The footnote in Section 2.1.1.2 of the Hatch-1 Technical Specifications restricts the applicability of the Safety Limit for MCPR [minimum critical power ratio] (SLMCPR) [safety limit minimum critical power ratio] to Cycle 18 only. By applying the same NRC- approved methods used to calculate the Cycle 18 SLMCPR it has been determined that the current value is bounding for Cycle 19 as well. However, because of the footnote, it [cannot] be applied to Cycle 19 without a Technical Specifications amendment. In order to eliminate future Technical Specifications revisions that do not change the SLMCPRs values, SNC [Southern Nuclear Operating Company, Inc.] proposes to delete the footnote which ties those values to a specific operating cycle. Removing the footnote does not change the method of calculating SLMCPR for other cycles, nor does it eliminate the requirement to revise the Technical Specifications if a different value is used for future cycles. Deletion of the cycle- specific footnote does not change the operation of any plant structure, system or component; therefore, it has no affect on the probability or consequences of an accident previously evaluated.

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

   Deleting the cycle-specific footnote in Section 2.1.1.2 of the Technical Specifications does not result in any new methods of operating the facility and does not involve any facility modifications. No new initiating events or transients result from this change.

   Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

   The purpose of the SLMCPR in the Technical Specifications is to ensure at least 99.9% of the fuel pins in the core are expected to avoid transition boiling during the worst anticipated operational occurrence (AOO) throughout an operating cycle. The footnote in Section 2.1.1.2 of the Hatch-1 Technical Specifications is intended to ensure the correct SLMCPR is used each cycle. Prior to the Spring of 1996, the Safety Limits had been calculated for each fuel type, independently of operating cycle. As long as the limiting fuel type in the core did not change from cycle to cycle, the Safety Limit did not change. It was discovered in 1996, however, that generic SLMCPRs based on fuel type alone may not be bounding for all cycles for all reactors. In response to this discovery GE committed to evaluating SLMCPRs based on cycle-unique information as a more accurate method of ensuring 99.9% of the fuel pins in the core are expected to avoid transition boiling during AOOs. The new methodology, which is now applied each cycle, is based on NRC-approved methods and incorporates implementing procedures that model cycle-specific parameters. This methodology was used to calculate the Cycle 18 value that is currently in the Technical Specifications. The same procedure was also employed to determine that the Hatch-1 Cycle 19 SLMCPR and it was determined the Cycle 19 value is bounded by the Cycle 18 value. Thus, except for the footnote in Section 2.1.1.2, there is no need to revise the Hatch-1 Technical Specifications in order to ensure the correct SLMCPR is implemented for Cycle 19. As a way of avoiding similar changes in the future, SNC proposes that the footnote be deleted. Since NRC-approved methodology will still be used to determine the cycle-specific SLMCPRs to ensure that [ ] 99.9% of the fuel rods are expected to avoid transition boiling during AOOs, there will be no reduction of margin of safety as a result of this change.

   Basis for Proposed Change 2

   The change does not involve a significant hazards consideration for the following reasons:

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

   Section 5.6.5.b.2) no longer describes NRC-approved methods for analyzing fuel in the Unit 1 and Unit 2 reactors because the ANF

   [advanced nuclear fuel] LUAs [lead use assemblies] have been permanently discharged. Deleting Section 5.6.5.b.2) from the Administrative Controls portion of the Technical Specifications does not change the operation of any structure, system, or component in the facility. Therefore, this amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   Deleting Section 5.6.5.b.2), which describes the use of ANF methods for analyzing LUAs, from the Technical Specifications does not result in any new methods of operating the facility and does not involve any facility modifications. No new initiating events or transients result from this change. Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

   ANF LUAs are no longer used as fuel in the Plant Hatch reactors, therefore, ANF NRC-approved methods described in Technical Specifications Section 5.6.5.b.2) are not used to determine power distribution limits which appear in the COLR [Core Operating Limit Report]. GE's [General Electric's] reload licensing methodology described in Section 5.6.5.b.1) will be incorporated into Section 5.6.5.b. and will continue to be used to analyze the GE fuel in both units. Therefore, this change does not involve a significant reduction in the margin of safety.

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

   Local Public Document Room location: Appling County Public Library, 301 City Hall Drive, Baxley, Georgia.

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

   NRC Project Director: Herbert N. Berkow.

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

   Date of amendment request: November 4, 1998.

   Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) Sections 4.6.A.1.b and Basis 3.16 for Units 1 and 2 to revise the start/load time testing and ratings for emergency diesel generators (EDGs). The changes will bring the TS into conformance with the Updated Final Safety Analysis Report.

   [[Page 4162]]

   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:

   Criterion 1--Operation of the Surry Units 1 and 2 in accordance with the proposed Technical Specification change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

   The currently specified ``less than 30 seconds'' time to be replaced has no specific safety significance or design basis regarding EDG starting. The proposed time change to ``less than or equal to 10 seconds'' is more conservative and in agreement with current accident analysis and surveillance testing. These changes do not, in any way, affect the as-built conditions of the plant and do not affect the initiators of analyzed events or the assumed mitigation of accident or transient events. Analyzed events are initiated by the failure of plant structures, systems, or components. The proposed changes do not impact the condition or performance of these structures, systems or components. Consequences of analyzed events are the result of the plant being operated within assumed parameters at the onset of any event, and the successful functioning of at least one train or division of the equipment credited with mitigating the event. There is no impact on the capability of the credited equipment to perform, nor is there any change in the likelihood that credited equipment will fail to perform. As a result, there is no significant increase in the probability or consequences of any accident previously evaluated and Criterion 1 is, thereby, satisfied.

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

   The proposed change does not involve a physical alteration of the plant, or a change in the methods used to operate the plant or to respond to plant transients. No new or different equipment is being installed and no installed equipment is being removed or operated in a different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints, which initiate protective or mitigative actions. Consequently, no new failure modes are introduced and the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated and Criterion 2 is, thereby satisfied.

   Criterion 3--The proposed Technical Specifications change does not involve a significant reduction in a margin of safety.

   Margin of safety is established through the design of the plant structures, systems and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The replacement of the ``less than 30 seconds'' requirement for loading the EDGs with the more stringent ``less than or equal to 10 seconds'' requirement makes no change to the condition or performance of equipment or system used in accident mitigation or assumed for any accident analysis that could reduce a margin of safety as described in the basis for any TS. Therefore, the proposed changes do not involve a significant reduction in any margin of safety described in the bases for the Technical Specifications and Criterion 3 is, thereby, satisfied.

   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: Swem Library, College of William and Mary, Williamsburg, Virginia 23185.

   Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 23219.

   NRC Project Director: Herbert N. Berkow.

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

   Date of amendment request: September 28, 1998 (TSCR 208).

   Description of amendment request: The proposed amendments will clarify the notation definition of ``R'' in the Technical Specifications (TS) and add a new frequency of ``A.'' The revision of ``R'' would specify the refueling frequency as 18 months and ``A'' would be defined as an annual or 12-month frequency.

   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 Point Beach Nuclear Plant [PBNP] in accordance with the proposed amendments will not result in a significant increase in the probability or consequences of an accident previously evaluated.

   These changes do not involve a significant increase in the probability of an accident previously evaluated because no such accidents are affected by the proposed revisions to clarify that the provisions of TS 15.4.0.2 apply to notation ``R'' in TS Table 15.4.1-1. The proposed TS changes do not introduce any new accident initiators since no accidents previously evaluated have as their initiators anything related to the change in the frequency of surveillance testing.

   The increased time potential between surveillance frequencies does not significantly increase the probability [of] failure of the instrumentation contained in TS Table 15.4.1-1. As noted above, instrument drift studies concluded that the magnitude of the instrument drift (for instrumentation affected by drift) that could occur over a 22.5-month interval was bounded by the uncertainty allowances used in determining safety system setpoints, and the review of historical calibration data concluded that the as-found and as-left data has not exceeded acceptable limits for the calibration intervals reviewed, except on rare occasions.

   In addition, initiating conditions and assumptions are unchanged and remain as previously analyzed for accidents in the PBNP Final Safety Analysis Report. The proposed TS changes do not involve any physical changes to systems or components, nor do they alter the typical manner in which the systems or components are operated. Therefore, these changes do not increase the probability of previously evaluated accidents.

   These changes do not involve a significant increase in the consequences of an accident previously evaluated because the source term, containment isolation or radiological releases are not being changed by these proposed revisions. Existing system and component redundancy and operation is not being changed by these proposed changes. The assumptions used in evaluating the radiological consequences in the PBNP Final Safety Analysis Report are not invalidated; therefore, these changes do not affect the consequences of previously evaluated accidents.

2. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated.

   These changes do not introduce nor increase the number of failure mechanisms of a new or different type than those previously evaluated since there are no physical changes being made to the facility. The surveillance test requirements and the way they are performed will remain unchanged. The design and design basis of the facility remain unchanged. The plant safety analyses remain unchanged. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not introduced.

3. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendments does not involve a significant reduction in a margin of safety.

   The proposed changes do not involve a significant reduction in the margin of safety because existing component redundancy is not being changed by these proposed changes. There are no new or significant changes to the initial conditions contributing to accident severity or consequences, and safety margins established through the design and facility license including the Technical Specifications remain unchanged. Therefore, there are no significant reductions in a margin of safety introduced by [these] proposed amendment[s].

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

   [[Page 4163]]

   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: The Lester Public Library, 1001 Adams Street, Two Rivers, Wisconsin 54241.

   Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

   NRC Project Director: Cynthia A. Carpenter.

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

   Date of amendment request: October 5, 1998 (TSCR 200).

   Description of amendment request: The proposed change modifies Technical Specifications Section 15.4.1, ``Operational Safety Review,'' by removing the requirement to check environmental monitors on a monthly basis.

   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 Point Beach Nuclear Plant [PBNP] in accordance with the proposed amendments does not result in a significant increase in the probability or consequences of an accident previously evaluated.

   The proposed change eliminates a surveillance requirement for environmental monitors. The environmental monitors referred to by this surveillance were eliminated from the Radiological Environmental Monitoring Program and from the Technical Specifications by previous amendments. Therefore, this change is administrative in nature in that it corrects a previous administrative oversight. The requirement is not related to any accident initiator or accident mitigation structures, systems or components for any previously evaluated accident. Therefore, no increase in the probability or consequences of a previously evaluated accident can result.

2. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendment[s] does not create a new or different kind of accident from any accident previously evaluated.

   The amendments remove a surveillance requirement from the Technical Specifications related to environmental monitors. The environmental monitors were removed from the environmental monitoring program by previously approved amendments. The surveillance requirement is not related to an existing design feature of PBNP. Therefore, elimination of the surveillance requirement cannot create a new or different kind of accident from any accident previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with the proposed amendment[s] does not result in a significant reduction in a margin of safety.

   Margins of safety are defined by the safety limits and design limits for PBNP. The surveillance is not related to, nor does it affect, these limits. Monitoring of the environment continues under an approved Radiological Environmental Monitoring Program which ensures that any changes in radiation levels in the environs is detected, thus ensuring the impact of PBNP operation on the environment is minimized. Therefore, the proposed change cannot 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: The Lester Public Library, 1001 Adams Street, Two Rivers, Wisconsin 54241.

   Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

   NRC Project Director: Cynthia A. Carpenter.

   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.

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

   Date of application for amendment: December 22 and revision dated December 23, 1998.

   Brief description of amendment: In order to prevent a potential shutdown due to sporadic grounds encountered on an annunciator circuit used to confirm operability of an ice condenser inlet door position monitoring system, the proposed amendment would provide a temporary, optional method of satisfying the requirements for the channel check until the next operating Mode, planned in late February 1999, for the next refueling outage. Date of publication of individual notice in the Federal Register: December 31, 1998 (63 FR 72339).

   Expiration date of individual notice: February 1, 1999.

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

   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

   [[Page 4164]]

   local public document rooms for the particular facilities involved.

   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: November 11, 1998.

   Brief description of amendments: The amendments revise Technical Specification Surveillance Requirements (SRs) 3.6.11.6 AND 3.6.11.7, regarding the Containment Pressure Control System (CPCS), of the units' joint Technical Specifications. The revision brings the SRs into conformity with the current design of the CPCS.

   Date of issuance: January 14, 1999.

   Effective date: As of the date of issuance to be implemented concurrently with implementation of Amendment Nos. 173 (Unit 1) and 165 (Unit 2).

   Amendment Nos.: 174--Unit 1; Unit 2--166.

   Facility Operating License Nos. NPF-35 and NPF-52: Amendments revise the Technical Specifications.

   Date of initial notice in Federal Register: December 2, 1998 (63 FR 66591). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 14, 1999.

   No significant hazards consideration comments received: No.

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

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

   Date of amendment request: September 17, 1998.

   Brief description of amendment: The amendment incorporates the use of a range rather then a specific setpoint for the automatic removal of the operating bypasses for the core power calculator (CPC) generated trips and the high logarithmic power level trip to accommodate the design of the plant protection system (PPS) which uses a single bistable to control both of these functions.

   Date of issuance: December 31, 1998.

   Effective date: December 31, 1998.

   Amendment No.: 196.

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

   Date of initial notice in Federal Register: October 21, 1998 (63 FR 56247).

   The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 31, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Tomlinson Library, Arkansas Tech University, Russellville, AR 72801.

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

   Date of application for amendment: June 29, 1998.

   Brief description of amendment: The amendment modifies the TS surveillance requirements for SR 4.8.2.3.b.2, SR 4.8.2.3.c.4 and the Bases for TS 3.8.2.3 Action b. The licensee is planning to modify the 120 volt vital alternating current (ac) electrical distribution system by installing new inverters during the 2R13 refueling outage. Normally, the present inverters for ANO-2 are ac powered and automatically shift to direct current (dc) power on a loss of the ac source. The new inverters will be powered from the 125 dc system at all times.

   Date of issuance: January 13, 1999.

   Effective date: January 13, 1999, with implementation following completion of the required modifications but prior to restart from the 2R13 outage.

   Amendment No.: 198.

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

   Date of initial notice in Federal Register: October 21, 1998 (63 FR 56244).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 13, 1999.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Tomlinson Library, Arkansas Tech University, Russellville, AR 72801.

   Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie Plant, Unit No. 1, St. Lucie County, Florida

   Date of application for amendment: October 29, 1998.

   Brief description of amendment: The amendment revised the terminology used in the St. Lucie Plant Technical Specifications (TS) relative to the implementation and automatic removal of certain protection system trip bypasses to ensure that the meaning of explicit terms used in the TS are consistent with the intent of the stated requirements.

   Date of Issuance: January 5, 1999.

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

   Amendment No.: 159.

   Facility Operating License No. DPR-67: Amendment revised the TS.

   Date of initial notice in Federal Register: December 2, 1998 (63 FR 66594) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 5, 1999.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

   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: August 4, 1998.

   Brief description of amendment: The amendment changes the Technical Specifications (TSs) relating to the condensate storage tank (CST) relating to the required minimum water volume and also adds a new TS which establishes requirements for the atmospheric steam dump valves (ASDVs) to assure their operability. The applicable TS Bases for the CST is updated to reflect the proposed changes and a new TS Bases section is added to discuss the new TS for the ASDVs.

   Date of issuance: December 31, 1998.

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

   Amendment No.: 223.

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

   Date of initial notice in Federal Register: August 26, 1998 (63 FR 45526).

   The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 31, 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.

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

   Date of amendment request: March 26, 1997, as supplemented by letters dated March 18, 1998, and November 17, 1998.

   Brief description of amendment: The amendment revises Technical

   [[Page 4165]]

   Specifications (TS) 2.1.6 and its associated Basis to restrict the number of inoperable main steam safety valves when the reactor is critical.

   Date of issuance: December 31, 1998.

   Effective date: December 31, 1998.

   Amendment No.: 189.

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

   Date of initial notice in Federal Register: July 16, 1997 (62 FR 38137). The March 18, 1998, and November 17, 1998, supplemental letters provided additional clarifying information and did not change the original no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 31, 1998.

   No significant hazards consideration comments received: No.

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

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

   Date of application for amendment: April 14, 1997, as supplemented October 17, 1997, March 20, 1998, May 18, 1998, and August 17, 1998.

   Brief description of amendment: The amendment changes the Technical Specifications to allow for a Safety Review Committee review of plant performance as opposed to an audit of plant performance and replaces the position title of Vice President Regulatory Affairs and Special Projects with Director Regulatory Affairs and Special Projects.

   Date of issuance: December 30, 1998.

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

   Amendment No.: 186.

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

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

   The October 17, 1997, March 20, 1998, May 18, 1998, and August 17, 1998, letters provided clarifying information that did not change the proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 30, 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. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

   Date of application for amendment: March 22, 1996, as revised and supplemented on February 6, 1998, April 17, 1998, and October 30, 1998.

   Brief description of amendment: The amendment provides function- specific actions and allowed outage times for certain instrumentation, and relocates some instrumentation requirements to licensee-controlled documents.

   Date of issuance: January 12, 1999.

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

   Amendment No.: 250.

   Facility Operating License No. DPR-59: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: May 8, 1996 (61 FR 20855).

   The revision and supplemental information provided on February 6, 1998, April 17, 1998, and October 30, 1998, provided clarifying information that did not change the initial proposed no significant hazards consideration. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 12, 1999.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Reference and Documents Department, Penfield Library, State University of New York, Oswego, New York 13126.

   Southern California Edison Company, et al., Docket Nos. 50-361 and 50- 362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego County, California

   Date of application for amendments: June 30, 1997.

   Brief description of amendments: The amendments delete License Condition 2.C(19)b for San Onofre Nuclear Generating Station (SONGS) Unit 2 and revises TSs 3.3.1, 3.3.2, 3.3.5, 3.3.10, 3.3.11, 3.4.7, 3.4.12.1, 3.7.5, 5.5.2.10 and 5.5.2.11 for both SONGS units. These changes reinstate provisions of the SONGS Units 2 and 3 TS previously revised as part of NRC Amendment Nos. 127 and 116, respectively, make corrections to the TS, or remove information inadvertently added to the TS that are not applicable to the SONGS units design.

   Date of issuance: December 22, 1998.

   Effective date: December 22, 1998, to be implemented within 30 days from the date of issuance.

   Amendment Nos.: Unit 2--147; Unit 3--139.

   Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised Facility Operating License No. NPF-10 and the technical specifications for both licenses.

   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 December 22, 1998.

   No significant hazards consideration comments received: No.

   Local Public Document Room location: Main Library, University of California, P. O. Box 19557, Irvine, California 92713.

   Dated at Rockville, Maryland, this 20th day of January 1999.

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

   [FR Doc. 99-1705Filed1-26-99; 8:45 am]

   BILLING CODE 7590-01-P