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 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from April 23, 2001, through May 4, 2001. The last biweekly notice was published on May 2, 2001 (66 FR 22021).

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 NRC Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By June 15, 2001, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). 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 Start Printed Page 27175following 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 participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.

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

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

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

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Branch, or may be delivered to the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the attorney for the licensee.

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

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

Date of amendment request: April 16, 2001.

Description of amendment request: Energy Northwest is requesting approval for a change to the facility as described by the Final Safety Analysis Report (FSAR). The change allows for an unisolable drain line between the reactor core isolation cooling (RCIC) and control rod drive/condensate (CRD/COND) pump rooms. This change will modify the requirements that the RCIC and CRD/COND pump rooms be water-resistant or watertight, and connected by an isolable drain line.

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

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

The proposed change to allow operation of the plant with an open drain line between the RCIC and CRD/COND pump rooms does not increase the chances of a flooding event occurring in the RCIC or CRD/COND pump rooms. Also, operating the plant with an open drain line between the RCIC and CRD/COND pump rooms does not increase the radiological consequences of any previously evaluated accidents. A conservative revision to the flooding safe shutdown analysis, which combines the effects on equipment of both rooms flooding simultaneously, shows that sufficient safe shutdown equipment remains available and safe plant shutdown can be accomplished. Remaining systems are the same as the equipment providing the safe shutdown path approved by the NRC for Appendix R post-fire safe shutdown scenarios. Furthermore, the effects of the postulated flood from a pipe crack plus other normal leakage spread over the large floor area of the combined RCIC and CRD/COND rooms results in a flood event that develops slowly. If credited, safety-related leak detection instrumentation is available to provide plant operators more time to terminate the flood and limit the amount of equipment potentially lost from the event. With consideration of operator action and mitigation, the flood could be terminated quickly with minimal components affected. Plant procedures provide direction for operators to take actions to mitigate floods.

The proposed change to remove the requirement that pump room wall penetrations and doors located in the Reactor Building be “water-resistant” or “watertight” does not contribute to the likelihood that a flooding event will occur, nor does it increase the radiological dose received in any previously evaluated accidents. Reactor Building pump room doors and penetrations will exhibit a minimal amount of leakage during a flooding event, and have seals that can leak yet still minimize flooding between rooms even with significant hydrostatic pressure generated from flooding water levels. The minimal water leakage past these seals is consistent with assumptions documented in the existing flooding analysis.

Therefore, operation of Columbia Generating Station in accordance with the proposed amendment will 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 Start Printed Page 27176accident from any accident previously evaluated.

The proposed change to allow an unisolable drain line between the RCIC and CRD/COND pump rooms is accounted for in a revised and conservative flooding safe shutdown analysis. The flooding safe shutdown analysis documents the impact of flooding on equipment in the pump rooms, and on electrical circuits routed through, but not terminated in, the RCIC and CRD/COND pump rooms. From this analysis no link could be established between affected systems and mechanisms that could create a new or different kind of accident. The analysis also concluded that the effects of the unisolable drain line and subsequent flood would not cause a transient that would be imposed on the current analysis that assumes a flood with a single active failure. Therefore, the unisolable line between the RCIC and CRD/COND rooms will not create the possibility of a new or different kind of accident. The proposed change to allow minimal water leakage past ECCS [Emergency Core Cooling System], RCIC and CRD/COND pump room doors and penetrations is consistent with and documented in existing flooding analysis assumptions. These rooms do not need to be water-resistant or watertight.

Therefore, operation of Columbia Generating Station in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change to allow an unisolable drain line between the RCIC and CRD/COND pump rooms does not result in a significant reduction in the margin of safety. The change is of very low risk significance, with an increase in core damage frequency of less than 1E-10. Furthermore, a revised and conservative flooding safe shutdown analysis has concluded, as with previous flooding analysis for the ECCS, RCIC and CRD/COND pump rooms, that the ability to safely shutdown the plant has been preserved when considering a flooding scenario which impacts both the RCIC and CRD/COND pump rooms. In addition, safety-related leak detection instrumentation is available and could be credited to provide plant operators more time to terminate the flood and limit the amount of equipment potentially lost from the event. With consideration of operator action and mitigation, the flood could be terminated quickly with minimal components affected. Plant procedures provide direction for operators to take actions to mitigate flooding in ECCS, RCIC and CRD/COND pump rooms.

The proposed change to allow minimal water leakage past ECCS, RCIC and CRD/COND pump room doors and penetrations does not result in a significant reduction in the margin of safety because it does not prevent the plant from achieving safe shutdown during a flooding event. Minimal water leakage is consistent with and documented in existing flooding analysis assumptions.

Therefore, the operation of Columbia Generating Station in accordance with the proposed amendment will 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: Thomas C. Poindexter, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Stephen Dembek.

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

Date of amendment request: February 19, 2001.

Description of amendment request: The proposed change to the Technical Specifications (TS) modifies TS 3.6.5, “Vacuum Relief Valves,” limiting condition for operation and extends the allowed outage time from 4 hours to 72 hours for returning an inoperable primary containment to annulus relief valve to OPERABLE status. In addition, Entergy proposes to delete Attachment 1 to the Waterford Steam Electric Station, Unit 3 Operating License and revise Condition 2.C.1 to reflect the deletion.

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. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes do not create any new system interactions and have no impact on operation or function of any system or equipment in a way that could cause an accident. The primary containment to annulus vacuum relief valves are part of the containment vacuum relief system and are not initiators of any events nor affect any accident initiators of any events previously analyzed in Chapter 15 of the FSAR [Final Safety Analysis Report].

The primary containment to annulus vacuum relief valves are designed to mitigate the consequences of an inadvertent containment spray system actuation during normal plant operation. The FSAR analysis determined that with one of the two containment vacuum lines failed, the resultant peak calculated external pressure load of 0.49 psi [pounds per square inch] on the containment was less than the design external pressure loading of 0.65 psi. These proposed changes do not affect any of the assumptions used in the analysis. Hence, the consequences of the design basis accident previously evaluated do not change.

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

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes do not alter the design, configuration, or the method of operation of the plant. There is no change being made to the parameters within which the plant is operated. The setpoints at which the protective or mitigative actions are initiated are unaffected by this change. As such, no new failure modes are being introduced that would involve any potential initiating events that would create any new or different kind of accident.

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

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

The proposed changes do not affect the bases used in or the results of the analysis to establish the margin of safety. The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. None of these are impacted by the proposed change. The proposed change is acceptable because it assures at least one vacuum relief line will remain available in the event of a single failure. This further assures the ability to actuate upon demand for the purpose of mitigating the consequences of the design basis accident (inadvertent actuation of the containment spray system during normal operation). The remaining vacuum relief line provides sufficient vacuum relief capacity to prevent exceeding the design external pressure loading on containment of 0.65 psi. The resultant calculated peak external pressure loading on containment is 0.49 psi.

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

-

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

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: April 2, 2001.

Description of amendment request: Pursuant to 10 CFR 50.59, Entergy Operations, Inc. (the licensee) requests Start Printed Page 27177review and approval of changes to the Waterford Steam Electric Station, Unit 3, design basis as described in the Updated Final Safety Analysis Report (UFSAR) for which it has been determined that an unreviewed safety question exists. The change concerns design requirements for the alignment of the Refueling Water Storage Pool (RWSP) boundary isolation valves to the RWSP Purification System.

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

The proposed change will allow the manual valves (FS-423 and FS-404) that isolate the RWSP from the RWSP Purification System to be maintained open. The RWSP Purification System is aligned to the RWSP to maintain the purity and clarity of the borated water contained in the pool. The RWSP is also one of two means of makeup to the Spent Fuel Pool (SFP), with the Condensate Storage Pool being the primary makeup source. These manual valves provide the boundary between the seismically qualified safety related RWSP and the non-seismic, non-safety related RWSP Purification System.

(1) The proposed activity does not involve a significant increase in the probability or consequences of any accident previously evaluated.

The RWSP is not involved in any initiating event that could result in any accident. The RWSP has a safety function that assists in accident mitigation.

The proposed change has been reviewed against Engineering Standards and Licensing requirements contained in the Updated Final Safety Analysis Report (UFSAR). This review has concluded that use of operator action to isolate the RWSP Purification System Boundary Isolation valves or secure the RWSP Purification pump, as necessary, will allow the RWSP to perform its safety function in any plant mode. The RWSP, however, is not required to perform a safety function concurrent with a seismic event. The highest estimated annual probability of a small Loss of Coolant Accident (LOCA) while the purification system is aligned to the SFP, and operator failure to isolate the purification system, causing a diversion of RWSP water that could affect the Emergency Core Cooling System (ECCS) pumps is about 2.5E-8 per year, which is considered a negligible risk.

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

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

The RWSP Purification System was intended to be aligned periodically to the RWSP. The proposed change will allow the RWSP Purification System to be normally aligned to the RWSP through manually operated open valves. It has been shown that operator action can be credited to isolate the RWSP Purification System in a sufficient time to ensure the safety function of the RWSP is maintained. If Recirculation Actuation Signal (RAS) occurs the isolation valves will become inaccessible due to high dose rates in the general area. However if the RAS occurs before an operator can isolate the RWSP (i.e. 54 minutes), the RWSP Purification System would not have to be isolated because the RWSP would have fulfilled its required safety function.

The proposed alignment to maintain the RWSP Purification System isolation valves open introduces a new system interaction during a LOCA. However, it has been demonstrated the the safety function of the RWSP is assured assuming the new system interaction.

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

(3) The proposed activity does not involve a significant reduction [in] a margin of safety.

It has been evaluated that it will take no more than 54 minutes for operations personnel to isolate the RWSP. During this time, approximately 1.4% of the RWSP level could be depleted assuming maximum leakage. This volume will be incorporated into the analytical limit. The proposed analytical limit will continue to assure the safety limits evaluated in the Design Basis Accident (DBA) analyses are maintained.

Therefore, this proposed activity 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: N. S. Reynolds, Esquire, Winston & Strawn 1400 L Street NW., Washington, DC 20005-3502.

NRC Section Chief: Robert A. Gramm.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 2, 2001.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to relocate TSs 3/4.9.4, “Refueling Operations, Decay Time;” 3/4.9.5, “Refueling Operations, Communications;” 3/4.9.6, “Refueling Operations, Refueling Platform;” and 3/4.9.7, “Refueling Operations, Crane Travel-Spent Fuel Storage Pool;” to the Hope Creek Updated Final Safety Analysis Report (UFSAR). The proposed amendment would also modify the associated Bases pages and index pages.

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

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

The requested amendments will not involve an increase in the probability or consequences of an accident previously evaluated. Relocation of the affected Technical Specification sections and their Bases to the Hope Creek UFSAR will have no affect on the probability that any accident will occur. Additionally, the consequences of an accident will not be impacted because the affected systems and components will continue to be utilized in the same manner as before. No impact on the plant response to accidents will be created.

Based on the above, the proposed changes do not significantly increase the probability or consequences of any accident previously evaluated.

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

The proposed amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms will be created as a result of the relocation of the affected Technical Specification requirements and their Bases to the Hope Creek UFSAR. Plant operation will not be affected by the proposed amendments and no new failure modes will be created. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed amendments will not involve a reduction in the margin of safety. Relocation of the affected Technical Specification requirements to the Hope Creek UFSAR is consistent with NUREG 1433, [“]Standard Technical Specifications, General Electric Plants, BWR/4,[”] Revision 1, dated April 1995, and with the NRC's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132), dated July 22, 1993, which encourages utilities to propose amendments consistent with NUREG 1433. The margin of safety is unchanged; therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, Start Printed Page 27178P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

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

Date of amendment request: April 17, 2001.

Description of amendment request: The amendment would remove unnecessary details for certain secondary post-accident monitoring instrumentation from Technical Specification Table 3.2.6.

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

The post accident monitoring (PAM) instrumentation is not considered as an initiator or contributor to any previously evaluated accident. The proposed change will not impact the ability of the PAM instrumentation to perform its intended function, nor does it impact any Final Safety Analysis Report safety analysis. Therefore, the proposed change will not increase the probability of any accident previously evaluated.

Additionally, while the PAM instrumentation provides information to the control room operator that may be used to mitigate an accident, this change does not affect the ability of the PAM instrumentation to perform this function. This change does not modify any parameters of previously analyzed events.

Therefore, the proposed change will not increase the consequences of any accident previously evaluated.

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

The proposed change does not involve any physical modification to the plant, change in Technical Specification setpoints, plant operation, or design basis of the plant. The PAM instrumentation provides information to the plant operator to assist in the mitigation of an accident, and the means for accomplishment of this function are unchanged. Under the proposed change, operability of the PAM instrumentation is not impacted. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change would delete instrument identification numbers and instrument ranges from Technical Specifications for certain PAM instrumentation. These details are not necessary to ensure the PAM instrumentation is maintained operable. The requirements of Technical Specification Limiting Condition for Operation and associated Surveillance Requirements are adequate to ensure the required instrumentation is maintained operable. The proposed change will not impact the ability of the PAM instrumentation to perform its intended function. 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: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

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

Date of amendment request: March 22, 2001 (ET 01-0007).

Description of amendment request: The amendment would delete the inequality (≤) in front of the allowed temperature value and increase the allowed methyl iodide penetration values in item c of Technical Specification 5.5.11, “Ventilation Filter Testing Program (VFTP),” for the engineered safety feature (ESF) control room emergency ventilation system and auxiliary/fuel building emergency exhaust ventilation system.

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

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

The proposed changes revise the allowable methyl iodide penetration percent for the carbon in the Control Room Emergency Ventilation System and the Auxiliary/Fuel Building Emergency Exhaust System when tested in accordance with ASTM D3803-1989. The proposed change is based on the values that would be derived using a safety factor of 2 between credited and tested carbon efficiencies. This use of a safety factor of 2 is discussed in Generic Letter 99-02. Generic Letter 99-02 allows the reduction of the safety factor between the credited and tested carbon efficiencies from 5 (for systems with heaters) and 7 (for systems without heaters) to 2 (for systems with or without heaters) when tested in accordance with ASTM D3803-1989. Analyses of design-basis accidents assume a particular charcoal filter adsorption efficiency when calculating offsite and control room operator doses. A test of the charcoal filter samples determines whether the filter adsorber efficiency is greater than that assumed in the design-basis accident analysis. The laboratory test acceptance criteria contain a safety factor to ensure that the efficiency assumed in the accident analysis is still valid at the end of the operating cycle. Because ASTM D3803-1989 is a more accurate and demanding test, the use of a safety factor of 2 provides an acceptable adsorption efficiency greater than that assumed in the safety analysis.

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 revise the allowable methyl iodide penetration percent for the carbon in the Control Room Emergency Ventilation System and the Auxiliary/Fuel Building Emergency Exhaust System when tested in accordance with ASTM D3803-1989. The change in the allowable methyl iodide penetration percent is based on the values that would be derived using a safety factor of 2 as provided in Generic Letter 99-02. Generic Letter 99-02 allows the reduction of the safety factor between the credited and tested carbon efficiencies from 5 (for systems with heaters) and 7 (for systems without heaters) to 2 (for systems with or without heaters) when tested in accordance with ASTM D 3803-1989. No new or different accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of using a safety factor of 2 and deletion of the inequality sign associated with the temperature at which testing occurs.

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

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

The charcoal adsorber sample laboratory testing protocol accurately demonstrates the required performance of the adsorbers in the Control Room Emergency Ventilation System and the Auxiliary Building Emergency Exhaust System following a design basis accident or in the Fuel Building Emergency Exhaust System following a fuel handling accident. The change in safety factor and deletion of the inequality sign associated with the temperature at which testing occurs will not affect system performance or operation. This use [of] a safety factor of 2 is discussed in Generic Letter 99-02. Generic Letter 99-02 allows the reduction of the safety factor between the credited and tested carbon efficiencies from 5 (for systems with heaters) and 7 (for systems without heaters) to 2 (for systems with or without heaters) when tested in accordance with ASTM D3803-1989. Analyses of design-basis accidents assume a particular charcoal filter adsorption efficiency when calculating offsite Start Printed Page 27179and control room operator doses. A test of the charcoal filter samples determines whether the filter adsorber efficiency is greater than that assumed in the design-basis accident analysis. The laboratory test acceptance criteria contain a safety factor to ensure that the efficiency assumed in the accident analysis is still valid at the end of the operating cycle. Because ASTM D3803-1989 is a more accurate and demanding test, the use of a safety factor of 2 ensures the charcoal filter adsorption efficiency is greater than that assumed in the safety analysis when the penetration acceptance criterion is met. The offsite and control room dose analyses are not affected by this change and will remain within the limits of 10 CFR 100 and 10 CFR 50, Appendix A.

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

NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: August 29, 2000.

Brief description of amendment: The amendment revised the Technical Specifications to change the “Administrative Controls” section regarding certain position titles and the Shift Technical Advisor (STA) staffing requirement to allow one of the required on-shift Senior Reactor Operator (SRO) positions to be combined with the required STA position so as to serve in a dual SRO/STA position.

Date of Issuance: April 27, 2001.

Effective date: April 27, 2001 and shall be implemented within 30 days of issuance.

Amendment No.: 220

Facilit Operating License No.DPR-16: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 29, 2000 (65 FR 71133).

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 27, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendments: April 26, 2000, as supplemented November 6, 2000.

Brief Description of amendments: The amendments change the Technical Specifications Ultimate Heat Sink maximum 24-hour average temperature from 89 degrees F to 90.5 degrees F and increase the lower temperature of the condition temperature range. In addition, they change a surveillance requirement to require verification that the temperature is less than or equal to 90.5 degrees F.

Date of issuance: April 20, 2001.

Effective date: April 20, 2001.

Amendment Nos.: 213 and 240.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: May 17, 2000 (65 FR 31356), superseded on March 21, 2001 (66 FR 15916). The November 6, 2000, supplement contained additional information that expanded the scope of the initial application. Subsequently, the supplemented application was renoticed.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 20, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 7, 2000.

Brief description of amendment: The amendment changes the Technical Specifications regarding the Limiting Conditions for Operation for the containment cooling systems, the component cooling water system, and the service water system to be similar to changes to the “Standard Technical Specifications, Combustion Engineering Plants,” NUREG-1432, Revision 1, made by the Nuclear Energy Institute Technical Specifications Task Force Change Number 325, “ECCS Conditions and Required Actions with < 100% Equivalent ECCS Flow.”

Date of issuance: May 3, 2001.

Effective date: As of the date of issuance and shall be implemented within 90 days.

Amendment No.: 199.

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

Date of initial notice in Federal Register: January 24, 2001 (66 FR 7677).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 3, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 7, 2000.

Brief description of amendment: The amendment changes Technical Start Printed Page 27180Specification (TS) 3.7.5 regarding the Limiting Conditions for Operation (LCOs) for the auxiliary feedwater system to be similar to changes to the “Standard Technical Specifications, Combustion Engineering Plants,” NUREG 1432, Revision 1, made by the Nuclear Energy Institute Technical Specifications Task Force Change Number 325, Revision 0.

Date of issuance: May 3, 2001.

Effective date: As of the date of issuance and shall be implemented within 90 days.

Amendment No.: 200.

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

Date of initial notice in Federal Register: January 24, 2001 (66 FR 7674).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 3, 2001.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station (formerly known as WNP-2), Benton County, Washington

Date of application for amendment: September 5, 2000, as supplemented December 14, 2000.

Brief description of amendment: The amendment revised Technical Specification (TS) Tables 3.3.5.1-1, 3.3.6.1-1 and 3.3.6.2-1. The changes add notes to the tables listing instrument channels that are common to, or support the operability of interrelated systems as governed by these technical specifications. Specifically:

(1) Added note “(e)” to Table 3.3.5.1-1, “Emergency Core Cooling System Instrumentation,” Functions 1c, 1d, 2c and 2d in the column entitled “Required Channels Per Function,” indicating the applicability of the new footnote which reads as follows: “(e) Also supports OPERABILITY of 230 kV offsite power circuit pursuant to LCO 3.8.1 and LCO 3.8.2.”

(2) Added note “(e)” to Table 3.3.6.1-1, “Primary Containment Isolation Instrumentation,” in Functions 2b and 2c, in the column entitled “Required Channels Per Trip System,” indicating the applicability of the new footnote which reads as follows: “(e) Also required to initiate the associated LOCA Time Delay Function pursuant to LCO 3.3.5.1.”

(3) Added note “(c)” to Table 3.3.6.2-1, “Secondary Containment Isolation Instrumentation,” in Functions 1 and 2, in the column entitled “Required Channels Per Trip System,” indicating the applicability of the new footnote which reads as follows: “(c) Also required to initiate the associated LOCA Time Delay Function pursuant to LCO 3.3.5.1.”

Date of issuance: April 30, 2001.

Effective date: April 30, 2001, to be implemented within 30 days of the date of issuance.

Amendment No.: 172.

Facility Operating License No. NPF-21: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 15, 2000 (65 FR 69059).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 30, 2001.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: September 7, 2000, as supplemented on April 2, 2001.

Brief description of amendment: The amendment revises Technical Specification 3.7.3 to reflect planned modifications to the main feedwater system.

Date of issuance: April 18, 2001.

Effective date: April 18, 2001.

Amendment No.: 207.

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

Date of initial notice in Federal Register: November 15, 2000 (65 FR 69062).

The April 2, 2001, submittal contained clarifying information only, and did not change the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 18, 2001.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: August 16, 2000.

Brief description of amendment: The amendment (1) removes the “Offgas Treatment System Explosive Gas Mixing Instrumentation” Technical Specification (TS) 3.7 from the Radiological Effluent Technical Specifications contained in Appendix B and adds a reference to the Offgas Treatment System Explosive Gas Monitoring Program to Administrative Section 6 of the TSs contained in Appendix A; (2) replaces the position title of Radiological and Environmental Services Manager contained in the Administrative Section 6 of Appendix A with Radiation Protection Manager; and (3) revises Plant Staff organization requirements contained in Administrative Section 6 to require either the Operations Manager or the Assistant Operations Manager to hold a Senior Reactor Operator license.

Date of issuance: April 18, 2001.

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

Amendment No.: 270.

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

Date of initial notice in Federal Register: September 20, 2000 (65 FR 56956).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 18, 2001.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of application for amendment: November 10, 2000, as supplemented by letters dated February 15 and March 22, 2001.

Brief description of amendment: The amendment changes the safety limit minimum critical power ratio (SLMCPR) for Cycle 12 operation with a mixed core of Siemens Power Corporation (now known as Framatome ANP Richland, Inc.) ATRIUM-10 reload fuel, and General Electric GE11 reactor fuel. The amendment reflects a decrease of the two recirculation loop SLMCPR from 1.09 to 1.08, with the single recirculation loop SLMCPR remaining unchanged at 1.10. The amendment also revises Technical Specification 5.6.5 to update the list of references that are currently used to determine core operating limits.

Date of issuance: April 26, 2001.

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

Amendment No: 146.

Facility Operating License No. NPF-29: The amendment revises the Technical Specifications.

Date of initial notice in Federal Register: December 27, 2000 (65 FR 81917).

The Commission's related evaluation of the amendment is contained in a Start Printed Page 27181Safety Evaluation dated April 26, 2001. The February 15 and March 22, 2001, supplements did not change the scope of the amendment or the original proposed no significant hazards consideration determination.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: October 24, 2000, as supplemented on March 26, 2001.

Brief description of amendments: The amendments revise the technical specifications to reference the generically approved Westinghouse Best-Estimate large break loss of coolant accident methodology for the plants.

Date of issuance: April 18, 2001.

Effective date: April 18, 2001.

Amendment Nos.: 118, 118, 112 and 112

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revise the Technical Specifications.

Date of initial notice in Federal Register: February 21, 2001 (66 FR 11052).

The March 26, 2001, letter provided clarifying information that did not change the scope of October 24, 2000, application or the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 18, 2001.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: July 5, 2000, as supplemented by letters dated November 27, 2000, December 21, 2000, January 31, 2001, February 20, 2001, February 28, 2001, March 26, 2001, April 5, 2001, and April 16, 2001.

Brief description of amendments: The amendments revise the licenses and technical specifications to reflect approval of an increase in maximum thermal power from 3411 megawatts-thermal (MWt) to 3586.6 MWt.

Date of issuance: May 4, 2001.

Effective date: May 4, 2001.

Amendment Nos.: 119, 119, 113 and 113.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revised the License and the Technical Specifications.

Date of initial notice in Federal Register: December 13, 2000.

The supplements to the application provided clarifying information that did not change the scope of the July 5, 2000, application or the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 4, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 17, 1999, as supplemented June 14, November 13, and December 4, 2000, and February 21, 2001.

Brief description of amendment: Increased the allowed outage time to restore an inoperable Emergency Diesel Generator to operable status from 72 hours to 14 days.

Date of Issuance: April 26, 2001.

Effective Date: Date of issuance, to be implemented within 60 days.

Amendment No.: 115.

Facility Operating License No. NPF-16: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 15, 1999 (64 FR 70089). The June 14, November 13, and December 4, 2000, and February 21, 2001, supplements did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 26, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 30, 2000, as supplemented February 28, 2001.

Brief description of amendments: The amendments revised Technical Specification 5.3.2 for Turkey Point Units 3 and 4 to extend the residual heat removal (RHR) pump allowed outage time (AOT) from 72 hours to 7 days to restore an inoperable RHR pump to operable status.

Date of issuance: April 25, 2001.

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

Amendment Nos. 212 and 206.

Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 27, 2000 (65 FR 81922). The February 28, 2001, supplemental letter provided clarifying information which did not change the proposed no significant hazards consideration determination or expand the scope of the request as noticed.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 25, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 6, 2000.

Brief description of amendments: These amendments revised Technical Specification Surveillance Requirement 4.6.1.3.c to extend the interval for testing the containment air lock interlock mechanisms from 6 months to 24 months. Additionally, the amendments corrected an unrelated administrative error in TS Table 3.3-2, “Engineered Safety Features Actuation System Instrumentation.”

Date of issuance: April 26, 2001.

Effective date: April 26, 2001.

Amendment Nos. 213 and 207.

Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 7, 2001 (66 FR 9385).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 26, 2001.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of application for amendment: October 16, 2000, as supplemented December 22, 2000.

Brief description of amendment: The amendment revises the TSs to incorporate new pressure and temperature (P-T) limit curves. The reactor pressure vessel P-T limit curves are updated for inservice leakage and hydrostatic testing, non-nuclear heatup and cooldown, and criticality. The revised P-T limit curves are approved for an interim period not to exceed September 1, 2003. Start Printed Page 27182

Date of issuance: April 30, 2001.

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

Amendment No.: 238.

Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 13, 2000 (65 FR 77921).

The December 22, 2000, letter was within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 30, 2001.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

Date of application for amendment: June 2, 2000, as supplemented by letters dated December 15, 2000, and February 14, 2001.

Brief description of amendments: The amendment revised Technical Specification (TS) Section 3.5.2, “ECCS—Operating,” Action A to allow a one-time increase in the allowed outage time for centrifugal charging pump (CCP) 2-1 during Unit 2's Cycle 10 from 72 hours to 7 days. This change will allow for a potential on-line repair or a potential replacement of CCP 2-1. This pump is currently experiencing elevated vibration levels due to a structural resonance in the outboard bearing support structure and has been on an increased testing frequency since May 1996 due to high vibration.

Date of issuance: April 20, 2001.

Effective date: April 20, 2001, and shall be implemented within 30 days from the date of issuance.

Amendment No.: 146.

Facility Operating License No. DPR-82: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 12, 2000 (65 FR 43051).

The December 15, 2000, and February 14, 2001, supplemental letters provided additional clarifying information, 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 amendments is contained in a Safety Evaluation dated April 20, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 30, 2000.

Brief description of amendments: The license amendments revised Technical Specifications (TS) Section TS 3.5.1, “Accumulators” to (1) reflect the values of the accumulator pressure and volume consistent with the analyses assumptions documented in the current DCPP Final Safety Analysis Report (FSAR) Update, and (2) align the DCPP TS with the standard TS for Westinghouse plants.

Date of issuance: May 3, 2001.

Effective date: May 3, 2001, and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: Unit 1-147; Unit 2-147.

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

Date of initial notice in Federal Register: December 27, 2000 (65 FR 81928).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 3, 2001.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County, Georgia

Date of application for amendment: March 9, 2001.

Brief description of amendment: The amendment revises the Technical Specifications only until the Fall 2001 refueling outage and allows Mode 2 (startup) operation with two required intermediate range monitor channels per trip system.

Date of issuance: April 27, 2001.

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

Amendment No.: 166.

Facility Operating License No. NPF-5: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 20, 2001 (66 FR 15768).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 27, 2001.

No significant hazards consideration comments received: No.

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

Date of amendment request: March 17, 2000.

Brief description of amendments: The amendments revise Technical Specification (TS) 3/4.7.4, “Essential Cooling Water System,” to delete Surveillance Requirement 4.7.b.3, and to change Surveillance Requirements 4.7.4.b.1 and 4.7.4.b.2 to incorporate the wording from the Standard Technical Specifications for Westinghouse Plants (NUREG-1431). Surveillance Requirement 4.7.4.b.3 requires verifying at least once per 18 months that each screen wash booster pump and the traveling screen start automatically on a Safety Injection test signal. The Bases for TS 3/4.7.4 were also changed.

Date of issuance: April 30, 2001.

Effective date: The amendments are effective as of the date of their issuance.

Amendment Nos.: Unit 1—126; Unit 2—115.

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

Date of initial notice in Federal Register: April 19, 2000 (65 FR 21039).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 30, 2001.

No significant hazards consideration comments received: No.

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

Date of amendment request: November 18, 1999, as supplemented by letters dated November 29, 1999, and November 22, 2000.

Brief description of amendments: Changes to Technical Specifications surveillance testing to satisfy the actions requested in Generic Letter 99-02. The November 29, 1999, and November 22, 2000, letters provided additional clarifying information that was within the scope of the original application and Federal Register notice and did not change the NRC staff's initial proposed no significant hazards consideration determination.

Date of issuance: May 1, 2001.

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

Amendment Nos.: Unit 1—127; Unit 2—116.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications. Start Printed Page 27183

Date of initial notice in Federal Register: December 29, 1999 (64 FR 73099).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 1, 2001.

No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: May 17, 2000, as supplemented by letters dated August 31, 2000, and January 31, 2001.

Brief description of amendments: The amendments revise the Allowable Values specified in Technical Specification (TS) Table 3.3.5-1, “Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation” to ensure that the 6.9 kiloVolt and 480 Volt undervoltage relays initiate the necessary actions when required. In addition, a change is made to Condition D of TS 3.3.5, “Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation,” to eliminate the term “undervoltage.” This change is consistent with a change to TS Table 3.3.5-1.

Date of issuance: April 20, 2001.

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

Amendment Nos.: 85 and 85.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 21, 2001 (66 FR 15930).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 20, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 30, 2000, as supplemented by letters dated March 8 and 12, 2001.

Brief description of amendment: The changes revise the operability requirements for the refueling interlocks contained within TS 3.12.A as well as the surveillance requirements specified within 4.12.A. Clarifying changes are made to TS 3.12.D and 3.12.E to indicate that only the required interlocks need to be operable. In addition, TS 3.12.F will be clarified to articulate that there must be a minimum of 24 hours fission product decay prior to fuel handling. Some editorial changes were made in TS 3.12.B.

Date of Issuance: April 20, 2001.

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

Amendment No.: 200.

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

Date of initial notice in Federal Register: March 20, 2001 (66 FR 15770).

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 20, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 27, 2000.

Brief description of amendment: The amendment eliminates the specifications associated with the 24 Vdc Emergency Core Cooling System (ECCS) instrumentation batteries and chargers. The 24 Vdc ECCS instrumentation loads will be transferred to the 125 Vdc main station batteries.

Date of Issuance: April 20, 2001.

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

Amendment No.: 201.

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

Date of initial notice in Federal Register: January 10, 2001 (66 FR 2024).

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 20, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 1, 2000.

Brief description of amendment: The amendment revises the operability requirement for high pressure coolant injection (HPCI) and reactor core isolation cooling low steam line pressure isolation instrumentation to coincide with system operability requirements. The proposed change eliminates the need to open manual containment isolation valves under administrative control during reactor heatup, reduces the potential for operator error when closing these valves (potential for leaving valve mispositioned), and clarifies the steam line low pressure isolation function description. An administrative change to correct the HPCI High Steam Line d/p instrument component numbers was also made to ensure the accuracy of isolation instrumentation information.

Date of Issuance: April 20, 2001.

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

Amendment No.: 202.

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

Date of initial notice in Federal Register: December 13, 2000 (65 FR 77928)

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 20, 2001.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 19, 2000 as supplemented on February 13 and 23, 2001, and March 29, 2001.

Brief description of amendment: The amendment revises the Technical Specifications (TSs) by changing the reactor vessel pressure/temperature limit curves specified in TS 3.6.A.1, “Reactor Coolant Systems—Pressure and Temperature Limitations,” as graphically represented in Figures 3.6.1, Hydrostatic Pressure and Leak Tests, Core Not Critical, Figure 3.6.2, Normal Operation, Core Not Critical, and 3.6.3, Normal Operation/Core Critical.

Date of Issuance: May 4, 2001.

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

Amendment No.: 203.

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

Date of initial notice in Federal Register: January 24, 2001 (66 FR 7687).

The February 13 and 23, and March 29, 2001, supplements provided clarifying information that did not expand the scope of the application as published in the Federal Register, or change the proposed no significant consideration determination.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated May 4, 2001.

No significant hazards consideration comments received: No. Start Printed Page 27184

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia

Date of application for amendment: June 22, 2000, as supplemented September 19, 2000, and January 4, February 14, March 13, March 22, and April 11, 2001.

Brief description of amendment: These amendments revise Technical Specification (TS) Figures 3.4-2 and 3.4-3, and the associated Bases. These amendments approve new pressure-temperature limits, low-temperature overpressure protection (LTOP) system setpoints, and LTOP system effective temperature (Tenable) in the TS to a maximum of 32.3 effective full-power years (EFPY) for Unit 1 and 34.3 EFPY for Unit 2. These changes were based, in part, on the use of the American Society of Mechanical Engineers Code Case N-641.

Date of issuance: May 2, 2001.

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

Amendment Nos.: 226 and 207.

Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the TS.

Date of initial notice in Federal Register: February 23, 2001 (66 FR 11334). The January 4, 2001, submittal expanded the scope of the original June 22, 2000, application, which was noticed at 65 FR 48760. The February 14, March 13, March 22, and April 11, 2001, supplements contained clarifying information only, and did not change the February 23, 2001, initial no significant hazards consideration determination or expand the scope of the Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 2, 2001.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland this 8th day of May 2001.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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

End Signature End Preamble

[FR Doc. 01-12192 Filed 5-15-01; 8:45 am]

BILLING CODE 7590-01-P