Operating licenses, amendments; no significant hazards considerations; biweekly notices,

[Federal Register: January 12, 2000 (Volume 65, Number 8)]

[Notices]

[Page 1918-1934]

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

[DOCID:fr12ja00-104]

NUCLEAR REGULATORY COMMISSION

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

  1. 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 December 13, 1999, through December 31, 1999. The last biweekly notice was published on December 29, 1999 (64 FR 73083).

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

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4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By February 11, 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 filedin accordance with the Commission's ``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and 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 filedby the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

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

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

A request for a hearing or a petition for leave to intervene must be filedwith the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 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. Where petitions are filedduring the last 10 days of the notice period, it is requested that the petitioner promptly so inform the Commission by a toll-free telephone call to Western Union at 1-(800) 248-5100 (in Missouri 1-(800)-342-6700). The Western Union operator should be given Datagram Identification Number N1023 and the following message addressed to (Project Director): petitioner's name and telephone number, date petition was mailed, plant name, and publication date and page number of this Federal Register notice. 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).

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Date of amendment request: October 25, 1999.

Description of amendment request: The proposed amendment would revise the Technical Specification allowable values for the reactor protection system electric power monitoring assembly overvoltage and undervoltage trip setpoints.

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

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    consequences of any accident previously evaluated.

    The proposed Technical Specification (TS) change revises the Reactor Protection System (RPS) Electric Power Monitoring Assembly overvoltage and undervoltage Allowable Values. The new Allowable Values and setpoints will continue to provide adequate margin to the normal operating voltage range for the RPS and MSIV [main steam isolation valve] solenoids, thus minimizing the potential for inadvertent trips. The proposed change does not have a detrimental impact on the condition or performance of any plant structure, system, or component that may initiate an analyzed event. The proposed change does not physically impact the plant nor does it impact any design or functional requirements of the associated system. That is, the proposed change does not degrade the performance or increase the challenges of any safety systems assumed to function in the accident analysis. Further, the proposed change does not impact the Surveillance Requirements themselves nor the way in which the Surveillances are performed. Consequently, the probability of an accident previously evaluated is not significantly increased.

    Additionally, the proposed change does not effect the affect the availability of equipment or systems required for mitigating the consequences of an accident. The revision of the overvoltage and undervoltage setpoints will ensure that the associated trip functions continue to protect the RPS scram solenoids and main steam isolation valve (MSIV) solenoids so that these devices will perform their intended safety function. Thus, the affected equipment is still required to be maintained Operable and capable of performing the accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected.

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

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

    The proposed TS change revises the Reactor Protection System (RPS) Electric Power Monitoring Assembly overvoltage and undervoltage Allowable Values. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The revised setpoints will continue to ensure that the RPS bus would be disconnected from its power supply under specified conditions that could damage the RPS bus powered equipment. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

    The proposed TS change revises the Reactor Protection System (RPS) Electric Power Monitoring Assembly overvoltage and undervoltage Allowable Values. The proposed change provides necessary conservatism in the Allowable Values in the RPS Surveillance Requirement to ensure that the equipment used to meet the Limiting Condition for Operation (i.e., each of the two electric power monitoring assemblies) can continue to perform its required functions. At the same time, the revised setpoint/Allowable Values continue to provide adequate margin to the expected operating voltage range to prevent inadvertent or unnecessary tripping of the electric power monitoring assemblies (thus preventing unnecessary or excessive transfer to the alternate power source). The affected equipment will thus continue to be tested (calibrated and functionally tested) in a manner that gives confidence that the equipment can perform its assumed safety 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: Kevin P. Gallen, Morgan, Lewis & Bockius LLP, 1800 M Street, NW, Washington, DC 20036.

    NRC Section Chief: Anthony J. Mendiola.

    AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

    Date of amendment request: October 25, 1999.

    Description of amendment request: The proposed amendment would revise the Technical Specification definitions for channel calibrations, channel functional tests, and logic system functional tests in accordance with Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-205, Revision 3, ``Revision of Channel Calibration, Channel Functional Test, and Related Definitions.''

    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:

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

    The proposed change clarifies the Technical Specification requirements for performance of channel calibrations, channel functional test, and logic system functional tests. Specifically, the proposed change incorporates the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-205, Revision 3, ``Revision of Channel Calibration, Channel Functional Test, and Related Definitions.'' The change approved per this TSTF is not expected to adversely affect the performance and effectiveness of required testing as testing appropriate to the associated Surveillance Requirements will continue to be performed. The proposed change does not have a detrimental impact on the condition or performance of any plant structure, system, or component that initiates an analyzed event. Consequently, the probability of an accident previously evaluated is not significantly increased. The equipment being tested is still required to be operable and capable of performing the accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected. Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

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

    The scope of the proposed change is limited to the clarification of existing test requirements. As such, the proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

    As noted above, the proposed change clarifies requirements for the performance of channel calibrations, channel functional tests, and logic system functional tests. Specifically, the proposed change incorporates the NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-205, Revision 3, ``Revision of Channel Calibration, Channel Functional Test, and Related Definitions.'' No changes or setpoints to plant process limits are involved. The surveillance requirements as revised will continue to ensure that affected equipment is tested in a manner that gives confidence that the equipment can perform its appropriate safety 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: Kevin P. Gallen, Morgan, Lewis & Bockius LLP, 1800 M Street, NW, Washington, DC 20036.

    NRC Section Chief: Anthony J. Mendiola.

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    AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

    Date of amendment request: December 16, 1999.

    Description of amendment request: The proposed amendment would allow a one-time extension of some Technical Specification surveillance intervals to support elimination of a planned spring 2000 mid-cycle outage (PO-8). For the applicable surveillances, the licensee proposes to extend their current surveillance intervals to November 30, 2000, the scheduled startup date from refueling outage 7 (RF-7).

    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:

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

    The proposed Technical Specification (TS) changes involve a one- time only change in the surveillance test intervals of selected Surveillance Requirements (SRs). The proposed TS changes do not impact the TS surveillance performance requirements themselves nor the way in which the surveillances are performed. The proposed TS changes do not physically involve any changes to the plant, nor do they impact any design or functional requirements of the associated systems. Thus, the proposed TS changes do not increase the challenges of any safety systems assumed to function in the accident analysis.

    In addition, the proposed TS changes do not significantly affect the availability of equipment or systems required to mitigate the consequences of an accident because (1) extension of the test intervals to the extent requested is not expected to have a significant impact on availability (i.e., no extended test interval would exceed 30 months), and (2) other or more frequent testing performed for the affected systems or components, as well as for redundant systems or components, supports continued availability of the affected function. The equipment subject to testing per the affected SRs is still required to be operable and capable of performing any accident mitigation functions assumed in the accident analysis. Furthermore, a historical review of surveillance test results identified no failures that would invalidate these conclusions.

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

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

    The proposed TS changes involve a one-time only change in the surveillance testing intervals of selected SRs. Such changes do not introduce any failure mechanisms of a different type than those previously evaluated since there are no physical changes being made to the facility. In addition, the surveillance test requirements themselves, and the way surveillance tests are performed, will remain unchanged. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated.

  9. The proposed changes do not involve a significant reduction in a margin of safety.

    The one-time extended surveillance frequencies do not result in a significant reduction in the margin of safety. Although the proposed TS changes will result in an increase in the interval between surveillance tests, the impact, if any, on system availability is small. This is because, as noted previously, extension of the test intervals to the limited extent proposed would not be expected to have a significant impact on availability. Other or more frequent testing performed for the affected systems or components, as well as the testing performed for redundant systems or components, supports continued availability of the affected functions.

    In addition, the proposed changes do not involve any physical changes to the affected systems or components, nor do they involve any changes to setpoints, operating limits, or safety limits.

    Based on the above, the assumptions in the licensing basis are not impacted, and the proposed TS changes do not significantly reduce 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: Kevin P. Gallen, Morgan, Lewis & Bockius LLP, 1800 M Street, NW, Washington, DC 20036.

    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: September 17, 1999.

    Description of amendment request: The proposed change to the Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Technical Specifications would lower the maximum limit for contents of the gaseous radioactive system from 300,000 curies (Ci) to 78,782 Ci and 82,400 Ci for ANO-1 and ANO-2, respectively. This limit would ensure that, upon an uncontrolled release of the tank's contents over a 2-hour period, the resulting total whole body exposure to a member of the public at the nearest exclusion area boundary would not exceed 0.5 roentgen equivalent man (rem).

    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:

    An evaluation of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92(c). A discussion of these standards as they relate to this amendment request follows:

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

    The proposed change to lower the current technical specification (TS) gas storage tank activity limits does not require new hardware or physical equipment modifications to the plant design. By lowering the setpoint, the resultant exposure at the exclusion area boundary upon an inadvertent release of a gas storage tank's content will be limited to 0.5 rem. Therefore the consequences of such an uncontrolled release of activity are effectively reduced. Additionally, no new accident is introduced by the proposed reduction in activity limits associated with the gas storage tanks.

    Therefore, reducing the gas storage tank limits from 300,000 Curies (Ci) to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2, respectively) does not involve a significant increase in the probability or consequences of any accident previously evaluated.

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

    The proposed change affects the consequences of an event associated with the loss of gas storage tank radioactive contents on either ANO-1 or ANO-2. Since this event has been previously evaluated, no new or different accident can be associated with the proposed change. Decreasing the present TS activity limits results in an exposure at the exclusion area boundary to be limited to 0.5 rem in the event of an inadvertent release of a gas storage tank's content.

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

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

    The proposed change conservatively lowers the existing TS GRW

    [Gaseous Radwaste] System gas storage tank activity limits from 300,000 Ci to 78,782 Ci and 82,400 Ci (ANO-1 and ANO-2 respectively). In doing so, the resultant exposure to a member of the public at the exclusion area boundary during an inadvertent release of gas storage tank contents over a two-hour period is reduced to 0.5 rem or less. The proposed change, therefore, retains the margin of safety for both ANO-1 and ANO-2.

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

    Therefore, based on the reasoning presented above and the previous discussion

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    of the amendment request, Entergy Operations, Inc. has determined that the requested change does not involve a significant hazards consideration.

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

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

    NRC Section Chief: Robert A. Gramm.

    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: December 16, 1999.

    Description of amendment request: The proposed license amendment request would revise Fuel Handling Accident (FHA) dose calculations for 3 scenarios documented in the River Bend Station, Unit 1 (RBS), Updated Safety Analysis Report (USAR). The first is a FHA in the fuel building, assumed to occur 24 hours post-shutdown. A second FHA analysis was prepared to support Amendment 35 to RBS Technical Specifications (TS) which assumed a FHA occurs in the primary containment 80 hours post- shutdown during Local Leakage Rate Testing (LLRT). A third analysis was prepared in support of Amendment 85 to the RBS TS which assumed the containment is open at 11 days.

    These analyses are being updated to account for several changes. The primary reason for the revisions, as stated by the licensee, was to update the analyses to reflect current RBS operating strategies and make the analyses consistent with each other. Specifically, Cases 1 and 2 of the three analyses assumed a Radial Peaking Factor (RPF) of 1.5 consistent with Regulatory Guide (RG) 1.25. However, current core design strategies could lead to an RPF as high as 1.65. In addition, to account for the potential impact of extended burnup fuel in future operating cycles, an increased iodine-131 gap fraction of 0.12 was more conservatively assumed in lieu of the 0.10 recommended by RG 1.25. The revised analysis also includes a change to the control room atmospheric dispersion factors (/Q) for the Main Control Room (MCR) ventilation system. Credit is taken for Standard Review Plan (SRP) Section 6.4 guidance for manual dual control room air intakes in that the /Q's are divided by 4. The revised FHA analyses also credit this action at a 20 minute delay to be consistent with the Loss of Coolant Accident (LOCA) analysis.

    Furthermore, an error was discovered in one of the FHA calculations. The release rate assumed in the analysis did not ensure that the RG 1.25 assumption of a 2-hour release was preserved. The error is the result of an inherent bias in the secondary mixing effects in the dose calculation. The results continue to be bounded by the guidance contained in SRP 15.7.4 and RG 1.25.

    Reanalysis showed that the release rate error, compounded with the other changes discussed above, resulted in calculated doses greater than those currently found in the RBS USAR. In addition, some of the doses were also greater than those presented in the Amendment 85 submittal. However, the licensee has stated that the results of the revised analyses remain ``well within'' 10 CFR 100, the guidance contained in SRP 15.7.4, and RG 1.25. Since the analyses results are above those reported in the RBS USAR, the criterion of 10 CFR 50.59(a)(2)(i) is, therefore, satisfied. Accordingly, the licensee has concluded that these changes involve an unreviewed safety question.

    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:

  10. The proposed changes do not significantly increase the probability or consequences of an accident previously evaluated.

    The analyses changes described by this proposed change to the USAR are not initiators to events, and, therefore, do not involve the probability of an accident. The changes to the FHA calculations for radiological doses following a FHA reflect the current operating strategies and make the analyses consistent. These changes included:

    Accounting for the impact of extended burnup fuel,

    Addressing a change to the control room atmospheric dispersion factors assumed in the analysis, and

    Revising the Radial Peaking Factor (RPF) used in the analysis. Current core design strategies could lead to a RPF higher

    [than] that assumed in Regulatory Guide 1.25.

    The TRANSACT code is used for offsite dose and control room dose calculations. The TRANSACT code is derived from the TACT V code documented in

    NUREG/CR-5109. RBS has benchmarked the TRANSACT code as discussed in the request dated August 17, 1995, (RBG-41728) which resulted in the NRC granting Amendment 85.

    The revisions to the FHA are used to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. These operational conditions include:

    Initial fuel movement in the Fuel Building 24 hours after shutdown,

    Fuel movement in Primary Containment after 80 hours with leakrate testing being conducted, and

    Fuel movement in Primary Containment with the Primary Containment open.

    Because the analyses affected by the changes are not considered an initiator to any previously analyzed accident, these changes cannot increase the probability of any previously evaluated accident. Therefore, this change does not increase the probability of occurrence of an accident evaluated previously in the safety analysis report (SAR).

    This proposed change to the USAR does increase the consequences of an accident, but the increase is within all regulatory limits and guidance. While the calculated off-site and control room doses of a FHA did increase, the dose consequences remain below the regulatory limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC [General Design Criterion]-19 as approved per NUREG-0989, and the guidance contained in SRP 15.7.4 of less than 25% of the 10 CFR 100 limits. The cause of these events remains the failure of the fuel assembly lifting mechanism. These analyses demonstrated that for the worst case bundle drop, the regulatory dose guidelines of SRP 15.7.4 continue to be satisfied for the required decay periods.

    This change accounts for the potential effects of current fuel design and operating strategies including increased burnup of fuel, increased iodine-131 fraction released, Main Control Room ventilation system operation, and release rate timing assumptions. Reanalysis of the off-site dose calculation demonstrates that the revised doses are increased but remain less than the regulatory limits of 10 CFR 100 and within the guidance of SRP 15.7.4. Therefore, this change does not significantly increase the consequences of an accident previously evaluated in the SAR.

    The proposed changes, in conjunction with existing administrative controls, bound the conditions of the current design basis fuel handling accident analysis. The analysis also concludes the limiting offsite radiological consequences are well within the acceptance criteria of NUREG[-]0800, Section 15.7.4 and 10 CFR 50, Appendix A, GDC[-]19. The analysis is also conducted in a conservative manner containing margins in the calculation of mechanical analysis, iodine inventory, and iodine decontamination factor. Each of these conservatisms will further decrease the consequences. Therefore, the proposed changes do not significantly increase the probability or consequences of any previously evaluated accident.

  11. The proposed changes would not create the possibility of a new or different kind of accident from any previous[ly] analyzed.

    This change does not involve initiators to any events in the SAR, nor does the activity create the possibility for any new accidents. Rather, this change is a result of the

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    evaluation of the most limiting FHA, which can occur at River Bend.

    The proposed changes to the dose analyses are consistent with previous limits, only revising previous evaluations to account for current operating strategies and assumptions. These changes included:

    Accounting for the impact of extended burnup fuel,

    addressing a change to the control room atmospheric dispersion factors assumed in the analysis, and

    Revising the Radial Peaking Factor (RPF) used in the analysis. Current core design strategies could lead to a RPF higher

    [than] that assumed in Regulatory Guide 1.25.

    The radiological consequences remain within accepted limits of 10 CFR 100 and guidance of the Standard Review Plan (NUREG-0800) Section 15.7.4. Therefore, these changes are consistent with the design basis analysis. The proposed changes do not introduce any new modes of plant operation and do not involve physical modifications to the plant. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previous[ly] analyzed.

  12. The proposed changes do not involve a significant reduction in a margin of safety.

    The dose consequences are calculated in accordance with regulatory guidance found in Regulatory Guide 1.25 and the SRP

    [S] ection 15.7.4. The RBS analyses conservatively assumed that failures are consistent with those in the standard General Electric GESTAR II. These analyses result in a bounding number of fuel failures. The RBS analyses are also consistent with those approved by the NRC [Nuclear Regulatory Commission] in support of Technical Specification Amendments 35 and 85 to the River Bend Station license (NPF-47). The radiological dose consequences resulting from these failures are therefore analyzed using accepted methods and criteria. In addition, the analyses contain known conservatisms and margins to ensure the results will remain bounding.

    The revised limits are used to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis and are established such that the radiological consequences are at or below the current regulatory limits and guidance. Safety margins and analytical conservatisms have been evaluated and are well understood. Conservative methods of analysis are maintained through the use of accepted methodology and benchmarking the proposed methods to previous analysis. Margins are retained to ensure that the analysis adequately bounds all postulated event scenarios. The proposed change only eliminates some excess conservatism from the analysis.

    In addition, EOI [Entergy Operations, Inc.] has implemented NUMARC [Nuclear Management and Resources Council (now NEI)] 91-06 guidelines for shutdown operations at RBS. Shutdown Operations Protection Plan and Primary-Secondary Containment Integrity procedures presently include guidance for closure of the containment hatch and other significant openings in containment, in addition to the requirements contained in the license and design basis. This additional protection will enhance the ability to limit offsite effects.

    Acceptance limits for the fuel handling accident are provided in 10 CFR 100 with additional guidance provided in NUREG[-] 0800, Section 15.7.4. The proposed changes continue to ensure that the whole-body and thyroid doses at the exclusion area and low population zone boundaries, as well as control room doses, are below the corresponding regulatory limits. These margins are unchanged, therefore, the proposed changes do not involve a significant reduction in a margin of safety.

    The commission has provided guidance concerning the application of the standards of 10 CFR 50.92 by providing certain examples (51 FR 7751, March 6, 1986) of amendments that are not considered likely to involve a significant hazards consideration. This proposed amendment is very similar to example (vi):

    (vi) A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan: for example, a change resulting from the application of a small refinement of a previously used calculational model or design method.

    As we have shown in the preceding discussion, this refinement to the FHA dose calculation results in a small increase to the consequences of a previously analyzed accident, but the results of the change remain clearly within the guidelines of 10 CFR 100, Appendix A, GDC[-]19, and the guidance of SRP [S]ection 15.7.4, without reducing 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: Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005.

    NRC Section Chief: Robert A. Gramm.

    Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: November 17, 1999 (L-99-241)

    Description of amendment request: The proposed amendment would revise the St. Lucie Unit 1 and Unit 2 Technical Specifications (TS) to require laboratory testing of activated charcoal samples for applicable engineered safety feature ventilation systems using the American Society for Testing and Materials (ASTM) D3803-1989 protocol. The affected TS are Units 1 and 2 shield building ventilation system, TS 4.6.6.1; Unit 1 emergency core cooling system area ventilation system, TS 4.7.8.1; Unit 1 control room emergency ventilation system, TS 4.7.7.1; Unit 2 control room emergency air cleanup system, TS 4.7.7; and Unit 1 fuel pool ventilation system--fuel storage, TS 4.9.12.

    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:

  13. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

    The proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated. The new charcoal testing protocol is performed offsite on samples extracted from the safety-related ventilation systems. Therefore, there is no impact on any accident initiator and therefore, no changes on the probability. The proposed testing protocol is more conservative than previous tests; therefore, the efficiency of charcoal for the affected safety-related systems would not be overestimated. With the new testing protocol, more conservative testing results are expected since the temperature at which testing is performed is lower and the charcoal retention capability is more consistent with actual accident conditions. The proposed change thus ensures that the charcoal in service will comply with the penetration requirements to meet the design basis accident conditions.

    Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

  14. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

    The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed new charcoal testing protocol only affects surveillance testing requirements for ventilation systems. The functions of these systems remain unchanged and unaffected. No new system interactions have been introduced by the proposed amendment, which would create a new or different type of accident than previously analyzed. No physical changes are being made to any structure, system or component. The operation of the facility has not been altered by the proposed amendment. The systems involved are not considered to initiate any accidents as previously evaluated.

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    The proposed amendment will not change the physical plant or the modes of operation defined in the facility license. The changes do not involve the addition of new equipment or the modification of existing equipment, nor do they alter the design of St. Lucie plant systems. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

  15. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

    The proposed amendment does not involve a reduction in the margin of safety. The margin of safety of the Technical Specifications, its bases, the Final Safety Analysis Report, the Safety Evaluation Report or in any other design document has not been affected by the proposed amendment. The change provided in this proposed amendment is related to introducing an improved testing protocol for the activated charcoal in safety related ventilation systems. The change consists of testing the charcoal with a new testing protocol and with lower test temperatures to resemble more closely accident conditions and to eliminate potential overestimation of charcoal efficiency.

    Therefore, operation of the facility in accordance with the proposed amendment would 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

    NRC Section Chief: Richard P. Correia

    GPU Nuclear, Inc., et al., Three Mile Island Nuclear Station, Unit 2 (TMI-2), Docket No. 50-320, Dauphin County, Pennsylvania

    Date of amendment request: November 5, 1999.

    Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) Sections 6.8.1.4, 6.5.4.6, 6.13, 6.14, and 6.8.3. Specifically, Sections 6.13, 6.14 and 6.8.3 would be revised to eliminate the requirement to notify the Nuclear Regulatory Commission (NRC) of exceeding environmental limits and changes to environmental permits such as National Pollution Discharge Elimination System (NPDES). The requirements contained in the individual environmental permits and program regulations administered by the U.S. Environment Protection Agency (EPA), Pennsylvania Department of Environmental Protection (PADEP), and other regulatory agencies with program jurisdiction for reporting are included in plant procedures and data base tracking systems. Sections 6.8.1.4 and 6.5.4.6 are changes to the amendment that are administrative in nature and reflect a streamlining of the GPU Nuclear, Inc. management structure.

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

  16. The proposed changes to the TMI-2 [Three Mile Island, Unit 2] Technical Specifications do not involve a significant increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously analyzed in the safety analysis report. The changes have no impact on plant operations or the release of radioactive materials.

  17. The proposed changes to the TMI-2 Technical Specifications will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report because no plant configuration or operational changes are involved.

  18. The changes will not involve a significant reduction in the margin of safety as defined in the basis for any technical specification for TMI-2 because no change to operational limits will be made.

    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: Ernest L. Blake, Jr Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N. Street, N.W., Washington, DC 20037.

    NRC Section Chief: Mike Masnik.

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

    Date of amendment request: November 10, 1999.

    Description of amendment request: The proposed amendment would revise Technical Specification (TS) 5.5.7.c, to commit to the American Society for Testing and Materials (ASTM) D3803-1989 test protocol for the ventilation filter testing program. The proposed changes are consistent with Attachment 2, Sample Technical Specifications, in Generic Letter 99-02. Because the current TS penetration limits do not reflect a safety factor in excess of that assumed in the dose calculations of the accident analysis, the TS change request would also revise the allowable penetration values to correspond to a safety factor of 2.

    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 for the administrative changes:

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

    The ESF [engineered safety feature] ventilation systems are not initiators of any accident previously evaluated and the change in testing protocol to ASTM D3803-1989 as requested by the NRC will be more accurate and realistic and provide greater assurance of consistency. The acceptance criteria will be more conservative than those currently used in TS 5.5.7.c.

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

    No new types of accidents are being introduced because no modifications or changes in operations are being proposed for the ESF [engineered safety feature] ventilation systems. The proposed changes to TS 5.5.7.c impact acceptance criteria and test protocols only.

  21. The proposed amendment will not involve a significant reduction in a margin of safety.

    The margin of safety is not reduced. The proposed change in ESF ventilation testing protocol includes a safety factor of two (2) for the penetration limit in excess of that assumed in the dose calculations of the DAEC [Duane Arnold Energy Center] accident analysis.

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

    Attorney for licensee: Al Gutterman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: November 22, 1999.

    Description of amendment request: The proposed amendment would adopt selected NRC-approved generic changes to the Improved Technical Specifications (ITS) NUREGs. The 16 changes come from the Technical Specification Task Force (TSTF) process

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    developed by the industry and the NRC. Three of these changes are Bases-only changes but are included for completeness relative to the TSTF process.

    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 for the administrative changes:

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

    The proposed change involves reformatting, renumbering, and rewording the existing Technical Specifications. The reformatting, renumbering, and rewording process involves no technical changes to the existing Technical Specifications. As such, this change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accident or transient events. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

  23. 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 a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose any new or eliminate any old requirements. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

    The proposed change will not reduce a margin of safety because it has no effect on any safety analyses assumptions. This change is administrative in nature. Therefore, the change does not involve a significant reduction in a margin of safety.

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

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

    The proposed change provides more stringent requirements for operation of the facility. These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

  26. 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 a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change does impose different requirements. However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

    The imposition of more restrictive requirements either has no effect on or increases the margin of plant safety. As provided in the justification, each change in this category is, by definition, providing additional restrictions to enhance plant safety. The change maintains requirements within the safety analyses and licensing basis. Therefore, the change does not involve a significant reduction in a margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below for less restrictive changes--removed detail:

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

    The proposed change relocates certain details from the Technical Specifications to other documents under regulatory control. The Bases, UFSAR [updated final safety analysis report], and Technical Requirements Manual will be maintained in accordance with 10 CFR 50.59. In addition to 10 CFR 50.59 provisions, the Technical Specification Bases are subject to the change control provisions in the Administrative Controls Chapter of the Technical Specification. The UFSAR is subject to the change control provisions of 10 CFR 50.71(e). Other documents are subject to controls imposed by Technical Specifications or regulations. Since any changes to these documents will be evaluated, no significant increase in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

  29. 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 a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not impose or eliminate any requirements and adequate control of the information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

    The proposed change will not reduce a margin of safety because it has no effect on any safety analyses assumptions. In addition, the details to be moved from the Technical Specifications to other documents are the same as the existing Technical Specifications. Since any future changes to these details will be evaluated, no significant reduction in a margin of safety will be allowed. A significant reduction in the margin of safety is not associated with the elimination of the 10 CFR 50.92 requirement for NRC review and approval of future changes to the relocated details. The proposed change is consistent with the BWR [Boiling Water Reactor]/4 Standard Technical Specifications, NUREG-1433, issued by the NRC Staff, revising the Technical Specifications to reflect the approved level of detail, which indicates that there is no significant reduction in the margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below for less restrictive changes--category 3, relaxation of completion time:

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

    The proposed change relaxes the Completion Time for a Required Action. Required Actions and their associated Completion Times are not initiating conditions for any accident previously evaluated and the accident analyses do not assume that required equipment is out of service prior to the analyzed event. Consequently, the relaxed Completion Time does not significantly increase the probability of any accident previously evaluated. The consequences of an analyzed accident during the relaxed Completion Time are the same as the consequences during the existing Completion Time. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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

    The relaxed Completion Time for a Required Action does not involve a significant reduction in the margin of safety. As provided in the justification, the change has been evaluated to ensure that the allowed Completion Time is consistent with the safe operation under the specified Condition, considering the operability status of the

    [[Page 1926]]

    redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA [design basis accident] occurring during the repair period. Therefore, this change does not involve a significant reduction in a margin of safety.

    As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below for less restrictive changes--category 4, relaxation of required action.

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

    The proposed change relaxes Required Actions. Required Actions and their associated Completion Times are not initiating conditions for any accident previously evaluated and the accident analyses do not assume that required equipment is out of service prior to the analyzed event. Consequently, the relaxed Required Actions do not significantly increase the probability of any accident previously evaluated. The Required Actions in the change have been developed to provide assurance that appropriate remedial actions are taken in response to the degraded condition, considering the operability status of the redundant systems of required features, and the capacity and capability of remaining features while minimizing the risk associated with continued operation. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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

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

    As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below for less restrictive changes--category 6, relaxation of surveillance requirement acceptance criteria:

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

    The proposed change relaxes the acceptance criteria of Surveillance Requirements. Surveillances are not initiators to any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The equipment being tested is still required to be Operable and capable of performing the accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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

    The relaxed acceptance criteria for Surveillance Requirements do not result in a significant reduction in the margin of safety. As provided in the justification, the relaxed Surveillance Requirement acceptance criteria have been evaluated to ensure that they are sufficient to verify that the equipment used to meet the LCO

    [limiting condition for operation] can perform its required functions. Thus, appropriate equipment continues to be tested in a manner that gives confidence that the equipment can perform its assumed safety 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: Al Gutterman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment requests: September 3, 1998, as supplemented by letters dated January 22, February 5, March 17, and November 24, 1999. The September 3, 1998, amendment application was previously noticed in the Federal Register on December 16, 1998 (63 FR 69345).

    Description of amendment requests: The amendment would revise Section 5.6.6, ``Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR),'' of the improved Technical Specifications (TSs), that were issued in Amendment Nos. 135 and 135 on May 28, 1999. The amendment would add the phrase ``and LTOP'' (low-temperature overpressure protection) to the first sentence of item 5.6.6.b that identifies the limits that can be determined by the licensee in the PTLR, and (2) replace the current list of documents listed in item 5.6.6.b by the NRC letter that would approve this amendment and Westinghouse WCAP-14040-NP-A, ``Methodology Used to Develop Cold Overpressure Mitigation System Setpoints and RCS Heatup and Cooldown Limit Curves,'' dated January 1996. WCAP-14040-NP-A is the NRC-approved topical report which provides a methodology for developing the LTOP setpoints and RCS heatup and cooldown limit curves for Westinghouse plants, such as Diablo Canyon Nuclear Power Plant, Units 1 and 2.

    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:

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

    The proposed changes to Figures 3.4-2 and 3.4-3 of Technical Specification (TS) 3.4.9.1 and the associated Bases adjust the reactor coolant system (RCS) heatup and cooldown pressure/ temperature (P/T) limits to permit operation through 16 effective full power years (EFPY). The 16 EFPY P/T limits are more restrictive than the current limits; this accounts for an expected incremental increase in reactor vessel embrittlement, and assures the reactors will continue to be operated within acceptable stresses and at temperatures for which the reactor vessel metal exhibits ductile properties. The P/T limits developed for 16 EFPY were determined in accordance with 10 CFR 50, Appendix G, and maintain the same margins of safety as the current limits. The proposed changes will not impact the probability of overpressurization or brittle fracture of the vessel, and therefore will not impact the consequences of an accident.

    The present low temperature overpressure protection (LTOP) pressure and enable temperature setpoints were reviewed and found to be acceptable and conservative for

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    use through 16 EFPY, based on use of ASME [American Society of Mechanical Engineers] Code Case N-514, which provides acceptable margins to the prevention of vessel overpressurization and brittle fracture. Therefore, there is no change to the consequences of accidents previously analyzed. Since no changes are proposed in the actual LTOP setpoints, nor any physical alteration of the LTOP system, nor a change to the method by which the LTOP system performs its function, there would be no change to the probability of an accident previously evaluated. The proposed change to the Bases incorporates use of ASME Code Case N-514, which will benefit DCPP

    [Diablo Canyon Power Plant] by not resulting in a reduced RCS P/T window and reduced power-operated relief valve (PORV) pressure setpoint for LTOP. This maintains the current level of operator flexibility during heatup and cooldown, and prevents an increase in the probability of an accident associated with an inadvertent PORV actuation.

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

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

    The proposed changes to TS 3.4.9.1, ``Reactor Coolant System-- Pressure/Temperature Limits,'' do not involve any physical alteration to any plant system or change the method by which any safety-related system performs its function. The changes to TS 3.4.9.1 account for the effects of an incremental increase in reactor vessel embrittlement and are requested in order to restrict future reactor operation to within acceptable stress levels and temperature regimes in accordance with 10 CFR 50, Appendix G, requirements. These changes are needed to maintain the current P/T limit margins of safety as defined by 10 CFR 50, Appendix G, and ASME XI, Appendix G, for operation through 16 EFPY. The possibility of a new kind of accident such as catastrophic failure of the reactor vessel is prevented by maintaining acceptable margins of safety.

    The present LTOP pressure setpoint was reviewed and found to be acceptable and conservative for the extension of the P/T curves to 16 EFPY.

    Additionally, the proposed changes will not affect the ability of the LTOP system to provide pressure relief at low temperatures, thereby maintaining the LTOP design basis.

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

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

    The proposed changes to TS 3.4.9.1 adjust the RCS heatup and cooldown P/T limits to permit operation through 16 EFPY. The P/T limits have been determined in accordance with 10 CFR 50, Appendix G, and include the safety margins with regard to brittle fracture required by the ASME Section XI, Appendix G, which maintain the same margins of safety as the current limits.

    The LTOP setpoints were reevaluated using the requirements of ASME Code Case N-514. This code case was developed to provide the necessary margins of safety for the prevention of reactor vessel overpressurization and brittle fracture. The LTOP evaluation results conclude the current LTOP setpoints are conservative for operation through 16 EFPY. In addition, avoiding an unnecessary reduction in the LTOP, the PORV pressure setpoint prevents an increase in the likelihood of an inadvertent PORV actuation

    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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration.

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

    NRC Section Chief: Stephen Dembek.

    Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power Plant, Unit 3, Humboldt County, California

    Date of amendment request: December 1, 1999.

    Description of amendment request: The proposed amendment would revise the Humboldt Bay Power Plant (HBPP) Unit 3 Technical Specifications (TS) related to fire protection, administrative controls, and quality assurance audits. The fire protection requirements would be relocated verbatim from the TS to the HBPP Defueled Safety Analysis Report (DSAR). The administrative controls requirements would be revised to (1) refer to the DSAR for a description of the plant organization, (2) modify information pertaining to plant staff titles and qualifications to reflect the current organization, and (3) replace a reference to the Final Hazards Summary Report with a reference to the DSAR. Quality assurance audit requirements would be relocated from the TS to the Quality Assurance Plan.

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

    For the proposed changes to the fire protection requirement, the licensee's analysis states:

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

    The FPP [Fire Protection Program] and FPS [Fire Protection System] are not being changed. Operability requirements and procedural controls of the FPP and FPS are not being changed. The proposed changes involve only where the FPP and FPS description is located and how changes can be made. Consequently, the changes will not affect the probability or consequences of an accident occurring.

    Future changes to the FPP and FPS as described in the Defueled Safety Analysis Report would be made in accordance with 10 CFR 50.59. This ensures that adequate controls will remain in place so that the public health and safety will be protected.

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

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

    The FPP and FPS are not being changed. Operability requirements and procedural controls of the FPP and FPS are not being changed. The proposed changes involve only where the FPP and FPS description is located and how changes can be made. Consequently, the changes will not affect the probability or consequences of an accident occurring.

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

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

    The FPP and FPS are not being changed. Operability requirements and procedural controls of the FPP and FPS are not being changed. The proposed changes involve only where the FPP and FPS description is located and how changes can be made. Consequently, the changes will not affect the probability or consequences of an accident occurring.

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

    For the proposed changes to the administrative controls requirements, the licensee's analysis states:

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

    The organization title and responsibility changes update the Technical Specification (TS) to reflect the current organization and have no impact on the function or operability of plant systems, structures, or components, or the ability of the plant to safely maintain SAFSTOR status. Consequently, the changes will not affect the probability or consequences of an accident occurring.

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

  47. The proposed change does not create the possibility of a new or different kind of

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    accident from any accident previously evaluated.

    The organization title and responsibility changes update the TS to reflect the current organization and have no impact on the function or operability of plant systems, structures, or components, or the ability of the plant to safely maintain SAFSTOR status. Consequently, the changes will not affect the probability or consequences of an accident occurring.

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

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

    The organization title and responsibility changes update the TS to reflect the current organization and have no impact on the function or operability of plant systems, structures, or components, or the ability of the plant to safely maintain SAFSTOR status. Consequently, the changes will not affect the probability or consequences of an accident occurring.

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

    For the proposed changes to the quality assurance audit requirements, the licensee's analysis states:

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

    The proposed changes simplify the Technical Specifications (TS), meet regulatory requirements for relocated TS, and implement: (1) The recommendations of NRC's letter dated October 25, 1993, from William T. Russell to the chairpersons of the industry owners groups; (2) the Commission's Final Policy Statement on TS Improvements; and (3) the current revision of 10 CFR 50.36. Future changes to these requirements will be controlled by 10 CFR 50.54. This ensures that adequate controls will remain in place so that the public health and safety will be protected. The proposed changes are administrative in nature and do not involve any modifications to any plant equipment or affect plant operation.

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

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

    The proposed changes are administrative in nature, do not involve any physical alterations to any plant equipment, and cause no change in the method by which any safety-related system performs its function.

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

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

    The proposed changes do not alter implementation of the basic regulatory requirements and do not affect any safety analyses. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

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

    NRC Section Chief: Michael Masnik.

    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: October 12, 1999 (TS 99-15).

    Brief description of amendments: The proposed amendments would change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79 (Unit 2) by revising the Technical Specification (TS) to provide for unisolation of containment penetrations under administrative controls. This revision will add a footnote to Specification 3.9.4.c indicating this allowance and the necessary Bases addition for this section to clarify the use of this allowance.

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

      The proposed revision will allow the opening of specific containment penetrations during the movement of irradiated fuel or core alterations provided administrative controls are implemented. These controls will establish the proper awareness of the unisolated penetration condition, designate individuals to isolate the penetration in the event of an FHA [fuel handling accident], and

      [to] ensure the auxiliary building gas treatment system (ABGTS) is available. The status of containment penetrations does not impact the generation of an accident nor does the ability to unisolate penetrations affect this potential. The proposed revision does not alter any plant equipment or operating practices other than penetration isolation such that the probability of an accident is increased.

      The administrative controls provide adequate requirements to provide timely identification and closure of penetrations opened under this allowance should a fuel handling event occur. Designated individuals ensure that adequate resources are available to isolate the penetration such that the offsite dose consequences are not significantly impacted. The lack of motive force in containment during fuel movement to expel the radioactive material allows a more flexible isolation interval. The exception for the containment ventilation isolation valves is based on being exposed to a motive force and the flow paths outside the auxiliary building secondary containment enclosure (ABSCE) is based on being exposed to an unfiltered atmosphere. Timely isolation of the specified flow paths is required to ensure that the unlikely transmission of radioactive material does not occur. Interactions that may occur during the period of time before isolation will be controlled by operation of the ABGTS and will not significantly increase the consequences of an accident as previously evaluated. Completion of penetration isolation and operation of the ABGTS, as required by the administrative controls, will maintain the offsite dose consequences well within the 10 CFR 100 limits.

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

      The proposed allowance to open penetrations in Mode 6 will not alter plant functions or equipment operating practices other than penetration isolation. Containment penetration status is not considered to be the source of an accident. Therefore, since the plant functions and equipment are not altered and the isolation status of containment penetrations do not contribute to the initiation of postulated accidents, the proposed revision will not create a new or different kind of accident.

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

      The isolation requirements for containment penetrations ensure that the release of radioactivity is minimized to maintain the 10 CFR 100 limits for offsite dose consequences in the event of an FHA. The proposed change to allow penetrations to be unisolated does not significantly affect the expected dose consequence because of the absence of containment pressurization potential during fuel movement or core alterations. The most significant offsite dose contributor to the fuel handling event is the containment purge system that generates a motive force for the radioactive material. This flow path is excluded from the proposed allowance because of this motive force potential along with flow paths outside the ABSCE. Without this motive force, as is the case with other penetrations during fuel movement or core alterations, the potential for additional offsite dose consequence is unlikely. As an additional measure, this allowance applies to flow paths that can be filtered by the ABGTS. Therefore, the margin of safety provided by the containment building penetration requirements is not significantly impacted by the proposed allowance to open penetrations under administrative controls. With the timely provision to identify and isolate affected penetrations and the provision for ABGTS operability, the margin of safety is maintained without a significant reduction.

      [[Page 1929]]

      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.

      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: November 24, 1999 (TS 99-16).

      Brief description of amendments: The proposed amendments would change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79 (Unit 2) by updating the Technical Specification (TS) surveillance requirements for penetration efficiency tests of charcoal adsorbers to comply with American Society for Testing and Materials (ASTM) test standard ASTM D3803-1989 as directed by NRC Generic Letter (GL) 99-02.

      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:

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

      The proposed revision will require laboratory tests of safety- related charcoal filter adsorbers to tighter specifications. NRC research indicates that the new test protocols yield more accurate measures of filter efficiency and better reproducibility of test results. No physical change is made to the filter by these expanded timeframes of testing and tighter controls; therefore, no change to the filter behavior is expected. Current methods for selecting and obtaining charcoal samples for testing will be retained without change. The proposed revision does not alter any plant equipment or operating practices other than filter tests that are conducted away from the plant site, and as such the probability of an accident is not increased.

      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 than older tests, upgrading TSs to the ASTM D3803-1989 protocol allows use of a safety factor of 2 for determining the acceptance criteria for charcoal filter efficiency. This safety factor can be used for systems with or without humidity control because the lack of humidity control is already accounted for in the test conditions.

      Applying the ASTM D3803-1989 test methodology and using the new safety factor is expected to yield a net improvement in safety. The ASTM D3803-1989 test protocol is expected to improve the identification of degraded charcoal filters and lead to their timely replacement without any adverse effects on filter performance. Therefore, the change in testing does not significantly increase the consequences of an accident.

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

      The proposed change in laboratory tests performed on charcoal filters will not alter plant functions or equipment operating practices other than possibly resulting in more frequent replacement of charcoal filters. As stated previously, current methods for selecting and obtaining charcoal samples for testing will be retained without change. The ASTM D3803-1989 test methodology is not expected to alter the filters; therefore, it will not adversely alter the resulting filter performance. Since the plant functions and equipment are not altered, the proposed revision will not create a new or different kind of accident.

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

      Analyses of design-basis accidents assume a particular ESF

      [Engineered Safety Feature] charcoal filter adsorption efficiency when calculating offsite and control room operator doses. Charcoal filter samples are tested to determine whether the filter adsorber efficiency is greater than that assumed in the design-basis accident analysis. The laboratory test acceptance criteria contains 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 than older tests, NRC indicated in GL 99-02 that licensees upgrading their TS to this new protocol will be able to use a safety factor as low as 2 for determining the acceptance criteria for charcoal filter efficiency. This safety factor can be used for systems with or without humidity control because the lack of humidity control is already accounted for in the test conditions. As stated in the GL, the new test protocol and associated safety factors have been reviewed and found to not significantly decrease the margin of safety.

      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.

      Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

      Date of amendment request: December 16, 1999.

      Brief description of amendment: The proposed change would amend Technical Specification 4.18.5.b to allow tube 110/60 to remain in service through the current operating cycle (cycle 16) with two axial indications that have potential through-wall depths greater than the plugging limit. The axial indications are located in the roll transition region and are contained within the upper tubesheet.

      Date of publication of individual notice in Federal Register: December 29, 1999 (64 FR 73080).

      Expiration date of individual notice: Comments on no significant hazards considerations by January 12, 2000; requests for hearing by January 28, 2000. Clarification: The December 29, 1999, notice indicated that requests for a hearing with respect to issuance of this amendment must be filedby January 12, 2000. The correct deadline for this action is January 28, 2000.

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

      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: December 16, 1999.

      Description of amendment request: The proposed changes would revise the River Bend Station (RBS) Technical Requirements Manual, Section TR

      [[Page 1930]]

      3.9.14, and add an exception to the current prohibition for travel of loads in excess of 1200 pounds over fuel assemblies in the spent fuel storage pool. The exception would allow the licensee to move the spent fuel pool (SFP) watertight gates, which separate the SFP from the cask and lower transfer pools, to perform maintenance and repairs on the gates and watertight seals. Related sections of the RBS Updated Safety Analysis Report would also be revised to be consistent with the exception. The licensee determined that movement of the gate, with its associated rigging, over spent fuel would involve an unreviewed safety question in accordance with Title 10 of the Code of Federal Regulations, Section 50.59.

      Date of publication of individual notice in Federal Register: December 21, 1999 (64 FR 71511).

      Expiration date of individual notice: January 20, 2000. Correction: The December 21, 1999, notice indicated that requests for a hearing with respect to issuance of this amendment must be filedby January 28, 2000. The correct deadline for this action is January 20, 2000.

      Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005.

      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).

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

      Date of application for amendments: December 16, 1998, as supplemented July 16, September 29, and December 21, 1999.

      Brief description of amendments: The amendments revise Technical Specifications 3.8.1 and 3.37 to ensure that the appropriate actions are taken to prevent double sequencing of safety-related loads and that the setpoint allowable values for the degraded voltage relays reflect the required function of the relays.

      Date of issuance: December 29, 1999.

      Effective Date: December 29, 1999, to be implemented within 90 days.

      Amendment Nos: Unit 1-123, Unit 2-123, Unit 3-123.

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

      Date of initial notice in Federal Register: March 24, 1999 (64 FR 14279) The July 16, September 29, and December 21, 1999, letters provided additional clarifying information that was written within the scope of the original application and Federal Register notice 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 December 29, 1999.

      No significant hazards consideration comments received: No.

      Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

      Date of application for amendments: September 1, 1999.

      Brief description of amendments: The amendments revised the Technical Specifications as follows:

  52. Technical Specification 1.1 is changed to replace the definition of Azimuthal Power Tilt with a new definition.

  53. Technical Specification 2.1.1.2 is changed by replacing the peak linear heat rate safety limit with less than or equal to 22 kW/ft.

  54. Technical Specification Surveillance Requirement (SR) 3.3.6.2 is changed by replacing the degraded voltage function with transient degraded voltage and steady-state degraded voltage functions.

  55. Technical Specification SRs 3.8.1.9 and 3.8.1.15 are changed by replacing the steady-state voltage range with the range of greater than or equal to 4060 volts and less than or equal to 4400 volts.

  56. Technical Specification 5.6.5.a is changed by adding Technical Specifications 3.1.4 and 3.3.1 to the list.

  57. Technical Specification Figure 2.1.1-1 is changed by removing the reference to Figure B2.1-1.

  58. Various Technical Specifications and Figures 2.1.1-1a are changed by removing references to Unit 2, Cycle 12, and deleting Figure 2.1.1-1a.

  59. Technical Specification 5.6.5.b, Item 41.ii is changed by correcting CEN-99(B)-P to CEN-119(B)-P.

    Date of issuance: December 15, 1999.

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

    Amendment Nos.: 232 and 208.

    Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Technical Specifications.

    Date of initial notice in Federal Register: October 6, 1999 (64 FR 54372).

    The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated December 15, 1999.

    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 6, 1999, as supplemented on November 15, 1999.

    Brief description of amendments: The amendments revised Technical

    [[Page 1931]]

    Specification 3/4.4.6, ``Vacuum Relief'' to remove specific operability requirements related to position indication for the suppression chamber-drywell vacuum breakers. The amendments also reformat the action statement for inoperable vacuum breakers, increase the surveillance interval for verifying that the vacuum breakers are closed, and delete the requirement to verify that the manual isolation valves are closed for an inoperable and open vacuum breaker.

    Date of issuance: December 21, 1999.

    Effective date: Immediately, to be implemented within 30 days.

    Amendment Nos.: 138 and 122.

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

    Date of initial notice in Federal Register: August 25, 1999 (64 FR 46428).

    The November 15, 1999, submittal provided additional clarifying information that 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 December 21, 1999.

    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 16, 1999.

    Brief description of amendments: The amendments revise Technical Specification 4.7.D.6 by replacing the leakage limit of 11.5 standard cubic feet per hour (scfh) for each main steam isolation valve (MSIV) with a limit of 46 scfh on the total combined leakage for the MSIVs of all four main steam lines.

    Date of issuance: December 21, 1999.

    Effective date: Immediately, to be implemented within 30 days.

    Amendment Nos.: 192 and 188.

    Facility Operating License Nos. DPR-29 and DPR-30: The amendments revised the Technical Specifications.

    Date of initial notice in Federal Register: August 25, 1999 (64 FR 46429).

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 21, 1999.

    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 15, 1999.

    Brief description of amendments: The amendments revise Section 5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the Technical Specifications by adding a new paragraph. The existing single paragraph of Section 5.5.7 requires that inspection of each reactor coolant pump flywheel be done per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14. The amendments add a new paragraph which specifies that in lieu of Regulatory Positions C.4.b(1) and C.4.b(2), alternative inspection techniques may be used. Date of issuance: December 21, 1999.

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

    Amendment Nos.: 182 (Unit 1); 174 (Unit 2).

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

    Date of initial notice in Federal Register: November 17, 1999 (64 FR 62705).

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 21, 1999.

    No significant hazards consideration comments received: No.

    Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: October 15, 1999.

    Brief description of amendments: The amendments revise Section 5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of the Technical Specifications by adding a new paragraph. The existing single paragraph of Section 5.5.7 requires that inspection of each reactor coolant pump flywheel be done per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14. The amendments add a new paragraph which specifies that in lieu of Regulatory Positions C.4.b(1) and C.4.b(2), alternative inspection techniques may be used.

    Date of issuance: December 21, 1999.

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

    Amendment Nos.: 190 (Unit 1); 171 (Unit 2).

    Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Technical Specifications.

    Date of initial notice in Federal Register: November 17, 1999 (64 FR 62706).

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 21, 1999.

    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: December 18, 1998, as supplemented September 13, 1999.

    Brief description of amendment: This amendment revises the St. Lucie, Unit 2 (SL-2), Technical Specifications (TS) Index Page III, TS 1.10, Dose Equivalent iodine-131; TS 2.1.1.2, Linear Heat Rate; TS 3.1.1.1/4.1.1.1.1, Shutdown Margin--TavgGreater than 200 deg.F; TS 3/4.1.1.2, Shutdown Margin--TavgLess Than or Equal to 200 deg.F; TS 3.1.2.2, Boration Systems Flow Paths--Operating; TS 3.1.2.4, Charging Pumps--Operating; TS 3.1.2.6, Boric Acid Makeup Pumps--Operating; TS 3.1.2.8, Borated Water Sources--Operating; and TS 6.9.1.11, Core Operating Limits Report (COLR). The amendment also relocates the core operating limits for shutdown margin to the SL-2 COLR. The following Bases have also been changed in connection with this amendment: TS Bases 2.1.1, Reactor Core; Bases Figure B2.1-1, Axial Power Distributions for Thermal Margin Safety Limits; TS Bases 2.2.1, Reactor Trip Setpoints (Variable Power Level-High); TS Bases 3/ 4.1.1.1 and 3/4.1.1.2, Shutdown Margin; and TS Bases 3/4.1.2, Boration Systems.

    Date of Issuance: December 21, 1999.

    Effective Date: As of date of issuance, to be implemented prior to fuel reload for Cycle 12.

    Amendment No.: 105.

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

    Date of initial notice in Federal Register: February 10, 1999 (64 FR 6697). The supplemental letter dated September 13, 1999, provided additional information that did not expand the scope of the amendment request as noticed or change the original proposed no significant hazards consideration determination.

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 21, 1999.

    No significant hazards consideration comments received: No.

    [[Page 1932]]

    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: April 26, 1999.

    Brief description of amendments: The amendments revise the Technical Specifications (TS) for Turkey Point Units 3 and 4 to correct the Technical Specification Index and to remove inconsistencies, and make administrative changes. A portion of the request, related to the proposed deletion of dates for the approved security plans, was denied.

    Date of issuance: December 20, 1999.

    Effective date: December 20, 1999.

    Amendment Nos.: 203 and 197.

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

    Date of initial notice in Federal Register: June 2, 1999 (64 FR 29711).

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 20, 1999.

    No significant hazards consideration comments received: No.

    GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 Dauphin County, Pennsylvania

    Date of application for amendment: December 3, 1998, as supplemented January 11, February 4, March 4, March 10, and March 15, 1999.

    Brief description of amendment: This amendment conforms the license to reflect the transfer of Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1, from GPU Nuclear, Inc., et al., to AmerGen Energy Company, LLC, as previously approved by Order dated April 12, 1999.

    Date of issuance: December 20, 1999.

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

    Amendment No.: 218.

    Facility Operating License No. DPR-50: Amendment revised the license and the Technical Specifications.

    Date of initial notice in Federal Register: December 21, 1998 (63 FR 70436).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 12, 1999.

    Comments received: Yes. See safety evaluation dated April 12, 1999.

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

    Date of application for amendment: April 12, 1999, as supplemented October 5 and 8, 1999.

    Brief description of amendment: The amendment revises Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.7 to allow a ``representative sample'' of reactor instrumentation line excess flow check valves (EFCVs) to be tested every 24 months, instead of testing each EFCV every 24 months.

    Date of issuance: December 29, 1999.

    Effective date: December 29, 1999.

    Amendment No.: 230.

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

    Date of initial notice in Federal Register: July 14, 1999 (64 FR 38028).

    The October 5 and 8, 1999, letters provided clarifying information that 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 December 29, 1999.

    No significant hazards consideration comments received: No.

    Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: December 3, 1998.

    Brief description of amendment: This amendment revised the Technical Specifications for sealed source leakage testing to specifically address testing requirements for fission detectors.

    Date of issuance: December 20, 1999.

    Effective date: December 20, 1999, with full implementation within 45 days.

    Amendment No.: 235.

    Facility Operating License No. DPR-58: Amendment revised the Technical Specifications.

    Date of initial notice in Federal Register: August 11, 1999 (64 FR 43773).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 20, 1999.

    No significant hazards consideration comments received: No.

    Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: September 17, 1999, as supplemented November 10 and 19, 1999.

    Brief description of amendments: The amendments would approve the licensee's revision of the Updated Final Safety Analysis Report and Emergency Operating Procedures to use methodology to credit the negative reactivity provided by insertion of the rod cluster control assemblies (RCCAs) into the reactor core following any design basis loss-of-coolant accident, during realignment from a cold leg recirculation to a hot leg recirculation configuration. This change to the licensing basis, when evaluated by the licensee in accordance with 10 CFR 59.59, resulted in an unreviewed safety question that requires prior approval by the NRC staff in accordance with the provisions of 10 CFR 50.90 prior to implementation. The amendments also change the Bases for Technical Specifications Section 3/4.5.5, Refueling Water Storage Tank.

    Date of issuance: December 28, 1999.

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

    Amendment Nos.: 236 and 218.

    Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Technical Specifications.

    Date of initial notice in Federal Register: October 20, 1999 (64 FR 56531).

    The licensee's letters of November 10 and 19, 1999, provided additional information that did not change scope of the application or the staff's proposed no significant hazards consideration determination.

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 28, 1999.

    No significant hazards consideration comments received: No.

    Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: November 5, 1999.

    Brief description of amendment: The amendments would revise Unit 1 and 2 Technical Specification (TS) 3.5.1, Action ``a'' and ``b,'' to reflect the monitoring of pressure from the Reactor Coolant System instead of the pressurizer. The amendment would also revise Unit 1 and 2 TS Surveillance Requirement 4.5.1.c to require verification that power is removed from each emergency core cooling system accumulator isolation valve operator instead of verification that each accumulator isolation valve breaker is physically removed from the circuit. Furthermore, the amendment would make administrative changes to Unit 1 and 2 TS Bases 3/4.5.1.

    Date of issuance: December 23, 1999.

    [[Page 1933]]

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

    Amendment No.: 237 and 219.

    Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Technical Specifications.

    Date of initial notice in Federal Register: November 23, 1999 (64 FR 65735).

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 23, 1999.

    No significant hazards consideration comments received: No.

    Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: August 17, 1999.

    Brief description of amendment: The amendment removes the steam generator voltage-based repair criteria, F* repair criteria, and sleeving methodologies from the Unit 1 Technical Specifications and clarifies the Bases sections accordingly.

    Date of issuance: December 22, 1999.

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

    Amendment No.: 238.

    Facility Operating License No. DPR-58: Amendment revises the Technical Specifications.

    Date of initial notice in Federal Register: October 6, 1999 (64 FR 54375).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 22, 1999.

    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: November 8, 1999.

    Brief description of amendment: The amendment changed action statements, definitions, and footnotes pertaining to the Technical Specifications for primary containment leakage and primary containment purge system to allow an alternative approach for isolating a bypass leakage path and/or purge system line.

    Date of issuance: December 16, 1999.

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

    Amendment No.: 87.

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

    Date of initial notice in Federal Register: November 16, 1999 (64 FR 62228).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 16, 1999.

    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: March 31, 1999.

    Brief description of amendment: Amendment changes Technical Specification Table 3.6.1.2-1 by adding two relief valves, and associated leak rate criteria, to be installed on the drywell equipment drain line and drywell floor drain line.

    Date of issuance: December 16, 1999.

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

    Amendment No.: 88.

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

    Date of initial notice in Federal Register: May 5, 1999 (64 FR 24197).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 16, 1999.

    No significant hazards consideration comments received: No

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

    Date of amendment request: March 5, 1998.

    Description of amendment request: This amendment revises the Technical Specifications (TSs) by relocating the procedural details of the Radiological Effluent Technical Specifications (RETS) to the Offsite Dose Calculation Manual. The TSs were also revised to relocate procedural details associated with solid radioactive wastes to the Process Control Program. In addition, the Administrative Controls section of the TSs was revised to incorporate programmatic controls for radioactive effluents and environmental monitoring. These changes are consistent with the guidance provided in Generic Letter 89-01, ``Implementation of Programmatic Controls for Radiological Effluent Technical Specifications 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.''

    Date of issuance: December 15, 1999.

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

    Amendment No.: 66.

    Facility Operating License No. NPF-86: Amendment revised the Technical Specifications.

    Date of initial notice in Federal Register: April 22, 1998 (63 FR 19972). The Commission received comments which were addressed in the staff's Safety Evaluation dated December 15, 1999.

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 15, 1999.

    No significant hazards consideration comments received: Yes.

    Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: October 20, 1999.

    Brief description of amendment: The amendment changes from December 31, 1999, to June 30, 2001, the date specified in TS 4.3.1.1.b Note associated with maintaining spent fuel pool boron concentration >2300 ppm at all times until a permanent resolution to the current criticality concerns is implemented.

    Date of issuance: December 21, 1999.

    Effective date: December 21, 1999.

    Amendment No.: 75.

    Facility Operating License No. DPR-18: Amendment revised the Technical Specifications.

    Date of initial notice in Federal Register: November 19, 1999 (64 FR 63345).

    The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 21, 1999.

    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: September 4, 1998, as supplemented on February 8, April 16, August 26, September 16, and November 17, 1999.

    Brief description of amendment: The amendment increases the spent fuel pool storage capacity from 2,870 to 3,353 fuel assemblies.

    Date of Issuance: December 21, 1999.

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

    [[Page 1934]]

    Amendment No.: 182

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

    Date of initial notice in Federal Register: October 1, 1998 (64 FR 52774). The supplemental information did not affect the staff's proposed no significant hazards consideration determination, and was within the scope of the original amendment application as published.

    The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated December 21, 1999.

    No significant hazards consideration comments received: No.

    Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50- 339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia

    Date of application for amendments: November 18, 1998, as supplemented by letter dated October 22, 1999.

    Brief description of amendments: The amendments change the North Anna Power Station Technical Specifications (TS) to increase the allowable groundwater elevation at the southeast section of the service water reservoir dike from 277 to 280 feet at the toe and from 280 to 295 feet at the crest. In addition, TS Table 3.7-6 has been reorganized to clarify zones of interest in the Service Water Reservoir, the location of piezometer devices, and piezometer device numbers. The proposal to eliminate device numbers from the TS was denied because the device number helps to indicate the location of the piezometer within the zone as well as the piezometer itself. Finally the column heading for Allowable Drain Flow Rate was clarified to be the total flow rate.

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

    Effective date: December 29, 1999.

    Amendment Nos.: 220 and 201.

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

    Date of initial notice in Federal Register: December 16, 1998 (63 FR 69349). The supplemental letter dated October 22, 1999, contained clarifying information only, and did not change the initial no significant hazards determination.

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 29, 1999.

    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: April 12, 1999.

    Brief description of amendments: These amendments update references in the Technical Specifications to information in the updated Final Safety Analysis Report (FSAR). The update is necessary to reflect relocation of the referenced information in the updated FSAR.

    Date of issuance: December 23, 1999.

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

    Amendment Nos.: Unit 1-192; Unit 2-197.

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

    Date of initial notice in Federal Register: May 5, 1999 (64 FR 24204).

    The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 23, 1999.

    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 5th day of January 2000.

    For the Nuclear Regulatory Commission. Suzanne Black, Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

    [FR Doc. 00-611Filed1-11-00; 8:45 am]

    BILLING CODE 7590-01-P

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