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

[Federal Register: May 28, 2002 (Volume 67, Number 102)]

[Notices]

[Page 36924-36939]

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

[DOCID:fr28my02-89]

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 May 3, 2002 through May 16, 2002. The last biweekly notice was published on May 14, 2002 (67 FR 34481).

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

[[Page 36925]]

action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC's Public Document Room (PDR), located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By June 27, 2002, 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 NRC's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/doc- collections/cfr/. If a request for a hearing or petition for leave to intervene is 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: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. 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 PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1- 800-397-4209, 304-415-4737 or by e-mail to pdr@nrc.gov.

[[Page 36926]]

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

Date of amendment request: April 26, 2002.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs), Section 2.3, ``Limiting Safety System Settings,'' Section 3.1, ``Protective Instrumentation,'' and Section 3.10, ``Core Limits,'' to reflect a methodology to assure coupled neutronic/thermal-hydraulic instabilities are adequately detected and suppressed. This methodology is identified as Option II by the Boiling Water Reactor (BWR) Owners Group. The proposed amendment includes technical (i.e., limiting safety system settings) and editorial changes, and is associated with the average power range monitoring (APRM) system.

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

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

   No. The APRM neutron monitoring system is not an initiator of, or a precursor to, a previously evaluated accident. The APRM system monitors the power level of the reactor core and provides automatic core protection signals in the event of a power transient. The revised requirements will result in a reactor scram, should one be needed, sooner than under the current requirements. These revised requirements do not lead to, and are not results of, physical design modifications. The APRM and other systems associated with the proposed TS requirements will thus continue to perform their functions as originally designed. Therefore, this amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   No. The proposed amendment does not affect accident initiators or precursors because it does not alter any design feature, reactor fuel safety limit, equipment configuration, or manner in which the unit is operated. Further, it does not alter or prevent the ability of structures, systems, or components to perform their intended safety or accident mitigating functions. Accordingly, the proposed amendment does not create a new or different kind of accident from any accident previously evaluated.

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

   No. The proposed amendment does not change any design feature, analysis methodology, safety limits or acceptance criteria. The APRM system under the revised requirements will continue to perform its design functions. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

   Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, LLP, 1800 M Street, NW., Washington, DC 20036-5869.

   NRC Section Chief: Richard J. Laufer.

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

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

   Exelon Energy Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

   Date of amendment request: April 10, 2002.

   Description of amendment request: The proposed changes to the Technical Specifications (TSs) will relocate the emergency diesel generator (EDG) 24-month maintenance inspection requirements to licensee-controlled documents, i.e., either the Updated Final Safety Analysis Report (UFSAR) or the Technical Requirements Manual as appropriate, either of which would be controlled in accordance with the requirements of 10 CFR 50.59.

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

   The proposed TS changes are administrative changes to relocate the current EDG maintenance inspection requirements.

   The EDGs are designed to provide a reliable alternative source of AC [alternating current] electrical power in the event of an accident coincident with the loss of offsite power. The failure of an EDG itself is not considered as an accident evaluated in the UFSAR. The proposed administrative changes to relocate the maintenance inspection requirements do not affect the current accident initiators or precursors that could lead to a previously evaluated accident.

   The failure of a single EDG to respond when required to mitigate the consequences of an accident has already been considered as a subsequent single failure in the current plant safety analyses. The proposed administrative changes to relocate the maintenance inspection requirements do not alter the EDG design features, operation, or accident analysis assumptions which could affect the ability of the EDGs to mitigate the consequences of a previously evaluated accident. Current TS testing requirements for the EDGs, e.g., starting, timing, loading, and sequencing will continue to ensure reliable EDG operation and are not being changed in this request.

   Since only the relocation of EDG maintenance inspection requirements is involved, the proposed changes will not increase the likelihood of the malfunction of another system, structure or component which has been assumed as an accident initiator or credited in the mitigation of an accident.

   Based on the above discussion, the proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

   [2.] The proposed amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated.

   The EDGs are designed to provide a reliable alternative source of AC electrical power in the event of an accident coincident with the loss of offsite power. The proposed TS changes are administrative changes to relocate the EDG maintenance inspection requirements.

   No change in the ability to perform the design function of the EDGs is involved. No change in the operation of the EDGs is required. Instrumentation setpoints, starting, sequencing, and loading functions associated with the EDGs are not affected by the proposed changes. No modifications to the EDGs are required to implement the proposed TS changes. Therefore, no new failure mechanism, malfunction, or accident initiator is considered credible.

   Additionally, the proposed TS changes do not affect other plant design, hardware, system operation, or procedures. Therefore, based on the above discussion, the proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

   [[Page 36927]]

   [3.] The proposed amendments do not involve a significant reduction in a margin of safety.

   The proposed TS changes are administrative changes to relocate the current EDG maintenance requirements.

   The consideration of safety margins for this amendment included a review of the acceptance criteria for emergency core cooling systems for light water nuclear power reactors in 10 CFR 50.46, and ECCS [emergency core cooling system] evaluation models in Appendix K to 10 CFR [Part] 50. The proposed amendments do not involve a relaxation of the criteria used to establish the safety limits, a relaxation of the bases for the limiting safety system setting, nor a relaxation of the bases for the limiting conditions for operation.

   Controlling values for the EDGs are included in current TS testing requirements, e.g., EDG starting, timing, loading, and sequencing. The proposed amendment will not modify these requirements or the accident analysis assumptions regarding the performance of the EDGs which could potentially challenge safety margins established to ensure fuel cladding integrity, as well as reactor coolant and containment system integrity.

   The safety analyses of the EDGs' ability to mitigate accidents do not require revision in order to implement the proposed amendment[s]. Modification of the existing margins is not required.

   Based on the above discussion, the proposed TS changes do not involve a significant reduction in a margin of safety.

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

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

   NRC Section Chief: Richard J. Laufer.

   Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50- 318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

   Date of amendments request: January 31, 2002.

   Description of amendments request: The proposed amendments would correct several administrative errors to Technical Specifications Sections 5.6.5.b and Appendix B. The changes would correct the title of a topical report, the date of issuance of a report, and the name of the state agency that issues pollution discharge elimination system permits.

   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. Would Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

   Technical Specification 5.6.5.b, Item 16 is being corrected to change the title of the publication for CENPD-382-P-A from ``C-E Methodology for Core Designs Containing Erbuim Burnable Absorbers'' to ``Methodology for Core Designs Containing Erbuim Burnable Absorbers.'' Correction of an administrative error does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   Technical Specification 5.6.5.b, Item 19 is being corrected to change the date of publication for CEN-161-(B)-P, Supplement 1-P, ``Improvements to Fuel Evaluation Model'' from April 1989 to April 1986. Correction of an administrative error does not involve a significant increase in the probability or consequences of an accident previously evaluated.

   Technical Specification Appendix B, Page 2-1 (both units) last paragraph is being corrected to change the State of Maryland Department of Health and Mental Hygiene to the Maryland Department of the Environment. The revision of a state organizational title to accurately reflect administrative changes made to that organization, does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Would Not Create the Possibility of a New or Different [Kind] of Accident From Any Accident Previously Evaluated

   Technical Specification 5.6.5.b, Item 16 is being corrected to change the title of publication for CENPD-382-P-A from ``C-E Methodology for Core Designs Containing Erbuim Burnable Absorbers'' to ``Methodology for Core Designs Containing Erbuim Burnable Absorbers.'' Correction of an administrative error will not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

   Technical Specification 5.6.5.b, Item 19 is being corrected to change the date of publication for CEN-161-(B)-P, Supplement 1-P, ``Improvements to Fuel Evaluation Model'' from April 1989 to April 1986. Correction of an administrative error will not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

   Technical Specification Appendix B, Page 2-1 (both units) last paragraph is being corrected to change the State of Maryland Department of Health and Mental Hygiene to the Maryland Department of the Environment. The revision of a state organizational title to accurately reflect administrative changes made to that organization will not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

3. Would Not Involve a Significant Reduction in [a] Margin of Safety

   Technical Specification 5.6.5.b, Item 16 is being corrected to change the title of publication for CENPD-382-P-A from ``C-E Methodology for Core Designs Containing Erbuim Burnable Absorbers'' to ``Methodology for Core Designs Containing Erbuim Burnable Absorbers.'' Correction of an administrative error will not involve a significant reduction in [a] margin of safety.

   Technical Specification 5.6.5.b, Item 19 is being corrected to change the date of publication for CEN-161-(B)-P, Supplement 1-P, ``Improvements to Fuel Evaluation Model'' from April 1989 to April 1986. Correction of an administrative error will not involve a significant reduction in [a] margin of safety.

   Technical Specification Appendix B, Page 2-1, (both units) last paragraph is being corrected to change the State of Maryland Department of Health and Mental Hygiene to the Maryland Department of the Environment. The revision of a state organizational title to accurately reflect administrative changes made to that organization will 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 amendments request involves no significant hazards consideration.

   Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

   NRC Section Chief: Richard J. Laufer.

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

   Date of amendments request: March 25, 2002.

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

   The NRC staff issued a notice of opportunity for comment in the Federal Register on June 14, 2001 (66 FR 32400),

   [[Page 36928]]

   on possible amendments concerning missed surveillances, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on September 28, 2001 (66 FR 49714). The licensee affirmed the applicability of the following NSHC determination in its application dated March 25, 2002.

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

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

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

   Criterion 2--The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

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

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

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

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

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

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

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

   NRC Section Chief: Thomas Koshy, Acting.

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

   Date of amendment request: March 26, 2002.

   Description of amendment request: The proposed amendment would revise Technical Specification (TS) 5.5.16, ``Containment Leakage Rate Testing Program,'' to require the performance of a Type A test within 15 years from the last Type A test, which was performed on April 9, 1992. The proposed change is supported by a plant-specific risk assessment.

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

1. The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

   The proposed change to TS 5.5.16 provides a one-time extension to the testing interval for Type A (containment integrated leak rate) testing. The existing 10-year test interval is based on past test performance. The proposed TS change provides a one-time extension of the Type A test interval to 15 years for HBRSEP, Unit No. 2. The proposed TS change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment vessel is designed to provide a leak tight barrier against the uncontrolled release of radioactivity to the environment in the unlikely event of postulated accidents. As such, the containment vessel is not considered as the initiator of an accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

   The proposed change involves only a one-time change to the interval between Type A containment leakage tests. Type B and C leakage testing will continue to be performed at the interval specified in 10 CFR Part 50, Appendix J, Option A, as currently required by the HBRSEP, Unit No. 2, TS. As documented in NUREG-1493, ``Performance-Based Containment Leakage-Test Program,'' industry experience has shown that Type B and C containment leakage tests have identified a very large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is very small. In fact, an analysis of 144 integrated leak rate tests results, including 23 failures, found that none of the failures involved containment liner breach. NUREG-1493 also concluded, in part, that reducing the frequency of Type A containment leakage rate testing to once per 20 years was found to lead to an imperceptible increase in risk. The HBRSEP, Unit No. 2, test history and risk-based evaluation of the proposed extension to the Type A test interval supports this conclusion. The design and construction requirements of the containment vessel, combined with the containment inspections performed in accordance with the American Society of Mechanical Engineers (ASME) Code, Section XI, and the Maintenance Rule (i.e., 10 CFR 50.65) provide a high degree of assurance that the containment vessel will not degrade in a manner that is detectable only by Type A testing. Therefore, the proposed Technical Specification change does not involve a significant increase in the consequences of an accident previously evaluated.

   Therefore, this change does not involve a significant increase in the probability or

   [[Page 36929]]

   consequences of an accident previously evaluated

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

   The proposed change to TS 5.5.16 provides a one-time extension to the testing interval for Type A (containment integrated leak rate) testing. The existing 10-year test interval is based on past test performance. The proposed TS change will provide a one-time extension of the Type A test interval to 15 years for HBRSEP, Unit No. 2. The proposed change to the Type A test interval does not result in any physical changes to HBRSEP, Unit No. 2. In addition, the proposed test interval extension does not change the operation of HBRSEP, Unit No. 2, such that a failure mode involving the possibility of a new or different kind of accident from any accident previously evaluated is created.

   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 Does Not Involve a Significant Reduction in the Margin of Safety

   The proposed change to TS 5.5.16 provides a one-time extension to the testing interval for Type A (containment integrated leak rate) testing. The existing 10-year test interval is based on past test performance. The proposed TS change will provide a one-time extension of the Type A test interval to 15 years for HBRSEP, Unit No. 2. The NUREG-1493 generic study of the effects of extending containment leakage testing found that a 20 year extension for Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 found that, generically, the design containment leakage rate contributes a very small amount to the individual risk, and that the decrease in Type A testing frequency would have a minimal affect on this risk, because most potential leakage paths are detected by Type B and C testing.

   The proposed change involves only a one-time extension of the interval for Type A containment leakage testing; the overall containment leakage rate specified by the HBRSEP, Unit No. 2, Technical Specifications is being maintained. Type B and C containment leakage testing will continue to be performed at the frequency required by the HBRSEP, Unit No. 2, Technical Specifications. The regular containment inspections being performed in accordance with ASME, Section XI, and the Maintenance Rule (i.e., 10 CFR 50.65) provide a high degree of assurance that the containment will not degrade in a manner that is only detectable by Type A testing. In addition, a plant-specific risk evaluation has demonstrated that the one-time extension of the Type A leakage test interval from 10 years to 15 years results in only a very small increase in risk for those accident sequences influenced by Type A testing.

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

   NRC Section Chief: Thomas Koshy, Acting.

   Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

   Date of amendment request: August 24, 2001.

   Description of amendment request: The proposed amendment would change Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.3, ``Control Room Emergency Filtration (CREF) System,'' to address a degraded CREF pressure boundary. Specifically, the amendment would (1) add a note to the LCO that would allow the CREF pressure boundary to be opened under administrative control; (2) add a new Condition (B) for two CREF subsystems inoperable due to an inoperable control room pressure boundary--the associated Required Action would be to restore the control room pressure boundary to operable status and the Completion Time would be 24 hours; (3) add the phrase, ``for reasons other than Condition (B),'' to the Condition requiring entry into LCO 3.0.3 for two CREF subsystems or a nonredundant component or portion of the CREF system inoperable in Mode 1, 2, or 3; and (4) renumber the remaining existing Conditions and Required Actions of LCO 3.7.3, as required.

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

1. The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

   No. The Control Room Emergency Filtration (CREF) System is not assumed to be an initiator of any analyzed accident. Therefore, the proposed change does not affect the probability of any accident previously evaluated. The proposed change to the CREF Technical Specifications would permit the control room pressure boundary to be opened intermittently under administrative control. Based on the proposed compensatory measures in the form of a dedicated individual who is in communication with the control room, and his ability to rapidly restore the pressure boundary, the capability to mitigate a design basis accident will be maintained. In addition, the proposed change adds a new Condition that would allow up to 24 hours to restore an inoperable control room pressure boundary to operable status and would modify existing Conditions to accommodate the new Condition (so as to maintain the requirements of the existing Conditions). The proposed change does not involve a significant increase in the consequences of an accident previously evaluated based on the availability of self-contained breathing apparatus equipment to minimize radiological dose due to iodine, and the ability to operate at least one CREF subsystem to maintain positive pressure or to at least minimize any inflow of air from outside of the control room.

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated

   No. The proposed change would permit the control room pressure boundary to be opened intermittently under administrative control. In addition, the proposed change would add a new Condition that would permit a 24-hour period to take action to restore an inoperable control room pressure boundary to operable status. The proposed change does not alter the operation of the plant or any of its equipment, introduce any new equipment, or result in any new failure mechanisms or single failures. Therefore, this change does not create the possibility of a new accident, and does not change the way that an analyzed accident will progress.

3. The Change Does Not Involve a Significant Reduction in the Margin of Safety

   No. The proposed change would permit the control room pressure boundary to be opened intermittently under administrative control. In addition, the proposed change would add a new Condition that would permit a 24-hour period to take action to restore an inoperable control room pressure boundary to operable status. The proposed change does not adversely affect the ability of the fission product barriers to perform their functions. The only safety-related equipment affected by the proposed change is the CREF system. Adequate compensatory measures are available to mitigate a breach in the CREF control room pressure boundary. The probability of a design basis accident that would place demands on the CREF System occurring during a period that the control room pressure boundary would be allowed to be inoperable have been shown to be negligible for this limited period of time. In addition, the proposed change would avoid the potential for placing the unit in TS Limiting Condition for Operation (LCO) 3.0.3, due solely to a breach in the control room pressure boundary. Therefore, this change does not involve a significant reduction in the margin of safety.

   The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the

   [[Page 36930]]

   amendment request involves no significant hazards consideration.

   Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.

   NRC Section Chief: L. Raghavan.

   Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone Nuclear Power Station, Unit No. 3, New London County, Connecticut

   Date of amendment request: February 5, 2002, as supplemented on March 6, 2002.

   Description of amendment request: This request changes the term in the Technical Specifications ``once each REFUELING INTERVAL'' to ``once per 24 months'' in several surveillance requirements.

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

1. Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

   The proposed change is to revise the term ``once each REFUELING INTERVAL'' to ``once per 24 months'' in several technical specification surveillance requirements. These surveillances were previously approved for once per 24 months when the term REFUELING INTERVAL was defined as once per 24 months. The proposed change does not revise any surveillance requirements. The change does not alter any regulatory requirement or any acceptance criteria for any design-basis accidents described in the Millstone Unit No. 3 Final Safety Analysis Report (FSAR). The proposed change does not alter the method by which the surveillances are conducted, does not involve any physical changes to the plant, and does not modify the manner in which the plant is operated. Since the change does not change the frequency of surveillance, it cannot affect the likelihood or consequences of accidents. Therefore, the change will not increase the probability or consequences of an accident previously evaluated.

2. Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated

   The proposed change does not involve a physical alteration of the plant or change the plant configuration (no new or different type of equipment will be installed). The proposed change does not require any new or unusual operator actions. The change does not alter the way any structure, system, or component functions and does not alter the manner in which the plant is operated. The change does not introduce any new failure modes and does not change the surveillance frequency. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a Significant Reduction in a Margin of Safety

   The proposed change does not change any analyses for the current design-basis accidents described in the Millstone Unit No. 3 FSAR. Therefore, the proposed change will not result in a reduction in a margin of safety.

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

   Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.

   NRC Section Chief: James W. Clifford.

   Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

   Date of amendment request: April 15, 2002.

   Description of amendment request: The proposed amendments would modify the allowable value and surveillance requirements for reactor protection system instrumentation for the reactor vessel steam dome pressure--high function.

   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 TS Changes Do Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

   The proposed Technical Specifications (TS) changes support a design change that upgrades the existing reactor vessel steam dome pressure--high instrumentation from pressure switches to analog trip units. Analog trip units use a proven technology that is more reliable than the existing equipment. The proposed design is consistent with a generic design that has been previously reviewed and approved by the NRC. Analog trip units are currently used in various applications at Dresden Nuclear Power Station (DNPS), including the reactor protection system (RPS) low water level scram function.

   The proposed TS changes add new channel check and trip unit calibration surveillance requirements (SRs), and modify other SRs in keeping with the use of pressure transmitters for the reactor vessel steam dome pressure--high function. The new SRs are not applicable to the existing instrumentation because the current pressure switches are non-indicating and do not employ trip units.

   TS requirements that govern operability or routine testing of plant instruments are not assumed to be initiators of any analyzed event because these instruments are intended to prevent, detect, or mitigate accidents. Therefore, these changes will not involve an increase in the probability of an accident previously evaluated. Additionally, these changes will not increase the consequences of an accident previously evaluated because the proposed change does not adversely impact structures, systems, or components. The planned instrument upgrade is a more reliable design than existing equipment. The proposed change establishes requirements that ensure components are operable when necessary for the prevention or mitigation of accidents or transients. Furthermore, there will be no change in the types or significant increase in the amounts of any effluents released offsite.

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

2. The Proposed TS Changes Do Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated

   The proposed changes support a planned instrumentation upgrade by incorporating SRs required to ensure operability. The change does not adversely impact the manner in which the instrument will operate under normal and abnormal operating conditions. Therefore, these changes provide an equivalent level of safety and will not create the possibility of a new or different kind of accident from any accident previously evaluated. The changes in allowable values and surveillance requirements do not affect the current safety analysis assumptions. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The Proposed TS Changes Do Not Involve a Significant Reduction in a Margin of Safety

   The proposed TS changes support a planned instrumentation upgrade. The proposed changes do not affect the probability of failure or availability of the affected instrumentation. The revised allowable values, addition of a channel check and trip unit calibration, and revision of other SRs do not affect the analytical limit assumed in the safety analyses for the actuation of the instrumentation. Therefore, the proposed changes do not result in a reduction in the margin of safety.

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

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

   NRC Section Chief: Anthony J. Mendiola.

   [[Page 36931]]

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

   Date of amendment request: May 1, 2002.

   Description of amendment request: The proposed amendments revise the required emergency diesel generator start time limit specified in Technical Specification (TS) Section 3.8.1, ``AC Sources--Operating,'' Surveillance Requirements from ``[le]10 seconds'' to ``[le]13 seconds.''

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

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

   The accidents previously evaluated in the Updated Final Safety Analysis Report (UFSAR) involving emergency diesel generator (EDGs) are failure of one EDG to start during a loss-of-offsite power (LOOP), with or without, simultaneous occurrence of the design basis loss-of-coolant accident (LOCA). New evaluations were necessary for the General Electric GE-14 (GE-14) fuel to be used at Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The proposed change increases the time limit allowed for an EDG to start by 3 seconds. However, the increased EDG start time span allowance is still within the time delay assumed in these newly evaluated accidents. Thus, the probability for a successful EDG start is unchanged by this proposed change. A change in the start time of an EDG, but still within the bounds of the time delay assumed in analyzed accidents, does not affect previously evaluated accidents.

   In either accident specified above (i.e., failure of one EDG with either a LOOP or a LOOP plus LOCA), the UFSAR accident analysis assumes the limiting single failure, as required by 10 CFR 50 Appendix K, ``ECCS Evaluation Models,'' which is the complete failure of the unit EDG to start. The limiting single failure is unchanged by the 3-second increase in EDG start time. For this limiting single failure, the redundant ``swing'' EDG starts within 13 seconds and powers the essential loads delivering Emergency Core Cooling System (ECCS) flow to the core within the GE-14 LOCA analysis assumptions. This GE-14 analysis meets all of the same 10 CFR 50.46, ``Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,'' requirements as the previously evaluated LOCA analysis assuming a 10-second EDG start time. Therefore, the consequences of an EDG start failure are not impacted by the proposed increase in the allowed EDG start time limit. Based on the above, the proposed TS change does not involve a significant increase in the consequences of an accident previously evaluated.

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

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

   The proposed change revises the required EDG start time limit utilized by the six Surveillance Requirements (SRs) that verify EDG start capability. No other changes in requirements are being proposed. The revised EDG start time limit is consistent with the EDG and ECCS start time delay assumed in the design basis accident analysis. Therefore, the proposed EDG start time utilized by the six SRs is still bounded by analyzed evaluations of a LOOP or a LOOP in conjunction with a LOCA. No new failure modes are introduced by this proposed change, In addition, the proposed change does not physically alter the plant and will not alter the operation of the structures, systems, or components of QCNPS. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated will not be created.

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

   The consequences of a LOOP or a LOOP in conjunction with a LOCA have been previously evaluated. New evaluations were performed for the GE-14 fuel to be used at QCNPS. These new evaluations assume a bounding and longer EDG start time delay, following detection of the LOOP condition, prior to powering permanent loads fed off its associated emergency bus. Since the longer EDG start time delay was assumed in these new evaluations, any EDG start within the longer start time is bounded. The currently specified TS EDG start time limit is based on the existing analyses for the fuel utilized by QCNPS. New analysis for GE-14 fuel has assumed more conservative and bounding time delays for the integrated ECCS delivery timing sequence. All of the acceptance criterion of 10 CFR 50.46 continue to be met with the new GE-14 conservative EDG and ECCS sequences analyzed. The proposed change does not alter the basis upon which the start time limit specified in the TS is derived. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

   NRC Section Chief: Anthony J. Mendiola.

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

   Date of amendment request: April 15, 2002.

   Description of amendment request: The proposed amendment would change the Technical Specifications (TSs) to relocate boron concentration limits contained in certain TSs to the Core Operating Limits Report (COLR). The proposed amendment would change TS 2.1, ``Safety Limits,'' to relocate Figure 2.1-1, ``Reactor Core Safety Limits-Four Loops in Operation,'' to the COLR; and would revise TSs 2.1.1 and 2.1.2 limiting conditions and actions to be consistent with the improved Standard Technical Specifications (ITS). The proposed amendment also would relocate the Departure from Nucleate Boiling (DNB)-related parameters, specified in TS 3/4.2.5, to the COLR. TS 6.8.1.6, ``Core Operating Limits Report,'' and the associated TS Bases, would be revised to reflect the above changes. Editorial and administrative changes, consistent with the ITS, would also be made to TS 6.8.1.6.

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

1. The Proposed Changes Do Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

   The proposed changes to relocate cycle-specific parameters from the TSs to the COLR are administrative in nature and do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which it is operated. The proposed changes do not alter or prevent the ability of structures, systems, or components to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR).

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

2. The Proposed Changes Do Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

   The proposed changes do not alter the design assumptions, conditions, or configuration of the facility or the manner in which it is operated. The proposed changes have no adverse impact on component or

   [[Page 36932]]

   system interactions. Since there are no changes to the design assumptions, parameters, conditions and configuration of the facility, or the manner in which the plant is operated and surveilled, the proposed changes do not create the possibility of a new or different accident from any previously analyzed.

3. The Proposed Changes Do Not Involve a Significant Reduction in the Margin of Safety.

   There is no adverse impact on equipment design or operation and there are no changes being made to the TSs themselves that would adversely affect any current margin of safety. The proposed changes are administrative in nature and impose alternative procedural and programmatic controls on these parameter limits.

   Therefore, relocation of the subject cycle-specific parameter limits and other proposed editorial changes, to be reflective of the relocated parameters, do not involve a significant reduction in the margin of safety.

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

   Attorney for licensee: William J. Quinlan, Esq. Assistant General Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270.

   NRC Section Chief: James W. Clifford.

   Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin

   Date of amendment request: April 17, 2002.

   Description of amendment request: The proposed amendment would revise the Kewaunee Nuclear Power Plant Technical Specification (TS) Section 6.3, ``Plant Staff Qualifications,'' to reflect the title change from Superintendent Plant Radiation Protection to Radiation Protection Manager. In addition, the licensee informed the United States Nuclear Regulatory Commission of its intention to reformat TS Section 6.3 to Microsoft WORD format.

   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. Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

   The proposed changes will not alter the intent of the TS. Reformatting TS Section 6.3 is administrative. Changing the title from Superintendent Radiation Protection to Radiation Protection Manager is also administrative in nature. There is no impact on accident initiators or plant equipment, and thus does not affect the probability or consequences of an accident.

2. Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated.

   The proposed changes do not involve a change to the physical plant or operations. Since these are administrative changes they do not contribute to accident initiation. Therefore, they do not produce a new accident scenario or produce a new type of equipment malfunction.

3. Involve a Significant Reduction in the Margin of Safety

   Since these are administrative changes, they do not involve a significant reduction in the margin of safety. The proposed changes do not affect plant equipment or operation. Safety limits and limiting safety system settings are not affected by this change.

   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: Bradley D. Jackson, Esq., Foley and Lardner, P.O. Box 1497, Madison, WI 53701-1497.

   NRC Section Chief: L. Raghavan, Section Chief

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

   Date of amendment request: April 2, 2002.

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

   The NRC staff issued a notice of opportunity for comment in the Federal Register on June 14, 2001 (66 FR 32400), on possible amendments concerning missed surveillances, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line-item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on September 28, 2001 (66 FR 49714). The licensee affirmed the applicability of the following NSHC determination in its application dated April 2, 2002.

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

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

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

   Criterion 2--The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

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

   [[Page 36933]]

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

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

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

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

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

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

   NRC Section Chief: Richard J. Laufer.

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

   Date of amendment request: April 9, 2002.

   Description of amendment request: The proposed amendment would revise the Ginna Station Improved Technical Specification (ITS) associated with Safety Limits, Instrumentation Setpoints, and the Core Operating Limits Report. The purpose of this license amendment is to provide a clear and consistent identification of instrumentation setpoints and their operability basis.

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

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

   Operation of Ginna Station in accordance with the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The Reactor Trip System (RTS) Instrumentation, Engineered Safety Feature Actuation System (ESFAS), Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation, and Containment Ventilation Isolation trip functions are part of the accident mitigation response and are not themselves an initiator for any transient. Therefore, the probability of an accident previously evaluated is not significantly affected, including the probability of a spurious actuation. This proposed amendment includes changes to Allowable Values that have been determined with the use of an accepted methodology. The new values ensure that all automatic protective actions will be initiated at or before the condition assumed in the safety analysis. This change will allow the nominal trip setpoints to be adjusted within the calibration tolerance band allowed by the setpoint methodology. Plant operation with these revised values will not cause any design or analysis acceptance criteria to be exceeded. The structural and functional integrity of plant systems is unaffected. There will be no adverse effect on the ability of the channels to perform their safety functions as assumed in the safety analyses. Since there will be no adverse effect on the trip setpoints or the instrumentation associated with the trip setpoints, there will be no significant increase in the consequences of any accident previously evaluated.

   Other changes in trip system function, content and format are proposed based on the current configuration of the trip system hardware. Similarly, since the ability of the instrumentation to perform its safety function is not adversely affected, there will be no significant increase in the consequences of any accident previously evaluated. The proposed editorial, administrative and format changes do not affect plant safety and are in accordance with NUREG-1431.

   The proposed change to relocate core safety limits and trip setpoint parameter values to the Core Operating Limits Report (COLR) is a programmatic and administrative change that does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. Because the design of the facility and system operating parameters are not being changed, the proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated. The cycle-specific values relocated into the COLR will continue to be controlled by the Ginna Station programs and procedures. Accident analyses addressed in the UFSAR [Updated Final Safety Analysis Report] will be examined with respect to changes in the cycle-dependent parameters, which are obtained from the use of Nuclear Regulatory Commission (NRC) approved reload design methodologies, to ensure that the transient evaluation of new reloads are bounded by previously accepted analyses. This examination, which will be conducted per the requirements of 10 CFR 50.59, will ensure that future reloads will not involve a significant increase in the probability or consequences of an accident previously evaluated. Therefore, the probability or consequences of an accident previously evaluated is not significantly increased.

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

   Operation of Ginna Station in accordance with the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment includes changes to the format and magnitudes of nominal trip setpoints and allowable values that preserve all safety analysis assumptions related to accident mitigation. The protection system will continue to initiate the protective actions as assumed in the safety analysis.

   The proposed change will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis. Plant operation will not be changed.

   Other proposed changes are made so that the technical specifications more accurately reflect the installed plant specific trip system hardware. Furthermore, the proposed changes do not alter the functioning of the protection systems. No new mode of failure has been created and no new equipment performance requirements are imposed. The proposed amendment has no affect on any previously evaluated accident.

   The proposed change to relocate core safety limits and trip setpoint parameter values to the COLR is a programmatic and administrative change and does not result in any change in the manner in which the plant is operated or the way in which the Reactor Trip System provides plant protection. All of the accident transients analyzed in the UFSAR will continue to be protected by the same trip functions with the required trip setpoints. Removal of the cycle specific variables has no influence or impact on, nor does it contribute in any way to the probability or consequences of an accident. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. The cycle specific variables are calculated using the NRC approved methods, and submitted to the NRC to allow the staff to continue to review the values of these limits. The technical specifications will continue to require operation within the core operating limits, and appropriate actions will be required if these limits are exceeded. Therefore, the possibility for a new or different kind of accident from any accident previously evaluated is not created.

   The proposed change does not involve a significant reduction in

   [a] margin of safety.

   Operation of Ginna Station in accordance with the proposed changes does not involve

   [[Page 36934]]

   a significant reduction in a margin of safety. The proposed trip setpoint Allowable Values are calculated with an accepted methodology. The proposed changes will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis. The response of the protection system to accident transients reported in the Updated Final Safety Analysis Report (UFSAR) is unaffected by this change. Therefore, accident analysis acceptance criteria are not affected.

   Other proposed changes are made so that the protection system technical specifications more accurately reflect the plant-specific trip system hardware. The proposed change does not involve revisions to any safety limits or safety system setting that would adversely impact plant safety. The proposed change does not alter the functional capabilities assumed in a safety analysis for any system, structure, or component important to the mitigation and control of design bases accident conditions within the facility. Nor does this change revise any parameters or operating restrictions that are assumptions of a design basis accident. In addition, the proposed change does not affect the ability of safety systems to ensure that the facility can be placed and maintained in a shutdown condition for extended periods of time.

   The proposed change to relocate core safety limits and trip setpoint parameter values to the COLR represents an administrative change and no hardware changes are involved; therefore, no accident analysis acceptance criteria are affected. The margin of safety is not affected by the removal of cycle specific core operating limits from the technical specifications. The margin of safety presently provided by current technical specifications remains unchanged. Appropriate measures exist to control the values of these cycle specific limits. The proposed amendment continues to require operation within the core limits as obtained from NRC approved methodologies, and the actions to be taken if a limit is exceeded. The development of the limits for future reloads will continue to conform to those methods described in NRC approved documentation. In addition, each future reload will involve a 10 CFR 50.59 review. The proposed amendment is a programmatic and administrative change that provides assurance that plant operations continue to be conducted in a safe manner. The proposed amendment does not result in any change in the manner in which the plant is operated or the way in which the Reactor Trip System (RTS) provides plant protection. The proposed relocation does not alter the manner in which safety limits, limiting safety system setpoints or limiting conditions for operation are determined. Therefore, the response of the RTS to accident transients described in the UFSAR is unaffected by this change. As stated previously, this portion of the proposed amendment does not physically alter safety-related systems, nor does it affect the way in which safety-related systems perform their functions. The accident transients are unaffected and the safety analysis acceptance limits are unaffected. The design of the facility and system operating parameters are not being changed. 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: Daniel F. Stenger, Ballard Spahr Andrews & Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 20005.

   NRC Section Chief: Richard J. Laufer.

   Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

   For further details with respect to the action see (1) The applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC web site, http://www.nrc.gov/ reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301- 415-4737 or by e-mail to pdr@nrc.gov.

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

   Date of application for amendments: March 1, 2002.

   Brief description of amendments: The amendments revise Surveillance Requirement (SR) 3.0.3 to extend the delay period before entering a limiting condition for operation following a missed SR from the current limit of ``* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less'' to ``* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.'' In addition, the following requirement is added to SR 3.0.3: ``A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.''

   Date of issuance: May 7, 2002.

   Effective date: May 7, 2002, and shall be implemented within 60 days of the date of issuance.

   Amendment Nos.: Unit 1--141, Unit 2--141, Unit 3--141.

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

   Date of initial notice in Federal Register: April 2, 2002 (67 FR 15621). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 7, 2002.

   No significant hazards consideration comments received: No.

   Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London County, Connecticut

   Date of application for amendment: June 4, 2001.

   Brief description of amendment: These amendments modify the Millstone Nuclear Power Station, Unit No. 2 (MP2) and Unit No. 3 (MP3) Technical Specifications (TSs) to relocate selected MP2 and MP3 technical specifications related to the reactor coolant system to the respective Technical Requirements Manual (TRM),

   [[Page 36935]]

   with the exception of MP3 Technical Specification Section 4.4.10, which will be relocated to Section6 of MP3's TS. The Bases of the affected TSs will be modified to address the proposed changes.

   Date of issuance: May 8, 2002.

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

   Amendment Nos.: 266, 204.

   Facility Operating License Nos. DPR-65 and NPF-49: Amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: October 31, 2001 (66 FR 55011). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 8, 2002.

   No significant hazards consideration comments received: No.

   Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, Millstone Nuclear Power Station, Unit No. 3, New London County, Connecticut Date of application for amendment: August 27, 2001.

   Brief description of amendment: The amendment changes the Millstone Nuclear Power Station, Unit No. 3 Technical Specifications (TSs) action and surveillance requirements associated with the containment airlock. The Bases of the affected TSs will be modified to address the proposed changes.

   Date of issuance: May 15, 2002.

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

   Amendment No.: 205.

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

   Date of initial notice in Federal Register: November 14, 2001 (66 FR 57119). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 15, 2002.

   No significant hazards consideration comments received: No.

   Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

   Date of application for amendment: August 24, 2001, as supplemented December 20, 2001, and February 15, 2002.

   Brief description of amendment: The amendment makes changes to the license and technical specifications to reflect the transfer of operating authority to Entergy Nuclear Operations, Inc.

   Date of issuance: May 5, 2002.

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

   Amendment No.: 193.

   Facility Operating License No. DPR-35: Amendment revised the Technical Specifications and License.

   Date of initial notice in Federal Register: October 4, 2001 (66 FR 50694). The supplemental information received after the initial notice did not expand the application beyond the scope of the notice or affect the applicability of the Commission's generic no significant hazards consideration determination pursuant to 10 CFR 2.1315. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 5, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: September 28, 2001.

   Brief description of amendment: The amendment revises the Anticipated Transient Without Scram Recirculation Pump Trip Reactor Pressure High setpoint by replacing the current conditional setpoints, which are based upon the number of Safety Relief Valves out of service, with a single setpoint.

   Date of issuance: May 8, 2002.

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

   Amendment No.: 273.

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

   Date of initial notice in Federal Register: December 12, 2001 (66 FR 64294). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 8, 2002.

   No significant hazards consideration comments received: No.

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

   Date of amendment request: January 31, 2002.

   Brief description of amendment: The amendment revises Technical Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to include an additional reference to Entergy Operations, Inc. (Entergy) Topical Report ENEAD-01-P, ``Qualification of Reactor Physics Methods for Pressurized Water Reactors in the Entergy System.''

   Date of issuance: May 15, 2002.

   Effective date: As of the date of issuance and shall be implemented within 60 days from the date of implementation of Amendment No. 215.

   Amendment No.: 216.

   Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: February 19, 2002 (67 FR 7416). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 15, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: January 25, 2001, as supplemented by letter dated February 20, 2002.

   Brief description of amendment: This amendment revises Technical Specification 3.1.4, Control Rod Scram Times, to increase the control rod scram time testing interval from 120 days to 200 days of full power operation.

   Date of issuance: May 14, 2002.

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

   Amendment No: 150.

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

   Date of initial notice in Federal Register: March 21, 2001 (66 FR 15923). The supplemental letter provided clarifying information that did not change the original application nor expand the scope of the Federal Register notice as published. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 14, 2002.

   No significant hazards consideration comments received: No.

   Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

   Date of application for amendments: November 30, 2001.

   Brief description of amendments: The amendments revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period is

   [[Page 36936]]

   extended from the current limit of ``* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less'' to ``* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.'' In addition, the following requirement is added to SR 3.0.3: ``A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.''

   Date of issuance: April 30, 2002.

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

   Amendment Nos.: 128 and 123.

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

   Date of initial notice in Federal Register: February 19, 2002 (67 FR 7417). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 30, 2002.

   No significant hazards consideration comments received: No.

   Exelon Generation Company, LLC Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

   Date of application for amendments: June 1, 2001, as supplemented on February 8, 2002.

   Brief description of amendments: These amendments revise Limiting Condition for Operation 3.6.1.7 concerning drywell average air temperature.

   Date of issuance: May 9, 2002.

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

   Amendment Nos.: 159, 121.

   Facility Operating License Nos. NPF-39 and NPF-85: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: August 8, 2001 (66 FR 41619). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 9, 2002.

   No significant hazards consideration comments received: No.

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

   Date of amendment request: May 9, 2001, as supplemented by electronic mail dated February 28, 2002.

   Brief description of amendment: The amendment consists of changes to the Technical Specifications (TSs) Table 3.3.1.1-1, Function 2.b due to an error in Amendment 184. The supplementary information in the electronic mail dated February 28, 2002, provided clarifications, and did not alter the Commission's conclusions regarding significant hazards consideration.

   Date of issuance: May 9, 2002.

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

   Amendment No.: 191.

   Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: November 28, 2001 (66 FR 59509). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 9, 2002.

   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: February 21, 2002.

   Description of amendment request: The amendment relocates specific pressure, differential pressure, and flow values, as well as specific test methods, associated with certain Engineered Safeguards Features (ESF) pumps from the Technical Specifications to the Seabrook Station Technical Requirements Manual.

   Date of issuance: May 2, 2002.

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

   Amendment No.: 83.

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

   Date of initial notice in Federal Register: March 19, 2002 (67 FR 12604). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 2, 2002.

   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: December 21, 2001, as supplemented March 25 and April 8, 2002.

   Description of amendment request: The amendment revises Technical Specification Surveillance Requirements 4.3.1.2 and 4.3.2.2 to allow verification in place of demonstration of response time associated with certain pressure sensors, differential pressure sensors, process protection racks, nuclear instrumentation, and logic systems.

   Date of issuance: May 2, 2002.

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

   Amendment No.: 84.

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

   Date of initial notice in Federal Register: January 22, 2002 (67 FR 2925). The supplements dated March 25 and April 8, 2002, provided additional information that clarified the application, did not expand the scope of the application as originally published, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 22, 2002, (67 FR 2925). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 2, 2002.

   No significant hazards consideration comments received: No.

   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 application for amendments: November 13, 2001, as supplemented by letters dated February 26, 2002, March 11, 2002, and April 18, 2002.

   Brief description of amendments: The amendments revise the technical specifications to incorporate a new alternate repair criteria (ARC) for steam generator (SG) tubes with axial primary water stress corrosion cracking (PWSCC) at dented tube support plate (TSP) intersections. These amendments will apply to future operating cycles of Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2.

   Date of issuance: May 1, 2002.

   Effective date: May 1, 2002, to be implemented within 30 days from the date of issuance.

   Amendment Nos.: Unit 1--152; Unit 2--152.

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

   Date of initial notice in Federal Register: January 8, 2002 (67 FR 929). The February 26, 2002, March 11, 2002, and April 18, 2002, supplemental letters provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the original proposed no significant hazards consideration determination. The Commission's related evaluation of the

   [[Page 36937]]

   amendments is contained in a Safety Evaluation dated May 1, 2002.

   No significant hazards consideration comments received: No.

   Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, Unit 2, Hamilton County, Tennessee

   Date of application for amendment: October 9, 2001, as supplemented March 13 and April 11, 2002.

   Brief description of amendment: The amendment revised Technical Specification (TS) 6.8.4.h, to extend the requirement to perform a containment integrated leak test from once every 10 years to 11.5 years for the current interval only.

   Date of issuance: May 7, 2002.

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

   Amendment No.: 265.

   Facility Operating License No. DPR-79: Amendment revises the technical specifications.

   Date of initial notice in Federal Register: November 14, 2001 (66 FR 57127). The supplemental letters provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 7, 2002.

   No significant hazards consideration comments received: No.

   Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

   Date of application for amendment: December 6, 2001.

   Brief description of amendment: The amendment revises several of the Required Actions in the Callaway Plant Technical Specifications (TSs) that require suspension of operations involving positive reactivity additions or suspension of operations involving reactor coolant system (RCS) boron concentration reductions. In addition, the proposed amendment revises several Limiting Condition for Operation (LCO) Notes that preclude reductions in RCS boron concentration. This amendment revises these Required Actions and LCO Notes to allow small, controlled, safe insertions of positive reactivity, but limits the introduction of positive reactivity such that compliance with the required shutdown margin or refueling boron concentration limits will still be satisfied. This amendment is based on an NRC-approved traveler, Technical Specification Task Force (TSTF)-286, Revision 2.

   Date of issuance: May 1, 2002.

   Effective date: May 1, 2002, and shall be implemented within 60 days from the date of issuance.

   Amendment No.: 149.

   Facility Operating License No. NPF-30: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: February 5, 2002 (67 FR 5340). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 1, 2002.

   No significant hazards consideration comments received: No.

   Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri

   Date of application for amendment: December 13, 2001, as supplemented by letter dated April 8, 2002.

   Brief description of amendment: The amendment revises the Limiting Condition for Operation (LCO) 3.5.5, Required Action A.1 for the LCO, and Surveillance Requirement 3.5.5.1 in Technical Specification (TS) 3.5.5, ``Seal Injection Flow.'' The revision replaces the flow and differential pressure limits that were stated for the reactor coolant pump seal injection flow by limits provided in Figure 3.5.5-1, which has been added to the TSs.

   Date of issuance: May 2, 2002.

   Effective date: May 2, 2002, and shall be implemented prior to entering Mode 3 ascending during the restart from Refueling Outage 12, which is scheduled for the Fall of 2002, subject to the note above Surveillance Requirement 3.5.5.1.

   Amendment No.: 150.

   Facility Operating License No. NPF-30: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: February 5, 2002 (67 FR 5341). The supplemental letter of April 8, 2002, does not expand the scope of the application as noticed, clarifies the proposed changes given in the application, and does not change the staff's proposed no significant hazards consideration determination.

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

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: November 20, 2001.

   Brief description of amendment: The amendment changes Technical Specifications Table 3.2.6 by revising the Allowed Outage Times (AOTs) and associated action requirements for certain post-accident monitoring instrumentation.

   Date of Issuance: May 10, 2002.

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

   Amendment No.: 207.

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

   Date of initial notice in Federal Register: January 8, 2002 (67 FR 934). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated May 10, 2002.

   No significant hazards consideration comments received: No.

   Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

   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 for the amendment 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.

   Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual 30-day Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

   For exigent circumstances, the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of

   [[Page 36938]]

   telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

   In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

   Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

   The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective 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 application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Assess and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-415-4737 or by e-mail to pdr@nrc.gov.

   The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By June 27, 2002, 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, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852, 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. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

   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 20555-001, Attention: Rulemakings and Adjudications Staff, or may be delivered to the Commission's

   [[Page 36939]]

   Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-001, 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 the factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

   Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, Unit 2, Hamilton County, Tennessee

   Date of amendment request: May 6, 2002, as supplemented on May 8, 2002 (TSC 02-05).

   Description of amendment request: The amendment changed the Technical Specifications (TSs) for Sequoyah Nuclear Plant, Unit 2. The proposed change would modify TS Surveillance Requirement 4.4.5.4.a.8 to clarify the scope of the steam generator (SG) tube inspections required in the SG tubesheet region.

   Date of issuance: May 10, 2002.

   Effective date: May 10, 2002.

   Amendment No.: 266.

   Facility Operating License No. DPR-79: The amendment revises the TSs.

   Public comments requested as to proposed no significant hazards consideration (NSHC): No. The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated May 10, 2002.

   Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

   NRC Section Chief: Thomas Koshy, Acting.

   Dated at Rockville, Maryland, this 20th day of May, 2002.

   For the Nuclear Regulatory Commission. John A. Zwolinski, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

   [FR Doc. 02-13082Filed5-24-02; 8:45 am]

   BILLING CODE 7590-01-P