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###### Nuclear Regulatory Commission

          - [Docket No. 50-440]

The U.S. Nuclear Regulatory Commission (the Commission) is considering issuance of an amendment to Facility Operating License No. NPF-58, issued to FirstEnergy Nuclear Operating Company (the licensee), for operation of the Perry Nuclear Power Plant, Unit 1, located in Lake County, Ohio. 

The proposed amendment would permit changes to the Perry Nuclear Power Plant Updated Safety Analysis Report (USAR) to incorporate descriptions (in the form of text, tables, and drawings) of modifications to the Emergency Service Water (ESW) alternate intake sluice gate. The modifications will include: (1) Installation of a safety-related Class 1E selector switch that will be used to disable the automatic opening function of the sluice gate during warm weather and (2) installation of a non-safety inflatable sealing device on the gates between the ESW forebay and the alternate intake tunnel. The modifications are designed to increase overall reliability of the ESW system and to eliminate undesired operation of the ESW pumps. 

Before issuance of the proposed license amendment, the Commission will have made findings required by the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations. 

The Commission has made a proposed determination that the amendment request involves 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. 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 new selector switch is procured as safety-related Class 1E, is fully qualified environmentally and seismically, and is also qualified in regard to mechanical and electrical operational cycles. Based on these characteristics, the switch is deemed to be highly reliable and will not introduce any new failure modes to the gate control circuit. In addition, the key operated feature of the selector switch ensures that inadvertent positioning of the switch,                         *i.e.,*                         an operator error, is not possible. Re-positioning of the switch will be procedurally controlled and will require conscious operator action along with use of a key. Therefore, it is concluded that addition of the new selector switch will not introduce any new failure modes and it will not cause or create any malfunctions of equipment.                     

                        The new inflatable seal and supporting mechanical equipment was procured as non-safety. The frequent verification of sluice gate seal integrity assures that the seals will be                               ( print page 37415)                            functional during accident and transient mitigation. This is supported by the probabilistic analysis that determined that the inflatable seal use results in a negligible increase in the Core Damage Frequency (7.4 E-8). Therefore, it is concluded that the new inflatable seals will not introduce any new failure modes and it will not cause or create any malfunctions of equipment.                     

The effect of disabling the automatic opening of the sluice gates with the proposed selector switch was evaluated and determined that the requirements of Regulatory Guide (RG) 1.27, Ultimate Heat Sink For Nuclear Power Plants, are met which ensures compliance with General Design Criteria (GDC) 44, Cooling Water. 

Analyzed events are initiated by the failure of plant structures, systems or components. The ESW system is an accident mitigating system that provides a reliable source of cooling water during accident conditions and is not an accident initiator. The proposed change does not have a detrimental impact on the integrity of any plant structure, system or component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. As a result, the probability of any accident previously evaluated is not significantly increased. 

Sufficient water is available to the ESW pumps to satisfy requirements for all modes of operation, accounting for minimum lake levels. The alternate intake tunnel that branches from the main discharge tunnel is isolated from the ESW pumphouse by the normally closed sluice gates. The alternate intake tunnel and sluice gates are not relied upon for mitigation of a Loss of Coolant Accident (LOCA) or other accidents with radiological consequences analyzed in the Updated Safety Analysis Report (USAR). The probabilistic analysis demonstrates that the unavailability of the alternate intake tunnel is acceptable during the time period that the sluice gate manual open/close circuit and the automatic opening signal is defeated, due to the extremely low probability of normal intake failure. The modifications do not result in changes to initial conditions of an accident nor alter assumptions used in any consequence determinations. This activity cannot increase the dose to the public nor on-site radiation doses such that actions to mitigate the radiological consequences of an accident would be impeded; nor does this modification directly or indirectly affect the ability of any other plant system to mitigate the radiological consequences of an accident. The proposed change will not alter the operation of any plant equipment assumed to function in response to the aforementioned analyzed events. Therefore, the probability of occurrence or the consequences of an accident previously evaluated remains unchanged. 

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

The proposed modification of the opening circuit has been designed, and will be procured and installed in accordance with the original ESW system design codes and standards. ESW system functions as required by GDC 44 and RG 1.27, have not been impacted by the change. Systems supporting the operation of the ESW system have not been affected. Failure of the modification to perform its design function due to electrical or mechanical failure would be identical to the current ESW system performance. 

Inflating the seals and defeating the automatic gate-opening signal results in the availability of only one intake path from the ultimate heat sink. Availability of only one intake during the time that the automatic opening function is disabled has been demonstrated to be acceptable because a water supply from the normal intake to the ESW pumps will be available. Cooling water supply from only one intake path cannot initiate an accident of a different type than previously evaluated because the cooling water supply paths cannot create or initiate an accident. 

The ESW system is an accident mitigating system and is not an accident initiator. Consequently, the alternate intake tunnel, the sluice gates, the sluice gate seals, and the sealing system are all components contained in the ESW system and are therefore not accident initiators. The operational change to the sluice gates, i.e., inflation of the seals and disabling of the sluice gate automatic opening feature, does not result in any interactions or interfaces with other plant systems, structures, or components that could create the possibility of an accident of a different type. The operational change prevents leakage past the sluice gates. Similarly, the sluice gates in the closed position does not result in any interactions or interfaces with other plant systems, structures, or components that could create the possibility of an accident of a different type. Performance of these isolation functions cannot initiate an accident. 

This change will not affect any known accident initiators or contributors; therefore, it will not increase the probability of an accident previously thought to be incredible. The proposed modifications do not affect any system or component that could initiate an accident. Therefore, this change will not create any different type of accident than previously evaluated in the USAR. Therefore, the probability of occurrence or the consequences of an accident previously evaluated remains unchanged. 

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

The design of the ESW system includes suitable redundancy and reliability to assure that an adequate supply of cooling water is provided and that no single failure will prevent safe shutdown of the unit. The normal cooling water supply to the ESW pump house is provided by a branch tunnel from the main intake tunnel, while a backup supply is available by means of a branch tunnel (alternate intake) from the main discharge tunnel. Currently, the sluice gates automatically open upon receipt of a signal from low water level switches in the ESW pump house forebay. Opening of a sluice gate ensures the necessary cooling water is available to the ESW pumps from the alternate intake tunnel. The licensing basis assumes that two supply paths are available and that automatic initiation would restore the cooling supply from the alternate path if the normal cooling supply were lost. The proposed modification will disable the manual/automatic-opening feature of the sluice gates during the summer months and will thus isolate the alternate supply path. A probabilistic study has demonstrated compliance with the requirements of RG 1.27. The study determined that an alternate source is not required due to having demonstrated that there is extremely low probability that a single aqueduct can functionally fail as the result of natural or site-related phenomena. Therefore, the proposed modification does not involve a significant reduction in the margin of safety. 

                    The closed sluice gates and the non-safety sluice gate seals prevent recirculation of plant discharge water to the ESW forebay and therefore maintain the forebay at or below its design temperature limit. The ESW system must be capable of providing cooling water at a temperature such that the heat exchangers serviced by ESW can remove their design heat loads for safe plant shutdown and for accident and transient mitigation. In order to prevent a reduction in the margin of safety associated with the ESW inlet temperature, the ESW forebay must not exceed 85°F. With the seals inflated, the closed sluice gates will prevent recirculation and subsequent increase of the forebay temperature above 85°F and therefore the closed sluice gates do not reduce the margin of safety associated with the ESW inlet temperature. The back-up air supply for the sluice gate seals, the frequent verification of the integrity of the sluice gate seals provided via administrative controls, and the functional and leak testing of the air system isolation check valves provides assurance that the inflated non-safety sluice gate seals can be credited during accident and transient mitigation and normal plant operation. Therefore, the margin of safety associated with ESW inlet temperature will not be reduced since the seals will be available to prevent leakage and subsequent increase of the forebay temperature above 85°F. Further, a probabilistic study supports this conclusion by demonstrating that seal failure, when needed, is highly improbable and would result in a negligible increase to core damage frequency. Therefore, it is concluded that inflation of the non-safety seals and                           ( print page 37416)                        reliance on them to prevent sluice gate leakage during all modes of operation does not represent a reduction to 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. 

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. 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 Administrative 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 6D59, 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 hearing and petitions for leave to intervene is discussed below. 

By July 14, 2000, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and accessible electronically through the ADAMS Public Electronic Reading Room link at the NRC Web site (http://www.nrc.gov). If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of 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 party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above. 

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

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

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

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

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

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to Mary E. O'Reilly, FirstEnergy Corporation, 76 South Main St., Akron, OH 44308, attorney for the licensee. 

                    Nontimely filings of petitions for leave to intervene, amended petitions,                           ( print page 37417)                        supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Commission, the presiding officer or the presiding Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).                 

For further details with respect to this action, see the application for amendment dated June 1, 2000, which is available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and accessible electronically through the ADAMS Public Electronic Reading Room link at the NRC Web site (http://www.nrc.gov). 

Dated at Rockville, Maryland, this 7th day of June, 2000. 

Anthony J. Mendiola, 

Chief, Section 2, Project Directorate III, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

[FR Doc. 00-15001 Filed 6-13-00; 8:45 am] 

BILLING CODE 7590-01-P