E5-1343. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and 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 March 4, 2005, through March 17, 2005. The last biweekly notice was published on March 15, 2005 (70 FR 12743).

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. Within 60 days after the date of publication of this notice, 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.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place 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 O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, 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 2.309, 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 within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include 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/requestor to relief. A petitioner/requestor who fails to satisfy 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.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, [e-mail]; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at [phone], verification number is [phone]. 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 it is requested that copies be transmitted either by means of facsimile transmission to [phone] or by e-mail to [e-mail]. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

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, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the 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 PDR Reference staff at 1 [phone], [phone] or by e-mail to [e-mail].

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

Date of amendment request: February 24, 2005.

Description of amendment request: The licensee proposed to revise Table 3.1.1, “Protective Instrumentation Requirements,” of the Technical Specifications to clarify the conditions under which the reactor building closed cooling water (RBCCW) pumps and the service water (SW) pumps will trip during a loss-of-coolant accident (LOCA). The revised wording would state that the RBCCW and SW pumps will trip during a LOCA only if offsite power is unavailable. The licensee also proposed to editorially move a footnote on page 3.6-1 to its correct place on page 3.6-2.

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

Response: No.

The proposed revision to Technical Specification (TS) Table 3.1.1 to clarify the tripping of the Service Water (SW) and Reactor Building Closed Cooling Water (RBCCW) pumps documents the as-built controls for these loads. Amendment No. 42 to the Oyster Creek Licensing Application concluded that these pumps are not required to perform any functions related to safe plant shutdown. During a loss of coolant accident (LOCA) condition, with offsite power available, the plant electrical busses have enough capacity and capability to supply the SW and RBCCW pumps. This proposed change is an administrative change only, and is being made to align the Oyster Creek Technical Specifications with the design of the plant. No physical changes are being made to the plant. Also, the footnote on TS page 3.6-1 would be relocated to TS page 3.6-2 to appear on the same TS page as the Specification to which it applies. The proposed changes do not alter the physical design or operational procedures associated with any plant structure, system, or component.

Therefore, the proposed changes do 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?

Response: No.

The proposed revision to Technical Specification Table 3.1.1 to clarify the tripping of the SW and RBCCW pumps documents as-built controls for these loads. These pumps provide cooling to various non-safety related plant equipment. Following a LOCA condition, with offsite power available, these pumps will help in removing plant heat loads. This clarification that the SW and RBCCW pumps do not trip during a LOCA, with offsite power available, does not affect the Emergency Diesel Generator time delayed loading sequence. The relocation of the footnote applicable to Specification 3.6.A.4.1 is editorial in nature and has no impact on any accident previously evaluated. Accordingly, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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 change involve a significant reduction in a margin of safety?

Response: No.

The proposed revision to Technical Specification Table 3.1.1 to clarify the tripping of the SW and RBCCW pumps documents as-built controls for these loads. The NRC Safety Evaluation Report (SER) for Amendment 42 to the Oyster Creek Licensing Application concluded that it is acceptable to automatically trip the SW and RBCCW pumps during a loss of coolant accident. The NRC SER for Technical Specification Amendment 60 concluded that the immediate tripping of the RBCCW pump and the time delayed tripping of the SW pumps during a LOCA was also acceptable. The clarification that the SW and RBCCW pumps do not trip during a loss of coolant accident when offsite power is available does not reduce any margin of safety because these pumps are not required to mitigate the consequences of any postulated accident. The relocation of the footnote applicable to Specification 3.6.A.4.1 is editorial in nature and has no impact on any accident margin of safety.

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

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

Attorney for licensee: Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: February 25, 2005.

Description of amendment request: The proposed changes would amend Operating License DPR-65 for Millstone Power Station, Unit No. 2 (MPS2) and Operating License NPF-49 for Millstone Power Station, Unit No. 3 (MPS3) by incorporating certain administrative changes into the MPS2 and MPS3 Technical Specifications (TSs).

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

Response: No.

The proposed changes are administrative in nature and do not alter any of the requirements of the affected TS[s]. The proposed changes do not modify any plant equipment and do not impact any failure modes that could lead to an accident. Additionally, the proposed changes have no effect on the consequence of any analyzed accident since the changes do not affect any equipment related to accident mitigation. Based on this discussion, the proposed amendment does not increase the probability or 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?

Response: No.

The proposed changes are administrative in nature. They do not modify any plant equipment and there is no impact on the capability of the existing equipment to perform their intended functions. No system setpoints are being modified and no changes are being made to the method in which plant operations are conducted. No new failure modes are introduced by the proposed changes. The proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident. 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 proposed amendment involve a significant reduction in a margin of safety?

Response: No.

These changes are administrative in nature and do not alter any of the requirements of the affected TS[s]. The proposed changes do not affect any of the assumptions used in the accident analysis, nor do they affect any operability requirements for equipment important to plant safety. Therefore, the proposed changes will not result in a significant reduction in the margin of safety as defined in the bases for technical specifications covered in this license amendment request.

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: Darrell J. Roberts.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: March 8, 2005.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 5.5.13, Primary Containment Leakage Rate Testing Program, for the Integrated Leak Rate Testing (ILRT) program to add an exception to the commitment to follow the guidelines of Regulatory Guide 1.163, “Performance-Based Containment Leak-Test Program.” The effect of this request would be a one-time extension of the interval since the last ILRT from 15 years to 15 years and 4 months (i.e., from August 2007 to December 2007).

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. Will operation of the facility in accordance with this proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment to TS 5.5.13 allows a one-time extension to the current interval for the ILRT. The current interval of fifteen years, based on past performance, would be extended on a one-time basis to 15-years and 4 months from the date of the last test. The proposed extension to the ILRT cannot increase the probability of an accident since there are no design or operating changes involved and the test is not an accident initiator. The proposed extension of the test interval does not involve a significant increase in the consequences since analysis has shown that, the proposed extension of the ILRT and DWBT [Drywell Bypass Test] frequency has a minimal impact on plant risk. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with this proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed extension to the interval for the ILRT does not involve any design or operational changes that could lead to a new or different kind of accident from any accidents previously evaluated. The tests are not being modified, but are only being performed after a longer interval. 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. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Will operation of the facility in accordance with this proposed amendment involve a significant reduction in a margin of safety?

Response: No.

An evaluation of extending the ILRT DWBT surveillance frequency from once in 10 years to once in 15 years and 4 months has been performed using methodologies based on the approved ILRT methodologies. This evaluation assumed that the DWBT frequency was being adjusted in conjunction with the ILRT frequency. This analysis used realistic, but still conservative, assumptions with regard to developing the frequency of leakage classes associated with the ILRT and DWBT. The results from this conservative analysis indicates that the proposed extension of the ILRT frequency has a minimal impact on plant risk and therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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

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

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: December 20, 2004.

Description of amendment request: Entergy Operations, Inc. is proposing that the Arkansas Nuclear One Unit 2 (ANO-2) Facility Operating License be amended to revise the requirements for ensuring containment structural integrity. The proposed changes modify the Containment Structural Integrity Technical Specification (TS) 3.6.1.5 to delete the existing Surveillance Requirements (SR) and add a new SR to verify containment structural integrity in accordance with a new Containment Tendon Surveillance Program. A new Containment Tendon Surveillance Program is added to TS 6.5.6 and a new reporting requirement is being added to TS 6.6.6. The proposed changes are generally consistent with NUREG 1432, “Standard Technical Specifications Combustion Engineering Plants,” Revision 3. This request for amendment also contains proposed administrative changes related to page numbering.

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

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

The containment building is not considered to be the initiator of any accident previously evaluated, but serves to mitigate accidents that could allow a release to the environment. The proposed TS change will provide for containment tendon inspections as required by 10 CFR 50.55a and prevent or inhibit release from the containment building as designed. Through appropriate inspections and implementation of corrective actions for any degradation discovered during the inspections that might lead to containment structural failures, the probability or consequences of accidents will not be increased.

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

The proposed change does not change the design, configuration, or method of operation of the plant. By implementing corrective actions for any degradation discovered during the required inspections of the containment, the possibility of a new or different kind of accident will not be created. Implementation of the requirements of Subsection IWL of the ASME code [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] and those of 10 CFR 50.55a(b)(2) provide an equally acceptable containment inspection program.

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

The proposed change to incorporate the applicable requirements of Subsection IWL of the ASME Code and of 10 CFR 50.55a(b)(2) into the ANO-2 containment inspection program has no impact on any safety analysis assumptions. The addition of structural integrity requirements to ANO-2 TS Specification 3.6.1.5 imposes consistent requirements with those previously specified in the ANO-2 TSs. The requirements of ASME IWL are more restrictive than those currently provided in the existing ANO-2 technical specifications. As a result, the margin of safety is not reduced by the proposed change.

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

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

NRC Section Chief: Allen G. Howe.

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

Date of amendment requests: February 25, 2005.

Description of amendment requests: The proposed amendments would modify the Technical Specifications by revising the near-end-of-life Moderator Temperature Coefficient (MTC) Surveillance Requirement by placing a set of conditions on core performance, which, if met, would allow conditional exemption from the required MTC measurement.

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?

Response: No.

The probability or consequences of accidents previously evaluated in the Updated Final Safety Analysis Report (UFSAR) are unaffected by this proposed change because there is no change to any equipment response or accident mitigation scenario. There are no additional challenges to fission product barrier integrity.

Therefore, the proposed 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?

Response: No.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system.

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

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

Response: No.

The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect on the availability, operability, or performance of the safety-related systems and components. A change to a surveillance requirement is proposed, but the limiting conditions for operation required by the Technical Specifications (TS) are not changed.

The Technical Specifications Bases are founded in part on the ability of the regulatory criteria to be satisfied assuming the limiting conditions for operation are met for the various systems. Conformance to the regulatory criteria for operation with the conditional exemption from the near-end of life moderator temperature coefficient (MTC) measurement is demonstrated and the regulatory limits are not exceeded. Therefore, the margin of safety as defined in the TS is not reduced.

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 requests involve no significant hazards consideration.

Attorney for licensee: James M. Petro, Jr., Esquire, One Cook Place, Bridgman, MI 49106.

NRC Section Chief: L. Raghavan.

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

Date of amendment request: March 8, 2005.

Description of amendment request: The proposed amendment would revise Technical Specification 2.1.1.2 for the single recirculation loop Safety Limit Minimum Critical Power Ratio (SLMCPR) value to reflect results of a cycle-specific calculation for Cycle 23 operations.

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

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

Response: No.

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. Changing the SLMCPR does not increase the probability of an evaluated accident. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected.

The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established, consistent with NRC approved methods, to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change conservatively establishes the safety limit for the minimum critical power ratio for CNS Cycle 23 such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.

The proposed change revises the SLMCPR to protect the fuel during normal operation as well as during any transients or anticipated operational occurrences. Operational limits Minimum Critical Power Ratio (MCPR) are established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences) is met. Since the operability of plant systems designed to mitigate any consequences of accidents has not changed, the consequences of an accident previously evaluated are not expected to increase.

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

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

Response: No.

Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration or changes in allowable modes of operation. The proposed change does not involve any modifications of the plant configuration or allowable modes of operation. The proposed change to the SLMCPR assures that safety criteria are maintained for Cycle 23.

Based on the above, NPPD concludes that the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1% of the rods are expected to be in boiling transition if the MCPR limit is not violated. The proposed change will ensure the appropriate level of fuel protection is maintained. Additionally, operational limits are established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences) are met.

Based on the above, NPPD concludes that the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: February 3, 2005.

Description of amendment request: The proposed amendments would modify the Technical Specifications (TSs) by revising TS 6.16.b.1, “Radioactive Effluent Controls Program,” to be consistent with the intent of 10 CFR 20 and NUREG-1431, “Standard Technical Specifications Westinghouse Plants” (STS).

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

NMC [Nuclear Management Company, LLC] Response:

No. Updating the specification to be consistent with 10 CFR 20 and the STS has no impact on plant structures, systems, or components, does not affect any accident initiators, and does not change any safety analysis. Therefore, the changes do not involve an increase in the probability or 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?

NMC Response:

No. Updating the specification to be consistent with 10 CFR 20 and the STS will not change any equipment, require new equipment to be installed, or change the way current equipment operates. No credible new failure mechanisms, malfunctions, or accident initiators are created by the proposed changes. Therefore, the 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?

NMC Response:

No. Updating the specification to be consistent with 10 CFR 20 and the STS has no impact on inputs to the safety analysis or to automatic plant actions. It also does not impact plant equipment or operation. Therefore, the change does not reduce the margin of safety.

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

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

NRC Section Chief: L. Raghavan.

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

Date of amendment request: October 15, 2004.

Description of amendment request: The proposed amendment revises TS 5.5.6, “Reactor Coolant Pump Flywheel Inspection Program,” to extend the allowable inspection interval to 20 years.

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated October 15, 2004.

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 to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's acceptance guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits; or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed 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 accident previously evaluated.

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed 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 a margin of safety.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.

NRC Section Chief: L. Raghavan.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: September 27, 2004.

Description of amendment request: The proposed amendment would revise the reactor coolant pump (RCP) flywheel inspection surveillance requirements to extend the allowable inspection interval to 20 years.

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated September 27, 2004.

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 to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's acceptance guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits; or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed 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 accident previously evaluated.

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed 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 a margin of safety.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: Darrell J. Roberts.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: January 11, 2005. Description of amendment request: The proposed amendment would delete the Technical Specification (TS) requirements to submit monthly operating reports and occupational radiation exposure reports.

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in licensing amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated January 11, 2005.

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

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

Response: No.

The proposed change eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed 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?

Response: No.

The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed 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?

Response: No.

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

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

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: Darrell J. Roberts.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: February 15, 2005.

Description of amendment request: The proposed amendment will revise the Salem, Unit Nos. 1 and 2 Technical Specifications to reflect the deletion of Reactor Coolant System (RCS) volume from design features Section 5.4.2. This design feature information will continue to be maintained in the plant's updated final safety analysis report.

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

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

Response: No.

The proposed change to remove this information from T/S [technical specifications] does not affect any accident initiators or precursors. Elimination of the RCS volume information from the T/S does not change the methods for plant operation or actions to be taken in the event of an accident. The quantity of radioactive material available for release in the event of an accident is not increased.

Barriers to release of radioactive material are not eliminated or otherwise changed. More detailed RCS component and piping volume information is included in the Salem UFSAR [updated final safety analysis report], and changes to that information would be evaluated prior to implementation in accordance with 10 CFR 50.59.

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

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

Response: No.

The deletion of the RCS volume information from the T/S does not change the methods of plant operation or modify plant systems, structures, or components. No new methods of plant operation are created. As such, the proposed change does not affect any accident initiators or precursors or create new accident initiators or precursors. More detailed and complete RCS component and piping volume information is included in the Salem UFSAR, and any changes to that information would be evaluated prior to implementation in accordance with 10 CFR 50.59.

Therefore, the proposed 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?

Response: No.

The deletion of the RCS volume information from the T/S does not affect safety limits or limiting safety system settings. Plant operational parameters are not affected. The proposed change does not modify the quantity of radioactive material available for release in the event of an accident. As such, the change will not affect any previous safety margin assumptions or conditions. The actual volume of the RCS is not affected by the change, only the location of the text describing the volume. More detailed and complete RCS component and piping volume information is included in the Salem UFSAR, and any changes to that information would be evaluated prior to implementation in accordance with 10 CFR 50.59.

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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: Darrell J. Roberts.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco Nuclear Generating Station, Sacramento County, California

Date of amendment request: January 24, 2005.

Description of amendment request: The proposed license amendment removes unnecessary and obsolete information from the facility license. The proposed changes are editorial and administrative in nature and will remove inappropriate and unnecessary information from the license given that the facility is permanently shutdown and defueled.

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

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

No. The proposed changes are administrative and involve deleting unnecessary and obsolete information from the facility operating license. These changes do not affect possible initiating events for accidents previously evaluated or alter the configuration or operation of the facility. Safety limits, limiting safety system settings, and limiting control systems are no longer applicable to Rancho Seco in the permanently defueled mode, and are therefore not relevant.

The proposed changes do not affect the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and have no impact on plant operations. Therefore, the proposed license amendment 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 accident from any previously evaluated.

No. As described above, the proposed changes are administrative. The safety analysis for the facility remains complete and accurate. There are no physical changes to the facility and the plant conditions for which the design basis accidents have been evaluated are still valid.

The operating procedures and emergency procedures are not affected. The proposed changes do not affect the emergency planning zone, the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and have no impact on plant operations. Consequently, no new failure modes are introduced as the result of the proposed changes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

No. As described above, the proposed changes are administrative. There are no changes to the design or operation of the facility. The proposed changes do not affect the emergency planning zone, the boundaries used to evaluate compliance with liquid or gaseous effluent limits, and have no impact on plant operations. Accordingly, neither the design basis nor the accident assumptions in the Defueled Safety Analysis Report (DSAR), nor the Technical Specification Bases are affected. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's significant hazards 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: Arlen Orchard, Esq., General Counsel, Sacramento Municipal Utility District, 6201 S Street, P.O. Box 15830, Sacramento, CA 95817-1899.

NRC Section Chief: Claudia M. Craig.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama

Date of amendment request: August 16, 2004 (TS-433).

Description of amendment request: The proposed amendment extends the frequency of “once-per-cycle” from 18 months to 24 months in several Technical Specification Surveillance Requirements. This change will allow the adoption of a 24-month refueling cycle.

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

No. The proposed amendment changes the surveillance frequency from 18 months to 24 months for Surveillance Requirements in the Unit 1 Technical Specification[s] that are normally a function of the refueling interval. Under certain circumstances, Surveillance Requirement 3.0.2 would allow a maximum surveillance interval of 30 months for these surveillances. TVA's evaluations have shown that the reliability of protective instrumentation and equipment will be preserved for the maximum allowable surveillance interval. The proposed changes do not involve any change to the design or functional requirements of plant systems and the surveillance test methods will be unchanged. The proposed changes will not give rise to any increase in operating power level, fuel operating limits, or effluents. The proposed change does not affect any accident precursors. In addition, the proposed changes will not significantly increase any radiation levels. Based on the foregoing considerations and the evaluations completed in accordance with the guidance of Generic Letter 91-04, it is concluded that the proposed amendment does not involve a significant increase in the probability or 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?

No. The proposed amendment does not require a change to the plant design, nor the mode of plant operation. The proposed changes do not create the possibility of any new failure mechanisms. No new external threats or release pathways are created. Therefore, the implementation of the proposed amendment will not create a possibility for an accident of a new or different type than those previously evaluated.

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

No. The proposed amendment changes the surveillance frequency from 18 months to 24 months for Surveillance Requirements in the Unit 1 Technical Specification[s] that are normally a function of the refueling interval. Under certain circumstances, Surveillance Requirement 3.0.2 would allow a maximum surveillance interval of 30 months for these surveillances. Although the proposed Technical Specification changes will result in an increase in the interval between surveillance tests, the impact on system availability is small based on other, more frequent testing or redundant systems or equipment. There is no evidence of any failures that would impact the availability of the systems. This change does not alter the existing setpoints, Technical Specification allowable values or analytical limits. The assumptions in the current safety analyses are not impacted and the proposed amendment does not reduce a margin of safety. Therefore, the proposed license amendment 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Section Chief: Michael L. Marshall, Jr.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear Plant (BFN), Unit 1, Limestone County, Alabama

Date of amendment request: October 12, 2004 (TS-438).

Description of amendment request: The proposed amendment request changes the frequency requirement for Technical Specification Surveillance Requirement (SR) 3.6.1.3.8 by allowing a representative sample (approximately 20 percent) of excess flow check valves (EFCVs) to be tested every 24 months, so that each EFCV is tested once every 120 months. The current SR requires testing of each EFCV every 24 months.

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

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

No. The current EFCV frequency requires that each reactor instrument line EFCV be tested every 24 months. The EFCVs are designed to automatically close upon excessive differential pressure including failure of the down stream piping or instrument and will reopen when appropriate. This proposed change will allow a reduction in the number of EFCVs that are verified tested every 24 months, to approximately 20 percent of the valves each cycle. BFN and industry operating experience demonstrates high reliability of these valves. Neither the EFCVs nor their failure is capable of initiating a previously evaluated accident. Therefore, there is no increase in the probability of occurrence of an accident previously evaluated.

The instrument lines going to the Reactor Coolant Pressure boundary with EFCVs installed have flow restricting devices upstream of the EFCV. The consequences of an unisolable failure of an instrument line have been previously evaluated and meet the intent of NRC Safety Guide 11. The offsite exposure has been calculated to be substantially below the limits of 10 CFR 50.67. The total control room Total Effective Dose Equivalent (TEDE) doses are less than the 5 REM limit and the offsite TEDE doses are less than 10% of the 25 REM limit. Additionally, coolant lost from such a break is inconsequential compared to the makeup capabilities of normal and emergency makeup systems. Although not expected to occur as a result of this change, the affects of a postulated failure of an EFCV to isolate and [sic] instrument line break as a result of reduced testing are bounded by TVA analysis.

Therefore, the proposed change does not involve a significant increase 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?

No. The proposed reduction in EFCV test frequency is bounded by previous evaluation of a line rupture. The proposed change does not introduce new equipment, which could create a new or different kind of accident. No new external threats, release pathways, or equipment failure modes are created. Therefore, the implementation of the proposed change will not create a possibility for an accident of a new or different type than those previously evaluated.

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

No. The consequences of an unisolable rupture of an instrument line have been previously evaluated and meet the intent NRC Safety Guide 11. The proposed change does not involve a significant reduction in a margin of safety. Therefore, the proposed revised surveillance frequency does not adversely affect the public health and safety, and does not involve any significant safety hazards.

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

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Section Chief: Michael L. Marshall, Jr.

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.

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

Date of amendment request: February 10, 2005.

Brief description of amendment request: The proposed amendment would extend the allowed outage time for the Emergency Generator Load Sequencer (Technical Specification 3/4.3.2, Table 3.3-3, Functional Unit 10) from 6 hours to 12 hours.

Date of publication of individual notice in Federal Register: February 22, 2005 (70 FR 8641).

Expiration date of individual notice: March 24, 2005 (public comments) and April 25, 2005 (hearing requests).

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: July 23, 2004, and January 6, 2005.

Brief description of amendment request: The proposed revision would modify the Technical Specification (TS) definition of OPERABILITY with respect to requirements for availability of normal and emergency power. Additionally, the proposed revision would modify the required actions for shutdown power TSs.

Date of publication of individual notice in Federal Register: March 1, 2005.

Expiration date of individual notice: March 31, 2005 (public comments), and May 2, 2005 (hearing requests).

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 (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 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 PDR Reference staff at 1 [phone], [phone] or by e-mail to [e-mail].

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

Date of application for amendment: April 30, 2004.

Brief description of amendment: The amendment modifies requirements in the Technical Specifications (TS) to adopt the provisions of Industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints.”

Date of issuance: March 2, 2005.

Effective date: As of the date of issuance and shall be implemented within 180 days.

Amendment No.: 163.

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

Date of initial notice in Federal Register: October 26, 2004 (69 FR 62469).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 2, 2005.

No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-324, Brunswick Steam Electric Plant, Unit 2, Brunswick County, North Carolina

Date of application for amendment: August 16, 2004.

Brief Description of amendment: The amendment adds topical report NEDE-32906P-A, “TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses,” to the documents listed in Technical Specification 5.6.5 describing the approved methodologies used to determine the core operating limits.

Date of issuance: March 4, 2005.

Effective date: March 4, 2005.

Amendment No.: 262.

Facility Operating License No DPR-62: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: October 26, 2004 (69 FR 62470).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 4, 2005.

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: May 27, 2004.

Brief description of amendments: The amendments revised the Technical Specifications by eliminating the requirements associated with hydrogen recombiners and hydrogen monitors.

Date of issuance: March 1, 2005.

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

Amendment Nos.: 219 and 214 .

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

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57982).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 1, 2005.

No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of application for amendment: September 27, 2004.

Brief description of amendment: The amendment eliminated the technical specification requirements to submit a monthly operating report and an annual occupational radiation exposure report.

Date of issuance: March 9, 2005.

Effective date: March 9, 2005.

Amendment No.: 190.

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

Date of initial notice in Federal Register : October 26, 2004 (69 FR 62472).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 9, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 15, 2004, as supplemented January 20, 2005.

Brief Description of amendments: The licensee has proposed to change the existing reactor coolant system (RCS) cooldown curve to a single 32 effective full power year pressure/temperature limit curve that is applicable for cooldowns at a rate of 100 °F/hour or 50 °F in any half-hour step. The licensee's proposed curve is applicable to RCS cold-leg temperatures ranging from 50 °F to 560 °F.

Date of issuance: March 7, 2005.

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

Amendment No.: 256.

Facility Operating License No. NFP-6: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register : May 11, 2004 (69 FR 26188). The supplemental letter provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 7, 2005.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: April 14, 2004.

Brief description of amendment: The amendment revised the Pilgrim Nuclear Power Station Technical Specifications (TSs) by adding a new limiting condition for operation (LCO) 3.0.7 to Section 3.0, “Limiting Condition for Operation (LCO) Applicability,” a new TS Section 3.14, “Special Operations,” and a new LCO 3.14.A, “Inservice Leak and Hydrostatic Testing Operation,” to the TSs. These changes permit the licensee to perform inservice hydrostatic testing and system leakage pressure testing of the reactor coolant system at temperatures greater than 212 °F with the reactor shut down.

Date of issuance: March 16, 2005.

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

Amendment No.: 211.

Facility Operating License No. DPR-35: The amendment revised the TSs.

Date of initial notice in Federal Register : December 21, 2004 (69 FR 76489).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 2005.

No significant hazards consideration comments received: No.

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

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: April 30, 2004.

Brief description of amendments: The amendments modify Technical Specifications (TS) requirements to adopt the provisions of Industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints.”

Date of issuance: March 10, 2005.

Effective date: As of the date of issuance and shall be implemented within 180 days.

Amendment Nos.: 212/204/223/218.

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. The amendments revised the Technical Specifications.

Date of initial notice in Federal Register : October 26, 2004 (69 FR 62474).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 10, 2005.

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: June 10, 2004, and supplemented July 19 and July 21, 2004 and January 21, 2005.

Brief description of amendments: The amendments revise the Quad Cities Nuclear Power Station Technical Specifications to change the allowable value and add Surveillance Requirements for the Main Steam Line Flow-High initiation of Group 1 Primary Containment Isolation System and Control Room Emergency Ventilation System isolation.

Date of issuance: March 15, 2005.

Effective date: As of the date of issuance and shall be implemented within 90 days for Unit 1 and no later than 90 days after the start of the Unit 2 refueling outage currently scheduled for March 2006 for Unit 2.

Amendment Nos.: 224, 219

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

Date of initial notice in Federal Register : August 31, 2004 (69 FR 53107). The supplemental letters 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 amendments is contained in a Safety Evaluation dated March 15, 2005.

No significant hazards consideration comments received: No.

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

Date of amendment request: June 1, 2004, as supplemented July 23, 2004, and February 18, 2005.

Description of amendment request: These amendments lowered the BVPS-2 overpressure protection system enable temperature, allowed one inoperable residual heat removal loop during surveillance testing, removed the BVPS-1 list of figures and list of tables from the Index of the BVPS-1 Technical Specifications (TSs), and made minor changes to achieve consistency between units and with the Standard TSs for Westinghouse plants and with some TS Task Force changes.

Date of issuance: March 11, 2004.

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

Amendment Nos.: 265 and 146.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. February 25, 2005 (70 FR 9391). The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination by March 11, 2005. No comments have been received. The notice also provided an opportunity to request a hearing by April 26, 2005, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated March 11, 2005.

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

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

Date of application for amendments: April 13, 2004.

Brief description of amendments: The amendments change the design basis as described in the Updated Final Safety Analysis Report to allow the use in control rod drive missile shield structural calculations of a reinforcing bar (rebar) yield strength value based on measured material properties, as documented in the licensee rebar acceptance tests.

Date of issuance: March 11, 2005.

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

Amendment Nos.: 286, 268.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the design basis.

Date of initial notice in Federal Register: October 12, 2004 (69 FR 60682).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 11, 2005.

No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: September 7, 2004.

Brief description of amendment: The amendment revised Technical Specification (TS) 5.9.5, “Core Operating Limits Report,” to be consistent with Specification 5.6.5 of NUREG-1432, “Standard Technical Specifications Combustion Engineering Plants.” In addition, the list of core reload analysis methodologies contained in TS 5.9.5b used to determine the core operating limits, has been updated. Many of these references were moved to the Omaha Public Power District core reload analysis methodology documents OPPD-NA-8301, 8302, and 8303, which are also listed in TS 5.9.5b. However, OPPD-NA-8302 has been revised to incorporate use of the code CASMO-4 in lieu of the previously approved CASMO-3 code.

Date of issuance: March 11, 2005.

Effective date: March 11, 2005, and shall be implemented within 90 days from the date of issuance.

Amendment No.: 233.

Renewed Facility Operating License No. DPR-40: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 12, 2004 (69 FR 60683)

The Commission's related evaluation of the amendment is contained in a safety evaluation dated March 11, 2005.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendments request: May 21, 2004.

Brief description of amendments: The amendments revised the Technical Specifications to delete the requirements to maintain hydrogen recombiners and hydrogen analyzers.

Date of issuance: March 8, 2005.

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

Amendment Nos.: 167 and 159.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57994)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 8, 2005.

No significant hazards consideration comments received: No

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendments request: July 28, 2004.

Brief description of amendments: The amendments delete the technical specification requirements to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: March 8, 2005.

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

Amendment Nos.: 168 and 160.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: October 12, 2004 (69 FR 60686)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 8, 2005.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of application for amendments: July 28, 2004.

Brief description of amendments: The amendments revised the Technical Specifications by deleting the requirements for monthly operating reports and occupational radiation exposure reports.

Date of issuance: March 8, 2005.

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

Amendment Nos.: 245 and 189.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 12, 2004 (69 FR 60686).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 8, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 13, 2003, as supplemented by letters dated April 12 and October 28, 2004.

Brief description of amendments: The amendments revised the Technical Specifications (TS) limiting conditions for operation 3.8.4, 3.8.5, and 3.8.6, on direct current sources, operating and shutdown, and battery cell parameters. The proposed amendments creates TS 5.5.19, for a battery monitoring and maintenance program. The TS Bases are revised to be consistent with these changes. The proposed amendments are based on Technical Specification Task Force (TSTF) Traveler, TSTF-360, Revision 1.

Date of issuance: March 2, 2005.

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

Amendment Nos.: 133 and 112.

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

Date of initial notice in Federal Register: January 20, 2004 (69 FR 2746). The supplements dated April 12 and October 28, 2004, provided clarifying information that did not change the scope of the October 13, 2003, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 2, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 21, 2004.

Brief description of amendments: The amendments revised the Technical Specifications to delete the requirements to maintain hydrogen recombiners and change requirements for hydrogen analyzers.

Date of issuance: March 7, 2005.

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

Amendment Nos.: 134 and 113.

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

Date of initial notice in Federal Register : September 28, 2004 (69 FR 57995).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 7, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 28, 2004.

Brief description of amendments: The amendments delete the technical specification requirements to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: March 8, 2005.

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

Amendment Nos.: 135 and 114.

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

Date of initial notice in Federal Register : October 12, 2004 (69 FR 60686)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 8, 2005.

No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: May 13, 2003, as supplemented by letters dated October 6, 2004, November 30, 2004, and January 20, 2005.

Brief description of amendments: The amendments approve revisions to the RETRAN-02 methodology that is used to evaluate certain design basis transients and accidents.

Date of issuance: March 7, 2005.

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

Amendment Nos.: Unit 1—171; Unit 2—159.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the RETRAN-02 methodology.

Date of initial notice in Federal Register: November 12, 2003 (68 FR 64138). The supplements dated October 6, 2004, November 30, 2004, and January 20, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 7, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 8, 2004, as supplemented in a letter dated November 24, 2004 (TS-448).

Brief description of amendments: The amendments modify Technical Specification Section 5.5.12 “Primary Containment Leakage Rate Testing Program” to allow a one-time 5-year extension to the 10-year frequency of the performance-based leakage rate testing program for Type A tests. The first Unit 2 Type A test performed after the November 6, 1994, Type A test shall be performed no later than November 6, 2009, and the first Unit 3 Type A test performed after the October 10, 1998, Type A test shall be performed no later than October 10, 2013. The local leakage rate tests (Type B and Type C), including their schedules, are not affected by this request.

Date of issuance: March 9, 2005.

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

Amendment Nos.: 293 and 252.

Facility Operating License Nos. DPR-52 and DPR-68: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register : August 3, 2004 (69 FR 46592).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 9, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 18, 2004.

Brief description of amendments: The amendments revised Technical Specification (TS) 3/4.4.2, “Safety Valves—Shutdown,” TS 3/4.4.3, “Safety and Relief Valves—Operating,” and TS 3/4.5.2, “ECCS Subsystems—T _avg Greater Than or Equal to 350°F.” TS 3/4.4.2 is eliminated because overpressure protection of the reactor coolant system does not rely upon the pressurizer safety valves during plant operation in Modes 4 and 5. TS 3/4.4.3 is revised to remove redundancy and add improvements consistent with NUREG-1431, Revision 3, “Standard Technical Specifications for Westinghouse Plants.” TS 3/4.5.2 is revised by adding a note to the Limiting Condition for Operation (LCO) supporting transition to and from LCO 3.4.12, “Low Temperature Overpressure Protection (LTOP) System.”

Date of issuance: March 9, 2005.

Effective date: As of the date of issuance. Unit 1 shall be implemented by May 15, 2005, and Unit 2 shall be implemented by completion of the 2005 Cycle 13 Refueling Outage.

Amendment Nos.: 299 and 288.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the TSs.

Date of initial notice in Federal Register: November 9, 2004 (69 FR 64991)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 9, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 15, 2004.

Brief description of amendment: The amendment modifies technical specification (TS) requirements for mode change limitations in Limiting Condition for Operation 3.0.4 and Surveillance Requirement 3.0.4 consistent with Industry/TS Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-359, Revision 9, “Increased Flexibility in Mode Restraints.” In addition, the amendment modifies TS requirements consistent with TSTF-153, Revision 0, “Clarify Exception Notes to be Consistent with the Requirement Being Excepted,” in part, and TSTF-285, Revision 1, “Charging Pump Swap LTOP (Low Temperature-Overpressurization) Allowance.”

Date of issuance: March 3, 2005.

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

Amendment No.: 55.

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

Date of initial notice in Federal Register: January 18, 2005 (70 FR 2901) and February 1, 2005 (70 FR 5226).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 3, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: September 8, 2003, as supplemented by letter dated September 11, 2003.

Brief description of amendment: The amendment revised the Updated Final Safety Analysis Report (UFSAR) by modifying the design and licensing basis to increase the postulated primary-to-secondary leakage in the faulted steam generator following a main steamline break accident from 1 to 3 gallons per minute.

Date of issuance: March 10, 2005.

Effective date: As of the date of issuance and shall be implemented as part of the next UFSAR update made in accordance with 10 CFR 50.71(e).

Amendment No.: 56

Facility Operating License No. NPF-90: Amendment revised the UFSAR.

Date of initial notice in Federal Register: September 18, 2003 (68 FR 54745).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 10, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 27, 2004.

Brief description of amendment: The amendment revised the Technical Specifications (TSs) by eliminating the requirements in TSs 5.6.1 and 5.6.4 to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: March 8, 2005.

Effective date: March 8, 2005, and shall be implemented within 90 days of the date of issuance.

Amendment No.: 166.

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

Date of initial notice in Federal Register: January 4, 2005 (70 FR 406).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 8, 2005.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas

Date of amendment request: July 22, 2004.

Brief description of amendment: The amendment revises Technical Specification Figure 3.5.5-1, “Seal Injection Flow Limits,” to reflect flow limits that allow a higher seal injection flow for a given differential pressure between the charging pump discharge header and the reactor coolant system.

Date of issuance: March 16, 2005.

Effective date: March 16, 2005, and shall be implemented prior to startup from Refueling Outage 14.

Amendment No.: 160.

Facility Operating License No. NPF-42: The amendment revises the Technical Specifications.

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53115).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 16, 2005.

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 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 (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 PDR Reference staff at 1 [phone], [phone] or by e-mail to [e-mail].

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. Within 60 days after the date of publication of this notice, 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 2.309, 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, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1 [phone], [phone], or by e-mail to [e-mail]. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for 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. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact.^[1] 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/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Each contention shall be given a separate numeric or alpha designation within one of the following groups:

1. Technical—primarily concerns/issues relating to technical and/or health and safety matters discussed or referenced in the applications.

2. Environmental—primarily concerns/issues relating to matters discussed or referenced in the environmental analysis for the applications.

3. Miscellaneous—does not fall into one of the categories outlined above.

As specified in 10 CFR 2.309, if two or more petitioners/requestors seek to co-sponsor a contention, the petitioners/requestors shall jointly designate a representative who shall have the authority to act for the petitioners/requestors with respect to that contention. If a petitioner/requestor seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/requestor a representative who shall have the authority to act for the petitioners/requestors with respect to that contention.

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. 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 filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, [e-mail]; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at [phone], verification number is [phone]. 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 it is requested that copies be transmitted either by means of facsimile transmission to [phone] or by e-mail to [e-mail]. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer or the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear Station Unit 2, York County, South Carolina

Date of amendment request: February 5, 2005, as supplemented by letter dated February 7, 2005.

Description of amendment request: The amendment revises the system bypass leakage acceptance criterion for the charcoal adsorber in the 2B Auxiliary Building Filtered Ventilation Exhaust System train as listed in Technical Specification 5.5.11, “Ventilation Filter Testing Program.”

Date of issuance: February 7, 2005.

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

Amendment No.: 213.

Renewed Facility Operating License No. NPF-52: Amendments revised the Technical Specifications.

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 7, 2005.

Attorney for licensee: Ms. Anne Cottingham, Esquire.

NRC Section Chief: John A. Nakoski.

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 7, 2005.

Attorney for licensee: Ms. Anne Cottingham, Esquire.

NRC Section Chief: John A. Nakoski.

Dated at Rockville, Maryland, this 21st day of March 2005.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

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

Footnotes

1.  To the extent that the applications contain attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant's counsel and discuss the need for a protective order.