00-22779. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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 August 14, 2000, through August 25, 2000. The last biweekly notice was published on August 23, 2000 (65 FR 51346).

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

By October 6, 2000, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and 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 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 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: Docketing and Services Branch, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the attorney for the licensee.

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

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

Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: August 10, 2000.

Description of amendment request: The requested amendment proposes to change the Technical Specifications for operations involving positive reactivity addition. The proposed changes revise the Required Actions and Limiting Condition for Operation (LCO) Notes to limit the introduction of reactivity such that the required SHUTDOWN MARGIN (SDM) or refueling boron concentration will remain satisfied.

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:

Carolina Power & Light (CP&L) Company has evaluated the proposed Technical Specifications change and has concluded that it does not involve a significant hazards consideration. The CP&L conclusion is in accordance with the criteria set forth in 10 CFR 50.92. The bases for the conclusion that the proposed change does not involve a significant hazards consideration are discussed below.

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

The proposed change does not involve any physical alteration of plant systems, structures or components. The proposed change revises ACTIONS in the H. B. Robinson Steam Electric Plant (HBRSEP) Unit No. 2 Technical Specifications (TS) that require suspending operations involving positive reactivity additions and several Limiting Condition For Operation (LCO) Notes that preclude reduction in boron concentration. The change revises these ACTIONS and LCO Notes to limit the introduction of reactivity such that the required SHUTDOWN MARGIN (SDM) or refueling boron concentration will still be satisfied. The proposed change ensures that the SDM of LCO 3.1.1 and minimum boron concentration requirements of LCO 3.9.1 are met. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated in the Safety Analysis Report (SAR) because the accident analysis assumptions and initial conditions will continue to be maintained.

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

The proposed change does not involve any physical alteration of plant systems, structures or components. The proposed change, which allows positive reactivity additions that do not result in SDM or the refueling boron concentration being exceeded, does not introduce new failure mechanisms for systems, structures or components not already considered in the SAR [Safety Analysis Report]. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created because no new failure mechanisms or initiating events have been introduced.

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

The proposed change will allow positive reactivity additions, but the reactivity additions will not result in a[n] SDM or refueling boron concentration outside of the associated design basis limits. Allowing positive reactivity additions that do not result in the SDM or the refueling boron concentration being exceeded will not significantly reduce the margin of safety.

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

Attorney for licensee: William D. Johnson, Vice President and Corporate Secretary, Carolina Power & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Section Chief: Richard P. Correia

Commonwealth Edison Company, 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 amendment request: June 19, 2000

Description of amendment request: The proposed amendment would revise the technical specifications to remove their applicability related to the Boron Dilution Protection System (BDPS) after the next refueling outage for each unit. During the refueling outages, modifications are scheduled to be made which will permit the licensee to mitigate a boron dilution event without the use of the BDPS.

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 only accident potentially impacted by the proposed changes is the inadvertent boron dilution event.

The Boron Dilution Protection System (BDPS) is not considered an initiator of any analyzed event. The BDPS performs detection and mitigative functions for the inadvertent boron dilution event. Therefore, the proposed changes have no impact on the probability of an event previously analyzed. Therefore, the proposed change does not involve a significant increase in the probability of occurrence of an accident previously evaluated.

The proposed changes impact the consequences of an inadvertent dilution event due to the new requirement to manually reposition the Chemical and Volume Control System (CVCS) valves that isolate the boron dilution sources and that re-start boration of the Reactor Coolant System (RCS) in Modes 3, 4, and 5 (i.e., Hot Standby, Hot Shutdown, and Cold Shutdown, respectively). The revised detection and mitigation methodology being proposed achieves the same basic function as the existing BDPS, i.e., to prevent a return to critical during an inadvertent boron dilution event. The proposed changes will provide an improved response to the inadvertent boron dilution event compared to the BDPS, and thereby will prevent a return to critical. Therefore, the proposed changes do not involve a significant increase in the 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 changes to manually isolate potential dilution sources and to re-start boration of the RCS do not create the potential for a new or different kind of accident because the change results in plant configurations that have always been allowed. In conjunction with these proposed changes, enhancements to plant hardware, revisions to procedures, and administrative controls will be implemented. The proposed enhancements to plant hardware include the addition of two new redundant Volume Control Tank (VCT) high level alarms, which are passive in nature (i.e., do not provide any control function), and therefore do not create the possibility of a new or different kind of accident. Administrative controls and revisions to procedures will increase the operator's awareness of a potential boron dilution event and will provide the steps necessary to respond to a boron dilution event. As a result, the administrative controls and revisions to procedures do not create the possibility of a new or different kind of accident.

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

The design criterion and margin of safety for the existing BDPS is that the inadvertent boron dilution event is terminated within a specified period prior to the complete loss of shutdown margin. This criterion will continue to be satisfied following implementation of the proposed changes. The proposed changes were evaluated to ensure that the plant operators prevent criticality in Modes 3, 4 and 5 following an inadvertent boron dilution event, based on the revised analytical methodology previously discussed with the NRC and found to be feasible as documented in a letter from L. R. Wharton (U.S. NRC) to Licensees (Commonwealth Edison, Texas Utilities Electric, Union Electric, Wolf Creek Nuclear Operating Corporation, and Westinghouse), “Utility Subgroup Technical Approach to Modify or Delete the Boron Dilution Mitigation System,” dated February 8, 1993. The proposed method of detecting and mitigating this event has been shown by the analysis supporting this Technical Specifications change request to prevent a return to critical following an inadvertent boron dilution event, and meets the same NRC acceptance criteria as specified in the Standard Review Plan (SRP), NUREG-0800, Section 15.4.6, “Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant (PWR),” dated July 1981, as applicable to the existing BDPS. 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 requested amendments involve no significant hazards consideration.

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

NRC Section Chief: Anthony J. Mendiola

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

Date of amendment request: June 1, 2000

Description of amendment request: The amendments would revise Technical Specification (TS) 3.6.16 Reactor Building; and TS 5.5.11 Ventilation Filter Testing Program. It will also revise Bases Sections 3.6.10, 3.6.16, 3.7.12, and 3.7.13. The amendments will: (1) Enhance the ability to determine that reactor building annulus outside air inleakage is within the maximum assumed design value used in the dose analyses. Administrative limits are currently imposed at Catawba to limit inleakage in order to ensure that the dose analyses remain conservative. The amendments also request changes for the Unit 2 Annulus Ventilation System (AVS) in-place penetration and bypass leakage criteria in TS 5.5.11. This portion of the amendments affects TS Bases 3.6.10, TS 3.6.16 and Bases, and TS 5.5.11; (2) Describe the alignment the Auxiliary Building Filtered Ventilation Exhaust System (ABFVES) filtered exhaust units should be tested in and request appropriate TS 5.5.11 limits in order to ensure that the ABFVES will continue to meet its design basis functions. Similar to Item 1 above, the amendments also request changes for the Unit 2 ABFVES in-place penetration and bypass leakage criteria in TS 5.5.11. This portion of the amendments affects TS Bases 3.7.12 and TS 5.5.11; and (3) Modify the TS Bases for the Fuel Handling Ventilation Exhaust System (FHVES) and similar to Items 1 and 2 above, the amendments also request changes for the Unit 2 FHVES in-place penetration and bypass leakage criteria in TS 5.5.11. This portion of the amendments affects TS Bases 3.7.13 and TS 5.5.11.

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 following discussion is a summary of the evaluation of the changes contained in this proposed amendment against the 10 CFR 50.92(c) requirements to demonstrate that all three standards are satisfied. A no significant hazards consideration is indicated if 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.

First Standard

Implementation of this amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. Neither the AVS, nor the ABFVES, nor the FHVES is capable of initiating any accident. The AVS, ABFVES, and FHVES, which are responsible for maintaining an acceptable environment in the annulus, the auxiliary building, and the fuel building during normal and accident conditions, will continue to function as designed, and in accordance with all applicable TS. The design and operation of the systems are not being modified by this proposed amendment. Therefore, there will be no impact on any accident probabilities or consequences.

Second Standard

Implementation of this amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the plant which will introduce any new accident causal mechanisms. This amendment request does not impact any plant systems that are accident initiators and does not impact any safety analyses.

Third Standard

Implementation of this amendment would not involve a significant reduction in a margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these fission product barriers will not be impacted by implementation of this proposed amendment. The performance of the AVS, the ABFVES, and the FHVES in response to normal and accident conditions will not be impacted by this proposed amendment. The changes to the AVS surveillances will provide for a better method to ensure that the assumptions of the dose analyses are met. There is no risk significance to this proposed amendment, as no reduction in system or component availability will be incurred. No safety margins will be impacted.

Based upon the preceding discussion, Duke has concluded that the proposed amendment 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 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. Lisa F. Vaughn, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

NRC Section Chief: Richard L. Emch, Jr.

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

Date of amendment request: August 10, 2000.

Description of amendment request: The proposed amendment would revise the Technical Specifications to allow an alternate storage configuration of fuel assemblies adjacent to the walls within Region 1 of the spent fuel pool.

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

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

The probability of fuel handling accidents (dropped assemblies, misplaced/misloaded assemblies, etc.) is not changed by utilizing the previously described vacant spaces that are face adjacent to the SFP [spent fuel pool] walls in Region I [Region 1] to store design basis assemblies that are less reactive than RI A [Region 1 Configuration A] type assemblies. Fuel assemblies of different types are presently stored face adjacent to these walls. This proposal will allow additional assemblies to be located face adjacent to the Region I SFP walls and does not effect the precursors to any postulated spent fuel pool accidents.

The consequences of an accident different than that previously analyzed additionally remains unchanged. Evaluations have demonstrated that the fuel handling accident reactivity values will remain less than the 0.95 K_eff acceptance criteria in the event of a fuel handling accident, assuming an initial SFP boron concentration of 1000 ppm. The boron concentration limit is additionally bounded by ANO-2 [Arkansas Nuclear One, Unit 2] TS [Technical Specification] Limiting Condition for Operation (LCO) 3.9.12.c which limits SFP boron to greater than 1600 ppm at all times.

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

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

As discussed previously, the proposed SFP configuration will not result in exceeding the acceptance criteria of 0.95 K_eff during normal or accident conditions. Since fuel assemblies are currently located along the Region I SFP walls, no new or different kind of accident than that previously evaluated exists. Locations required to be vacant will remain physically blocked. In the event that a “misloading” type accident occurs in this region, evaluations have shown that the fuel handling accident reactivity values will remain well below 0.95 K_eff when initial SFP boron concentrations are at or above 1000 ppm, which is significantly less than the TS boron limit of 1600 ppm.

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

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

As previously discussed, the proposed configuration will not result in exceeding the 0.95 K_eff acceptance criteria during normal operations that assume zero concentration of boron at the maximum water density in the SFP or during accident conditions that assume an initial SFP boron concentration of at least 1000 ppm. Furthermore, ANO-2 TS 3.9.12.c requires SFP boron to be maintained greater than 1600 ppm at all times. Fuel assemblies are presently stored along the Region I SFP walls; therefore, storing additional assemblies along these same walls will not significantly reduce the margin to safety since it has been shown that the current CSA [criticality safety analysis] remains valid.

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

NRC Section Chief: Robert A. Gramm

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

Date of amendment request: May 12, 2000, as supplemented June 19, 2000.

Description of amendment request: The proposed amendment would revise the Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2), calculated doses and associated descriptions/information listed in the Updated Final Safety Analysis Reports (UFSARs) for the Design Basis Accidents (DBAs). An evaluation of all of the BVPS-1 and 2 dose calculations was completed which reviewed the input parameter values, the input assumptions, and the methodologies used. Some of the input parameter values, input assumptions and methodologies used in the DBA dose calculations were revised. The resultant DBA dose calculation revisions necessitate associated revisions to the UFSARs. Additionally, some changes would be made in response to Generic Letter 99-02. For BVPS-1, the requested amendment would affect the analyses for the following DBAs: loss of offsite AC power, fuel-handling accident, accidental release of waste gas, steam generator tube rupture, major secondary system pipe rupture, rod cluster control assembly ejection, single reactor coolant pump locked rotor, and loss of reactor coolant from small ruptured pipes/loss-of-coolant accidents. For BVPS-2, the requested amendment would affect the analyses for the following DBAs: steam system piping failures, loss of AC power, reactor coolant pump shaft seizure, rod cluster control assembly ejection, failure of small lines carrying primary coolant outside containment, steam generator tube rupture, loss-of-coolant accidents, waste gas system failure, and fuel-handling accidents.

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

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

Following a reevaluation of the calculated dose values for BVPS Unit 1 and Unit 2 design basis accidents (DBAs) as described in their respective [Updated Final Safety Analysis Report] UFSAR, several calculated dose values were identified to be increased. These increases were small and remained within the applicable DBA previously approved regulatory limit.

The increases for each DBA were as a result of revised plant data being used in the dose calculation, revised calculation assumptions, or new methodology. These changes were not the result of plant hardware changes. The changes were intended to ensure that accurate, current and conservative licensing basis information and assumptions were used for DBA dose analyses. The UFSAR changes are proposed to reflect the revised analyses results for the Unit 1 and Unit 2 UFSAR.

Since the calculated DBA radiological doses remain within the applicable DBA previously approved regulatory limit, these calculated dose values do not involve a significant increase in the probability or consequences of an accident as previously evaluated in the BVPS Unit 1 and Unit 2 UFSAR.

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

BVPS Unit 1 and Unit 2 calculations which are used to determine DBA calculated dose values were revised. The changes were as a result of revised plant data being used in the dose calculation, revised calculation assumptions or new methodology. The changes were intended to ensure that accurate, current and conservative licensing basis information and assumptions were used for DBA dose analyses. The DBA events themselves remain the same postulated events as previously described within the BVPS Unit 1 and Unit 2 UFSARs. The revised dose calculations do not create the possibility of a new or different kind of accident from the DBA accidents previously evaluated in the UFSAR. These changes were not the result of plant hardware changes. The changes were only in the calculations. The UFSAR changes are proposed to reflect the revised analyses['] results for the Unit 1 and Unit 2 UFSAR.

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

This amendment request addresses only proposed changes to the Unit 1 and Unit 2 UFSAR, which was determined to involve an Unreviewed Safety Question pursuant to 10 CFR 50.59. This request does not propose modifying any Technical Specification criteria. This request proposes that several calculated dose values for BVPS Unit 1 and Unit 2 DBAs be increased following a reevaluation of their design basis calculations. These proposed increases are small and remained within the applicable DBA previously approved regulatory limit. Thus, the proposed changes to the UFSAR which originated from revised BVPS DBA dose calculations [do] not involve a significant reduction in the margin of safety for BVPS Unit 1 and Unit 2 because the Technical Specifications will not be altered and the increase in calculated dose values is small and remains within regulatory approved limits.

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: Mary O'Reilly, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Marsha Gamberoni

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

Date of amendment request: July 31, 2000

Description of amendment request: The amendment would change Technical Specifications 3.8.1.1, “Electrical Power Systems—A.C. Sources—Operating,” and 3.8.1.2, “Electrical Power Systems—A.C. Sources—Shutdown.” The index and the Bases for these Technical Specifications will be modified as a result of the proposed changes. The proposed changes will allow certain emergency diesel generator (EDG) surveillance requirements to be performed when the plant is operating instead of shut down as currently required. Additional changes will remove EDG accelerated testing and special reporting requirements, and the surveillance requirement to perform EDG inspections. EDG inspections will still be performed as recommended by the manufacturer. The proposed changes will not adversely impact the type and amounts of effluents that may be released off site.

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's analysis is presented below:

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

The proposed Technical Specification changes are associated with the surveillance requirements for the Emergency Diesel Generators (EDGs) and will not affect the ability of the EDGs to perform their intended safety function. Therefore, the proposed Technical Specification changes will not result in a significant increase in the probability or consequences of an accident previously evaluated.

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

Since there are no changes in components, component operation, or system operation, this change does not create the possibility of an accident of a different type.

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

The proposed changes will have no adverse effect on plant operation or equipment important to safety. The plant response to the design basis accidents will not change and the accident mitigation equipment will continue to function as assumed in the design basis accident analysis. Therefore, there will be no significant reduction in a margin of safety.

Based on the staff's analysis, 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: Lillian M. Cuoco, Senior Nuclear Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, Connecticut

NRC Section Chief: James W. Clifford

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: June 14, 2000

Description of amendment request: The proposed amendments would revise Vogtle's Surveillance Requirements (SR) 3.8.1.9 and 3.8.1.14 to reduce the emergency Diesel Generator (EDG) loading requirements from ≥6800 kW and ≤7000 kW to ≥6500 kW and ≤7000 kW. These changes will make the above SRs consistent with SR 3.8.1.3 and 3.8.1.13 which are in the current Technical Specifications (TS). In addition, the proposed amendments would revise TS section 5.6.7, “EDG Failure Report”, to correct a typographical error.

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?

No. The proposed change to section 5.6.7 is administrative only since it does nothing more than correct a typographical error. The proposed changes to the DG loading requirements specified in SRs 3.8.1.9 and 3.8.1.14 have no impact on or relationship to the probability of any of the initiating events assumed for the accidents previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability of any accident previously evaluated. Furthermore, since the proposed loading requirements bound the maximum expected loading for the DGs, SRs 3.8.1.9 and 3.8.1.14 will continue to demonstrate that the DGs are capable of performing their safety function. Since the proposed changes do not adversely affect the capability of the DGs to perform their safety function, the outcome of the accidents previously evaluated (i.e., radiological consequences) will not be affected. Therefore, the proposed changes do not involve a significant increase in the consequences of any accident previously evaluated.

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

No. The proposed change to section 5.6.7 is administrative only since it does nothing more than correct a typographical error. The proposed changes to the DG loading requirements specified in SRs 3.8.1.9 and 3.8.1.14 will not introduce any new equipment or create new failure modes for existing equipment. Other than the reduced loading requirements for the DGs, the proposed changes will not affect or otherwise alter plant operation. The DGs will remain capable of performing their safety function. No other safety related or important to safety equipment will be affected by the proposed changes. Therefore, the proposed changes will 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 a margin of safety?

No. The proposed change to section 5.6.7 is administrative only since it does nothing more than correct a typographical error. The proposed changes reduce the loading requirements of SRs 3.8.1.9 and 3.8.1.14. The new loading requirements bound the maximum expected loading of the DGs under the worst case scenario, and they are consistent with the regulatory guidance found in Regulatory Guide (RG) 1.9, Revision 3, “Selection, Design, and Qualification of Diesel-Generator Units Used as Standby (Onsite) Electric Power Systems at Nuclear Power Plants,” July 1993. Reduction in wear and tear should inherently increase the reliability of the DGs. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Conclusion

Based on the above evaluation, the proposed changes do not involve a significant hazard as defined in 10 CFR 50.92.

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. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-2216

NRC Section Chief: Richard L. Emch, Jr.

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

Date of amendment request: August 11, 2000 (TS-400).

Description of amendment request: The proposed amendment would change the Units 2 and 3 Technical Specifications to revise the testing frequency for certain isolation valves of a type known as excess flow check valves (EFCV). The proposed testing frequency would allow a representative sample to be tested every 24 months, such that each EFCV is tested at least once every 120 months. The current specification requires that each EFCV be tested at least once every 24 months.

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:

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

The current excess flow check valve (EFCV) frequency requires that each reactor instrument line EFCV be tested every 24 months. The EFCVs are designed to automatically close upon excessive differential pressure including failure of the down stream piping or instrument and will reopen when appropriate. This proposed change will allow a reduction in the number of EFCVs that are verified tested every 24 months, to approximately 20 percent of the valves each cycle. BFN and industry operating experience demonstrates high reliability of these valves. Neither the EFCVs or their failure is capable of initiating a previously evaluated accident. Therefore, there is no increase in the probability of occurrence of an accident previously evaluated.

The instrument lines going to the Reactor Coolant Pressure boundary with EFCVs installed have flow restricting devices upstream of the EFCV. The consequences of a unisolable failure of an instrument line has been previously evaluated and meets the intent of NRC Safety Guide 11. The offsite exposure has been calculated to be substantially below the limits of 10 CFR 100. Additionally, coolant lost from such a break is inconsequential compared to the makeup capabilities of normal and emergency makeup systems. Although not expected to occur as a result of this change, the effects of a postulated failure of an EFCV to isolate and instrument line break as a result of reduced testing are bounded by TVA analysis.

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

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

This proposed change reduces the number of EFCVs tested each operating cycle. No other changes to the TS are being proposed. BFN and industry operating experience demonstrates that these valves are highly reliable, a proposed reduction in test frequency is bounded by previous evaluation of a line rupture. The change will not alter the operation of process variables, structures, systems or components described in the BFN Updated Final Safety Analysis Report. Therefore, reduction in the number of EFCVS tested each cycle does not result in the possibility of a new or different kind of accident.

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

The consequences of an unisolable rupture of an instrument line has been previously evaluated and meets the intent of NRC Safety Guide 11. The proposed amendment does not involve a significant reduction in a margin of safety.

Therefore, the proposed revised surveillance frequency does not adversely affect the public health and safety, and does not involve any significant safety hazards.

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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET I0H, Knoxville, Tennessee 37902.

NRC Section Chief: Richard P. Correia.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: August 4, 2000 (TS 99-20)

Brief description of amendments: The proposed amendments would change the Sequoyah Nuclear Plant (SQN) Technical Specifications (TS), Section 6.2.2, to change the title of various shift members and to change the Shift Technical Advisor requirements.

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

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

The title change of Shift Operations Supervisor to Shift Manager is administrative. The elimination of TS 6.2.2.b and Table 6.2-1 is considered an administrative change. These two items contain similar requirements as those contained in 10 CFR 50.54(m)(2)(iii), 10 CFR 50.54(m)(2)(i), and 10 CFR 50.54(k). These sections are considered a duplicate of the requirements contained in the Code of Federal Regulations. This request also eliminates the title of Shift Technical Advisor (STA) but will not eliminate or reduce licensee responsibilities in this area. This request is based on an NRC policy statement, contained in Generic Letter 86-04, that supports the transition of engineering expertise from the STA position to another individual on shift who possesses the mandated education qualifications. The proposed administrative and organizational changes do not result in any increase in the probability or consequences of an accident previously evaluated.

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

As described above, the proposed changes are administrative and organizational in nature and cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

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

As described above, the proposed changes are administrative and organizational in nature. The proposed changes are based on approved NRC guidance. The margin of safety is, therefore, not reduced.

The NRC 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Richard P. Correia.

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

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

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

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

Date of amendment request: July 21, 2000.

Description of amendment request: The amendment requests approval to remove a shutdown requirement with regard to the relief valve position indication system in Section 3.13 of the Technical Specifications.

Date of publication of individual notice in Federal Register: August 2, 2000 (65 FR 47520).

Expiration date of individual notice: September 1, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

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

Date of amendment request: July 21, 2000.

Description of amendment request: The proposed amendment revises the Oyster Creek Nuclear Generating Station Technical Specifications Section 3.13 to remove a shutdown requirement with regard to the relief valve position indication system.

Date of issuance: August 21, 2000.

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

Amendment No.: 214.

Facility Operating License No. DPR-16: This amendment revised the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration: Yes (65 FR 47520) August 2, 2000. That notice provided an opportunity to submit comments on the Commission's proposed no significant hazards consideration determination. No comments have been received. The notice also provided for an opportunity to request a hearing by September 1, 2000, but indicated that if the Commission makes a final no significant hazards consideration determination any such hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment finding of exigent circumstances, state consultation, and final determination of no significant hazards consideration determination are contained in a Safety Evaluation dated August 21, 2000.

Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, LLP, 1800 M Street, N.W., Washington, D.C. 20036-5869.

NRC Section Chief: Marsha Gamberoni.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 0-530, Palo Verde Nuclear Generating Station, Units Nos. 1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: June 6, 2000.

Brief description of amendments: The amendments revise the information in Figure 3.5.5-1, “Minimum Required RWT Volume in TS 3.5.5, Refueling Water Tank (RWT),” for the three units. The amendments relocate design information to the Bases of the TSs, truncate the lower end of the RWT limit curve at 210 °F, retitle the right-hand ordinate from “minimum useful volume required in the RWT” to “RWT Volume,” and delete the two footnotes and the references to the footnotes.

Date of issuance: August 18, 2000.

Effective date: August 18, 2000, to be implemented within 45 days of the date of issuance.

Amendment Nos.: Unit 1-127, Unit 2-127, Unit 3-127.

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

Date of initial notice in Federal Register: July 12, 2000 (65 FR 43043).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 18, 2000.

No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power Station, Unit 2, Grundy County, Illinois

Date of application for amendment: April 30, 1999.

Brief description of amendment: The amendment revised the expiration date of the operating license to allow 40 years of operation from the original date of issuance of the Provisional Operating License.

Date of issuance: August 24, 2000.

Effective date: August 24, 2000.

Amendment No.: 178.

Facility Operating License No. DPR-19: The amendment revised the Facility Operating License. Date of initial notice in Federal Register: March 22, 2000 (65 FR 15376).

The Commission's related evaluation of the amendment is contained in an Environmental Assessment dated June 1, 2000, and a Safety Evaluation dated August 24, 2000.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

Date of application for amendment: November 18, 1999, as supplemented by letter dated June 7, 2000.

Brief description of amendment: The amendment changes Technical Specification 5.5.7, “Ventilation Filter Testing Program (VFTP)” to include the requirement for laboratory testing of engineered safety feature ventilation system charcoal samples per American Society for Testing and Materials D3803-1989 and the application of a safety factor of 2.0 to the charcoal filter efficiency assumed in the plant design-basis dose analyses.

Date of issuance: August 25, 2000.

Effective date: August 25, 2000.

Amendment No.: 167.

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

Date of initial notice in Federal Register: December 29, 1999 (64 FR 73088).

The June 7, 2000, supplemental letter 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 amendment is contained in a Safety Evaluation dated August 25, 2000.

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: April 9, 1998, as supplemented by letters dated January 13, 1999, and June 28, 2000.

Brief description of amendment: The amendment consists of changes to the River Bend Station (RBS) Facility Operating License, paragraph 2.C(13). The amendment allows RBS to operate with final feedwater temperature reduction in order to extend the fuel cycle by maintaining the core thermal power at or close to rated power, thus delaying the start of normal coastdown. The January 13, 1999, letter provided a revised proprietary version of the licensee's analysis submitted in its original April 9, 1998, application and the June 28, 2000, letter provided additional information to support staff review of the original application, and did not affect the initial finding of no significant hazards consideration determination dated May 20, 1998 (63 FR 27762).

Date of issuance: August 22, 2000.

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

Amendment No.: 112.

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

Date of initial notice in Federal Register: May 20, 1998 (63 FR 27762).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 22, 2000.

No significant hazards consideration comments received: No.

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

Date of amendment request: July 14, 1999, as supplemented by letters dated February 24, 2000, and July 17, 2000.

Brief description of amendments: The proposed amendments delete requirements from the Technical Specifications to maintain a Post Accident Sampling System (PASS). Licensees were required to implement PASS upgrades as a result of NUREG-0737, “Clarification of TMI [Three Mile Island Nuclear Station] Action Plan Requirements,” and Regulatory Guide 1.97, Revision 3, “Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environmental Conditions During and Following an Accident.” Implementation of these upgrades were an outcome of the Nuclear Regulatory Commission's lessons learned from the accident that occurred at TMI, Unit 2. The staff has concluded that the information obtained using PASS is not required for the development of protective action recommendations or for core damage assessment.

Date of issuance: August 17, 2000.

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

Amendment Nos.: 208 and 218

Facility Operating License Nos. DPR-51 and NPF-6: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR 43773). The supplements dated February 24 and July 17, 2000, did not change the scope of the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 17, 2000.

No significant hazards consideration comments received: No

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: June 1, 2000, as supplemented by letter dated June 30, 2000.

Brief description of amendment: The amendment approves a proposed modification that changes the Perry Nuclear Power Plant as described in the Updated Safety Analysis Report by installing inflatable seals that surround the Emergency Service Water (ESW) alternate intake sluice gates. This modification is necessary so that the licensee may use inflatable seals to minimize leakage of warm water into the ESW forebay from the Service Water discharge and thus maintain the ESW temperature below the design limit.

Date of issuance: August 22, 2000

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

Amendment No.: 114

Facility Operating License No. NPF-58: This amendment authorizes revision of the Updated Safety Analysis Report.

Date of initial notice in Federal Register: June 14, 2000 (65 FR 37414) The supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register Notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 22, 2000.

No significant hazards consideration comments received: No

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of application for amendment: May 10, 1999, as supplemented April 6, April 26, and June 5, 2000.

Brief description of amendment: Changes Technical Specifications to establish the actions to be taken for an inoperable “Standby Filter Unit” (SFU) System due to a degraded control building boundary.

Date of issuance: August 11, 2000

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

Amendment No.: 233

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

Date of initial notice in Federal Register: July 14, 1999 (64 FR 38029). The April 6, April 26, and June 5, 2000, submittals provided additional clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the application beyond the initial notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 11, 2000.

No significant hazards consideration comments received: No

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: June 7, 2000

Brief description of amendment: The amendment revised the Technical Specifications, Section 3.10.8, “SHUTDOWN MARGIN (SDM) Test — Refueling,” correcting an administrative error introduced when Amendment No. 92, dated March 2, 2000, was issued.

Date of issuance: August 24, 2000

Effective date: As of the date of issuance to be implemented concurrently with Amendment No. 92.

Amendment No.: 93

Facility Operating License No. NPF-69: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 16, 2000 (65 FR 37807)

The staff's related evaluation of the amendment is contained in a Safety Evaluation dated August 24, 2000.

No significant hazards consideration comments received: No

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

Date of application for amendment: February 1, 2000, as supplemented on April 13, 2000

Brief description of amendment: The amendment temporarily suspends the technical (TSs) requirements for TSs 3.7.7 and 3.7.8 in order to conduct testing of the cable spreading room that will pressurize the area to a pressure that exceeds the adjacent control room envelope area.

Date of issuance: August 22, 2000

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

Amendment No.: 181

Facility Operating License No. NPF-49: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: May 31, 2000 (65 FR 34748)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 22, 2000.

No significant hazards consideration comments received: No

Northern States Power Company, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: October 29, 1999, as supplemented March 14 and April 25, 2000

Brief description of amendment: The amendment conforms the license to reflect the transfer of possession under Operating License No. DPR-22 to a newly formed utility operating company subsidiary of Northern States Power Company merged with New Century Energies, Inc., as approved by Order of the Commission dated May 12, 2000.

Date of issuance: August 18, 2000

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment No.: 111

Facility Operating License No. DPR-22. Amendment revised the Operating License.

Date of initial notice in Federal Register: February 10, 2000 (65 FR 6641)

The March 14 and April 25, 2000, supplements were within the scope of the initial application as originally noticed.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 12, 2000.

No significant hazards consideration comments received: No

Northern States Power Company, Docket No. 50-263, Monticello Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: February 29, 2000, as supplemented July 10, 2000

Brief description of amendment: The amendment (1) approves continued use of two exceptions previously granted by the Nuclear Regulatory Commission (NRC) to the American Society of Mechanical Engineers N510-1989 testing requirements for the emergency filtration train (EFT) system, (2) revises the Technical Specifications (TSs) to reflect modifications to the EFT system that eliminate the need for additional test exceptions, (3) revises the TSs to be consistent with the guidance of NRC Generic Letter 99-02, and (4) revises the TSs to include operability requirements for the EFT system during operations that could result in a fuel handling accident.

Date of issuance: August 18, 2000

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment No.: 112

Facility Operating License No. DPR-22. Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: April 5, 2000 (65 FR 17917)

The July 10, 2000, supplemental letter provided clarifying information that was within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 18, 2000.

No significant hazards consideration comments received: No

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, and Docket No. 72-10, Prairie Island Independent Spent Fuel Storage Installation, Goodhue County, Minnesota

Date of application for amendments: October 29, 1999, as supplemented March 14 and April 25, 2000.

Brief description of amendments: The amendments conform the licenses to reflect the transfer of possession under Operating Licenses Nos. DPR-42 and DPR-60 and Materials License No. SNM-2506 to a newly formed utility operating company subsidiary of Northern States Power Company merged with New Century Energies, Inc., as approved by Order of the Commission dated May 12, 2000.

Date of issuance: August 18, 2000.

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment Nos.: 154 and 145.

Facility Operating Licenses Nos. DPR-42 and DPR-60 and Materials License No. SNM-2506: Amendments revised the Operating Licenses and Materials License.

Date of initial notice in Federal Register: February 10, 2000 (65 FR 6642)

The March 14 and April 25, 2000, supplements were within the scope of the initial application as originally noticed.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 12, 2000.

No significant hazards consideration comments received: No

PECO Energy Company, Public Service Electric and Gas Company Delmarva Power and Light Company; and Atlantic City Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: August 11, 1999, as supplemented June 29, 2000.

Brief description of amendments: The Updated Final Safety Analysis Report (USFAR) was updated to reflect credit for use of a limited amount of containment overpressure in calculations of net positive suction head available for emergency core cooling pumps.

Date of issuance: August 14, 2000.

Effective date: As of Date of issuance.

Amendments Nos.: 233 and 237.

Facility Operating License Nos. DPR-44 and DPR-56: The amendments authorized changes to the UFSAR.

Date of initial notice in Federal Register: April 19, 2000 (65 FR 21038). The June 29, 2000, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 14, 2000.

No significant hazards consideration comments received: No

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of application for amendment: June 4, 1999, as supplemented October 22, 1999.

Brief description of amendment: The amendment revises the license and Technical Specifications to reflect changes related to the transfer of the license for the Hope Creek Generating Station, to the extent held by Public Service Electric and Gas Company, to PSEG Nuclear Limited Liability Company.

Date of issuance: August 21, 2000

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

Amendment No.: 129

Facility Operating License No. NPF-57: This amendment revised the License and the Technical Specifications.

Date of initial notice in Federal Register: June 30, 1999 (64 FR 35193). The October 22, 1999, supplement provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 16, 2000.

No significant hazards consideration comments received: No

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of application for amendments: April 13, 2000

Brief description of amendments: The amendments deleted Technical Specification (TS) 3/4.1.3.2.2 which is related to shutdown and control rod group demand position indication in Modes 3, 4, and 5.

Date of issuance: August 17, 2000

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

Amendment Nos.: 232 and 213

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 28, 2000 (65 FR 39960)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 17, 2000.

No significant hazards consideration comments received: No

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of application for amendments: June 4, 1999, as supplemented October 22, 1999.

Brief description of amendments: The amendment revises the license and Technical Specifications to reflect changes related to the transfer of the license for the Salem Nuclear Generating Station, Unit Nos. 1 and 2, to the extent held by Public Service Electric and Gas Company, to PSEG Nuclear Limited Liability Company.

Date of issuance: August 21, 2000

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

Amendment Nos.: 233 and 214

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the License and Technical Specifications.

Date of initial notice in Federal Register: June 30, 1999 (64 FR 35192). The October 22, 1999, supplement provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 16, 2000.

No significant hazards consideration comments received: No

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of application for amendments: January 24, 2000, as supplemented April 19 and May 31, 2000.

Brief description of amendments: The amendments revise the radiological effluent technical specifications (RETS) and administrative controls requirements (i.e., Sections 3/4.3, Instrumentation, 3/4.11, Radioactive Effluents, 3/4.12, Radiological Environmental Monitoring, 6.0, Administrative Controls, and the table of contents and definitions) in the Technical Specifications (TSs) by implementing programmatic controls for RETS in the administrative controls section and relocating procedural details of the RETS, with various changes, to the offsite dose calculation manual (ODCM) or to the process control program (PCP). The proposed changes follow the guidance and requirements in NRC Generic Letter 89-01, “Implementation of Programmatic Controls in the Technical Specifications for Radiological Effluent Technical Specifications (RETS) in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program,” that was issued in 1989. There is also the change to add the word “oxygen” to the title of “Radioactive Gaseous Effluent Monitoring Instrumentation.”

Date of issuance: August 24, 2000

Effective date: August 24, 2000

Amendment Nos.: 234 and 215

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 1, 2000 (65 FR 11094) The supplemental letters dated April 19 and May 31, 2000, provided clarification that did not alter the scope of the proposed action or the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 24, 2000.

No significant hazards consideration comments received: No

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Date of application for amendments: August 24, 1999, as supplemented on December 29, 1999, and June 16, 2000

Brief description of amendments: The amendments revised Technical Specification 3.3.2 “Engineered Safety Features Actuation System (ESFAS) Instrumentation” to relax the slave relay test frequency from quarterly to every refueling not to exceed 18 months.

Date of issuance: August 22, 2000

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

Amendment Nos.: 114 and 92

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 22, 2000 (65 FR 15386). The supplemental letters dated December 29, 1999, and June 16, 2000, provided clarifying information only, and did not change the scope of the August 24, 1999, 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 August 22, 2000.

No significant hazards consideration comments received: No

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

Date of application for amendment: March 6, 2000

Brief description of amendment: Revised the Technical Specification (TS) and associated Bases for Limiting Condition for Operation 3.9.4, “Refueling Operations—Containment Penetrations,” to allow the containment personnel airlock doors and certain containment penetrations to be open during refueling activities under appropriate administrative controls.

Date of issuance: August 24, 2000

Effective date: August 24, 2000

Amendment No.: 26

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

Date of initial notice in Federal Register: May 17, 2000 (65 FR 31361)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 24, 2000.

No significant hazards consideration comments received: No

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 amendments: June 22, 2000, as supplemented July 25, 2000

Brief description of amendments: The amendments revise the Technical Specifications Sections 3.4.1.4, 3.4.1.6, 4.4.1.4, and 4.4.1.6.1; add Sections 4.4.1.6.4 and 4.4.1.6.5; and revise Bases Section 3/4.4.1 for Units 1 and 2. These changes will allow for the implementation of a vacuum-assisted backfill technique when returning an isolated Reactor Coolant System (RCS) loop to service, and provide the necessary controls for temperature and boron concentration of the isolated RCS loop to ensure the required shutdown margin is maintained.

Date of issuance: August 25, 2000

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

Amendment Nos.: 223 and 204

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

Date of initial notice in Federal Register: July 26, 2000 (65 FR 46019). The letter dated July 25, 2000, contained clarifying information only, and did not change the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 25, 2000.

No significant hazards consideration comments received: No

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

Date of application for amendments: May 19, 2000, as supplemented August 3, 2000

Brief description of amendments: These amendments eliminate one of the license conditions and associated implementation dates from Appendix C to the licenses. The license condition required the licensee to submit a license amendment application and supporting radiological dose analyses demonstrating compliance with General Design Criterion 19 dose limits without reliance on potassium iodide.

Date of issuance: August 15, 2000

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment Nos.: 198 and 203

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

Date of initial notice in Federal Register: June 6, 2000 (65 FR 35966)

The August 3, 2000, supplemental letter provided clarifying information that was within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 15, 2000.

No significant hazards consideration comments received: No

Dated at Rockville, Maryland, this 30th day of August 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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