E6-9434. 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 May 25, 2006 to June 8, 2006. The last biweekly notice was published on June 6, 2006 (71 FR 32603).

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].

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

Date of amendment request: May 8, 2006.

Description of amendment request: The proposed change will add an NRC-approved topical report to the analytical methods referenced in Technical Specification (TS) Section 5.6.5, “Core Operating Limits Report (COLR).”

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.

Core operating limits are established each operating cycle in accordance with TS 3.2, “Power Distribution” and TS 5.6.5, “Core Operating Limits Report (COLR)”. These core operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO). In addition, the Average Planar Linear Heat Generation Rate (APLHGR) operating limits imposed by Technical Specification 3.2.1 also ensure that the Peak Cladding Temperature (PCT) during the postulated design[-]basis LOCA [loss-of-coolant accident] does not exceed the 2200 °F limit specified in 10 CFR 50.46. The APLHGR is a measure of the average linear heat generation rate of all the fuel rods in a fuel assembly at any axial location.

The methods used to determine the operating limits are those previously found acceptable by the NRC and listed in TS Section 5.6.5.b. A change to TS Section 5.6.5.b is requested to include an updated LOCA analysis method, EXEM BWR-2000. The updated method will be used to determine the APLHGR operating limits imposed by Technical Specification 3.2.1. EXEM BWR-2000 has been reviewed and approved by the NRC and is applicable to the GGNS [Grand Gulf Nuclear Station, Unit 1] plant design and the FRA-ANP [Framatome-Advance Nuclear Power] fuel being used at GGNS. The application of the LOCA analytical model will continue to ensure that the APLHGR operating limits are established to protect the fuel cladding integrity during normal operation, AOOs, and the design-basis LOCA. The requested TS changes concern the use of analytical methods and do not involve any plant modifications or operational changes that could affect any postulated accident precursors or accident mitigation systems and do not introduce any new accident initiation mechanisms.

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 TS amendment will not change the design function, reliability, performance, or operation of any plant systems, components, or structures. It does not create the possibility of a new failure mechanism, malfunction, or accident initiators not considered in the design and licensing bases. Plant operation will continue to be within the core operating limits that are established using NRC[-]approved methods that are applicable to the GGNS design and the GGNS fuel.

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 ECCS [emergency core cooling system] performance analysis methods are used to establish the APLHGR limits required by Technical Specification 3.2.1. The APLHGR limits are specified in the COLR and are the result of fuel design, design[-]basis accident (DBA), and transient analyses. Limits on the APLHGR are specified to ensure that the fuel design limits are not exceeded during anticipated operational occurrences (AOOs) and that the peak cladding temperature (PCT) during the postulated design[-]basis LOCA does not exceed the 2200 °F limit specified in 10 CFR 50.46.

The EXEM BWR-2000 evaluation model is an updated LOCA analytical method that has been approved by the NRC and is applicable to the GGNS plant design and the fuel being used at GGNS. A GGNS plant[-]specific ECCS performance analysis has been performed with the EXEM BWR-2000 evaluation model. This evaluation concluded that the resulting PCT still afforded adequate margin to the 2200 °F limit of 10 CFR 50.46.

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: Nicholas S. Reynolds, Esquire, Winston and Strawn LLP, 1700 K Street, NW., Washington, DC 20006

NRC Branch Chief: David Terao.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: May 24, 2006.

Description of amendment request: This amendment revises TS 1.0, Definitions, TS 3/4.4.5, Steam Generator Tube Integrity, TS 3/4.4.6.2, Reactor Coolant System (RCS) Operational LEAKAGE, adds a new specification TS 6.8.4.k for Steam Generator Program and adds a new TS 6.9.1.12, Steam Generator Tube Inspection Report. The proposed changes are necessary in order to implement the guidance for the industry initiative on NEI 97-06, “Steam Generator Program Guidelines.” 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 March 2, 2005, (70 FR 10298). The licensee affirmed the applicability of the model NSHC determination in its application dated May 24, 2006.

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 requires a SG Program that includes performance criteria that will provide reasonable assurance that the SG tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cooldown and all anticipated transients included in the design specification). The SG performance criteria are based on tube structural integrity, accident induced leakage, and operational LEAKAGE.

A SGTR event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. In the analysis of a SGTR event, a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in the licensing basis plus the LEAKAGE rate associated with a double-ended rupture of a single tube is assumed.

For other design basis accidents such as MSLB, rod ejection, and reactor coolant pump locked rotor the tubes are assumed to retain their structural integrity (i.e., they are assumed not to rupture). These analyses typically assume that primary to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to 1 gallon per minute as a result of accident induced stresses. The accident induced leakage criterion introduced by the proposed changes accounts for tubes that may leak during design basis accidents. The accident induced leakage criterion limits this leakage to no more than the value assumed in the accident analysis.

The SG performance criteria proposed change to the TS identify the standards against which tube integrity is to be measured. Meeting the performance criteria provides reasonable assurance that the SG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident. The performance criteria are only a part of the SG Program required by the proposed change to the TS. The program, defined by NEI 97-06, Steam Generator Program Guidelines, includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring. The proposed changes do not, therefore, significantly increase the probability of an accident previously evaluated.

The consequences of design basis accidents are, in part, functions of the DOSE EQUIVALENT 1-131 in the primary coolant and the primary to secondary LEAKAGE rates resulting from an accident. Therefore, limits are included in the plant technical specifications for operational leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure the plant is operated within its analyzed condition. The typical analysis of the limiting design basis accident assumes that primary to secondary leak rate after the accident is 1 gallon per minute with no more than 150 gallons per day in any one SG, and that the reactor coolant activity levels of DOSE EQUIVALENT 1-131 are at the TS values before the accident. The proposed change does not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. The proposed approach updates the current TSs and enhances the requirements for SG inspections. The proposed change does not adversely impact any other previously evaluated design basis accident and is an improvement over the current TSs.

Therefore, the proposed change does not affect the consequences of a SGTR accident and the probability of such an accident is reduced. In addition, the proposed changes do not affect the consequences of an MSLB, rod ejection, or a reactor coolant pump locked rotor event, or other previously evaluated accident.

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

The proposed performance based requirements are an improvement over the requirements imposed by the current technical specifications. Implementation of the proposed SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The result of the implementation of the SG Program will be an enhancement of SG tube performance. Primary to secondary LEAKAGE that may be experienced during all plant conditions will be monitored to ensure it remains within current accident analysis assumptions.

The proposed change does not affect the design of the SGs, their method of operation, or primary or secondary coolant chemistry controls. In addition, the proposed change does not impact any other plant system or component. The change enhances SG inspection requirements.

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

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

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of an SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change is expected to result in an improvement in the tube integrity by implementing the SG Program to manage SG tube inspection, assessment, repair, and plugging. The requirements established by the SG Program are consistent with those in the applicable design codes and standards and are an improvement over the requirements in the current TSs.

For the above reasons, the margin of safety is not changed and overall plant safety will be enhanced by the proposed change to the TS.

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

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

Attorney for licensee: J. Hagood Hamilton, South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Section Chief: Evangelos C. Marinos.

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

Date of amendment request: May 1, 2006 (TS-05-10).

Description of amendment request: The proposed amendment would extend the burnup limit of the Mark-BW fuel design with advanced alloy material referred to as M5 alloy. This proposed change affects Section 6.9.1.14.a of the Sequoyah Nuclear Plant Technical Specifications (TSs). The impact to Section 6.9.1.14.a includes adding an NRC-approved topical report (TR) associated with M5 alloy fuel assemblies. This TR will be utilized, among others, in the determination of core operating limits for each fuel cycle. In addition, the proposed amendment includes the adoption of Industry/Technical Specification Task Force (TSTF) Traveler, TSTF-363, Revision 0, “Revised Topical Report References in Improved Technical Specification (ITS) 5.6.5, Core Operating Limits Report (COLR),” which removes any references to dates, revision numbers, and supplements in the TS listing of TRs.

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.

In general, fuel assemblies and more specifically fuel rod cladding, of any burnup level, is not a precursor to accidents previously evaluated. An evaluation has been performed of the Mark-BW design fuel assembly for all loss-of-coolant accidents (LOCA) and non-LOCA transient events. This evaluation confirmed and justified the use of Mark-BW fuel for operation in Sequoyah Nuclear Plant (SQN) Units 1 and 2.

The ability of the M5 fuel rod cladding material to provide a barrier against the release of radioactive fuel material has not been reduced with respect to the Zircaloy-4 material. The approved TR evaluated postulated accidents that involved adverse core conditions and the release of radionuclides, and found that higher burnup limits have very little impact on the overall radiological consequences. Radiological consequences, as well as other safety limits, are evaluated on a cycle-to-cycle basis to confirm that the analyses of record remain bounding. If a proposed extended burnup core design exceeds bounding safety analysis values, then either the core design would be changed, or the safety values would be changed.

Rod cladding failures are assumed to occur in the fuel handling accident; however, the consequences of this event are independent of the properties of the fuel rod cladding. This is based on the fuel handling event assuming the rupture of all fuel rods regardless of the rod cladding material.

No change is proposed to the established safety analysis fuel assembly inputs, specifically fuel assemblies are still limited to a maximum 1500 effective full power day (EFPD) burnup and the reactor core average maximum burnup will remain at 1000 EFPD burnup ensuring the present accident analyses remain bounding. Based on above discussion, the proposed revision to extend the burnup limit of M5 fuel rod cladding material will not significantly increase the consequences of an accident and the potential for the release of radioactive material to the environment.

Removing revision numbers, dates, and parenthetical information from the listed TRs has no impact on the actual analytical methods used to determine the core operating limits, nor does the change have impact on the calculations performed for the current or future reloads. This change is administrative in nature. This change has no impact on plant equipment operation nor does it affect the likelihood or consequences of an accident previously evaluated.

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

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

Response: No.

Mark-BW fuel design with M5 alloy has been demonstrated to have similar characteristics to that of the Mark-B fuel design. Extended burnup of the M5 material has not been shown to alter the functions of the rod cladding, which is to provide a barrier against the release of radioactive material. Initial plant conditions, which are considered in the accident analysis, will also be maintained such that no new plant conditions will exist that could affect the analysis results. Since plant functions and conditions are not impacted by the proposed revision and the higher burnup limit of the Mark-BW fuel design with M5 alloy material is not postulated to become an accident initiator based on the similarity with Mark-B fuel design and Zircaloy-4 material, the possibility of a new or different kind of accident is not created.

The proposed changes will not alter the plant configuration or require any new or unusual operator actions. They do not alter the way any structure, system, or component functions and do not alter the manner in which the plant is operated. These changes do not introduce any new failure modes.

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 is established by the acceptance criteria used by NRC. Meeting the acceptance criteria assures that the consequences of accidents are within known and acceptable limits. The emergency core cooling system (ECCS) acceptance criteria are not exceeded. Testing has been performed on M5 alloy with respect to criteria for peak cladding temperature (PCT) and maximum cladding oxidation. These tests demonstrate that M5 alloy rod cladding remains within PCT of 2200 degrees Fahrenheit and conservatively bounded by the 17 percent limit for maximum cladding oxidation. M5 alloy oxidation rates are lower than that of Zircaloy at temperatures less than 2200 degrees Fahrenheit and have similar rates for temperatures up to about 2300 degrees Fahrenheit. High-temperature oxidation rates of M5 alloy remain equivalent to Zircaloy and, as such, respond as hydrogen generators to the same extent. Core geometry for amenable cooling is not directly related to rod cladding material; however, it applies equally well to all materials. The consequences of both thermal and mechanical deformation of fuel assemblies have been assessed, and the resultant deformations have been shown to maintain coolable core configurations. The ECCS is evaluated against the thermal power immediately after shutdown. The thermal power is largely a function of short-lived fission products which tend to saturate at relatively low burnup limits and are not appreciably affected by extended burnup. Therefore, with no system changes being proposed; long-term cooling is maintained. Additionally, the fuel storage cooling system is capable of supporting the long-term storage of the extended burnup fuel assemblies' decay heat.

The changes to burnup limit have been evaluated against Departure from Nucleate Boiling (DNB) events and all applicable acceptance criteria are met. In addition, the proposed revision to allow an increase in the burnup limit of the Mark-BW fuel design with M5 alloy will not impact plant setpoints that maintain the margin of safety. Based on these results, it is concluded that the margin of safety is not significantly reduced. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Removing revision numbers, dates, and parenthetical information from the listed TRs will not reduce a margin of safety because this information has no effect on any safety analysis assumption nor does it revise any setpoints assumed in the analysis of record. The proposed change is consistent with NUREG-1431, issued by the NRC staff, revising the TSs to reflect the approved level of detail, which indicates that there is no 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

NRC Branch Chief: Michael L. Marshall, Jr.

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

Date of amendment request: May 25, 2006 (TSC 06-02).

Description of amendment request: The proposed amendment would revise Section 6.2.1.6 of the Sequoyah Nuclear Plant (SQN) Updated Final Safety Analysis Report (UFSAR). This change would revise the methodology used for containment sump debris transport analysis and affects SQN's current design and licensing basis described in Section 6.2.1.6 of the SQN UFSAR.

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 design function of the sump during accident conditions is to support emergency core cooling systems (ECCS) and containment spray system operation for recirculation. The sump is a passive feature that does not act as an accident initiator, (i.e., failure of the sump would not initiate a design basis accident).

The proposed change to the UFSAR regarding debris transport analysis provides an overall improvement in the analysis for recirculation operation and does not change the consequences of accidents previously evaluated. The change in methodology is neutral with regard to probability. Consequently, the changes associated with the enclosed license amendment do not affect the frequency of occurrence for accidents previously evaluated in the UFSAR.

Accident dose as previously evaluated in the UFSAR is unaffected by the proposed license amendment.

Based on the above discussion, 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 sump is a passive component and is not an accident initiator; i.e., failure of the sump will not initiate a design basis accident. The sump transport methodology is used to confirm the ability of the sump to perform all safety functions during normal and accident conditions. Consequently, this activity does not create a possibility of a new or different type of accident than any previously evaluated.

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

Response: No.

The changes addressed in TVA's proposed amendment are associated with methodology for debris transport to the containment sump.

The change does not affect specific safety limits, design limits, set points, or other critical parameters. The transport methodology is used to confirm that the ECCS and containment spray systems will perform their safety functions for all accident conditions within existing equipment performance capability margins.

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

NRC Branch Chief: Michael L. Marshall, Jr.

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

Date of amendment request: May 9, 2006.

Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) 1.1, “Definitions,” and 3.4.16, “RCS [reactor coolant system] Specific Activity.” The revisions would replace the current Limiting Condition for Operation (LCO) 3.4.16 limit on RCS gross specific activity with limits on RCS Dose Equivalent I-131 and Dose Equivalent Xe-133 (DEX). The conditions and required actions for LCO 3.4.16 not being met, and surveillance requirements for LCO 3.4.16, are being revised. The modes of applicability for LCO 3.4.16 would be extended. The current definition of Ē—Average Disintegration Energy in TS 1.1 would be replaced by the definition of DEX. In addition, the current definition of Dose Equivalent I-131 in TS 1.1 would be revised to allow alternate, NRC-approved thyroid dose conversion factors.

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 change[s] involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes would add new thyroid dose conversion factor reference[s] to the definition of DOSE EQUIVALENT I-131, eliminate the definition of Ē—AVERAGE DISINTEGRATION ENERGY, add a new definition of DOSE EQUIVALENT XE-133, replace the Technical Specification (TS) 3.4.16 limit on reactor coolant system (RCS) gross specific activity with a limit on noble gas specific activity in the form of a Limiting Condition for Operation (LCO) on DOSE EQUIVALENT XE-133, increase the Completion Time for Required Action B.1, replace TS Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT I-131, extend the Applicability of LCO 3.4.16, and make corresponding changes to TS 3.4.16 to reflect all of the above. The proposed changes are not accident initiators and have no impact on the probability of occurrence of any design basis accidents.

The proposed changes will have no impact on the consequences of a design basis accident because they will limit the RCS noble gas specific activity to be consistent with the values assumed in the radiological consequence analyses. The changes will also limit the potential RCS [radio]iodine concentration excursion to the value currently associated with full power operation, which is more restrictive on plant operation than the existing allowable RCS [radio]iodine specific activity at lower power levels.

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

2. [Do] the proposed change[s] create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes do not alter any physical part of the plant nor do they affect any plant operating parameters besides the allowable specific activity in the RCS. The changes which impact the allowable specific activity in the RCS are consistent with the assumptions assumed in the current radiological consequence analyses. [The proposed changes are also not accident initiators.]

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

3. [Do] the proposed change[s] involve a significant reduction in a margin of safety?

Response: No.

The acceptance criteria related to the proposed changes involve the allowable control room and offsite radiological consequences following a design basis accident. The proposed changes will have no impact on the radiological consequences of a design basis accident because they will limit the RCS noble gas specific activity to be consistent with the values assumed in the radiological consequence analyses. The changes will also limit the potential RCS [radio]iodine specific activity excursion to the value currently associated with full power operation, which is more restrictive on plant operation than the existing allowable RCS [radio]iodine specific activity at lower power levels.

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: John O'Neill, Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: David Terao.

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

Date of amendment request: May 11, 2006.

Description of amendment request: The proposed amendment would revise Surveillance Requirements 3.7.2.1, 3.7.3.1, and 3.7.3.3 on verifying the closure time of the main steam isolation valves (MSIVs), main feedwater regulating valves (MFRVs), main feedwater regulating valve bypass valves (MFRVBVs), and main feedwater isolation valves (MFIVs) in the Technical Specifications (TSs). These valves are the Main Steam and Main Feedwater System isolation valves. The revisions would replace (1) the specified maximum acceptable valve closure time for the MSIVs, MFRVs, and MFRVBVs, and (2) TS Figure 3.7.3-1, which shows acceptable valve closure times for the MFIVs, by the reference to the valve closure time, is verified to be “within limits.” The maximum acceptable valve closure times for the MFRVs and MFRVBVs, and TS Figure 3.7.3-1 will be relocated to the TS Bases. The maximum acceptable valve closure time for the MSIV is already in the TS Bases.

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

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

Response: No.

Because the proposed change[s remove] specific isolation times from the TS and [relocate] the specific values to the TS Bases, there are no design or physical changes to the facility or to the Main Steam and Main Feedwater System isolation valves themselves. The design and functional performance requirements, operational characteristics, and reliability of these components remain unchanged. There is[,] therefore[,] no impact on the design safety function of the valves to close (as an accident mitigator), nor is there any change with respect to inadvertent closure (as a potential transient initiator). Since no failure mode or initiating condition that could cause an accident (including any plant transient) evaluated per the FSAR [Final Safety Analysis Report]-described safety analyses is created or affected, the change cannot involve a significant increase in the probability of an accident previously evaluated. The probability of an accident is not affected. The Main Steam and Main Feedwater System isolation valves are assumed to function to mitigate some accidents (for example, SLB [steam line break] and FWLB [main feedwater line break]). The proposed change[s] only [affect] the level of detail included in the TS. The TS requirements continue to provide the same level of assurance as before that the Main Steam and Main Feedwater System isolation valves are capable of performing their intended safety function. These isolation valves will continue to be verified operable in the same manner as before. As such, the proposed change[s do] not affect the ability of the isolation valves to perform their assumed mitigation function.

Therefore, 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.

The proposed change[s] only [affect] the level of detail included in the TS. The TS requirements [are not being changed and they will] continue to provide the same level of assurance as before that the Main Steam and Main Feedwater System isolation valves are capable of performing their intended safety function. The Main Steam and Main Feedwater System isolation valves will continue to be verified operable in the same manner. As such, the proposed change[s do] not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed change[s] will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure.

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

3. [Do] the proposed change[s] involve a significant reduction in a margin of safety?

Response: No.

The proposed change[s do] not reduce the margin of safety. The proposed change[s] only [affect] the level of detail included in the TS. The TS requirements [are not being changed and will] continue to provide the same level of assurance as before that the Main Steam and Main Feedwater System isolation valves will continue to be verified operable in the same manner as before. As such, the proposed change[s do] not affect the assumptions of any accident analysis or the availability or operability of any plant equipment.

Therefore, the proposed changes do 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: John O'Neill, Esq., Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: David Terao.

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.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: February 14, 2006.

Brief description of amendment request: The proposed amendments would add a requirement to the Title 10 of the Code of Federal Regulations, (10 CFR) part 50 license to restrict the minimum cooling time and burnup of spent fuel assemblies that will be placed into storage in the NUHOMS HD spent fuel dry storage system at Surry starting in the summer of 2006.

Date of publication of individual notice in Federal Register: May 16, 2006 (71 FR 28390).

Expiration date of individual notice: 30 day expiration date, June 15, 2006, and 60 day expiration date, July 17, 2006.

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 26, 2004, as supplemented April 18 and October 11, 2005, and May 19, 2006.

Brief description of amendment: The amendment revised Technical Specification 3.8.7, “Inverters—Operating” to change the completion time for restoration of an inoperable Division 1 or 2 inverter from the current 24 hours to 7 days.

Date of issuance: May 26, 2006.

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

Amendment No.: 174.

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

Date of initial notice in Federal Register: June 8, 2004 (69 FR 32072). The supplements dated April 18 and October 11, 2005, and May 19, 2006, provided additional information that clarified the application, but 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.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 26, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 3, 2005, as supplemented by letter dated March 7, 2006.

Brief description of amendments: The amendments revise the Updated Final Safety Analysis Report (UFSAR) to incorporate the description of the approved changes associated with the plant modifications made to the diesel generator cooling water system for each emergency diesel generator as described in the amendment application of June 3, 2005, as supplemented by letter dated March 7, 2006.

Date of issuance: May 25, 2006.

Effective date: As of the date of issuance to be implemented within 90 days from the date of issuance.

Amendment Nos.: Unit 1-160, Unit 2—160, Unit 3 -160.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revise the Operating Licenses and the UFSAR for all three units.

Date of initial notice in Federal Register: July 5, 2005 (70 FR 38715). The March 7, 2006, supplemental letter provided additional clarifying information, 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.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 25, 2006.

No significant hazards consideration comments received: No.

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

Date of application of amendments: August 20, 2004, supplemented January 31, 2006.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.3.8, “Post Accident Monitoring (PAM) Instrumentation,” to eliminate TS requirements associated with the reactor building spray flow instruments commensurate with the importance of their post-accident function.

Date of Issuance: June 1, 2006.

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

Amendment Nos.: 350/352/351.

Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the Licenses and the Technical Specifications.

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57983).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 1, 2006.

No significant hazards consideration comments received: No.

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

Date of amendment request: September 19, 2005.

Brief description of amendment: The proposed changes would revise Technical Specification (TS) 3.1.1.5, “Minimum Temperature for Criticality.” The request proposes to change the current Limiting Condition for Operation (LCO) for TS 3.1.1.5 by raising the minimum temperature for criticality from the current value of ≥ 525 °F to ≥ 540 °F; to change the current Action statement for LCO 3.1.1.5 to reflect this change; and to delete the current statement in Surveillance Requirement 4.1.1.5 and replace the statement with wording consistent with NUREG-1432, “Standard Technical Specifications Combustion Engineering Plants.” Also, changes will be made to the ANO-2 TS Bases in accordance with the Technical Specifications (TS) Bases Control Program (ANO-2 TS 6.5.14).

Date of issuance: May 30, 2006.

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

Amendment No.: 264.

Renewed Facility Operating License No. NPF-6: The amendment revised the Technical Specifications and Surveillance Requirements.

Date of initial notice in Federal Register: December 6, 2005, (70 FR 72672).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 30, 2006.

No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

Date of application for amendments: January 20, 2005, as supplemented July 5, 2005.

Brief description of amendments: The amendments revised several Technical Specifications (TSs) using six TS Task Force (TSTF) generic changes. The six TSTFs (nos. 5, 93, 258, 299, 308, and 361) delete redundant safety limit violation notification requirements; extend the pressurizer heater surveillance frequency from 92 days to 18 months; remove redundant requirements and add other requirements to the Administrative Controls section of the TSs; clarify the requirements regarding the frequency of testing for cumulative and projected dose contributions from radioactive effluents; and add a note to the residual heat removal requirements during Mode 6 low water level operations that allows one required residual heat removal (RHR) loop to be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is operable and in operation.

The amendments represent partial approval of the January 20, 2005, application for the proposed amendments. The Commission has granted the request of Florida Power and Light Company (the licensee) to withdraw portions of its January 20, 2005, application for the proposed amendment. The application also included TSTF-95, which would extend the completion time for reducing the Power Range High trip setpoint from 8 to 72 hours and TSTF-101, which would change the auxiliary feedwater pump test frequency to be consistent with the inservice test program frequency. However, by letter dated March 22, 2005, the licensee withdrew the request to adopt TSTF-95 and by letter dated October 13, 2005, the licensee withdrew the request to adopt TSTF-101.

Date of issuance: May 26, 2006.

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

Amendment Nos: 229 and 225.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the TSs.

Date of initial notice in Federal Register: March 15, 2005 (70 FR 12747). The supplement dated July 5, 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 May 26, 2006.

No significant hazards consideration comments received: No.

Nuclear Management Company, Docket No. 50-263, Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota

Date of application for amendment: June 29, 2005, as supplemented by letters dated April 25 (two letters), May 4, and May 12, 2006.

Brief description of amendment: The amendment converts the current Technical Specifications (CTSs) to the Improved Technical Specifications (ITSs) format and relocates certain requirements to other licensee-controlled documents. The ITSs are based on NUREG-1433, “Standard Technical Specifications General Electric Plants BWR/4,” Revision 3, dated June 2004; the Commission's Final Policy Statement, “NRC Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors,” dated July 22, 1993 (58 FR 39132); and 10 CFR 50.36, “Technical specifications.” The purpose of the conversion is to provide clearer and more readily understandable requirements in the TSs for MNGP to ensure safer operation of the unit. In addition, the amendment includes a number of issues that are considered beyond the scope of NUREG-1433.

Date of issuance: June 5, 2006.

Effective date: As of the date of issuance and shall be implemented by September 30, 2006.

Amendment No: 146.

Facility Operating License No. DPR-22: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: November 16, 2005 (70 FR 70889).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 5, 2006.

No significant hazards consideration comments received: No.

Amendment No: 146.

Facility Operating License No. DPR-22: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: November 16, 2005 (70 FR 70889).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 5, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 19, 2005, as supplemented by letter dated December 23, 2005.

Brief description of amendments: The amendments updated the Technical Specification (TS)5.3, “Unit Staff Qualifications,” operator minimum qualification requirements contained in the March 28, 1980, NRC letter to all licensees with the more recent NRC-approved operator qualification requirements contained in American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.1-1993. In addition, the changes removed the TS 5.3.1 plant staff retraining and replacement training program requirements, which have been superseded by requirements contained in 10 CFR 50.120.

Date of issuance: May 26, 2006.

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

Amendment Nos.: Unit 1—187 ; Unit 2—189.

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

Date of initial notice in Federal Register: December 20, 2005 (70 FR 75495). The December 23, 2005, supplemental letter provided additional information that clarified the application, and did not expand the scope of the application as originally noticed.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 26, 2006.

No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of application for amendments: February 28, 2006, as supplemented on April 7, 2006.

Brief description of amendments: The amendments revise the SSES 1 and 2 Technical Specification (TS) Surveillance Requirements 3.8.4.7 and 3.8.4.8 to clarify that Diesel Generator “E” (DG E) electrical power subsystem testing does not require a mode restriction when the DG E diesel is not aligned to the Class 1E distribution system.

Date of issuance: May 30, 2006.

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

Amendment Nos.: 235 and 212.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the TSs and license.

Date of initial notice in Federal Register: March 28, 2006 (71 FR 15485). The supplement dated April 7, 2006, 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.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 30, 2006.

No significant hazards consideration comments received: No.

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 application for amendments: February 10, 2005, as supplemented by letters dated July 14, 2005, and October 20, 2005.

Brief description of amendments: The amendments modified Technical Specification Surveillance Requirement 4.5.3.2 b to allow safety injection and charging pumps to run in a recirculation flow path, provided that two independent means are used to prevent injection into the reactor coolant system.

Date of issuance: May 31, 2006.

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

Amendment Nos.: 273 and 254.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 12, 2005 (70 FR 19116). The supplements dated July 14, 2005 and October 20, 2005 provided clarifying information only 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 May 31, 2006.

No significant hazards consideration comments received: No.

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 application for amendments: August 31, 2005, as supplemented by letters dated December 8, 2005, and April 10, 2006.

Brief description of amendments: The amendments changed the Technical Specifications (TSs) to move the requirements for the containment area high-range radiation monitors from TS 3/4.3.3.1, “Radiation Monitoring Instrumentation,” to TS 3/4.3.3.7, “Accident Monitoring Instrumentation,” and correct a typographical error in Surveillance Requirement 4.2.2.

Date of issuance: May 25, 2006.

Effective date: May 25, 2006.

Amendment Nos.: 272 and 253.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the TSs.

Date of initial notice in Federal Register: January 17, 2006 (71 FR 2594). The April 10, 2006 supplement did not expand the scope of the application, as originally noticed, and did not change the staff's original proposed no significant hazards consideration.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 25, 2006.

No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: April 29, 2005, as supplemented on August 15 and December 9, 2005, and January 11 and 25, and May 9, 2006.

Brief description of amendment: The amendment revises Technical Specification (TS) 3.5.1, “Accumulators,” and TS 3.5.4, “Refueling Water Storage Tank,” to reflect the results of revised analyses performed to accommodate the proposed extended power uprate and revises TS 5.6.4, “Core Operating Limits Report,” to permit the use of approved methodology for large-break and small-break loss-of-coolant accident analyses.

Date of issuance: May 31, 2006.

Effective date: As of the date of issuance to be implemented prior to restart from the fall 2006 refueling outage.

Amendment No.: 96.

Renewed Facility Operating License No. DPR-18: Amendment revised the Technical Specifications and the license.

Date of initial notice in Federal Register: June 7, 2005 (70 FR 33219). The August 15 and December 9, 2005, and January 11 and 25, and May 9, 2006, letters 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 amendment is contained in a Safety Evaluation dated May 31, 2006.

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: March 17, 2006, as supplemented on April 14, 2006. The supplemental letter dated April 14, 2006, provided clarifying information that did not change the scope of the March 17, 2006, application nor the initial proposed no significant hazards consideration determination.

Brief description of amendments: The amendments authorized the licensee to credit administering potassium iodide (KI) to reduce the 30-day post-accident thyroid dose to the occupants of the main control room for an interim period of 4 years. In addition, the design-basis accident analysis section of the Updated Final Safety Analysis Reports will be updated to reflect crediting of KI.

Date of issuance: May 25, 2006.

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

Amendment Nos.: 249 and 193.

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

Date of initial notice in Federal Register: March 27, 2006 (71 FR 15223). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 25, 2006.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: July 21, 2005.

Brief Description of amendments: These amendments revised the Technical Specifications (TSs) to change the accident monitoring instrumentation listing, allowed outage times, requirements, and surveillances to be consistent with the requirements of the Improved TSs for post-accident monitoring instrumentation.

Date of issuance: May 31, 2006.

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

Amendment Nos.: 247/246.

Renewed Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: January 3, 2006 (71 FR 155).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 31, 2006.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this June 12, 2006.

For the Nuclear Regulatory Commission.

Catherine Haney,

Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.