68 Fed. Reg. 74262 (Dec. 23, 2003)

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

Federal RegisterDec 23, 2003

68 Fed. Reg. 74262 (Dec. 23, 2003)

Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from November 27 through December 11, 2003. The last biweekly notice was published on December 9, 2003 (68 FR 68654).

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

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

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

Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By January 22, 2004, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: June 11, 2003, as supplemented by letter dated August 20, 2003, and October 13, 2003.

Description of amendment request: The proposed amendment would modify Technical Specification (TS) 5.5.16, “Containment Leakage Rate Testing Program” to allow a one-time extension of the containment Type A leak rate test interval from once in 10 years to once in 15 years.

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

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

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

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

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

The proposed change to TS 5.5.16 provides a one-time extension of the containment Type A test interval to 15 years for HBRSEP, Unit No. 2. The proposed change to the Type A test interval does not result in any physical changes to HBRSEP, Unit No. 2. In addition, the proposed test interval extension does not change the operation of HBRSEP, Unit No. 2, such that a failure mode involving the possibility of a new or different kind of accident from any accident previously evaluated is created.

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

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

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

The proposed change only involves a one-time extension of the interval for containment Type A testing; the overall containment leak rate specified by the HBRSEP, Unit No. 2, TS is being maintained. Type B and C testing will continue to be performed at the frequency required by the HBRSEP, Unit No. 2, TS. The regular containment inspections being performed in accordance with the ASME Code, Section XI, and the Maintenance Rule (10 CFR 50.65) provide a high degree of assurance that the containment will not degrade in a manner that is only detectable by Type A testing. In addition, a plant-specific risk evaluation demonstrates that the extension of the Type A test interval from 10 years to 15 years results in a “very small” increase in risk for those accident sequences influenced by Type A testing and a “small” increase in risk when compared to the test frequency of 3 tests per 10 years.

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

Based on the above discussion, Progress Energy Carolinas, Inc., has determined that the requested change does not involve a significant hazards consideration.

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

Attorney for licensee: Steven R. Carr, Associate General Counsel—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: March 20, 2003.

Description of amendment request: The amendments would revise the Technical Specifications to update the heatup, cooldown, criticality, and inservice test pressure and temperature limits for the reactor coolant system.

First Standard

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

Response: No.

The proposed changes to the reactor coolant system (RCS) pressure-temperature (P/T) limits are developed utilizing the methodology of ASME (American Society of Mechanical Engineers) XI, 10 CFR (part) 50 Appendix G, in conjunction with the methodology of Code Case N-640. Usage of these methodologies provides compliance with the underlying intent of 10 CFR (part) 50 Appendix G and provides operational limits that ensure failure of the reactor vessel will not occur. The proposed changes to allow operation with two pumps capable of injecting into the RCS and utilization of the residual heat removal (RHR) suction relief valves has been evaluated and determined to provide adequate protection of the RCS from the worst case pressure transient.

The probability of any design basis accident (DBA) is not affected by these changes, nor are the consequences of any DBA affected by these changes. The P/T limits, and low temperature overpressure protection (LTOP) setpoints, and T_enable value are not considered to be initiators or contributors to any accident analysis addressed in the Catawba UFSAR (updated final safety analysis report).

The proposed changes do not adversely affect the integrity of the RCS such that its function in the control of radiological consequences is affected. The changes do not alter any assumption previously made in the radiological consequence evaluations nor affect the mitigation of the radiological consequences of an accident previously evaluated. The proposed changes to the TS are consistent with the intent of the flexibility currently provided in NUREG-1431, Standard Technical Specifications for Westinghouse Plants, Revision 2.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated in the updated final safety analysis report (UFSAR) because the accident analysis assumptions and initial conditions will continue to be maintained.

Second Standard

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

Response: No.

The proposed change does not involve any physical alteration of plant systems, structures, or components. The requirements for the P/T limit curves and LTOP setpoints remain in place. The fundamental approach follows approved ASME and Westinghouse report methodology. The proposed curves and change to the enable temperature for LTOP system reflect changes in material properties acknowledged and managed by regulation and an upgrade in technology, which has been approved by ASME.

The proposed changes to allow operation with two pumps capable of injecting into the RCS and utilization of the RHR suction relief valves has been evaluated. The evaluation has shown that both the PORVs (power-operated relief valves) and RHR suction relief valves provide adequate relief protection of the RCS from the worst case pressure transient and provide equivalent protection to that already allowed by the current TS (technical specification).

The proposed changes do not introduce new failure mechanisms for system structures, or components not already considered in the UFSAR. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created because no new failure mechanisms or initiating events have been introduced.

Third Standard

Does operation of the facility in accordance with the proposed amendment involve a significant reduction in the margin of safety.

Response: No.

The proposed changes are developed utilizing the methodology of ASME XI, 10 CFR (part) 50 Appendix G, in conjunction with Code Case N-640 and Code Case N-641 methodology. Usage of these methodologies provides compliance with the underlying intent of 10 CFR (part) 50 Appendix G and provides operational limits that ensure failure of the reactor vessel will not occur. Although the Code Cases constitute relaxation from the current requirements of 10 CFR (part) 50 Appendix G, the alternative methodology allowed by the Code is based on industry experience gained since the inception of the 10 CFR (part) 50 Appendix G requirements for which some of the requirements have now been determined to be excessively conservative. The more appropriate assumptions and provisions allowed by the Code Cases maintain a margin of safety that is consistent with the intent of 10 CFR (part) 50 Appendix G, i.e., with regard to the margin originally contemplated by 10 CFR (part) 50 Appendix G for determination of RCS P/T limits.

The analyses completed for this proposed TS amendment demonstrate that established acceptance criteria continue to be met. Specifically, the P/T limit curves, LTOP setpoints, allowances for operating two pumps, utilization of RHR suction relief valves and LTOP T_enable values provide acceptable margin to vessel fracture under both normal operation and LTOPs design basis (mass addition and heat addition) accident conditions. The proposed changes to the TS are consistent with the intent of the flexibility currently provided in NUREG-1431, Standard Technical Specifications for Westinghouse Plants, Revision 2. Therefore, there will be no significant reduction in a margin of safety as a result of the proposed changes.

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

NRC Section Chief: John A. Nakoski.

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 amendment request: March 28, 2003, as supplemented by letter dated October 23, 2003.

Description of amendment request: The proposed amendments would revise the technical specifications to reduce the main steam line low pressure primary containment isolation allowable value.

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

Response: No.

Current licensing bases events remain bounding for ATWS, transient, and accident analyses. For the bounding events, a reduction in the allowable value for the MSL LPIS produces no significant change in the limiting results with respect to the acceptance criteria. The proposed change does not alter the response of plant equipment to transient conditions, nor does it introduce any new equipment, modes of system operation or failure mechanisms. The proposed change does not adversely impact structures, systems, or components.

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

ECCS-LOCA Performance

In the analyses used to evaluate the ECCS-LOCA performance, the MSIVs are assumed to close at the start of the accident for all break locations. Therefore, the low pressure isolation trip is not used in the LOCA analyses and the LOCA analysis results are not affected by the reduction in the LPIS.

If the steam line break is too small to result in a high flow isolation signal, MSIV closure may be initiated by another signal (e.g., high steam line tunnel temperature or low reactor water level) or it may occur due to the LPIS trip. In either case, steam line breaks of any size are not the limiting events with respect to ECCS performance, and a 40 psi reduction in the LPIS will not affect compliance with the acceptance criteria of 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.”

Based on the above discussions, the reduction of the MSIV LPIS has no adverse impact on the plant response to a LOCA or on compliance with the acceptance criteria of 10 CFR 50.46.

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

Containment System Response

In evaluating containment response to pipe breaks inside containment, the MSIVs are assumed to close at the start of the accident for all break locations in the containment system response analyses. Therefore, the low pressure isolation trip is not assumed and the analysis results are not affected by the reduction in the LPIS.

In the event that MSIV closure does not occur at the beginning of the accident, MSL isolation is effectively achieved as the pressure regulator closes the turbine control and bypass valves in an attempt to maintain turbine throttle pressure at the regulator setpoint of approximately 925 psig. Thus, for events other than breaks in the main steam line, isolation occurs before the LPIS is reached.

For large breaks in the MSL (both inside and outside containment), the MSIV closure is initiated by a high steam line flow signal at the beginning of the event, well before the LPIS is reached. For these cases, the containment system response is not affected by the reduction in the LPIS. For a steam line break too small to result in a high flow isolation signal, MSIV closure may be initiated by another signal (e.g., low reactor water level) or it may occur due to the LPIS trip. Small breaks do not determine the peak drywell shell temperature and equipment qualification (EQ) envelope. Large breaks, as characterized in Section 3.3.2 of Attachment 4, are large enough to depressurize the reactor irrespective of the MSIV closure. Hence, a 40-psi reduction in the LPIS will not affect the peak drywell shell temperature or the drywell temperature EQ envelope.

Based on the above discussions, the reduction of the MSIV LPIS has no adverse impact on the containment system response.

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

Subcompartment Pressurization

The MSL break mass and energy release used in the evaluation are based on steady-state reactor operating conditions. Therefore, the low pressure isolation trip is not used in the subcompartment pressurization analysis. In addition, the peak annulus pressurization loads occur at the beginning of the event, well before MSIV closure can occur.

The subcompartment pressurization results are not affected by the reduction in the MSL LPIS.

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

Appendix R Fire Protection

The reactor system response for the Appendix R fire protection analysis was performed during the Extended Power Uprate (EPU) project. The sequence of events for the analysis shows that closure of the MSIVs is initiated on low-low reactor water level. However, before the LPIS setpoint is reached, the turbine control valves closing on low inlet pressure effectively isolate steam flow following a scram. The revised LPIS has no adverse impact on the reactor system response to an Appendix R fire protection event.

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

Station Blackout

The initiating event for a station blackout, a loss of off-site power, results in MSIV closure at the beginning of the event. The reduction of the MSL LPIS has no adverse impact on the reactor system response during a station blackout.

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

High Energy Line Break

The steam line break analysis assumes closure of the MSIVs due to high steam line flow at the beginning of the event. Thus, the low pressure isolation trip is not used in the analyses and the results are not adversely affected by the reduced LPIS.

The steam line break case determines the short-term peak steam tunnel temperature. However, the range of break sizes for which the low pressure isolation trip initiates MSIV closure is limited. Such a break must be large enough to depressurize the vessel below the pressure regulator setpoint, approximately 925 psig, but small enough such that high steam line flow trip does not result. Although such cases could result in an increase in the mass and energy released, similar to a larger line break, isolation will still occur before the LPIS is reached. The isolation will occur as a result of Main Steam Line Tunnel Temperature—High for any leak greater than 1% rated steam flow. Thus, a 40 psig reduction in the LPIS will not adversely affect the peak temperature in the steam tunnel. In addition, the dynamic effects (e.g., pipe whip and jet impingement) on other structures, systems and components are unaffected by the reduced LPIS.

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

Radiological Consequences

The MSIVs are assumed to close due to high steam line flow at the start of an accident in the analysis. The low pressure isolation trip is not used in the mass release analysis and the radiological consequences are not affected by the reduction of the LPIS.

If the steam line break is too small to cause a high flow isolation signal, MSIV closure may be initiated by another signal (e.g., high steam tunnel temperature or low reactor water level) or it may result from the low pressure isolation trip. Thus, a 40 psig reduction in the LPIS will have no adverse impact on the radiological consequences. The radiological consequences of a reduction in the MSL LPIS are addressed further in Section 6 of this attachment.

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

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

Response: No.

General Electric Company (GE) evaluated the impact of reducing the LPIS analytical limit from 825 to 785 psig, including analysis of transient and safety related licensing bases for DNPS, Units 2 and 3, and QCNPS, Units 1 and 2. Current licensing bases events remain bounding for ATWS, transient, and accident analyses. The proposed change revises the allowable value of TS Table 3.3.6.1-1, Function 1.b, but does not alter the instrumentation or control logic of the Primary Containment Isolation 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 change involve a significant reduction in a margin of safety?

Response: No.

The revised LPIS does not change the current licensing bases events, which remain bounding for ATWS, transient and accident analyses. The conclusion that a reduction in the MSIV LPIS will not have an adverse impact on plant accident analyses is valid. The LPIS was analyzed by GE during the EPU project for impact on safety limits and safety margins and was determined to be a non-impacted item. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: August 19, 2003.

Description of amendment request: The proposed amendments would modify Technical Specification (TS) 5.5.13, “Primary Containment Leakage Rate Testing Program,” by identifying a specific exception to the testing guidance contained in Regulatory Guide (RG) 1.163, “Performance-Based Containment Leak-Test Program.”

LaSalle County Station (LSCS) Units 1 and 2 conduct their leakage rate testing of the primary containments to the requirements of 10 CFR 50.54(o) and 10 CFR part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” Option B as modified by approved exemptions. Additionally, the program is in accordance with the guidelines contained in RG 1.163. The proposed TS change would take exception to RG 1.163 guidance by allowing the testing of potential valve atmospheric leakage paths (e.g., valve stem packing), that are not exposed to reverse direction Type B or C leakage test pressure during the regularly scheduled Type A test. A list of the potential valve atmospheric leakage paths, the leakage rate measurement method and the acceptance criteria will be contained in the program. This exception will be applicable only to valves that are not isolable from the primary containment free air space.

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

The proposed change will revise LaSalle County Station, Units I and 2, Technical Specification (TS) 5.5.13, “Primary Containment Leakage Rate Testing Program” by identifying a specific exception to the testing guidance contained in Regulatory Guide (RG) 1.163, “Performance-Based Containment Leak-Test Program.”

The function of the primary containment is to isolate and contain fission products released from the reactor Primary Coolant System (PCS) following a design basis Loss of Coolant Accident (LOCA) and to confine the postulated release of radioactive material to within limits. The probability of an accident previously evaluated is not dependent on the test-frequency of the primary containment Type A, B or C testing. The test interval associated with primary containment testing is not a precursor of any accident previously evaluated. The proposed specific exception to the testing guidance contained in RG 1.163 will continue to test all potential valve atmospheric leakage paths and will not be a precursor to a Design Basis Accident (DBA). Containment testing does provide assurance that the LaSalle County Station primary containments will not exceed allowable leakage rate values specified in the Technical Specifications and will continue to perform their design function following an accident.

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

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not affect the control parameters governing unit operation or the response of plant equipment to transient conditions. The proposed change does not introduce any new equipment, modes of system operation or failure mechanisms.

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

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

The integrity of the primary containment is verified through Type B and Type C local leak rate tests (LLRTs) and the overall leak tight integrity of the primary containment is verified by a Type A integrated leak rate test (ILRT) as required by 10 CFR part 50, Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.” These tests are performed to verify the essentially leak tight characteristics of the primary containment at the design basis accident pressure. The proposed change for a specific exception to the testing guidance contained in Regulatory Guide (RG) 1.163 will continue to test all potential valve atmospheric leakage paths and does not effect the test acceptance criteria for Type A, B or C testing. Therefore, LSCS has determined that the proposed change provides an equivalent level of protection as that currently provided.

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

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

Attorney for licensee: Mr. Edward J. Cullen, Deputy General Counsel, Exelon BSC—Legal, 2301 Market Street, Philadelphia, PA 19101.

NRC Section Chief : Anthony J. Mendiola.

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

Date of amendment request: November 20, 2003.

Description of amendment request: The licensee proposes to revise the safety limit minimum critical power ratio (SLMCPR) values in section 2.1.1.2 of the Technical Specifications (TSs). The SLMCPR values are based on cycle-specific calculations done for the next fuel cycle, Cycle 10, using methodology previously approved by the Nuclear Regulatory Commission (NRC).

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

The first standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed SLMCPR values, calculated using an NRC-approved methodology, will be made in a manner such that conservatism is maintained through compliance with applicable NRC regulations and guidance. No hardware design change is involved with the proposed amendment, thus there will be no adverse effect on the functional performance of any plant structure, system, or component (SSC). All SSCs will continue to perform their design functions with no decrease in their capabilities to mitigate the consequences of postulated accidents. SLMCPR values were not previously factored into the probability of accidents, nor were they factored into scenarios of previously analyzed accidents. Accordingly, the revised SLMCPR values will lead to no increase in the consequences of an accident previously evaluated, and no increase of the probability of an accident previously evaluated.

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

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

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

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: November 21, 2003.

Description of amendment request: The proposed amendment would allow the position of a rod to be monitored by a means other than the movable incore detectors.

(1) Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed change provides an alternative method for the monitoring of the position of a rod once the position of the rod is verified using the moveable incore detector system. The proposed monitoring of stationary gripper coil parameters provides a reasonably similar approach to rod position monitoring as that provided by the movable incore detector system. In particular, the ability to immediately detect a rod drop or misalignment is not directly provided by the movable incore detector system or by the monitoring of stationary gripper coil parameters. Additionally, neither the movable incore detector system, nor the monitoring of stationary gripper coil parameters, provides the capability to verify rod position following a reactor trip or shutdown. Therefore, the monitoring of stationary gripper coil parameters, in lieu of the use of the movable incore detector system, provides an equivalent and acceptable method of monitoring rod position while a position indicator is inoperable.

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

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

No. As described above, the proposed change provides only an alternative method of monitoring the position of a rod. No new accident initiators are introduced by the proposed alternative manner of performing rod position monitoring. The proposed change does not affect the reactor protection system or the reactor control system. Hence, no new failure modes are created that would cause a new or different kind of accident from any accident previously evaluated.

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

(3) Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

No. The bases for TS (Technical Specification) 3.1.8 state that the operability of the rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. The proposed change does not alter the requirement to determine rod position but provides an alternative method for monitoring the position of the affected rod after the position of the rod is verified using the moveable incore detector system. As a result, the initial conditions of the accident analysis are preserved and the consequences of previously analyzed accidents are unaffected.

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

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

NRC Section Chief: Allen G. Howe.

Notice of Issuance of Amendments to Facility Operating Licenses

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

Notice of consideration of issuance of amendment to facility operating license, proposed no significant hazards consideration determination, and opportunity for a hearing in connection with these actions was published in the Federal Register as indicated.

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

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, 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 NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.

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

Date of application for amendments: December 13, 2002, as supplemented September 25, 2003.

Brief description of amendments: These amendments changed the Technical Specifications (TSs) by removing the requirement to have the charging pumps operable when thermal power is greater than 80% of rated thermal power. The change also removes Surveillance Requirement 3.5.2.4 for verifying the required charging pump flow rate. The change to TS 3.5.2 does not modify any other charging pump requirements in the Technical Requirements Manual (e.g., requirements of charging pump availability for boration and cooldown remain in effect).

Date of issuance: December 3, 2003.

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

Amendment Nos.: 260 and 237.

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

Date of initial notice in Federal Register: February 18, 2003 (68 FR 7812).

The September 25, 2003, supplemental letter provided clarifying information that did not enlarge the scope of the amendment as noticed in the original Federal Register notice or change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated December 3, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 28, 2003, as supplemented November 25, 2003.

Brief description of amendments: These amendments changed the reactor pressure vessel pressure-temperature limit cooldown curves in the Calvert Cliffs 1 and 2 Technical Specifications by incorporating a different range of temperatures for which a maximum cooldown rate of 100°F/hr is acceptable.

Date of issuance: December 9, 2003.

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

Amendment Nos.: 261 and 238.

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

Date of initial notice in Federal Register: July 8, 2003 (68 FR 40701).

The November 25, 2003, supplemental letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated December 9, 2003.

No significant hazards consideration comments received: No.

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 14, 2003, as supplemented by letter dated June 24, 2003.

Brief description of amendment: The amendment revises Technical Specification (TS) 3.8.1, “AC Sources—Operating,” Surveillance Requirements (SRs) pertaining to the testing of the Division 1 and 2 standby diesel generators (DGs). Specifically, the proposed changes eliminate mode restrictions that previously prevented performance of SRs during Modes 1 and 2 for the Division 1 and 2 DGs. The changes allow the performance of SR 3.8.1.9 and SR 3.8.1.10 for the Division 1 and 2 DGs during any plant operating mode.

Date of issuance: November 7, 2003.

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

Amendment No.: 137.

Facility Operating License No. NPF-47: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: (68 FR 18275). The June 24, 2003, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 7, 2003.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: May 28, 2003, as supplemented on June 24, 2003.

Brief description of amendment: The amendment revised Technical Specification (TS) Section 3.4.3, “RCS Pressure and Temperature (P/T) Limits,” and section 3.4.12, “Low Temperature Overpressure Protection (LTOP),” to incorporate revised reactor pressure vessel P/T limits and overpressure protection system limits to allow operation up to 20 effective full-power years. Specifically, the amendment changed TS Figures 3.4.3-1 to 3.4.3-3 and TS Figures 3.4.12-1 to 3.4.12-4.

Date of issuance: December 3, 2003.

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

Amendment No.: 220.

Facility Operating License No. DPR-64: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 22, 2003 (68 FR 43389).

The June 24 letter provided clarifying information that did not enlarge the scope of the amendment request or change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 3, 2003.

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: May 1, 2003, as supplemented by letter dated September 30, 2003.

Brief description of amendment: The amendment modifies the surveillance testing requirements for the containment spray system by deleting the requirement to verify the position of valves that are locked, sealed, or otherwise secured in their correct position (and by deleting wording regarding the verified valves being positioned to take suction from the refueling water tank), and replacing the quantitative allowable pump degradation value with a requirement to verify the pumps perform in accordance with the Inservice Testing Program.

Date of issuance: December 4, 2003.

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

Amendment No.: 252.

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

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28851).

The September 30, 2003, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 4, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 27, 2003, as supplemented on July 17, July 31, September 11, and November 25, 2003.

Brief description of amendments: The amendments revise Technical Specification Section 3.4.9, “Reactor Coolant System Pressure and Temperature (P/T) Limits,” incorporating revisions to the P/T limit curves. The amendment also deletes the license conditions specified in DNPS Unit 2 Facility Operating License Section 2.C(8) and DNPS Unit 3 Facility Operating License Section 3.P, “Pressure-Temperature Limit Curves.”

Date of issuance: November 26, 2003.

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

Amendment Nos.: 205/197.

Facility Operating License Nos. DPR-19 and DPR-25: The amendments revised the Facility Technical Specifications and license conditions specified in the Facility Operating Licenses.

Date of initial notice in Federal Register: August 5, 2003 (68 FR 46242).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 26, 2003.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 26, 2002.

Brief description of amendments: These amendments revised TS 3.1.3.1, “Control Rod Operability,” by adding new Limiting Condition for Operation criteria and applicable ACTION requirements for scram discharge volume (SDV) vent and drain valves. The changes also modified TS 3.6.3, “Primary Containment Isolation Valves,” to clarify the relationship between TS 3.1.3.1 and TS 3.6.3 regarding SDV vent and drain valves.

Date of issuance: November 26, 2003.

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

Amendment Nos.: 168 and 131.

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

Date of initial notice in Federal Register: January 7, 2003 (68 FR 803).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 26, 2003.

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, Beaver County, Pennsylvania

Date of application for amendments: March 26, 2003.

Brief description of amendments: These amendments modify Technical Specifications (TSs) 4.0.1 and 4.0.3 to be consistent with the Improved Standard Technical Specifications. The amendments also modify the TS requirements for missed surveillances in TS 4.0.3 to be consistent with the Nuclear Regulatory Commission-approved Technical Specification Task Force (TSTF), Standard Technical Specification Change TSTF-358, Revision 6.

Date of issuance: November 25, 2003.

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

Amendment Nos.: 258 and 140.

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

Date of initial notice in Federal Register: June 24, 2003 (68 FR 37577).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 25, 2003.

No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California

Date of application for amendments: December 9, 2002, as supplemented by letter dated August 28, 2003.

Brief description of amendments: The changes would revise Technical Specification (TS) 3.75, “Auxiliary Feedwater System,” Surveillance Requirement (SR) 3.7.5.2 Frequency. Specifically, the wording of the Frequency of SR 3.7.5.2 would change from “31 days on a Staggered Test Basis” to “In accordance with the Inservice Testing Program.” This change is requested to implement recommendations of the Standard Technical Specifications for Combustion Engineering Plants, NUREG-1432, Revision 2.

Date of issuance: November 25, 2003.

Effective date: November 25, 2003, to be implemented within 60 days of issuance.

Amendment Nos.: Unit 2—191; Unit 3—182.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 7, 2003 (68 FR 812).

The August 28, 2003, 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 no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 25, 2003.

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: November 14, 2002, as supplemented by letters dated October 30, and November 6, 2003.

Brief description of amendments: The amendments revise the Updated Final Safety Analysis Report (UFSAR) to eliminate the turbine missile design basis.

Date of issuance: December 2, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance. The UFSAR changes shall be implemented in the next periodic update to the UFSAR in accordance with 10 CFR 50.71(e).

Amendment Nos.: Unit 1—158; Unit 2—146.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the UFSAR.

Date of initial notice in Federal Register: February 18, 2003 (68 FR 7821).

The October 30, and November 6, 2003, supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on February 18, 2003 (68 FR 7821).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 2, 2003.

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 22, 2003, as supplemented by letters dated September 10 and September 30, 2003.

Brief description of amendments: The amendments change the pressurizer safety valve lift tolerance, as specified in Technical Specification (TS) 3.4.2.2, “Reactor Coolant System,” from plus/minus (±) 2 percent (%) to +2% and −3%.

Date of issuance: December 2, 2003.

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

Amendment Nos.: Unit 1—159; Unit 2—147.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the TSs.

Date of initial notice in Federal Register : June 24, 2003 (68 FR 37583).

The September 10 and September 30, 2003, supplemental letters provided clarifying information that was within the scope of the original Federal Register notice (68 FR 37583) and did not change the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 2, 2003.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, Alabama

Date of application for amendments: August 7, 2003.

Description of amendment request: The amendments modified Technical Specification (TS) requirements for mode change limitations to adopt Industry/TS Task Force (TSTF) change TSTF-359, “Increase Flexibility in Mode Restraints.”

Date of issuance: December 1, 2003.

Effective date: Date of issuance, to be implemented within 60 days.

Amendment Nos.: 249, 286 & 244.

Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. Amendments revised the TSs.

Date of initial notice in Federal Register : October 14, 2003 (68 FR 59221).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 1, 2003.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

Date of application for amendment: March 13, 2003, as supplemented July 30, 2003.

Description of amendment: The amendment revises the boron concentration requirements in Technical Specifications (TSs) 3.5.2, “Cold Leg Accumulators,” and 3.5.5, “Refueling Water Storage Tank.” The revised boron concentration requirement is a function of the number of tritium producing burnable absorber rods (TPBARs) in the core.

Date of issuance: December 1, 2003.

Effective date: As of the date of issuance to be implemented no later than startup from an outage in which TPBARs are loaded into the reactor.

Amendment Nos.: 289 & 279.

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

Date of initial notice in Federal Register : April 15, 2003 (68 FR 18286). The supplemental letter provided clarifying information only and did not change the scope of the original amendment request or the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated December 1, 2003.

No significant hazards consideration comments received: No.

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

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

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual 30-day notice of consideration of issuance of amendment, proposed no significant hazards consideration determination, and opportunity for a hearing.

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

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

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

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated.

For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Assess and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By January 22, 2004, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, 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 1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

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

Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Southern Nuclear Operating Company, Inc., et al., Docket No. 50-425, Vogtle Electric Generating Plant, Unit 2, Burke County, Georgia

Date of amendment request: November 5, 2003.

Description of amendment request: The proposed amendment would extend the surveillance interval for the Memories Test portion of the Actuation Logic Test for: (1) Power Range Block (Switch position 1), (2) Intermediate Range Block (Switch position 2), (3) Source Range Block (Switch positions 3 and 4), (3) Safety Injection (SI) Block, Pressurizer (Switch positions 5 and 6), (4) SI Block, High Steam Pressure Rate (Switch positions 7 and 8), (5) Auto SI Block (Switch position 9), and (6) Feedwater Isolation on P14 or SI (Switch positions 10 and 11). In addition to the functions listed above, the licensee is requesting an extension of the surveillance interval for the portions of the Actuation Logic Test for Feedwater Isolation on P14 or SI that pass through the memories circuits and the Power Range block of the Source Range Trip test for the Unit 2 Train B Solid State Protection System to the next refueling outage at the end of Cycle 10 or the next Unit 2 shutdown to MODE 5, whichever comes first.

Date of issuance: December 3, 2003.

Effective date: December 3, 2003.

Amendment No.: 108.

Facility Operating License No. NPF-81: Amendment revises the technical specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. November 18, 2003 (68 FR 65092). The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided an opportunity to request a hearing by December 18, 2003, 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 December 3, 2003.

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

NRC Section Chief: John A. Nakoski.

Dated in Rockville, Maryland, this 15th day of December, 2003.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

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

[FR Doc. 03-31314 Filed 12-22-03; 8:45 am]

BILLING CODE 7590-01-P