Skip to Content

Notice

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

Document Details

Information about this document as published in the Federal Register.

Published Document

This document has been published in the Federal Register. Use the PDF linked in the document sidebar for the official electronic format.

Start Preamble

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section Start Printed Page 78516189 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, November 25, through December 12, 2002. The last biweekly notice was published on December 10, 2002 (67 FR 75867).

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 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 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 Commission's Public Document Room (PDR), located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By January 23, 2003, 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,[1] which is available at the Commission's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

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

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to Start Printed Page 78517participate 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: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to hearingdocket@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and because of continuing disruptions in delivery of mail to United States Government offices, it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent 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 PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 304-415-4737 or by e-mail to pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, (TMI Unit 1) Dauphin County, Pennsylvania

Date of amendment request: November 8, 2002.

Description of amendment request: The proposed amendment would delete sections 3.15.3 and 4.12.3, “Auxiliary and Fuel Handling Building Air Treatment System,” of the TMI Unit 1 Technical Specifications (TSs) and their corresponding Bases. Various minor typographical corrections and other administrative corrections are also proposed.

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

Response: No.

This change will delete the existing Technical Specifications 3.15.3 and 4.12.3. It does not impact nor change the physical configuration of any system, structure or component, nor does it change the manner in which any system is operated. Any change to the system design will be evaluated in accordance with the requirements of [title 10 of the Code of Federal Regulations (10 CFR)] 10 CFR 50.59. Failure of the AFHBVS [Auxiliary and Fuel Handling Building Ventilation System] will neither initiate any type of accident nor increase the severity of the consequences of an accident.

Previously approved analyses of the dose consequences of the accidents described in the TMI Unit 1 UFSAR [Updated Final Safety Analysis Report] confirmed that potential dose consequences were below the limits of 10 CFR 100 or 10 CFR 50.67 without the operation of the AFHBVS. These analyses are not affected by the proposed Technical Specification change. Thus the AFHBVS is not required for mitigation of any accident as described in TMI Unit 1 UFSAR chapter 14.

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

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

Response: No.

This activity will delete sections of the Technical Specifications applicable to the AFHBVS. This change does not physically alter any system, structure or component. Any change to the system design will be evaluated in accordance with the requirements of 10 CFR 50.59. The proposed change will not cause the AFHBVS to operate outside its design basis. There will be no impact to any operational feature of the system or any procedures that control its operation. The design basis of the AFHBVS as described in the UFSAR is not revised.

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

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

Response: No.

The deletion of Technical Specification sections 3.15.3 and 4.12.3 will not impact the operation of the Auxiliary Fuel Handling Building Air Treatment System or the Fuel Handling Building ESF (engineered safety features) Ventilation system. The proposed change will not cause these systems to be placed in a configuration outside of their design basis nor will it reduce the margin of safety of these systems. The AFHBVS will continue to be operable in accordance with the applicable plant operating procedures. The AFHBVS will also continue to be tested and maintained under periodic operations surveillance and the TMI Unit 1 Preventive Maintenance Program.

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

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

Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice President, General Counsel and Secretary, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: October 24, 2002, as supplemented November 21, 2002.

Description of amendment request: The proposed amendments would Start Printed Page 78518revise the Technical Specifications to extend the completion time for an inoperable train of low pressure injection from 72 hours to seven days. The proposed amendments are risk-informed.

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:

Pursuant to 10 CFR 50.91, Duke Energy Corporation (Duke) has made the determination that this amendment request involves a No Significant Hazards Consideration by applying the standards established by the NRC regulations in 10 CFR 50.92. The specific responses to the criterion are discussed below:

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

Response: No.

The proposed change allows for one train of Low Pressure Injection to be inoperable for up to seven days. The Low Pressure Injection system is not an initiator for any accident previously evaluated and the consequences of an event during the extended Completion Time are no more severe than the consequences of the same event during the current Completion Time. Therefore, the consequences of an event previously analyzed are not increased. Consequently, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change allows for one train of Low Pressure injection to be inoperable for up to seven days. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods of governing normal plant operation. Therefore, the proposed changes does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change allows for one train of Low Pressure injection to be inoperable for up to seven days. An evaluation presented in Topical Report BAW-2295 and accepted by the NRC concluded that the extended Completion Time did not result in a significant reduction in the margin of safety. Therefore, the proposed changes 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.

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 17th Street, NW., Washington, DC 20005.

NRC Section Chief: John A. Nakoski.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of amendment request: October 22, 2002.

Description of amendment request: The proposed amendment deletes requirements from the technical specifications (TS) and other elements of the licensing bases to maintain a Post Accident Sampling System (PASS). Licensees were generally required to implement PASS upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI Unit 2. Requirements related to PASS were imposed by Order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. Lessons learned and improvements implemented over the last 20 years have shown that the information obtained from PASS can be readily obtained through other means or is of little use in the assessment and mitigation of accident conditions.

The changes are based on NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-413, “Elimination of Requirements for a Post Accident Sampling System (PASS).” The NRC staff issued a notice of opportunity for comment in the Federal Register on December 27, 2001 (66 FR 66949), on possible amendments concerning TSTF-413, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on March 20, 2002 (67 FR 13027). The licensee affirmed the applicability of the following NSHC determination in its application dated October 22, 2002.

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

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

The PASS was originally designed to perform many sampling and analysis functions. These functions were designed and intended to be used in post accident situations and were put into place as a result of the TMI-2 accident. The specific intent of the PASS was to provide a system that has the capability to obtain and analyze samples of plant fluids containing potentially high levels of radioactivity, without exceeding plant personnel radiation exposure limits. Analytical results of these samples would be used largely for verification purposes in aiding the plant staff in assessing the extent of core damage and subsequent offsite radiological dose projections. The system was not intended to and does not serve a function for preventing accidents and its elimination would not affect the probability of accidents previously evaluated.

In the 20 years since the TMI-2 accident and the consequential promulgation of post accident sampling requirements, operating experience has demonstrated that a PASS provides little actual benefit to post accident mitigation. Past experience has indicated that there exists in-plant instrumentation and methodologies available in lieu of a PASS for collecting and assimilating information needed to assess core damage following an accident. Furthermore, the implementation of Severe Accident Management Guidance (SAMG) emphasizes accident management strategies based on in-plant instruments. These strategies provide guidance to the plant staff for mitigation and recovery from a severe accident. Based on current severe accident management strategies and guidelines, it is determined that the PASS provides little benefit to the plant staff in coping with an accident.

The regulatory requirements for the PASS can be eliminated without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. The elimination of the PASS will not prevent an accident management strategy that meets the initial intent of the post-TMI-2 accident guidance through the use of the SAMGs, the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

Therefore, the elimination of PASS requirements from Technical Specifications (TS) (and other elements of the licensing bases) does not involve a significant increase Start Printed Page 78519in the consequences of any accident previously evaluated.

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

The elimination of PASS related requirements will not result in any failure mode not previously analyzed. The PASS was intended to allow for verification of the extent of reactor core damage and also to provide an input to offsite dose projection calculations. The PASS is not considered an accident precursor, nor does its existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radioisotopes within the containment building.

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

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

The elimination of the PASS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. Methodologies that are not reliant on PASS are designed to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The use of a PASS is redundant and does not provide quick recognition of core events or rapid response to events in progress. The intent of the requirements established as a result of the TMI-2 accident can be adequately met without reliance on a PASS.

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

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

Attorney for licensee: Thomas C. Poindexter, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Stephen Dembek.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: August 19, 2002.

Description of amendment request: The proposed amendment would modify the Technical Specifications (TSs) 3/4.2, “Protective Instrumentation,” and TS 3/4.7, “Containment Systems,” by changing requirements associated with post-accident monitoring (PAM) instrumentation. This will reflect the guidance of the U.S. Nuclear Regulatory Commission Regulatory Guide 1.97, and adopt standard TS requirements for PAM instrumentation. The proposed amendment would also modify the associated Bases.

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

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

Response: No.

Post-Accident Monitoring (PAM) Instrumentation is not an initiator of any previously evaluated accident because there is no credible failure of PAM instrumentation that could initiate previously evaluated accidents. Therefore, the proposed changes do not involve a significant increase in the probability of an accident previously analyzed.

The availability and use of PAM instrumentation help to ensure that the manual operator actions for mitigating an accident will be taken, and that the operator will be able to verify that automatic actions have occurred. The proposed changes make the requirements in the Technical Specifications more consistent with assumed operator actions. The proposed required actions, allowed out-of-service times, and surveillance intervals are appropriate based on operating experience, other instrumentation available, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation. Therefore, the proposed changes do not involve a significant increase in the consequences of an accident previously analyzed.

Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously analyzed.

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

Response: No.

The proposed change does not involve the physical modification of structures[,] systems, or components, plant design basis, or the manner in which the plant is operated. PAM instrumentation is passive and does not initiate automatic actions. As a result, there are no credible failures that could initiate a new or different kind of accident from any accident previously evaluated.

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

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

Response: No.

PAM instrumentation performs no automatic functions. PAM instruments help to ensure that operators take necessary manual actions to mitigate the consequences of an accident, and that operators have adequate information to confirm the operation of automatic accident mitigation functions have occurred. The proposed required actions, allowed out-of-service times, and surveillance intervals are appropriate based on operating experience, other instrumentation available, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation.

Therefore, the proposed amendment does not involve a significant reduction in 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.

Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts 02360-5599.

NRC Section Chief: James W. Andersen, Acting.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: February 26, 2002, as revised on October 9, 2002 and supplemented on October 30, 2002. This notice supersedes 67 FR 34495 published on May 14, 2002, which was based on the licensee's application dated February 26, 2002.

Description of amendment request: Revise the definition of Operable in Technical Specification (TS) 1.0.K with respect to support system requirements for AC power sources. Conforming changes are made to specific support system TSs in sections 3/4.5, “Core and Containment Cooling Systems,” 3/4.7, “Station Containment Systems,” and 3/4.10, “Auxiliary Electrical Power Systems,” and associated Bases.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below:

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

The revised definition of “Operable” redefines the AC power source requirements to allow either normal or emergency power available for equipment requiring AC power to be considered operable and provides conforming changes to specific supported system TSs. None of the proposed changes Start Printed Page 78520affects any parameters or conditions that could contribute to the initiation of any accident. The proposed change does not affect the ability of the AC power sources to perform their required safety functions nor does the proposed change affect the ability of the systems requiring AC power to perform their respective safety functions. As a result, the ability of these systems to mitigate accident consequences is unchanged. As such, these changes do not impact initiators of analyzed events, nor the analyzed mitigation of design-basis accident or transient events.

More stringent requirements for the inoperable AC power source action provisions that ensure availability of all TS required systems, subsystems, trains, components, and devices and the purely administrative changes do not affect the initiation of any event, nor do they negatively impact the mitigation of any event.

The elimination of some explicit requirements to verify the operability of remaining equipment (i.e., to verify which TS action is required to be entered and taken) does not affect the initiation of any event, nor does it negatively impact the mitigation of any event.

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

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

The proposed changes do not involve any physical modification to the plant, change in TSs setpoints, change in plant design basis, or a change in the manner in which the plant is operated. No new of different type of equipment will be installed. No safety-related equipment or safety functions are altered as a result of these changes. In addition, there are no changes in methods governing normal plant operation. No new accident modes are created since plant operation is unchanged. None of the proposed changes affects any parameters or conditions that could contribute to the initiation of any accident. The changes do not introduce any new accident or malfunction mechanism that could create a new or different kind of accident, thus, no new failure mode is created. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The manner in which plant systems relied upon in the safety analyses to provide plant protection is not changed. Plant safety margins continue to be maintained through the limitations established in the TSs Limiting Conditions for Operation and Actions. These changes do not impact plant equipment design or operation, and there are no changes being made to safety limits or safety system settings that would adversely affect the ability of the plant to respond as assumed in the accident analyses as a result of the proposed changes. Since the changes have no effect on any safety analysis assumptions or initial conditions, the margins of safety in the safety analyses are maintained.

In addition, administrative changes that do not change technical requirements or meaning, and the imposition of more stringent requirements to ensure operability, have no negative impact on margins of safety.

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

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Andersen, Acting.

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

Date of amendment request: November 22, 2002.

Description of amendment request: The proposed amendment would allow for a one-time change to revise the steam generator (SG) inservice inspection frequency requirements in Technical Specification 4.4.5.3.a to allow a 40-month inspection interval after one inspection, rather than after two consecutive inspections, based on the results falling into the C-1 classification.

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

Response: No.

There are no damage mechanisms that are active in the ANO-2 (Arkansas Nuclear One, Unit 2) SGs that would prematurely create an accident or increase SG leakage. The scope of inspections performed during 2R15, the first refueling outage following SG replacement, exceeded the TS (technical specification) requirements for ensuring that the ANO-2 steam generator[s] fell into the C-1 category. The ANO-2 steam generator[s] meet the current industry examination guidelines without performing inspections during the next refueling outage. The results of the Condition Monitoring Assessment performed during 2R15 demonstrated that all performance criteria were met. The results of the 2R15 Operational Assessment show that all performance criteria are being met over the proposed operating period.

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

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

Response: No.

The proposed change will not alter any plant design basis or postulated accidents resulting from potential SG tube degradation. The scope of inspections performed during the 2R15 outage, the first refueling outage following steam generator replacement, exceeded the TS requirements.

The proposed change does not affect the design of the SGs, the method of operation, or reactor coolant chemistry controls. No new equipment is being introduced and installed equipment is not being operated in a new or different manner. The proposed change involves a one-time extension to the SG tube inservice inspection frequency, and therefore will not give rise to new failure modes. In addition, the proposed change does not impact any other plant systems or components.

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

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

Response: No.

Steam generator tube integrity is a function of design, environment, and current physical condition. Extending the steam generator tube inservice inspection frequence by one operating cycle will not alter their function or design. Inspections conducted prior to placing the SGs into service and inspection during the first refueling outage following SG replacement demonstrate that the SGs do not have fabrication damage or an active damage mechanism. The scope of those inspections significantly exceeded those required by the TS. These inspection results were comparable to similar inspection results for the same model of RSGs (replacement steam generators) installed at other plants, and subsequent inspections at those plants yielded results that support this extension request. The improved design of the replacement SGs also provides reasonable assurance that significant tube degradation is not likely to occur over the proposed operating period.

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

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

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

NRC Section Chief: Robert A. Gramm. Start Printed Page 78521

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

Date of amendment request: August 16, 2002.

Description of amendment request: The proposed amendments would modify Technical Specification (TS) Surveillance section 4.0.3 to extend the delay time for completion of a missed surveillance to 24 hours or up to the surveillance frequency, whichever is greater. Additionally the proposed change would add a TS Bases Control Program.

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

The proposed change relaxes the time allowed to perform a missed surveillance. The time between surveillances is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The equipment being tested is still required to be operable and capable of performing the accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected. Any reduction in confidence that a standby system might fail to perform its safety function due to a missed surveillance is small and would not, in the absence of other unrelated failures, lead to an increase in consequences beyond those estimated by existing analyses. The addition of a requirement to assess and manage the risk introduced by the missed surveillance will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The relocation of two sentences from one specification to another in TS section 4.0, and the addition of a TS Bases Control Program in TS section 6.0, consistent with STS (Standard TS), is administrative in nature, does not affect the interpretation or execution of the TS, and has no effect on the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. A missed surveillance will not, in and of itself, introduce new failure modes or effects and any increased chance that a standby system might fail to perform its safety function due to a missed surveillance would not, in the absence of other unrelated failures, lead to an accident beyond those previously evaluated. The addition of a requirement to assess and manage the risk introduced by the missed surveillance will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The relocation of two sentences from one specification to another in TS section 4.0, and the addition of a TS Bases Control Program in TS section 6.0, consistent with STS, is administrative in nature, does not affect the interpretation or execution of the TS, and does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The extended time allowed to perform a missed surveillance does not result in a significant reduction in the margin of safety. As supported by the historical data, the likely outcome of any surveillance is verification that the LCO [limiting condition for operation] is met. Failure to perform a surveillance within the prescribed frequency does not cause equipment to become inoperable. The only effect of the additional time allowed to perform a missed surveillance on the margin of safety is the extension of the time until inoperable equipment is discovered to be inoperable by the missed surveillance. However, given the rare occurrence of a missed surveillance on inoperable equipment would be very unlikely. This must be balanced against the real risk of manipulating the plant equipment or condition to perform the missed surveillance. In addition, parallel trains and alternate equipment are typically available to perform the safety function of the equipment not tested. Thus, there is confidence that the equipment can perform its assumed safety function and this change does not involve a significant reduction in a margin of safety.

The relocation of two sentences from one specification to another in TS section 4.0, and the addition of a TS Bases Control Program in TS section 6.0, consistent with STS, is administrative in nature, does not affect the interpretation or execution of the TS, and does not involve a significant reduction in a margin of safety.

Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

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

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

NRC Section Chief: Allen G. Howe.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: November 15, 2002.

Description of amendment request: The proposed changes would revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) for both two recirculation (dual) loop operation and single recirculation loop operation in Technical Specification (TS) 2.1.1.2 to reflect results of a cycle specific calculation performed for Cycle 22.

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

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established, consistent with NRC approved methods, to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change conservatively establishes the safety limit for the minimum critical power ratio (SLMCPR) for Cooper Nuclear Station Cycle 22 such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.

Changing the SLMCPR does not increase the probability of an evaluated accident. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected.

The proposed change revises the SLMCPR to protect the fuel during normal operation as well as during any transients or anticipated operational occurrences. Operational limits (MCPR) are established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences) is met. Since the operability of plant systems designed to mitigate any consequences of accidents has not changed, the consequences of an accident previously evaluated are not expected to increase.

Based on the above NPPD [Nebraska Public Power District] concludes that the proposed changes do not involve a significant increase Start Printed Page 78522in the probability or consequences of an accident previously evaluated.

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

Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation. The proposed change does not involve any modifications of the plant configuration or allowable modes of operation. The proposed change to the SLMCPR assures that safety criteria are maintained for Cycle 22.

Based on the above NPPD concludes that the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1% of the rods are expected to be in boiling transition if the MCPR limits is violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences) are met.

Based on the above, NPPD concludes that the proposed changes do not involve a significant reduction in a margin of safety.

From the above discussions, NPPD concludes that the proposed amendment involves no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: Robert A. Gramm.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of amendment requests: September 24, 2002.

Description of amendment requests: The proposed license amendments would revise Technical Specifications (TS) 3.4.11, “Pressurizer Power Operated Relief Valves (PORVs),” and the licensing basis to credit automatic actuation of the Class 1 power operated relief valves (PORVs), instead of the pressurizer safety valves (PSVs), to limit reactor coolant system pressure changes for the spurious operation of the safety injection system at power event, and other design basis accidents. Also, TS 3.4.10, “Pressurizer Safety Valves,” would be revised to allow PSV loop seal temperatures to be less than the lower design temperature during plant heatup and cooldown in Mode 3 and in Mode 4 when any reactor coolant system cold leg temperature is greater than the low temperature overpressure protection arming temperature specified in the pressure temperature limits report, provided at least one Class I PORV is available and capable of providing automatic pressure relief. This would allow gradual stabilization of the loop seal temperatures, and avoid having to partially drain the loop seals to establish the proper PSV inlet temperature.

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.

Part of the instrumentation for automatic control of the Class 1 power operated relief valves (PORVs) during power operation is Instrument Class II. The automatic actuation circuitry will be upgraded to eliminate the Class II actuation circuitry, by providing output from the reactor protection system directly to the Class 1 PORVs. This upgrade does not adversely affect the ability of the Class 1 PORVs to function to mitigate a reactor coolant system (RCS) overpressure condition, and would not increase the probability of a spurious opening of a PORV.

The spurious operation of the safety injection (SI) system at power event is analyzed to assure that the RCS pressure limits are not exceeded, and that the departure from nucleate boiling ratio (DNBR) limits are met. The event is discussed in Final Safety Analysis Report (FSAR) Update Section 15.2.15. The current pressurizer overfill analysis takes credit for operation of the pressurizer safety valves (PSVs) to relieve a RCS overpressure condition. No credit is taken in the current analysis for automatic operation of the PORVs, which function to limit undesirable opening of the PSVs, since part of the automatic actuation circuitry is currently Instrument Class II. The current analysis that verifies that the DNBR limits are met remains bounding and was not reanalyzed.

The spurious operation of the SI system at power event was reanalyzed for pressurizer overfill using a RETRAN02/Mod005.2 computer code model of Diablo Canyon Power Plant. The analysis credits for automatic actuation of upgraded Class 1 PORVs to prevent water relief from the PSVs. Use of the Class 1 PORVs to perform any new safety related function would be evaluated in accordance with 10 CFR 50.59.

The RETRAN analysis demonstrates that the Class 1 PORVs can be expected to mitigate the consequences of a spurious operation of the SI system at power event, and that there is sufficient time for the operators to take action and open a PORV block valve(s) if closed.

Crediting the PORVs in the pressurizer overfill case for the spurious operation of the SI system at power event does not increase the probability of the occurrence of the transient since the automatic opening of the PORVs for RCS pressure control is not an initiator for the event. This change allows for the acceptance criteria to be met for the spurious operation of the SI system at power event, ensuring that the consequences of this event remain within acceptable levels.

The probability of a spurious operation of the SI system at power event is not affected by this proposed change and the above analysis demonstrates that the PORVs will adequately function in the automatic mode to mitigate the consequences of the transient. As such, there are no changes in the type or amount of any effluent released offsite as a result of this change.

The proposed change would allow the PSV loop seal temperatures to be less than the lower design temperature during plant heatup and cooldown in Mode 3, and in Mode 4 when any RCS cold leg temperature is greater than the low temperature overpressure protection (LTOP) arming temperature specified in the pressure temperature limits report (PTLR), provided at least one Class 1 PORV is available and capable of providing automatic pressure relief. An evaluation of the applicable events in these modes indicates one Class 1 PORV is capable of preventing water relief from the PSVs and maintaining the reactor coolant pressure below 110 percent of its design value.

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

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

The proposed changes would allow for automatic actuation of the Class 1 PORVs to be credited instead of the PSVs for the spurious operation of the SI system at power event. The proposed changes also allow the PSV loop seal temperatures to be less than the lower design temperature during plant heatup and cooldown in Mode 3, and in Mode 4 when any RCS cold leg temperature is greater than the LTOP arming temperature specified in the PTLR, provided at least one Class 1 PORV is available and capable of providing automatic pressure relief. Operation of the PORVs would prevent water relief from the PSVs, reducing the potential for a PSV not to properly reseat, and keep reactor coolant pressure below 110 percent of its design value. No new system interactions have been created, such that there is no increase in the possibility of a new or different kind of accident.

Therefore, the proposed changes do not create the possibility of a new or different Start Printed Page 78523kind of accident from any accident previously evaluated.

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

The proposed changes would allow for automatic actuation of the Class 1 PORVs to be credited instead of the PSVs for the spurious operation of the SI system at power event. The proposed changes allow the PSV loop seal temperatures to be less than the lower design temperature during plant heatup and cooldown in Mode 3, and in Mode 4 when any RCS cold leg temperature is greater than the LTOP arming temperature specified in the PTLR, provided at least one Class 1 PORV is available and capable of providing automatic pressure relief.

The spurious operation of the SI system at power event is analyzed to assure that the RCS pressure limits are not exceeded, and that the DNBR limits are met. The current pressurizer overfill analysis takes credit for operation of the PSVs to relief a RCS overpressure condition. No credit is taken in the current analysis for automatic operation of the PORVs, since part of the PORV automatic actuation circuitry is currently Instrument Class II. Since the PORV function would limit undesirable opening of the PSVs, the automatic actuation circuitry will be upgraded so that the PORVs can be credited for accident mitigation. This change would specifically allow for automatic actuation of the upgraded Class 1 PORVs to be credited instead of the PSVs in the accident analysis for the pressurizer overfill case.

A reanalysis for pressurizer overfill takes credit for the upgraded PORVs and shows that they can be expected to mitigate the consequences of a spurious operation of the SI system at power event, and that there is sufficient time for the operators to take action and open a PORV block valve(s) if closed. The current DNBR analysis remains bounding and was not reanalyzed.

The Class 1 PORVs will actuate to prevent water relief from the PSVs and keep reactor coolant pressure below 110 percent of its design value for a spurious operation of the SI system at power event. The conservative acceptance criteria for the current FSAR Update design analysis will continue to be met, and the margins of safety established in previous accident and transient analysis are not altered. The Class 1 PORVs will also provide overpressure protection during the period when the PSV loop seal temperature is less than the design limit.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: October 30, 2002.

Description of amendment request: The proposed amendments would decrease the Control Room Emergency Outside Air Supply System (CREOASS) maximum allowed filter train pressure drop from <9.1 inches water gage (wg), to <7.3 inches wg in Technical Specification (TS) 5.5.7.d to correct an error in the maximum allowed value. The proposed maximum allowed pressure drop across a filter train is consistent with current design analyses and test acceptance criteria.

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

Response: No.

The proposed change decreases the maximum acceptable pressure loss through the Control Room Emergency Outside Air Supply System (CREOASS) filter train. A limit is placed on the filter train pressure loss to assure that the CREOASS can deliver the design flowrate assumed in the control room radiological consequence analysis presented in the SSES Final Safety Analysis Report (FSAR). The proposed change assures the system design flowrate will be met. Thus, the consequences of any accident previously evaluated are not increased. [The proposed change does not involve a physical difference or alteration of plant equipment (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change does not change the design function or operation of the CREOASS.] The maximum allowable pressure drop through the CREOASS filter train is not an accident initiator thus, the probability of an accident previously evaluated is not increased. Therefore, the change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a physical modification or alteration of plant equipment (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change does not change the design function or operation of the CREOASS. Thus this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed action does not involve a significant reduction in a margin of safety. For the CREOASS, a lower maximum allowed pressure drop in TS does not adversely impact theoperation of any safety-related component or equipment. The proposed TS value is consistent with the design analysis and test acceptance criteria. Engineering evaluations concluded that there are no impacts on safety-related systems or accident analyses associated with the proposed change.

The margin of safety is established through the design of plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not impact the condition or performance of structures, systems, and components relied upon for accident mitigation.

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

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

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Section Chief: Richard J. Laufer.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: October 31, 2002.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to incorporate generic change (Technical Specification Task Force) TSTF-306, Revision 2 to NUREG 1433, “Standard Technical Specifications for General Electric Plants (BWR/4),” Revision 1, which has been approved by the NRC for adoption by licensees. Limiting Condition for Operation (LCO) 3.3.6.1, “Primary Containment Isolation Instrumentation,” would be revised to add an ACTIONS Note allowing intermittent opening, under administrative control, of penetration flow paths that are isolated to comply with ACTIONS, and to breakout Traversing Incore Probe (TIP) System isolation as a separate isolation function with an associated Required Action to Start Printed Page 78524isolate the penetration within 24 hours rather than immediately initiate a unit shutdown. The associated Bases would also be revised in accordance with TS 5.5.10, “TS Bases Control Program,” to be consistent with TSTF-306, Revision 2, and to document the proposed changes and provide supporting information.

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

The proposed change relaxes Required Actions. Required Actions and their associated Completion Times are not initiating conditions for any accident previously evaluated. Further, the Required Actions in this change have been developed to provide assurance that appropriate remedial actions are taken in response to the degraded condition considering the operability status of the redundant systems of required features, [and] the capacity and capability of remaining features, while minimizing the risk associated with continued operation. Therefore, the relaxed Required Actions do not significantly increase the probability or consequences of any accident previously evaluated.

2. Does the proposed change 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 (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The Required Actions and associated Completion Times in this change have been evaluated to ensure that no new accident initiators are introduced. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The relaxed Required Actions do not involve a significant reduction in a margin of safety. As provided in the justification, this change has been evaluated to minimize the risk of continued operation under the specified Condition, considering the operability status off the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repair or replacement of required features, and the low probability of a design basis accident occuring during the repair period. Therefore, this change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Section Chief: Richard J. Laufer.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: October 31, 2002.

Description of amendment request: The proposed amendment would revise the SSES Technical Specification (TS) requirements for OPERABILITY of the Main Turbine Bypass System (MTBS) bypass valves. Specifically, Surveillance Requirement (SR) 3.7.6.1 would be revised to verify one complete cycle of only each required turbine bypass valve every 31 days. Currently this TS assumes all five main turbine bypass valves are required to be operable.

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

The proposed change provides LCO [Limiting Condition for Operation] requirements for operation of the facility that are consistent with the safety analyses. Since the safety analyses do not take credit for any margin provided by the fifth main turbine bypass valve, these LCO requirements do not result in operation that will increase the probability of initiating an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event. The requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the current safety analyses and licensing basis. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change 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 (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change does impose different requirements. However, the change is consistent with the assumptions in the current safety analyses and licensing basis, and has been evaluated to ensure that no new accident initiators are introduced. Thus this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The imposition of less restrictive LCO requirements does not involve a significant reduction in a margin of safety. As provided in the justification, this change has been evaluated to ensure that the current safety analyses and licensing basis requirements are maintained. This change does not involve a significant reduction in a margin of safety since the required number of main turbine bypass valves will be the number assumed in the safety analysis.

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.

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Section Chief: Richard J. Laufer.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Units 1 and 2, Somerville County, Texas

Date of amendment request: November 19, 2002.

Brief description of amendments: The proposed amendments would revise Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, Operating Licenses, Appendix B, “Environmental Protection Plan,” to revise and replace references to the U.S. Environmental Protection Agency's (EPA's) National Pollutant Discharge Elimination System (NPDES) permit. The EPA delegated the provisions of the NPDES permit for CPSES to the State of Texas, Texas Natural Resource Conservation Commission (currently the Texas Commission on Environmental Quality), in accordance with the rules and regulations of both agencies. In addition, minor administrative changes to the Environmental Protection Plan's description are also proposed to be consistent with provisions of the current Texas Pollutant Discharge Elmination System (TPDES) permit and the Final Environmental Statement for the Operating License.

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:

Start Printed Page 78525

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

Response: No

The requested changes involve an administrative correction to the Comanche Peak Steam Electric Station (CPSES) Operating Licenses, Appendix B “Environmental Protection Plan” to replace references to the U.S. Environmental Protection Agency's (EPA's) National Pollutant Discharge Elimination System (NPDES) permit with references to the current Texas Pollutant Discharge Elimination System (TPDES) permit. The continuing environmental regulatory provisions of the NPDES permit are incorporated and renewed in the current State of Texas TPDES permit. The change in permit issuing authority was achieved in a manner consistent with the rules and regulations of both the EPA and the Texas Natural Resource Conservation Commission (TNRCC) (currently the Texas Commission on Environmental Quality).

Other minor changes proposed in the Environmental Protection Plan's description are administrative in nature and provide consistency with the provisions of the current TPDES permit and the NRC's [U.S. Nuclear Regulatory Commission] Final Environmental Statement—Operating License Stage.

This request involves administrative changes only. No actual plant equipment or accident analyses will be affected by the proposed change. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No

This request involves administrative changes only. No actual plant equipment or accident analyses will be affected by the proposed changes and no failure modes not bounded by previously evaluated accidents will be created. Therefore, the proposed changes do not create a new or different kind of accident from any accident previously evaluated.

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

Response: No

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel and fuel cladding, Reactor Coolant System pressure boundary, and containment structure) to limit the level of radiation dose to the public. This request involves administrative changes only.

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

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

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 5, 2002.

Description of amendment request: The proposed changes would revise the secondary coolant surveillance test requirements in table 4-2B, item 6, of the Technical Specifications (TS).

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

The proposed revision to Technical Specifications deletes the secondary coolant sampling requirements for the fifteen minute degassed beta and gamma activity test required once per 72 hours and for the semiannual dose equivalent I-131 analysis in TS Table 4.1-2B. The requirement for a dose equivalent I-131 analysis to be performed on a monthly basis remains in Table 4.1-2B. In accordance with the requirements of 10 CFR 50.92, the enclosed application is judged to involve no significant hazards based upon the following information:

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

The proposed change revises the sampling surveillance test requirements for the secondary coolant. Analyzed events are initiated by the failure of plant structures, systems, or components. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that could initiate an analyzed event. The proposed change will not alter the design and operation of, or otherwise increase the likelihood of failure of, any plant equipment that could initiate an analyzed accident.

The deletion of the 15 minute degassed beta and gamma activity test once every 72 hours is a less restrictive change, while the deletion of the semiannual equivalent dose I-131 analysis is more restrictive. In view of the higher sensitivity of the liquid gamma isotopic test used in calculating the dose equivalent I-131, the proposed deletion of the 15 minute degassed beta and gamma activity test and the proposed monthly performance of the dose equivalent I-131 analysis is appropriate. The dose equivalent I-131 analysis serves to confirm the validity of the safety analysis assumptions.

As a result, the probability or consequences of any accident previously evaluated are not significantly affected by the proposed change in surveillance frequencies.

2. Does the proposed license amendment 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 (i.e., no new or different type of equipment will be installed) or a change in the method of plant operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

A limit on the specific activity of the secondary coolant is required in order to limit the radiological consequences of a main steam line break to a small fraction of the 10 CFR 100 criteria. The proposed sampling surveillance test requirements for the secondary coolant will verify that the TS-required specific activity limit is satisfied and will serve to confirm the validity of the safety analysis assumptions. Hence, the proposed change in sampling surveillance test requirements 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.

Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Section Chief: John A. Nakoski.

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 Start Printed Page 78526connection 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, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: September 6, 2002.

Brief description of amendments: The amendments replace the peak linear heat rate safety limit, in TS 2.1.1.2, “Reactor Core SLs [Safety Limits],” by a peak fuel centerline temperature safety limit.

Date of issuance: December 2, 2002.

Effective date: December 2, 2002, and shall be implemented within 90 days of the date of issuance.

Amendment Nos.: Unit 1-145, Unit 2-145, Unit 3-145.

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

Date of initial notice in Federal Register: October 29, 2002 (67 FR 66007). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 2, 2002.

No significant hazards consideration comments received: No.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck Plant, Middlesex County, Connecticut

Date of amendment request: July 7, 2000, as supplemented by letters dated June 14, July 31, August 15, August 22, September 6, September 7, 2001, and May 9, June 26, August 15, August 20, and October 10, 2002.

Brief description of amendment: The amendment adds a license condition which approves the License Termination Plan (LTP) for the Haddam Neck Plant, and provides the criteria by which the licensee may make changes to the LTP without prior NRC approval.

Date of issuance: November 25, 2002.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment No.: 197.

Facility Operating License No. DPR-61: The amendment adds a condition to the Facility Operating License.

Date of initial notice in Federal Register: December 13, 2000 (65 FR 77915).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 25, 2002.

No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: July 10, 2002.

Brief description of amendment: This amendment revises Technical Specifications Surveillance Requirement 3.1.4.2 to extend the control rod scram time testing interval from 120 days to 200 days of full power operation.

Date of issuance: December 12, 2002.

Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance.

Amendment No.: 126.

Facility Operating License No. NPF-47: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: September 17, 2002 (67 FR 58641).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 12, 2002.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-10, Dresden Nuclear Power Station (DNPS), Unit 1, Grundy County, IL

Date of amendment request: August 1, 2002.

Brief description of amendments: The amendment revises the Operating License to update references to plant documents, deletes Technical Specification (TS) limiting conditions for required equipment and surveillance requirements that no longer apply or are being relocated to the Dresden Technical Requirements Manual, and deletes or revises TS administrative control and staffing requirements that either no longer apply or have changed due to the Unit 1 Fuel Storage Pool no longer containing spent fuel.

Date of issuance: December 3, 2002.

Effective date: December 3, 2002.

Amendment No.: 41.

Facility Operating License No. DPR-2: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: September 17, 2002 (67 FR 58642).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 3, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 19, 2002, as supplemented by letters dated October 21 and November 8, 2002.

Brief description of amendments: The amendments would extend the use of the current pressure and temperature (P/T) limit curves in Technical Specification (TS) 3.4.11, “RCS Pressure and Temperature (P/T) Limits,” until December 15, 2004. The change will allow sufficient time for the incorporation of the General Electric Topical Report NEDC-32983P, “General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation,” methodology into the P/T curves in TS 3.4.11.

Date of issuance: December 3, 2002.

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

Amendment Nos.: 156 & 142.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments revised the Technical Specifications.Start Printed Page 78527

Date of initial notice in Federal Register: October 30, 2002 (67 FR 66170).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 3, 2002.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile Point Nuclear Station, Unit 1, Oswego County, New York

Date of application for amendment: July 12, 2002.

Brief description of amendment: The amendment revised Technical Specifications sections 3.1.1 and 4.1.1, “Control Rod System,” by reducing the power level below which the rod worth minimizer or a second independent verification of rod position must be used from 20% to 10% rated thermal power.

Date of issuance: December 9, 2002.

Effective date: As of the date of issuance to be implemented before startup from Refueling Outage 17.

Amendment No.: 178.

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

Date of initial notice in Federal Register: August 6, 2002 (67 FR 50957).

The staff's related evaluation of the amendment is contained in a Safety Evaluation dated December 9, 2002.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of application for amendment: August 26, 2002.

Brief description of amendment: The amendment revises Surveillance Requirement (SR) 3.0.3 to extend the delay period before entering a Limiting Condition for Operation following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”

Date of issuance: December 12, 2002.

Effective date: As of the date of issuance and shall be implemented within 120 days.

Amendment No.: 210.

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

Date of initial notice in Federal Register: October 1, 2002 (67 FR 61683).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 12, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 30, 2002, as supplemented June 26, August 29, October 3, October 23, and November 11, 2002.

Brief description of amendments: These amendments increase the licensed reactor core power level by 1.4 percent from 1518.5 megawatts thermal (MWt) to 1540 MWt.

Date of issuance: November 29, 2002.

Effective date: As of the date of issuance and shall be implemented within 120 days.

Amendment Nos.: 207 and 212.

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

Date of initial notice in Federal Register: September 11, 2002 (67 FR 57630). The June 26, August 29, October 3, October 23, and November 11, 2002, supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 29, 2002.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 23, 2002, as supplemented by letters dated October 8 and 28, 2002.

Brief description of amendment: The amendment revises TS 2.5(1), “Steam and Feedwater Systems” to: (1) remove the requirement to demonstrate operability of redundant auxiliary feedwater system components, and (2) provide an allowed outage time to restore operability of the emergency feedwater storage tank. In addition to these revisions, TS 2.5 has been revised to be more consistent with NUREG-1432, “Improved Standard Technical Specification (ISTS) for Combustion Engineering Plants, Revision 2.”

Date of issuance: November 26, 2002.

Effective date: November 26, 2002, and shall be implemented within 120 days from the date of issuance.

Amendment No.: 212.

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

Date of initial notice in Federal Register: September 3, 2002 (67 FR 56327). The October 8 and 28, 2002, supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 26, 2002.

No significant hazards consideration comments received: No.

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

Date of amendments request: November 7, 2001, as supplemented by letter dated October 18, 2002.

Brief Description of amendments: The amendments revise the operating licenses by replacing the license conditions concerning spent fuel cask lifting devices with a commitment to the requirements in American National Standards Institute N14.6-1978, “Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 lbs (4500 kg) or More for Nuclear Materials,” in the Updated Final Safety Analysis Report.

Date of issuance: December 2, 2002.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: 158 and 149.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise the Operating License.

Date of initial notice in Federal Register: October 29, 2002 (67 FR 66013). The supplement dated October 18, 2002, provided clarifying information that did not change the scope of the November 7, 2001, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 2, 2002.

No significant hazards consideration comments received: No. Start Printed Page 78528

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendments request: May 23, 2002, as supplemented by letter dated October 31, 2002. The supplemental information provided clarification that did not change the scope or the initial no significant hazards consideration determination.

Brief description of amendments: The amendments revise the technical specifications for the end-of-life moderator temperature coefficient surveillance requirements.

Date of issuance: November 26, 2002.

Effective date: Amendments are effective on the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: Unit 1-144; Unit 2-132.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: July 9, 2002 (67 FR 45572). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 26, 2002.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 14, 2002, as supplemented July 22, 2002.

Brief Description of amendments: These amendments revise Technical Specifications section 4.5 and the associated Bases to change the surveillance frequency of the containment spray and recirculation spray header nozzles from a periodic surveillance of once every 10 years to a performance-based surveillance following maintenance that could cause nozzle blockage.

Date of issuance: December 10, 2002.

Effective date: December 10, 2002.

Amendment Nos.: 232 and 232.

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

Date of initial notice in Federal Register: June 25, 2002 (67 FR 42831). The July 22, 2002, supplement contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 10, 2002.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: September 27, 2001, as supplemented by letters dated June 27 and September 19, 2002.

Brief description of amendment: The amendment revises section 5.3.1.1, “Unit Staff Qualifications,” of the technical specifications to state new education and experience eligibility requirements for operator license applicants. As stated in the letter dated September 19, 2002, the new requirements are outlined by the National Academy for Nuclear Training in its “Guidelines for Initial Training and Qualification of Licensed Operators,” which were issued January 2000.

Date of issuance: November 26, 2002.

Effective date: November 26, 2002, and shall be implemented within 30 days of the date of issuance.

Amendment No.: 150.

Facility Operating License No. NPF-42. The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 23, 2002 (67 FR 48223).

The September 19, 2002, supplemental letter provided additional information that clarified the application, did not change the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 26, 2002.

No significant hazards consideration comments received: No.

Start Signature

Dated in Rockville, Maryland, this 16th day of December 2002.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

End Signature End Preamble

Footnotes

1.  The most recent version of title 10 of the Code of Federal Regulations, published January 1, 2002, inadvertently omitted the last sentence of 10 CFR 2.714 (d) and paragraphs (d)(1) and (d)(2) regarding petitions to intervene and contentions. For the complete, corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29, 2002.

Back to Citation

[FR Doc. 02-32081 Filed 12-23-02; 8:45 am]

BILLING CODE 7590-01-P