Agency information collection activities: Operating licenses, amendments; no significant hazards considerations; biweekly notices,

[Federal Register: March 5, 2002 (Volume 67, Number 43)]

[Notices]

[Page 10006-10022]

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

[DOCID:fr05mr02-86]

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 February 8, 2002 through February 21, 2002. The last biweekly notice was published on February 19, 2002 (67 FR 7410).

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.

[[Page 10007]]

However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC'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 April 4, 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

[[Page 10008]]

Electronic Reading Room on the internet at the NRC Web site, http:// www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 304-415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: September 10, 2001.

Description of amendment request: The proposed amendment would revise the requirements in Technical Specifications (TSs), Sections 3.4.A.7.c and 3.4.A.8.c, to determine operability of core spray pumps and system components by verification rather than testing.

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 TS changes are not associated with accident initiators. The proposed changes are, however, associated with emergency core cooling requirements for loss of coolant mitigation. This event is a loss of coolant from the reactor vessel when the plant is shutdown and was evaluated in the NRC [Nuclear Regulatory Commission] Safety Evaluation Report supporting License Amendment No. 12, dated January 21, 1976. The proposed changes contained in this request do not affect the assumptions or conclusions of that evaluation and do not impact the physical characteristics of the core spray and fire protection systems. Therefore, the proposed changes do not degrade the ability of the core spray and fire protection systems to perform their intended accident mitigation function. The proposed changes to core spray pump/component and fire protection system operability verification versus demonstration in TS 3.4.A.7.c and core spray pump/component operability verification versus demonstration in TS 3.4.A.8.c provide an alternate means of determining equipment operability without reliance on frequent testing. The clarification of the extent of core spray system operability verification in TS 3.4.A.7.c does not change any existing requirements. Therefore, the proposed changes to TS 3.4.A.7.c and 3.4.A.8.c 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 TS changes are not associated with accident initiators. They are changes that provide an alternate means of determining equipment operability while eliminating frequent testing.

   The proposed changes to TS 3.4.A.7.c and 3.4.A.8.c do not involve the addition of any new plant structure, system or component (SSC). Similarly, the proposed TS changes do not involve physical changes to an existing SSC nor do they modify any current operating parameters. Providing an alternate means of determining equipment operability does not alter the functional capability of any accident mitigation system. The clarification of the extent of core spray system operability verification in TS 3.4.A.7.c does not change any existing requirements. Therefore, the proposed changes do 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 changes to TS 3.4.A.7.c and 3.4.A.8.c are not associated with accident initiators and do not introduce new SSCs or physically impact existing SSCs. They are changes that provide an alternate means (i.e., verification) of determining core spray and fire protection system component operability. The capability of the necessary core spray and fire protection components to provide the required core cooling flow is demonstrated during surveillance testing. While the proposed changes revise the method of determining the operability of the core spray and fire protection system in the reduced availability mode, they do not degrade the ability of the systems to perform their intended function. The clarification of the extent of core spray system operability verification in TS 3.4.A.7.c does not change any existing requirements. Therefore, the proposed TS change does not involve a significant reduction in a margin of safety.

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

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

   NRC Section Chief: Joel Munday, Acting.

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

   Date of amendment request: September 11, 2001.

   Description of amendment request: The proposed amendment would revise the Technical Specifications, Section 3.9, ``Refueling,'' to incorporate compensatory provisions which permit fuel-handling operations without the refueling interlocks operable.

   Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. 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 proposed amendment involves refueling interlock operability requirements during refueling operations. The only design-basis accident described in the Oyster Creek Updated Final Safety Analysis Report (UFSAR) for cold shutdown or refueling conditions is a postulated fuel handling (dropped bundle) accident. The refueling interlocks are not postulated to cause, and are not involved in the mitigation of such an accident. Thus, the proposed amendment does not affect the safety function of the refueling interlocks. The proposed alternative actions provide an equivalent level of protection against inadvertent criticality during fuel handling operations. 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 parameter, condition, 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 parameter, analysis methodology, safety limits or acceptance criteria. The revised requirement (i.e., proposed alternative) will continue to ensure against inadvertent criticality during fuel handling operations. 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

   [[Page 10009]]

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

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

   NRC Section Chief: Joel Munday, Acting.

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

   Date of amendment request: August 14, 2001.

   Description of amendment request: The proposed amendment revises Section 6, ``Administrative Controls,'' of the Technical Specifications (TSs) to delete Section 6.5.4, ``Independent Onsite Safety Review Group,'' and all associated subsections. The licensee will revise its Operational Quality Assurance Plan to incorporate conforming changes to provide its proposed alternative independent nuclear safety oversight provisions.

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

   This change involves deletion of the TS requirements for the Independent Onsite Safety Review Group [IOSRG]. To satisfy the NUREG-0737 [``Clarification of TMI Action Plan Requirements,'' November 1980] guidance concerning organizational independence, the proposed IOSRG alternative provides for technical expertise by onsite engineering and licensing organizations. These site engineering and licensing organizations report through the Site Vice-President and are independent of the production reporting chain through the plant manager. Additionally, high-level management positions are located in the corporate and regional offices for these engineering and licensing organizations which set policy and have responsibility for governance and oversight of these functional areas. These corporate and regional high-level positions are not in the management chain for power production.

   Organizational and procedural changes at TMI Unit 1 [Three Mile Island Nuclear Station, Unit 1] following the issuance of NUREG-0737 have resulted in improvements to the review processes that meet the intent of the requirements [of] NUREG-0737 for an IOSRG. Therefore, inclusion of the IOSRG in the plant or plant support organization is unneccessary. In light of the considerable improvement in the processes listed above, the contribution of three full time engineers assigned as a separate group to address nuclear safety oversight is not significant in comparison to the contribution of the overall organization. This change does not affect assumptions contained in the plant safety analyses, the physical design and/or operation of the plant, nor does it affect Technical Specifications that preserve safety analysis assumptions. No Technical Specification Limiting Condition for Operation, Action Statement, or Surveillance Requirement is affected by this change. The proposed change does not alter design, function, operation, or reliability of any plant component. This change does not involve a physical modification to the plant, a mode of operation, or a change to the UFSAR [Updated Final Safety Analysis Report] transient analyses. Normal and accident dose to plant personnel or to the public are unaffected.

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

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

   This change to remove the IOSRG from the TS[s] is administrative in nature and does not affect the assumptions contained in the plant safety analyses, the physical design and/or modes of plant operation defined in the plant operating license that preserve safety analysis assumptions.

   This proposed change does not introduce a new mode of plant operation or surveillance requirement, nor involve a physical modification to the plant. The proposed change does not alter the design, function, or operation of any plant system or component.

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

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

   This change only involves Technical Specification Section 6, ``Administrative Controls,'' which does not include any margins of safety. None of the proposed changes involve a physical modification to the plant, a new mode of operation, an instrument setpoint, or a change to the UFSAR transient analyses. No Limiting Safety System Setting, Technical Specification Limiting Condition for Operation, Action Statement, or Surveillance Requirement is affected. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

   NRC Section Chief: Joel T. Munday (Acting).

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

   Date of amendment request: September 10, 2001.

   Description of amendment request: The proposed amendment would revise Technical Specification 3/4.9.7 and corresponding Bases to address use of a single-failure-proof handling system, as defined by NUREG-0612 (``Control of Heavy Loads at Nuclear Power Plants'') and NUREG-0554 (``Single-Failure-Proof Cranes For Nuclear Power Plants''). The modifications will allow handling loads in excess of 1,800 pounds near or over the Spent Fuel Pool. The anticipated types of heavy loads include the combination of a spent fuel storage canister and transfer cask.

   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.

   Concerning the application of a single-failure-proof handling system for handling heavy loads near or over the Spent Fuel Pool, NUREG-0612, ``Control of Heavy Loads at Nuclear Power Plants'' asserts that the probability of an accidental load drop while handling loads over the spent fuel is insignificant.

   Under the proposed amendment, the evaluation criteria of NUREG- 0612, Section 5.1 are satisfied by the combination of (a) the continued implementation of procedures and the practices for both the Fuel Handling Cranes and the Yard Crane that provide conformance with the guidelines of Section 5.1.1 of NUREG-0612, and (b) the application of a single-failure-proof handling system that satisfies the criteria of NUREG-0612, Sections 5.1.2(1) and 5.1.6 for the movement of any load with a weight greater than 1800 pounds either (i) over any spent fuel assembly in the Spent Fuel Pool or (ii) near or over any area of the Spent Fuel Pool, including the Spent Fuel Cask Laydown Area.

   The proposed amendment retains existing restrictions on crane travel for the Fuel Handling Cranes, which are not qualified to the single-failure-proof criteria of NUREG-0612. These retained restrictions continue to support the existing safety analysis of Section 15.2.2, ``Fuel Handling Accident'' of the

   [[Page 10010]]

   UFSAR [Updated Final Safety Analysis Report], Reference (9) [Haddam Neck Plant UFSAR Change 34 dated August 2, 2000].

   Additionally, the proposed amendment corresponds to the application of a single-failure-proof handling system to fulfill the NUREG-0612 Phase II condition that is required prior to the handling of a spent fuel cask near or over any area of the Spent Fuel Pool.

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

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

   The proposed changes will allow the handling by a single- failure-proof handling system of loads in excess of 1800 pounds over fuel assemblies in any region of the Spent Fuel Pool, including the Spent Fuel Cask Laydown Area.

   Additionally, the proposed changes correspond to the application of a single-failure-proof handling system for the fulfillment of the required condition for the handling of spent fuel casks near or over any area of the Spent Fuel Pool. This required condition is identified in the documentation for the NRC Issuance of License Amendment 125, Ref. (7) [Letter from US NRC to CYAPCO, dated April 26, 1990] and it is acknowledged in the CYAPCO submittal for the proposed license amendment that was issued as License Amendment 188, Ref. (5) [Letter from J. F. Opeka (CYAPCO) to US NRC, ``Haddam Neck Plant Proposed Revision to Technical Specifications Spent Fuel Pool Capacity Expansion,'' Letter Number B15136, dated March 31, 1995.] and the NRC Issuance of License Amendment 195, Ref. (6) [Letter from T. L. Fredrichs (NRC) to R. A. Mellor (CYAPCO), ``Haddam Neck Plant- Issuance of Amendment RE: Relocation of Requirements to Licensee-- Controlled Documents (TAC No. MA5756),'' dated October 19, 1999].

   NUREG-0612, Section 5.1.2 identifies that the capability of a single-failure-proof handling system to handle heavy loads has been identified as equivalent in risk to the capabilities of a non- single-failure-proof heavy load handling system that complies with the criteria of one of the other three alternative sets from NUREG- 0612 (including alternative criteria that include analyses concerning postulated heavy load drops).

   A structural evaluation of the heavy load interfaces within the Spent Fuel Cask Laydown Area and the Cask Transfer Bay was performed per the requirements of EDR-1 [(Reference 2) Generic Licensing Topical Report EDR-I (P)-A, ``EDERER's Nuclear Safety Related eXtra Safety And Monitoring (X-SAM) CRANES,'' Revision 3, Amendment 3, dated October 8, 1982] Appendix B and C (Attachments 2 and 3

   [attachments to this application] ). The results of the evaluation confirmed the design bases for the Spent Fuel Pool and the Spent Fuel Building are maintained.

   As such, use of a single-failure-proof handling system precludes the possibility of a heavy load drop which could cause an accident outside of the existing design bases.

   Additionally, the proposed changes retain existing restrictions on the travel of non-single-failure-proof cranes over fuel assemblies in the Spent Fuel Pool. These retained restrictions continue to support the existing safety analysis of Section 15.2.2, ``Fuel Handling Accident'' of the UFSAR, Reference (9).

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

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

   Section 5.1.2 of NUREG-0612 identifies that each of the four alternative sets of criteria for the handling of heavy loads near or over the Spent Fuel Pool, including over fuel assemblies, provides a level of safety that is essentially equivalent to the level of safety provided by any of the other three alternative sets of criteria.

   The proposed change corresponds to the application of the first of the four alternative sets of criteria, which is described in NUREG-0612 Section 5.1.2(1), implementation of a single-failure- proof handling system.

   Additionally, the proposed change includes the retention of existing crane travel restrictions for the Fuel Handling Cranes, therefore, maintaining the existing margin of safety concerning the operation of those other cranes.

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

   Attorney for licensee: Mr. Robert K. Gad, III, Ropes & Gray, One International Plaza, Boston, Massachusetts 02110-2624.

   NRC Section Chief: Stephen Dembek.

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

   Date of amendment request: December 20, 2001.

   Description of amendment request: The amendments would revise the Technical Specifications (TS) to eliminate the use of the term ``unreviewed safety question.'' The change is proposed by the licensee to reflect changes in the NRC's regulations in 10 CFR 50.59 as noticed in the Federal Register on October 4, 1999. The proposed changes in the license amendment request are consistent with an NRC approved Technical Specifications Task Force Standard TS Traveler (TSTF-364).

   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 implementation of the changes proposed in this LAR

   [license amendment request] involve a significant increase in the probability or consequences of an accident previously evaluated?

   No. This LAR makes an administrative change to the Technical Specifications [TS] made necessary as part of Duke's implementation of revised NRC regulations. The changes proposed to these TS have no substantive impact on the Catawba licensing bases, nor Duke's ability to conservatively evaluate changes to these licensing bases. Therefore, the proposed changes have no impact on any accident probabilities or consequences.

2. Would implementation of the changes proposed in this LAR create the possibility of a new or different kind of accident from any accident previously evaluated?

   No. This LAR makes administrative changes that have no impact on any accident analyses.

3. Would implementation of the changes proposed in this LAR involve a significant reduction in a margin of safety?

   No. The proposed changes are administrative, an implementation of the revised 10CFR50.59 regulation. Implementation of the revised 10CFR50.59 regulation provides the necessary regulatory requirements to ensure that nuclear plants' margin of safety is preserved.

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

   Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

   NRC Section Chief: Richard J. Laufer, Acting.

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

   Date of amendment request: December 20, 2001.

   Description of amendment request: The amendments would revise Technical Specification 5.6.5.b to eliminate the revision number and dates of the topical reports that contain the analytical methods used to determine the core operating limits. This proposed change is consistent with TSTF (Technical Specification Task Force)-363.

   [[Page 10011]]

   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 implementation of the changes proposed in this LAR

   [license amendment request] involve a significant increase in the probability or consequences of an accident previously evaluated?

   No. This LAR makes an administrative change to TS 5.6.5.b, Core Operating Limits Report (COLR), affecting a list of documents that are separately reviewed and approved by the NRC. The changes proposed to TS 5.6.5.b have no substantive impact on the Catawba licensing bases. Only NRC-approved methodologies will be used to generate the core operating limits. Based on these considerations, it has been determined that the proposed changes have no impact on any accident probabilities or consequences.

2. Would implementation of the changes proposed in this LAR create the possibility of a new or different kind of accident from any accident previously evaluated?

   No. This LAR makes administrative changes that have no impact on any accident analyses.

3. Would implementation of the changes proposed in this LAR involve a significant reduction in a margin of safety?

   No. The analytical methodologies used to generate the core operating limits are unchanged by this LAR. As such, this LAR has no affect on margins of safety. Future changes to these methodologies will remain subject to NRC review and approval. Therefore, this proposed amendment does not involve a reduction in any margin of safety.

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

   Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

   NRC Section Chief: Richard J. Laufer, Acting.

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

   Date of amendment request: December 20, 2001.

   Description of amendment request: The proposed amendments would revise the Technical Specification 5.6.5.b to eliminate the revision number and dates of the topical reports that contain the analytical methods used to determine the core operating limits. This proposed change is consistent with TSTF (Technical Specification Task Force)- 363. This notice supersedes in its entirety the previous notice issued on February 5, 2002 (67 FR 5326) for the Oconee December 20, 2001, application, which contained the incorrect licensee analysis.

   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 implementation of the changes proposed in this LAR

   [license amendment request] involve a significant increase in the probability or consequences of an accident previously evaluated?

   No. This LAR makes an administrative change to TS 5.6.5.b, Core Operating Limits Report (COLR), affecting a list of documents that are separately reviewed and approved by the NRC. The changes proposed to TS 5.6.5.b have no substantive impact on the Oconee licensing bases. Only NRC-approved methodologies will be used to generate the core operating limits. Based on these considerations, it has been determined that the proposed changes have no impact on any accident probabilities or consequences.

2. Would implementation of the changes proposed in this LAR create the possibility of a new or different kind of accident from any accident previously evaluated?

   No. This LAR makes administrative changes that have no impact on any accident analyses.

3. Would implementation of the changes proposed in this LAR involve a significant reduction in a margin of safety?

   No. The analytical methodologies used to generate the core operating limits are unchanged by this LAR. As such, this LAR has no affect on margins of safety. Future changes to these methodologies will remain subject to NRC review and approval. Therefore, this proposed amendment does not involve a reduction in any margin of safety.

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

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

   NRC Section Chief: Richard J. Laufer, Acting.

   Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

   Date of amendment request: January 8, 2002.

   Description of amendment request: The amendment would delete the Technical Specification (TS) requirements governing the reactor vessel material surveillance program; would change the TS Sections 4.2, ``Inservice Inspection and Testing,'' 5.2.C, ``Design Features-- Containment,'' and 6.4, ``Administrative Controls--Training,'' to correct errors; and would change TS Section 6.1, ``Responsibility,'' and 6.2, ``Organization,'' to reflect the organizational changes resulting from the license transfer to Entergy Nuclear Operations, Inc. (ENO).

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

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

   The proposed change to TS Section 3.1.B involves deleting specific TS requirements that duplicate the requirements of 10 CFR

   [Code of Federal Regulations] 50.60, 10CFR50 Appendix G, and 10CFR50 Appendix H. The proposed change does not result in a change to the design or operation of any plant structure, system or component. Therefore any assumptions of the operability or performance of any structure, system or component in accident evaluations are unchanged.

   The proposed change to TS 4.2.1 simply corrects an improper reference to the CFR. There are no physical changes to IP2 or to the operation of any system, structure, or component.

   The proposed change to TS 5.2.C makes the design feature description consistent with TS Limiting Condition for Operation 3.3.B wherein the requirements for the method of post-accident iodine removal are specified. Making the Design Feature consistent with the appropriate LCO has no effect on the assumptions and the results of the accident analyses.

   TS sections 6.1, 6.2, and 6.4 are administrative controls. Changing an administrative control has no affect on accident analyses.

   Therefore, there will be no increase in the probability or in the consequences of an accident previously evaluated.

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

   The proposed change to TS Section 3.1.B does not affect the effectiveness of ENO's implementation of the requirements of 10CFR50.60 that ensure the reactor vessel continues to be protected against non-ductile failure.

   There is no change to any system, structure, or component as a result of any of the proposed changes.

   [[Page 10012]]

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

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

   The proposed TS changes simplify the methods of controlling the schedule for the reactor vessel surveillance specimen withdrawal schedule in that a duplicative control is removed. The effectiveness of ENO compliance with 10CFR50.60 and 10CFR50 Appendices G and Appendix H is not adversely affected by this change. The level of regulatory control for the reactor vessel pressure/temperature limits is not changed.

   The effectiveness of IP2's [Indian Point 2] inservice testing program is not affected by the correction of the improper CFR reference in TS 4.2.1. ENO is required to comply with 10CFR55 at IP2. The effectiveness of ENO's compliance with 10CFR55 is not affected by deleting the improper CFR citation from TS 6.4. Similarly, ENO's compliance with the IP2 license and the all applicable laws and regulations is not affected by the proposed changes to the TS sections for responsibility and organization.

   The change to the Design Features to properly identify the method specified in TS 5.2.B for post-accident iodine removal does not affect the margin of safety.

   This change does not affect any design function for or the operation of any plant structure, system, or component.

   Therefore, the change [* * *] does not result in a change to any of the safety analyses or [a] margin of safety.

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

   Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

   NRC Section Chief: Joel Munday, Acting.

   Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

   Date of amendment request: January 8, 2002.

   Description of amendment request: The amendment would incorporate the use of a more conservative equation to calculate the power range high neutron trip setpoint when one or more main steam safety valves are inoperable during four loop operation--Technical Specification (TS) 3.4, ``Steam and Power Conversion System.''

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

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

   The proposed change to the setpoints will cause a reactor trip on high neutron flux for a decreased heat removal event at an earlier (more conservative) condition. The consequences of an accident with the proposed setpoints are less severe than those predicted with the use of the current setpoints.

   The main steam line code safety valves, in conjunction with the high neutron flux reactor trip mitigate the consequences of decreased heat removal and uncontrolled rod cluster assembly bank withdrawal events. The systems acting together do not initiate or cause any accident. Therefore, the probability of analyzed accidents is unchanged by the proposed TS change.

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

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

   There is no change to either the design or operation of the main steam line code safety valves. This proposed change only changes the high neutron flux trip setpoints in response to the inoperable main steam line code safety valves. This feature currently exists both in the plant and in the TS.

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

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

   The current TS setpoints have been determined to be non- conservative and insufficient to guarantee safety. The proposed change would impose limits that were anticipated in the original TS and are conservative with respect to the current TS. Therefore, the margin of safety as defined in the TS (protection of the secondary system from overpressurization so that it is available for decay heat removal) will be restored to that intended with the original TS.

   The ability to keep the core cooled in spite of the inoperability of some main steam line code safety valves is enhanced by the proposed change. Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in [a] margin of safety.

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

   Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

   NRC Section Chief: Joel Munday, Acting.

   Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

   Date of amendment request: January 8, 2002.

   Description of amendment request: The amendment would revise the Technical Specification (TS) Section 3.7.C, ``Gas Turbine Generators,'' and Section 4.6, ``Emergency Power System Periodic Tests,'' by changing the requirement to maintain a minimum amount of fuel oil stored on site from 54,200 gallons to 94,870 gallons.

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

   The Gas Turbine Generators only provide a Licensing Basis Event mitigating function. There is no previously evaluated accident or event that is initiated by the Gas Turbine Generators or their associated fuel storage system. The ability of the Gas Turbine Generators to provide power, as a backup to the Emergency Diesel Generators, is enhanced by the proposed change to increase the amount of fuel stored on site and dedicated to Gas Turbine Generator operation. The increase in minimum load has an insignificant affect because the Gas Turbine Generators are capable of loads far in excess of the proposed minimum load.

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

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

   There is no physical change to the plant. The currently existing fuel oil storage facilities will be used. The only change is to increase the minimum amount of fuel oil that must be maintained at the plant.

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

3. Operation of the facility in accordance with the proposed amendment would not

   [[Page 10013]]

   involve a significant reduction in [a] margin of safety.

   The proposed limit for Gas Turbine Generator fuel oil storage ensures compliance with the current licensing basis that the Gas Turbine Generators be able to power all the loads required by 10 CFR

   [Code of Federal Regulations] 50 Appendix R to place the plant into a safe shutdown condition following a fire and maintain safe shutdown for three days. The increase in the minimum load rating ensures that each Gas Turbine Generator will support operation of additional components to enhance operational flexibility in response to an event.

   Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in [a] margin of safety.

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

   Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

   NRC Section Chief: Joel Munday, Acting.

   Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

   Date of amendment request: January 8, 2002.

   Description of amendment request: The amendment would delete the Technical Specification (TS) requirements governing the Fuel Storage Building Air Filtration System. The proposed changes affect TS 3.8, ``Refueling, Fuel Storage and Operations with the Reactor Vessel Head Bolts Less Than Fully Tensioned,'' and TS 4.5.F, ``Fuel Storage Building Air Filtration System.''

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

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

   The fuel storage building air filtration system is not involved in the initiation of any accident nor does it function to prevent any accident. The fuel storage building air filtration system was an accident mitigating system. Therefore there is no affect on the probability of occurrence of a fuel handling accident in the fuel storage building.

   The fuel storage building air filtration system was designed to provide an accident mitigation function by filtering the radionuclides that might have been released from a damaged fuel assembly in the event of a fuel handling accident. The charcoal adsorber was the primary component that supported this filtration function. However, based on the recent IP2 [Indian Point 2] analyses to show compliance with 10CFR [Code of Federal Regulations] 50.67, it has been shown that the doses to the public and to control room operators due to a fuel handling accident remain well within regulatory limits even assuming no credit for either isolation or filtration. Therefore, the charcoal filtration function is not required in the event of a fuel handling accident.

   There would be no change to the radiological consequences of the fuel handling accident in the fuel storage building analysis as a result of the proposed change. The proposed changes ensure that the assumptions of the fuel handling accident analysis for the release of radioactivity from a damaged fuel assembly in the fuel storage building are maintained.

   Therefore, there will be no increase in the probability or in the consequences of an accident previously evaluated.

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

   The fuel storage building air filtration system is not an accident initiator. It was designed as an accident mitigation system to filter the radionuclides that may be released from a damaged fuel assembly during a fuel handling accident. The fuel storage building air filtration system does not affect any accident initiator.

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

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

   The margin of safety is defined by 10CFR50.67 and 10CFR50 Appendix A Criterion 19. The radiological consequences of a fuel handling accident in the fuel storage building have been shown to be well within the regulatory requirements even when assuming no credit for the fuel storage building air filtration system operation.

   The proposed change ensures that the assumptions of the current fuel handling analysis for the release of radioactivity from a damaged fuel assembly are maintained.

   Therefore, the change does not result in a change to any of the safety analyses or [a] margin of safety.

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

   Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear. Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

   NRC Section Chief: Joel Munday, Acting.

   Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear Power Station, Unit 2, Grundy County, Illinois

   Date of amendment request: September 5, 2001.

   Description of amendment request: The proposed amendments would revise the battery terminal voltage on float charge for the alternate battery.

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

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

   The change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change is to a SR 3.8.4.1 acceptance criterion that will continue to ensure equipment operability. By continuing to ensure equipment operability, the probability or consequences of an accident previously evaluated are not increased. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The 125 VDC alternate battery continues to provide reliable DC power for operation of the required equipment. The number of cells in the alternate battery was increased from sixty to sixty-three and the acceptable float voltage needed to be revised to reflect the additional cells. The addition of the three cells has been evaluated and documented in calculations. These calculations demonstrate that the batteries are appropriately sized to supply the required loads following a loss of offsite power. The ability of the battery to perform its intended function remains unchanged. In addition, the proposed change has no impact on any initial condition assumptions for accident scenarios. Onsite or offsite dose consequences resulting from an accident previously evaluated are not affected by this proposed amendment request.

   Accordingly, there is no significant change in the probability or consequences of an accident previously evaluated.

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

   The proposed license amendment provides a change in a TS Surveillance Requirement that continues to ensure equipment operability. The increase in terminal voltage specifically supports the increase in the number of cells for the battery. The operation of the safety-related equipment and components remains unchanged. As such, the relationship between the 125 VDC power

   [[Page 10014]]

   system and plant transient response is maintained. The change in the acceptance criterion ensures that the equipment remains operable.

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

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

   The proposed change does not involve a significant reduction in the margin of safety. The proposed change continues to ensure equipment operability. The increase in terminal voltage specifically supports the increase in the number of cells for the alternate battery. Since the change maintains the necessary level of system reliability, it does not involve a significant reduction in the margin of safety. The change in acceptance criterion is to reflect the increase in battery cells from sixty to sixty-three. This acceptance criterion ensures that the equipment remains operable.

   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.

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

   Date of amendment request: January 10, 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 January 10, 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 proposes to determined that the amendment requests involve no significant hazards consideration.

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

   NRC Section Chief: Stephen Dembek.

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

   Date of amendment request: December 10, 2001.

   Description of amendment request: The proposed amendments would revise the Technical Specifications (TSs) to incorporate the Nuclear Regulatory Commission (NRC)-approved generic change TSTF-287, Revision 5, to the ``Standard Technical Specifications for General Electric Plants (BWR/4),'' NUREG-1433, Revision 1. Specifically, the proposed changes would: (a) insert a note in the Limiting Condition for Operation (LCO) in TS 3.7.3 to state that the control room habitability envelope boundary may be opened intermittently under administrative control; (b) insert a new LCO Action B in TS 3.7.3 to allow 24 hours to restore the control room habitability envelope boundary to

   [[Page 10015]]

   operable status if two control room emergency outside air supply (CREOAS) subsystems should become inoperable due to an inoperable control room habitability envelope boundary in Modes 1, 2 and 3; (c) re-label the existing LCO Actions b, c, d, and e to c, d, e, and f respectively; and (d) revise the existing LCO Action D to require immediate entry into LCO 3.0.3 when two CREOAS subsystems are inoperable for situations other than when the inoperability is due to an inoperable control room habitability envelope boundary. Minor formatting and editorial changes are also made.

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

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

   The proposed change relaxes the Required Actions of LCO 3.7.3 by allowing 24 hours to restore an inoperable control room habitability envelope pressure boundary to OPERABLE status. Required Actions and their associated Completion Times are not initiating events for any accidents previously evaluated. The accident analyses do not assume that required CREOAS equipment is out of service prior to the analyzed event. Consequently, this change in Required Actions does not significantly increase the probability of occurrence of any accident previously evaluated. The Required Actions in the proposed change have been developed to provide assurance that appropriate remedial actions are taken in response to the degraded condition, considering the operability status of the CREOAS system and the capability of minimizing the risk associated with continued operation. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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

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

   The relaxed Required Actions do not involve a significant reduction in the margin of safety. The proposed change has been evaluated to minimize the risk of continued operation with the control room habitability envelope pressure boundary inoperable. The operability status of the CREOAS system, a reasonable time for repairs or replacement of required features, and the low probability of a design basis accident occurring during the repair period have been considered in the evaluation. Therefore, this change does not involve a significant reduction in a margin of safety.

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

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

   NRC Section Chief: Joel T. Munday, Acting.

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

   Date of amendment request: December 14, 2001.

   Description of amendment request: A change is proposed to Surveillance Requirement (SR) 3.0.3 to allow a longer period of time to perform a missed surveillance. The time is extended from the current limit of ``* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less'' to ``* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.'' In addition, the following requirement 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 December 14, 2001.

   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

   [[Page 10016]]

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

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

   NRC Section Chief: Richard J. Laufer, Acting.

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

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

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

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

   Date of amendment request: January 31, 2002.

   Brief description of amendment request: The proposed amendment would revise the technical specifications by replacing the peak linear heat rate safety limit with a peak fuel centerline temperature safety limit.

   Date of publication of individual notice in Federal Register: February 11, 2002 (67 FR 6279).

   Expiration date of individual notice: The comment period expires on February 25, 2002, and the hearing period expires on March 13, 2002.

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

   Date of amendment request: January 31, 2002.

   Brief description of amendment request: The proposed amendment would replace the Technical Specification (TS) Safety Limit 2.1.1.2, ``Peak Linear Heat Rate,'' (PLHR) with a Peak Fuel Centerline Temperature Safety Limit and update the Index accordingly. The associated TS Bases changes are also made to appropriately reflect the proposed new Safety Limit.

   Date of publication of individual notice in Federal Register: February 11, 2002 (67 FR 6281).

   Expiration date of individual notice: The comment period expires on February 25, 2002, and the hearing period expires on March 13, 2002.

   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.

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

   Date of application for amendment: August 21, 2001, as supplemented January 11, 2002.

   Brief description of amendment: The amendment revises the actions taken for an inoperable battery charger, revises the battery charger testing criteria, and relocates certain safety-related battery surveillance requirements from the Technical Specifications to a licensee-controlled program.

   Date of issuance: February 15, 2002.

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

   Amendment No.: 142.

   Facility Operating License No. NPF-62: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: November 14, 2001 (66 FR 57118). The letter of January 11, 2002, provided clarification and did not affect the NRC staff's proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 15, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: December 19, 2000, as supplemented on September 24, 2001.

   Brief description of amendment: The amendment revises the Oyster Creek Technical Specifications, Section 3.17, to remove reference to the current licensing basis control room calculated dose consequences and substitute the associated regulatory dose limits that apply for control room habitability in

   [[Page 10017]]

   accordance with General Design Criterion (GDC) 19 of 10 CFR part 50 and Standard Review Plan Section 6.4. Concurrent with this requested change, AmerGen Energy Company (AmerGen) recalculated control room relative concentration (X/Q) values using the ARCON96 methodology to demonstrate its capability to meet GDC-19 dose requirements. The NRC staff finds acceptable the use of the diffuse source X/Q values calculated by AmerGen because they appear to be more limiting than assuming a point source release through a building penetration.

   Date of Issuance: February 7, 2002.

   Effective date: February 7, 2002 and shall be implemented within 30 days of issuance.

   Amendment No.: 225.

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

   Date of initial notice in Federal Register: August 22, 2001 (66 FR 44163). The September 24, 2001, letter provided clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated February 7, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: September 20, 2000, as supplemented August 2 and September 28, 2001.

   Brief description of amendment: The Technical Specification (TS) changes deleted the specification for hydrogen monitoring instrumentation from TS Sections 3.5.5.2, 3.6, and Tables 3.5-3 and 4.1-4, corrected a typographical error in Item 8 of Table 4.1-4, deleted the specifications for hydrogen recombiners in TS Section 4.4.4, and changed the Bases for TS Section 4.12.2 to delete its reference to hydrogen purge and hydrogen recombiners.

   Date of issuance: February 8, 2002.

   Effective date: As of the date of issuance and shall be implemented within 30 days including the designation of hydrogen monitoring instrumentation as Category 3 variables as defined in Regulatory Guide 1.97.

   Amendment No.: 240.

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

   Date of initial notice in Federal Register: November 14, 2001 (66 FR 57118). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 8, 2002.

   No significant hazards consideration comments received: No.

   Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

   Date of application for amendment: November 19, 2001.

   Brief description of amendment: The amendment increases the allowed outage time of one train of the control room emergency ventilation system from 10 to 14 days (for the loss of the emergency power supply only). This is a one-time change to support corrective maintenance and inspections of the 1A Diesel Generator during the Unit 1 refueling outage.

   Date of issuance: February 13, 2002.

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

   Amendment No.: 223.

   Renewed License No. DPR-69: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: January 8, 2002 (67 FR 926). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 13, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: June 29, 2000, as supplemented on October 16, 2000, and January 25, April 4, and September 21, 2001.

   Brief description of amendment: The amendment modifies Technical Specification (TS) Sections 3.3.2, ``Instrumentation--Engineered Safety Features Actuation System Instrumentation;'' 3.7.7, ``Plant Systems-- Control Room Emergency Ventilation System;'' 3.7.8, ``Plant Systems-- Control Room Envelope Pressurization System;'' 3.7.9, ``Plant Systems-- Auxiliary Building Filter System;'' 3.9.1.1, ``Refueling Operations-- Boron Concentration;'' 3.9.1.2, ``Refueling Operations--Boron Concentration;'' 3.9.2, ``Refueling Operations--Instrumentation;'' 3.9.4, ``Refueling Operations--Containment Building Penetrations;'' 3.9.9, ``Refueling Operations--Containment Purge and Exhaust Isolation System;'' 3.9.10, ``Refueling Operations--Water Level--Reactor Vessel;'' and 3.9.12, ``Refueling Operations--Fuel Building Exhaust Filter System.'' The changes are associated with the revised fuel handling accident analyses, and integrity of the Control Room and the Fuel Building boundaries. Several administrative changes were also made to reflect Millstone Unit 3 terminology, remove unnecessary information, and eliminate confusion by providing consistency between LCO's (limiting conditions for operations), Action Requirements, and Surveillance Requirements. The TS changes are associated with the revised containment fuel handling accident analysis which results in an increase in the consequences of a containment fuel handling accident since the current analysis of a containment fuel handling accident does not assume the release of any radioactive material from containment. The revised analysis assumes a release of radioactive material because it assumes both personnel access hatch doors are open and at least one hatch door is closed within 10 minutes of a fuel handling accident inside containment.

   Date of issuance: February 20, 2002.

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

   Amendment No.: 203.

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

   Date of initial notice in Federal Register: November 29, 2000 (65 FR 71136). The letters dated October 16, 2000, January 25 April 4, and September 21, 2001, provided clarifying information and did not change the staff's initial proposed no significant hazards consideration determination or expand the scope of the application as published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 20, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendments: March 22, 2001, as supplemented by letter dated October 11, 2001.

   Brief description of amendments: The amendments revised the Technical Specifications (TS) surveillance requirement (SR) for the methodology and frequency for the chemical analyses of the ice condenser ice bed. Also, these amendments add a new TS SR to address sampling requirements for ice

   [[Page 10018]]

   additions to the ice bed. In addition, the amendments revise the current TS surveillance requirement acceptance criteria and surveillance frequency for the inspection of ice condenser ice basket flow channel areas.

   Date of issuance: February 11, 2002.

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

   Amendment Nos.: 195 and 188.

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

   Date of initial notice in Federal Register: July 11, 2001 (66 FR 36339). The supplement dated October 11, 2001, provided clarifying information that did not expand the scope of the original Federal Register notice or the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 11, 2002.

   No significant hazards consideration comments received: No.

   Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

   Date of application for amendment: July 13, 2001.

   Brief description of amendment: The amendment deleted Technical Specification (TS) Tables 3.6-1, ``Non-Automatic Containment Isolation Valves Open Continuously or Intermittently for Plant Operation,'' and 4.4-1, ``Containment Isolation Valves.'' The amendment also revised other TS sections that reference these tables. The removal of the tables is in accordance with the guidance in NRC Generic Letter 91-08, ``Removal of Component Lists from Technical Specifications.''

   Date of issuance: February 12, 2002.

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

   Amendment No.: 223.

   Facility Operating License No. DPR-26: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 22, 2001 (66 FR 44166). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 12, 2002.

   No significant hazards consideration comments received: No.

   Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2, Westchester County, New York

   Date of application for amendment: July 16, 2001, as supplemented January 11, 2002.

   Brief description of amendment: The amendment updates the pressure- temperature limit curves for Indian Point Nuclear Generating Unit No. 2.

   Date of issuance: February 15, 2002.

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

   Amendment No.: 224.

   Facility Operating License No. DPR-26: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 8, 2001 (66 FR 41613). The January 11, 2002, letter provided clarifying information that 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 February 15, 2002.

   No significant hazards consideration comments received: No.

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

   Date of amendment request: July 23, 2001, as supplemented by letters dated September 21, and November 8, 2001.

   Brief description of amendment: This submittal requests a change to administrative Technical Specification (TS) 6.15. The change postpones the next Type A test performed after May 12, 1991, to no later than May 11, 2006, resulting in an extended interval of 15 years for the performance of Integrated Leak Rate Test.

   Date of issuance: February 14, 2002.

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

   Amendment No.: 178.

   Facility Operating License No. NPF-38: The amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 22, 2001 (66 FR 44169). The September 21, and November 8, 2001, supplemental letters contained clarifying information that did not change the scope of the July 23, 2001, application nor the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 14, 2002.

   No significant hazards consideration comments received: No.

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

   Date of amendment request: July 23, 2001, as supplemented December 11, 2001.

   Brief description of amendment: As a follow-up response to a commitment identified in the Nuclear Regulatory Commission (NRC) staff letter dated December 22, 2000, ``Completion of Licensing Action for Generic Letter (GL) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions,'' Entergy Operations Inc., (Entergy, the licensee) has proposed to revise their Waterford Steam Electric Station, Unit 3 (Waterford 3) Final Safety Analysis Report (FSAR) to resolve the ten containment penetrations susceptible to thermally induced overpressurization through an evaluation, detailed analysis, or installation of physical modifications prior to startup from the spring 2002 refueling outage. Entergy determined a change to Waterford 3's license basis, through procedural controls, risk analysis, and engineering analysis, for seven penetrations, as discussed in this license basis change request was necessary. Permanent resolution to the GL 96-06 issues for the remaining three penetrations will be satisfied through the installation of physical modifications.

   Date of issuance: February 19, 2002.

   Effective date: This license amendment is effective as of its date of issuance and shall be implemented prior to startup from Refuel 11 scheduled for March 2002. Implementation of the amendment is the incorporation into the FSAR of the changes to the description of the facility as described in the licensee's application dated July 23, 2001, as supplemented by letter dated December 11, 2001.

   Amendment No.: 179.

   Facility Operating License No. NPF-38: The amendment revised the FSAR.

   Date of initial notice in Federal Register: September 19, 2001 (66 FR 48285). The December 11, 2001, supplement contained clarifying information that did not change the scope of the July 23, 2001, application nor the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 19, 2002.

   No significant hazards consideration comments received: No.

   [[Page 10019]]

   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 application for amendments: July 6, 2001, as supplemented by letters dated October 25, 2001, and December 17, 2001.

   Brief description of amendments: The amendments revise technical specifications (TS) Section 3.3.1.1, ``Reactor Protection System Instrumentation,'' to modify the description for Reactor Protection System (RPS) Function 7.a, ``Scram Discharge Volume Water Level-- High.'' This change supports a planned upgrade to the scram discharge volume level instrumentation from Fluid Components International thermal switches to Magnetrol float switches. These float switches are more reliable than the existing thermal switches, which are highly sensitive to a steam environment, since they respond to actual water level increases within the scram discharge volume. These types of Magnetrol float switches are used successfully in various applications at Quad Cities.

   Date of issuance: February 11, 2002.

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

   Amendment Nos.: 203 and 199.

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

   Date of initial notice in Federal Register: January 11, 2002 (67 FR 1520). The October 25, 2001, and December 17, 2001, supplements provided clarifying information that was within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards considerations determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 11, 2002.

   No significant hazards consideration comments received: No.

   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 application for amendments: August 13, 2001, as supplemented by letters December 17 and December 26, 2001, and January 10, 2002.

   Brief description of amendments: The amendments revise the Technical Specifications to support the licensee's planned upgrade of the reactor water level instrumentation, including changes to surveillance requirements frequencies, functional testing, and allowable values.

   Date of issuance: February 12, 2002.

   Effective date: As of the date of issuance and shall be implemented for Unit 1 prior to reaching Startup (i.e., Mode 2) following refueling outage 17, scheduled for completion in November 2002, and for Unit 2 prior to reaching Startup (i.e., Mode 2) following refueling outage 16, scheduled for completion in February 2002.

   Amendment Nos.: 204 and 200.

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

   Date of initial notice in Federal Register: October 17, 2001 (66 FR 52800). The December 17 and December 26, 2001, and January 10, 2002, supplements provided clarifying information that was within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards considerations determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 12, 2002.

   No significant hazards consideration comments received: No.

   FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

   Date of application for amendment: June 29, 2001, as supplemented by letters dated October 4, and December 1, 2001.

   Brief description of amendment: The amendment revised the pressure- temperature curves and the cold overpressure protection limits. The changes are supported by a new fluence determination based on evaluation of a surveillance capsule, and the use of the American Society of Mechanical Engineers Code Case N-640.

   Date of issuance: February 20, 2002.

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

   Amendment No: 249.

   Facility Operating License No. DPR-66: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: October 17, 2001 (66 FR 52801). The supplemental letters provided additional information but did not change the initial proposed no significant hazards consideration determination or expand the amendment beyond the scope of the initial notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 20, 2002.

   No significant hazards consideration comments received: No.

   FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

   Date of application for amendment: March 28, 2001, as supplemented by letter dated September 25, 2001.

   Brief description of amendment: The amendment revised the technical specifications associated with crediting soluble boron for reactivity control in the spent fuel pool.

   Date of issuance: February 11, 2002.

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

   Amendment No: 128.

   Facility Operating License No. NPF-73: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 8, 2001 (66 FR 41620). The September 25, 2001, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the original Federal Register notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 11, 2002.

   No significant hazards consideration comments received: No.

   FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

   Date of application for amendment: June 28, 2001, as supplemented September 13, 2001, December 19, 2001, and January 21, 2002.

   Brief description of amendment: The amendment revised the technical specification (TS), 3.1.1.4, upper limit for the moderator temperature coefficient (MTC), from 0 x 10 -4change in reactivity per degree Fahrenheit (k/k/ deg.F) to +0.2 x 10 -4 k/k/ deg.F for power levels up to 70 percent of rated thermal power (RTP), and ramping linearly to 0 x 10 -4k/ k/ deg.F from 70 percent to 100 percent RTP. The change is needed to address future core designs with higher energy requirements, associated with plant operation at higher capacity factors.

   Date of issuance: February 21, 2002.

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

   Amendment No: 129.

   Facility Operating License No. NPF-73. Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: October 31, 2001 (66 FR 55019). The September 13, 2001,

   [[Page 10020]]

   December 19, 2001, and January 21, 2002, letters provided clarifying information that 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 February 21, 2002.

   No significant hazards consideration comments received: No.

   FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

   Date of application for amendment: April 1, 2001.

   Brief description of amendment: FirstEnergy Nuclear Operating Company submitted License Amendment Request (LAR) 00-0003 for Davis- Besse Nuclear Power Station, Unit 1, for review and approval. This amendment request proposes to revise references in the Technical Specification (TS) to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code), Section XI, as the source of requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The TS will reference the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code).

   Date of issuance: January 31, 2002.

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

   Amendment No.: 250.

   Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: May 30, 2001 (66 FR 29375). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2002.

   No significant hazards consideration comments received: No.

   FirstEnergy Nuclear Operating Company, Docket No. 50346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

   Date of application for amendment: May 22, 2001, as supplemented by letters dated November 15, 2001, February 12, 2002, and electronic transmission dated February 19, 2002.

   Brief description of amendment: This amendment revised Davis-Besse Nuclear Power Station (DBNPS) Technical Specifications in accordance with Framatone Technologies Incorporated Topical Report BAW-2303P, Revision 4, ``OTSG Repair Roll Qualification Report.'' The changes revise the existing DBNPS Once-Through Steam Generators (OTSGs) repair roll requirements to (1) use updated limiting tensile tube loads, (2) define new exclusion zones within the steam generator in which application of the repair roll is prohibited, (3) allow the repair roll to be used in the lower tubesheet area, (4) remove the limitation of only one repair roll per OTSG tube, and (5) replace the requirement that the repair roll be one inch in length with a requirement that the repair roll be installed in accordance with Topical Report BAW-2303P, Revision 4.

   Date of issuance: February 20, 2002.

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

   Amendment No.: 252.

   Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: August 8, 2001 (66 FR 41621). The supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 20, 2002.

   No significant hazards consideration comments received: No.

   FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

   Date of application for amendment: May 15, 2001, as supplemented by letter dated February 8, 2002.

   Brief description of amendment: This amendment revises Technical Specification (TS) 3/4.9.4, ``Refueling Operations--Containment Penetrations,'' TS 3/4.9.12, ``Refueling Operations--Storage Pool Ventilation,'' and associated Bases. The changes will allow the containment equipment hatch cover to be removed during core alterations and movement of irradiated fuel inside containment provided the Emergency Ventilation System is operable with the ability to filter any radioactive release.

   Date of issuance: February 14, 2002.

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

   Amendment No.: 251.

   Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: June 27, 2001 (66 FR 34284). The supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 14, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendment: December 5, 2001.

   Brief description of amendment: This amendment revised Technical Specification 5.5.11, ``Technical Specifications (TSs) Bases Control Program,'' to reflect Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-364, Revision 0, ``Revision to Technical Specification Bases Control Program to Incorporate Changes to 10 CFR 50.59.''

   Date of Issuance: February 21, 2002.

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

   Amendment No.: 121.

   Facility Operating License No. NPF-58: This amendment revised the Technical Specifications.

   Date of initial notice in Federal Register: January 8, 2002 (67 FR 927). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 21, 2002.

   No significant hazards consideration comments received: No.

   Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York

   Date of application for amendment: March 29, 2001, as supplemented October 30, 2001.

   Brief description of amendment: The amendment revised the Technical Specifications (TS) and TS Bases to eliminate the requirements for certain engineered features operability during core alterations and movement of irradiated fuel which had decayed for at least 2 days.

   Date of issuance: February 11, 2002.

   Effective date: As of the date of issuance, to be implemented prior to Refueling Outage 8.

   Amendment No.: 101.

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

   Date of initial notice in Federal Register: May 30, 2001 (66 FR 29358). The licensee's October 30, 2001, supplement withdrew portions of the

   [[Page 10021]]

   original application, leaving the balance unchanged.

   The staff's related evaluation of the amendment is contained in a Safety Evaluation dated February 11, 2002.

   No significant hazards consideration comments received: No.

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

   Date of application for amendments: July 17, 2001, as supplemented July 26, and October 15, 2001.

   Brief description of amendments: The amendments update the reactor pressure vessel pressure-temperature limit curves for SSES-1 and 2.

   Date of issuance: February 7, 2002.

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

   Amendment Nos.: 200, 174.

   Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the Technical Specifications.

   Date of initial notice in Federal Register: October 3, 2001 (66 FR 50471). The July 26, and October 15, 2001, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 7, 2002.

   No significant hazards consideration comments received: No.

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

   Date of amendment request: November 8, 2001, as supplemented on December 14, 2001.

   Brief description of amendments: The amendments add the following to the Technical Specifications: (1) The phrase, ``or if open, capable of being closed'' to Limiting Condition for Operation 3.9.4 for the equipment hatch, during core alterations or movement of irradiated fuel assemblies inside containment, and (2) the requirement to verify the capability to install the equipment hatch in a new Surveillance Requirement (SR) 3.9.4.2. The previous SR 3.9.4.2 was renumbered SR 3.9.4.3, but not changed.

   Date of issuance: February 20, 2002.

   Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance, including the incorporation of the changes to the Technical Specification Bases as described in the licensee's application dated November 8, 2001.

   Amendment Nos.: 93 and 93.

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

   Date of initial notice in Federal Register: December 12, 2001 (66 FR 64307). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 20, 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 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

   [[Page 10022]]

   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 April 4, 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 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).

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

   Date of amendment request: February 8, 2002, as supplemented February 10, 2002.

   Description of amendment request: The amendment adds a license condition allowing a one-time limited duration exception from the Technical Specification (TS) Surveillance Requirement (SR) to verify that the opening, closing, and frictional torque of the ice condenser inlet doors are within specified limits as required by TS SRs 4.6.5.3.1.b.3, 4.6.5.3.1.b.4, and 4.6.5.3.1.b.5, respectively.

   Date of issuance: February 14, 2002.

   Effective date: As of the date of issuance, to be implemented immediately.

   Amendment No.: 265.

   Facility Operating License No. (DPR-58): Amendment revises the Operating License.

   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 February 14, 2002.

   Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

   NRC Section Chief: William D. Reckley, Acting Section Chief.

   Dated at Rockville, Maryland, this 25th day of February, 2002.

   For the Nuclear Regulatory Commission. Ledyard B. Marsh, Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

   [FR Doc. 02-5000Filed3-4-02; 8:45 am]

   BILLING CODE 7590-01-P