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

[Federal Register: May 27, 2003 (Volume 68, Number 101)]

[Notices]

[Page 28843-28864]

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

[DOCID:fr27my03-85]

NUCLEAR REGULATORY COMMISSION

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

1. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, May 2, 2003, through May 15, 2003. The last biweekly notice was published on May 13, 2003 (68 FR 25648).

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

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public

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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 Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By June 26, 2003, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.714, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (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 filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

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

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

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415- 1101 or by e-mail to hearingdocket@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and because of continuing disruptions in delivery of mail to United States Government offices, it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent 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

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public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 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: April 21, 2003.

Description of amendment request: The licensee proposed to revise Sections 3.7 and 4.7, ``Auxiliary Electrical Power,'' of the Technical Specifications (TSs) to make them generally consistent with Nuclear Regulatory Commission (NRC) guidance set forth in NUREG-1433, ``Standard Technical Specifications, General Electric Plants, BWR

[Boiling Water Reactor] /4,'' Revision 2, and with the NRC-approved industry guidance identified as Technical Specification Task Force (TSTF) traveler TSTF-360, Revision 1. The amendment would also add a new Section 6.8.5, ``Station Battery Monitoring and Maintenance Program.'' The resulting Sections 3.7, 4.7, and 6.8.5 will be explicitly applicable to station batteries B and C, both safety-related subsystems, and their associated battery chargers. The proposed amendment would revise requirements concerning surveillance, monitoring, and maintenance of the subject batteries and chargers.

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

The first standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes, if approved by the NRC, will be made in a manner such that conservatism is maintained through compliance with applicable NRC regulations and guidance. No hardware design change is involved with the proposed amendment, thus there will be no adverse effect on the functional performance of any plant structure, system, or component (SSC). Consequently, all SSCs will continue to perform their design functions with no decrease in their capabilities to mitigate the consequences of postulated accidents. Station battery surveillance, monitoring, and maintenance were not previously factored into the probability of accidents, nor were they factored into scenarios of previously analyzed accidents. Consequently, the proposed revision to Sections 3.7, 4.7, and 6.8.5 of the TSs will lead to no increase in the consequences of an accident previously evaluated, and no increase of the probability of an accident previously evaluated.

The second standard requires that operation of the unit in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment is not the result of a hardware design change, nor does it lead to the need for a hardware design change. There is no change in the methods the unit is operated. As a result, all SSCs will continue to perform as previously analyzed by the licensee, and previously evaluated and accepted by the NRC staff. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any previously evaluated.

The third standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. Since the licensee did not propose to exceed or alter a design basis or safety limit, the proposed amendment will not affect in any way the performance characteristics and intended functions of any SSC. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Richard J. Laufer.

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

Date of amendments request: April 15, 2003.

Description of amendments request: The amendments would revise Sections 2.2, ``SL [Safety Limits] Violations,'' for reporting such violations to positions in the plant organization; 5.2.1, ``Onsite and Offsite Organization,'' for the position responsible for overall safe plant operation; and 5.5.1, ``Offsite Dose Calculation Manual (ODCM),'' for the position that approves changes to the ODCM, of the Technical Specifications (TSs). The revisions would account for the elimination of the positions of Vice President, Nuclear Production, and Director, Site Chemistry, and the assignment of the responsibilities of these positions in the above TS sections to other positions in the plant organization. Also, there would be the format change of adding the title of Section 2.2 near the top of TS page 2.0-2.

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

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

   These changes involve minor changes in the organization of PVNGS. It is expected that the organizational changes will have a positive effect on the conduct of plant operations and safety- related work. Functions which are necessary to operate the facility safely and in accordance with the operating licenses, remain in the re-aligned organization and will not affect the safe operation of the plant and continue to ensure proper control of administrative activities. The Quality Assurance (QA) organization reporting structure has not been affected by these changes allowing the QA organization to maintain the required authority and organizational freedom to identify quality problems; to initiate, recommend, or provide solutions; and to verify implementation of solutions. The proposed changes will not affect the operation of structures, systems, [or] components, and will not reduce programmatic controls such that plant safety would be affected. (The changes in the plant organization are also not initiators of an accident.) Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

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   The proposed changes will not affect the operation of structures, systems, [or] components, and will not reduce programmatic controls such that plant safety would be affected. The changes in the organization will continue to provide necessary oversight and control of administrative processes. [The changes in the plant organization are also not initiators of an accident.] Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

   These changes are administrative and will not diminish any organizational or administrative controls currently in place. The proposed changes will not affect the operation of structures, systems, [or] components, and will not reduce programmatic controls such that plant safety would be affected. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

   Based on the above, APS concludes that the activities associated with the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92 ``Issuance of Amendment,'' (c) and, accordingly, a finding of ``no significant hazards consideration'' is justified.

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

   Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona Public Service Company, PO Box 52034, Mail Station 7636, Phoenix, Arizona 85072-2024.

   NRC Section Chief: Stephen Dembek.

   Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

   Date of amendment request: May 1, 2003.

   Description of amendment request: The proposed amendment would increase the maximum enrichment limit of the fuel assemblies that can be stored in the Unit 1 spent fuel pool by taking credit for soluble boron in maintaining acceptable margins of subcriticality. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below:

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

   The proposed change will increase the maximum enrichment limit of the fuel assemblies that can be stored in the Unit 1 spent fuel pool (SFP) by taking credit for soluble boron in maintaining acceptable margins of subcriticality. The proposed change will modify Technical Specification 4.3.1 ``Criticality'' and add Technical Specification 3.7.16 ``Spent Fuel Pool Boron Concentration.'' The postulated accidents for the SFP are basically four types: (1) dropped fuel assembly on top of the storage rack, (2) a misloading accident, (3) an abnormal location of a fuel assembly, and (4) loss-of-normal cooling to the SFP.

   There is no increase in the probability of a fuel assembly drop accident in the SFP when considering the higher enriched fuel or the presence of soluble boron in the SFP water. Dropping a fuel assembly on top of the SFP storage racks is not credible at Calvert Cliffs due to the design of the spent fuel handling machine and due to the height of the SFP storage racks. The handling of the fuel assemblies has always been performed in borated water and will not change as a result of crediting soluble boron in the SFP criticality analysis. The proposed change does not change the general design and characteristics of the fuel assemblies. Therefore, the proposed change does not increase the probability of a fuel assembly drop accident.

   There is no increase in the probability of the accidental misloading of irradiated fuel assemblies into the SFP storage racks when considering the higher enriched fuel or the presence of soluble boron in the SFP water for criticality control. Fuel assembly placement will continue to be controlled pursuant to approved fuel handling procedures.

   Due to the design of the SFP storage racks, an abnormal placement of a fuel assembly into the SFP storage racks is not possible. Also, the design of the SFP prevents an inadvertent placement of a fuel assembly between the outer most storage cell and the pool wall. The proposed change does not make any change to the design of SFP. Therefore, there is no increase in the probability of abnormal placement of a fuel assembly into the SFP storage racks.

   The proposed change will not result in any changes to the SFP cooling system, and the fuel assembly design and characteristics are not changed by an increase in fuel enrichment. Therefore, there is no increase in the probability of a loss of SFP cooling. Also, since a high concentration of soluble boron has always been maintained in the SFP water, there is no increase in the probability of the loss of normal cooling to the SFP water considering the presence of soluble boron in the pool water for criticality control.

   There is no increase in the consequences of an accidental drop or accidental misloading of a maximum enriched fuel assembly into the SFP storage racks, because the criticality analysis demonstrates that the pool will remain subcritical following either event, even if the pool contains a boron concentration less than the proposed Technical Specification limit. The proposed Technical Specification limit will ensure that an adequate SFP boron concentration will be maintained.

   There is no increase in the consequences of a loss-of-normal SFP cooling because the Technical Specification boron concentration provides significant negative reactivity. Loss of the SFP water via boiling will not result in a loss of soluble boron, since the soluble boron is not volatile. Therefore, loss of spent fuel pool cooling system without makeup flow is not a mechanism for boron dilution. Even in the unlikely event that soluble boron in the SFP is completely diluted via unborated makeup flow, a pool completely filled with maximum enriched unburned assemblies will remain subcritical by a design margin of k-effective not to exceed 0.986.

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

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

   The proposed change will increase the maximum enrichment limit of the fuel assemblies that can be stored in the Unit 1 SFP by taking credit for soluble boron in maintaining acceptable margins of subcriticality. Increasing the maximum enrichment limit does not create a new type of criticality accident.

   Soluble boron has been maintained in the SFP water and is currently required by procedures. Therefore, crediting soluble boron in the SFP criticality analysis will have no effect on normal pool operation and maintenance. Crediting soluble boron will only result in increased sampling to verify the boron concentration. This increased sampling will not create the possibility of a new or different kind of accident.

   A dilution of the SFP soluble boron has always been a possibility. However, the boron dilution event previously had no consequences, since boron was not previously credited in the accident analysis. The initiating events that were considered for having the potential to cause dilution of the boron in the SFP to a level below that credited in the criticality analyses fall into three categories: dilution by flooding, dilution by loss-of-coolant induced makeup, and dilution by loss-of-cooling system induced makeup. The spent fuel pool dilution analysis demonstrates that a dilution that could increase the rack k-effective greater than 0.95 is not a credible event. It is not credible that dilution could occur for the required length of time without operator notice, since this event would activate the high level alarm and initiate Auxiliary Building flooding. In addition, in excess of 1,043,000 gallons of unborated water must be added to the SFP to reach the minimum soluble boron concentration. This is more water volume than is contained in both pretreated water storage tanks and also more water volume than is contained in the demineralized water storage tank and both condensate storage tanks combined. Even in the unlikely event that soluble boron in the SFP is completely diluted, the SFP will remain subcritical by a design margin of k-effective will not exceed 0.986.

   The proposed change will not result in any other change in the plant configuration or equipment design. Therefore, the proposed

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

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

   The Technical Specification changes proposed by this license amendment request will provide an adequate safety margin to ensure that the stored fuel assembly array of maximum enriched fuel will always remain subcritical. Those limits are based on a plant specific criticality analysis performed for the Calvert Cliffs Unit 1 SFP, that include technically supported margins.

   While the criticality analysis utilized credit for soluble boron, the SFP rack k-effective will remain less than 0.986 with no soluble boron with a 95 percent probability at a 95 percent confidence level. This substantial reduction in the SFP soluble boron concentration was evaluated and shown not to be credible. Soluble boron is used to provide subcritical margin such that the spent fuel pool k-effective is maintained less than or equal to 0.95. Since k-effective is less than or equal to 0.95, the current margin of safety is maintained.

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

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

   NRC Section Chief: Richard J. Laufer.

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

   Date of amendments request: April 17, 2003.

   Description of amendments request: The proposed amendment would (1) make 19 specific changes to the Technical Specifications actions currently requiring suspension of operations involving positive reactivity additions, and (2) revise various notes precluding reduction in boron concentration. The proposed changes follow the guidance of Technical Specification Task Force (TSTF) Change Traveler 286, Revision 2.

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

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

   The intent of this change is to clarify those Technical Specifications involving positive reactivity additions to the shutdown reactor so that small, controlled, safe insertions of positive reactivity will be allowed where they are now categorically prohibited, posing operational difficulties. These controlled activities could result in a slight change in the probability of an event occurring as Reactor Coolant System (RCS) manipulations that are currently prohibited would now be allowed. However, RCS manipulations are rigidly controlled to minimize the possibility of a significant reactivity increase. In addition, there is sufficient shutdown margin available in these conditions to allow for these slight reactivity changes without significantly increasing the probability of an accident previously evaluated.

   The proposed change does not permit the shutdown margin required by the Technical Specifications to be reduced. While the proposed change will permit changes in the discretionary boron concentration above the technical specification requirements, this excess concentration is not credited in the Updated Final Safety Analysis Report safety analysis. Because the initial conditions assumed in the safety analysis are preserved, no increase in the consequence of an accident previously evaluated would occur. In addition, small temperature changes in the RCS impose reactivity changes by means of the moderator temperature coefficient of reactivity. These small changes are within the required shutdown margin, therefore, there is no increase in the consequence of an accident previously evaluated.

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

2. Would not create the possibility of a new or different kind of accident from any accident previously evaluated.

   This proposed amendment allows for minor plant operational adjustments without adversely impacting the safety analysis required shutdown margin. It does not involve any change to plant equipment or the shutdown margin requirements in the Technical Specifications.

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

3. Would not involve a significant reduction in [a] margin of safety.

   The margin of safety in Modes 3, 4, 5, and 6 is preserved by the calculated shutdown margin which prevents a return to criticality. The proposed change will permit reductions in the discretionary shutdown margin beyond the Technical Specification requirements. However, the shutdown margin required by the Technical Specifications is not changed. The proposed change only affects Reactor Coolant System temperature and boron concentration above the calculated shutdown margin. By not impacting the shutdown margin, the margin of safety is not affected.

   Therefore, the proposed change will not involve a significant reduction in [a] margin of safety.

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

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

   NRC Section Chief: Richard J. Laufer.

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

   Date of amendment request: February 13, 2003.

   Description of amendment request: The proposed amendment would allow the use of an alternative source term (AST) methodology in accordance with 10 CFR 50.67 based on a reevaluation of the loss-of- coolant accident (LOCA) design-basis accident (DBA). Using an approved AST, the licensee has also proposed changes to increase the allowable secondary containment bypass and main steam isolation valve (MSIV) leakage limits and eliminate the MSIV leakage control system. The licensee also proposed changes to the TS Bases.

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

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

   The implementation of AST assumptions has been evaluated in a revision to the analysis of the Loss of Coolant Accident (LOCA) and an update to the analysis of the Fuel Handling Accident (FHA).

   Based upon the results of the analyses, it has been demonstrated that, with the requested changes, the dose consequences of these limiting Design Basis Accidents (DBAs) are within the regulatory guidance provided by the NRC for use with the AST. This guidance is presented in 10 CFR 50.67, Regulatory Guide 1.183 [``Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors''], and Standard Review Plan (SRP) Section 15.0.1.

   The requirements for MSIV [main steam isolation valve] Leakage Control System operability for eliminating MSIV leakage to the environment are being eliminated. This is acceptable because, with the application of AST, this system is no longer credited in mitigating the consequences of a LOCA or any other DBA.

   The proposed changes also increase the limits on maximum allowable leakage from

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   secondary containment bypass and main steam isolation valves, and on unfiltered inleakage into the Control Room. This is acceptable due to the new assumptions used in calculating Control Room and offsite dose following the affected design basis accident using the AST methodology.

   The proposed changes do not affect the normal design or operation of the facility before the accident; rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequence. The radiological consequences of the analyzed DBAs have been evaluated with application of AST assumptions. The results conclude that the radiological consequences remain within applicable regulatory limits. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The application of AST does not affect the design, functional performance or normal operation of the facility. Similarly, it does not affect the design or operation of any component in the facility such that new equipment failure modes are created. Elimination of the MSIV Leakage Control System cannot create a new accident because it is used as a mitigation system to limit MSIV leakage after the accident has occurred. Similarly, the use of Standby Liquid Control System to buffer suppression pool pH to prevent iodine reevolution is another mitigation function credited after the accident has occurred and; therefore, cannot create a new accident.

   As such the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   This proposed license amendment involves changes from the original source term developed in accordance with Technical Information Document (TID) 14844 to a new AST, as described in Regulatory Guide 1.183. The results of the DBA analyses and the requested Technical Specification changes, are subject to revised acceptance criteria. The analyses have been performed using conservative methodologies.

   Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed events have been carefully selected and margin has been retained to ensure that the analysis adequately bounds postulated event scenario. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67, Regulatory Guide 1.183 and SRP Section 15.0.1.

   The margin of safety is that provided by meeting the applicable regulatory limits. The effect of relaxation of these design and Technical Specification requirements has been analyzed and doses resulting from the design basis accidents have been found to remain within the regulatory limits. The changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits.

   Therefore, operation of Fermi 2 in accordance with the proposed changes will not involve a significant reduction in a margin of safety.

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

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

   NRC Section Chief: L. Raghavan.

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

   Date of amendment request: February 13, 2003.

   Description of amendment request: The proposed amendment would revise the Technical Specification (TS) Section 5.5.10, ``Technical Specification (TS) Bases Control Program,'' to be consistent with changes made to 10 CFR 50.59, which were published in the Federal Register on October 4, 1999 (64 FR 53582), and which became effective March 13, 2001.

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

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

   The proposed change deletes the reference to ``unreviewed safety question'' as defined in 10 CFR 50.59. Deletion of the definition of ``unreviewed safety question'' was approved by the NRC with the revision of 10 CFR 50.59. This change is administrative in nature. Consequently, the probability of an accident previously evaluated is not significantly increased. Changes to the TS Bases are still evaluated in accordance with 10 CFR 50.59. As a result, the probability or consequences of any accident previously evaluated are not significantly affected. There is no increase in the radiological dose at the site boundary for any previously evaluated accident. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

   The proposed change does not involve a physical alteration of the plant (i.e., no new or different types of equipment will be installed) or a change to the methods governing normal plant operation. These changes are considered administrative in nature and do not modify, add, delete, or relocate any technical requirements in the TS. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

   The proposed change will not reduce a margin of safety because it has no direct effect on any of the safety analysis assumptions. Changes to the TS Bases that result in meeting the criteria in paragraph 10 CFR 50.59(c)(2) continue to require NRC approval pursuant to 10 CFR 50.59. This change is administrative in nature based on the revision to 10 CFR 50.59. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

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

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

   NRC Section Chief: L. Raghavan.

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

   Date of amendment request: March 31, 2003.

   Description of amendment request: The proposed amendment would revise Technical Specification (TS) Surveillance Requirement (SR) 3.7.3.6 associated with the verification of control room emergency filtration (CREF) system duct work unfiltered inleakage. This amendment request supercedes DECo's previous amendment request dated September 26, 2002, in its entirety. The September 26, 2002, amendment request was previously noticed in the Federal Register on November 26, 2002 (67 FR 70765).

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

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

   This license amendment proposes an alternative test for performing the (Control Room Emergency Filtration) CREF system surveillance associated with measuring the

   [[Page 28849]]

   Control Room Envelope (CRE) unfiltered inleakage. The CREF system provides a configuration for mitigating radiological consequences of accidents; however, it does not involve the initiation of any previously analyzed accident. Similarly, the implementation of compensatory measures to address the failure of the surveillance to meet the design basis unfiltered inleakage limits is required to mitigate the consequences of a radiological release. Therefore, the proposed changes cannot increase the probability of any previously evaluated accident.

   The CREF system provides a radiologically controlled environment from which the plant can be safely operated following a radiological accident. Design basis accident analyses conclude that radiological consequences are within the regulatory acceptance criteria. The current TS surveillance (SR 3.7.3.6) measures inleakage from four sections of CREF system duct work outside the CRE that are at negative pressure during accident conditions. The proposed Tracer Gas test provides a measurement of CRE inleakage from all potential sources including the four sections of duct work. Measuring the CRE inleakage using Tracer Gas testing has no effect on the CREF system function. The results of Tracer Gas testing will be evaluated against the assumptions in the approved Alternative Source Term (AST) design basis accident analyses and compensatory measures will be implemented, as necessary, to ensure compliance with 10 CFR 50.67. If compliance with 10 CFR 50.67 cannot be demonstrated or if compensatory measures have been in place for more than 18 months, a conservative plant shutdown will be required to minimize risk. Therefore, the proposed changes do not significantly increase the radiological consequences of any previously evaluated accident.

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

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

   The proposed changes do not alter the design function or operation of the system involved. The CREF system will still provide protection to control room occupants in case of a significant radioactive release. The revised TS surveillance requirements provide an alternative test method that has been widely accepted for the measurement of CRE unfiltered inleakage. The proposed changes do not introduce any new modes of plant or CREF system operation. Therefore, the proposed changes do not create the potential for a new or different kind of accident from any accident previously evaluated.

3. The changes do not involve a significant reduction in the margin of safety.

   The proposed changes to the Fermi 2 TS surveillance requirements do not affect the radiological release from a design basis accident nor the postulated dose to the control room occupants as a result of the accident. The alternate surveillance test requirements provide an acceptable approach for the measurement of CRE inleakage. Safety margins and analytical conservatisms are included in the analyses to ensure that all postulated event scenarios are bounded. The proposed TS requirements continue to ensure that the radiological consequences at the control room are below the corresponding regulatory guidelines and that compliance with 10 CFR 50.67 and GDC (General Design Criterion)-19 is not affected. Therefore, the proposed changes will not result in a significant reduction in the margin of safety.

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

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

   NRC Section Chief: L. Raghavan.

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

   Date of amendment request: April 7, 2003

   Description of amendment request: The proposed amendment would move selected Technical Specification (TS) parameters to the Core Operating Limits Reports (COLR). Specifically, the changes proposed affect TSs 2.2, ``Limiting Safety System Settings, Table 2.2-1;'' 3/4.1.1.1.1, ``Reactivity Control Systems, Boration Control, SHUTDOWN MARGIN--Modes 3, 4, and 5 Loops Filled;'' 3/4.1.1.2, ``Reactivity Control Systems, SHUTDOWN MARGIN--Cold Shutdown--Loops Not Filled;'' 3/4.2.5, ``Power Distribution Limits, DNB Parameters;'' 3/4.3.5, ``Instrumentation, SHUTDOWN MARGIN Monitor;'' 3/4.9.1.1, ``Refueling Operations, Boron Concentration;'' Section 6.9.1.6.a, ``Core Operating Limits Report, Core Operating Limits;'' and Section 6.9.1.6.b, ``Core Operating Limits Report, The Analytical Methods Used to Determine the Core Operating Limits,'' and the corresponding pages and Bases sections.

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

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

   The relocation of cycle-specific core operating limits from the technical specifications to the COLR has no influence or impact on the probability or consequences of a Design Basis Accident. Adherence to the COLR and methodologies acceptable for establishing COLR parameters continues to be controlled by Technical Specifications. The proposed amendment still requires exactly the same actions to be taken when or if limits are exceeded. Each accident analysis addressed in the Final Safety Analysis Report (FSAR) will be examined with respect to the changes in cycle- dependent parameters, which are obtained from application of the Nuclear Regulatory Commission (NRC) approved reload design methodologies, to ensure that the transient evaluation of new core designs are bounded by previously accepted analysis. This examination, which will be performed in accordance with the requirements of 10 CFR 50.59, ensures that future designs will not involve a significant increase in the probability or consequences of an accident previously evaluated.

   The proposed change to add new document references to Technical Specification Sections 6.9.1.6.b.16 and 6.9.1.6.b.17 are required to identify the most recent methodology to be used in the Millstone Unit No. 3 Small Break Loss of Coolant Accident (SBLOCA) analysis. Section 6.9.1.6.b.18 is added to describe NRC approved Overpower DT and Overtemperature DT trip function methodology. The use of these methodologies demonstrates that the acceptance criteria for SBLOCA events and Overpower DT and Overtemperature DT are met. This change has no impact on plant equipment operation. Since these changes only affect the method of analysis, they cannot affect the likelihood or consequences of accidents. Therefore, these changes will not increase the probability or consequences of an accident previously evaluated.

   Deleting the revision number and the date from the documents contained in Technical Specification Section 6.9.1.6.b.1 and in Technical Specification Sections 6.9.1.6.b.4 through 6.9.1.6.b.10 has no impact on the actual analytical methods used to determine the core operating limits, nor does it affect the likelihood or consequences of accidents. Therefore, this change will not increase the probability or consequences of an accident previously evaluated.

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

   As stated earlier, the relocation of the cycle-specific variables to the COLR, adding new document references and deleting the revision number and the date in Technical Specification Section 6.9.1.6.b have no influence or impact, nor does it contribute in any way to the probability or consequences of an accident. No safety related equipment, safety function, or plant operations will be altered as a result of this proposed change. The cycle specific variables are calculated using NRC-approved methods and submitted to the NRC to allow the Staff to continue to trend the values of these limits. The Technical Specifications will continue to require operation within the required core operating limits and appropriate actions will be taken when or if limits are exceeded. Therefore the proposed amendment does not in any way create the possibility of a new or

   [[Page 28850]]

   different kind of accident from any accident previously evaluated.

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

   The proposed changes have no impact on plant equipment operation. The proposed changes do not revise any setpoints assumed in the analyses and do not affect the acceptance criteria for SBLOCA analyses. Therefore, the proposed changes will not result in a 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: Lillian M. Cuoco, Senior Nuclear Counsel, Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.

   NRC Section Chief: James W. Clifford.

   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: April 10, 2003.

   Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) for the low temperature overpressure protection system. Currently, TS Surveillance Requirement (SR) 3.4.12.5 requires performance of a channel functional test for the power-operated relief valve within 12 hours of decreasing reactor coolant system (RCS) temperature to