E7-5342. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from March 2, 2007 to March 15, 2007. The last biweekly notice was published on March 13, 2007 (72 FR 11383).

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, Rulemaking, Directives and Editing 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: February 8, 2007.

Description of amendment request: The proposed amendment would modify Grand Gulf Nuclear Station, Unit 1 (GGNS) technical specification (TS) requirements for MODE change limitations in limiting condition for operation (LCO) 3.0.4 and surveillance requirement (SR) 3.0.4. The proposed TS changes are consistent with Revision 9 of Nuclear Regulatory Commission (NRC) approved Industry TS Task Force (TSTF) Standard TS Change Traveler, TSTF-359, “Increase Flexibility in MODE Restraints.” In addition, the proposed amendment would also change TS Section 1.4, Frequency, Example 1.4-1, “Surveillance Requirements,” to accurately reflect the changes made by TSTF-359, which is consistent with NRC-approved TSTF-485, Revision 0, “Correct Example 1.4-1.”

The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), as part of the Consolidated Line Item Improvement Process (CLIIP), on possible amendments to revise the plant-specific TS to modify requirements for MODE change limitations in LCO 3.0.4 and SR 3.0.4.

The NRC staff subsequently issued a notice of availability of the models for Safety Evaluation and No Significant Hazards Consideration Determination for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the CLIIP, including the model No Significant Hazards Consideration Determination, in its application dated February 8, 2007.

The proposed TS changes are consistent with NRC-approved Industry TSTF Standard TS change, TSTF-359, Revision 8, as modified by 68 FR 16579. TSTF-359, Revision 8, was subsequently revised to incorporate the modifications discussed in the April 4, 2003, Federal Register notice and other minor changes. TSTF-359, Revision 9, was subsequently submitted to the NRC on April 28, 2003, and was approved by the NRC on May 9, 2003.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the NRC staff's analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Changes Do Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed changes in TS Section 1.4, Frequency, Example 1.4-1, would accurately reflect the changes made by TSTF-359 in LCO 3.0.4 and SR 3.0.4, which are consistent with NRC-approved TSTF-485, Revision 0. These changes are considered administrative in that they modify the example to demonstrate the proper application of LCO 3.0.4 and SR 3.0.4. The requirements of LCO 3.0.4 and SR 3.0.4 are clear and are clearly explained in the associated Bases. As a result, modifying the example will not result in a change in usage of the TS.

The proposed changes in LCO 3.0.4 and SR 3.0.4 allow entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The proposed changes do not adversely affect accident initiators or precursors, the ability of structures, systems, and components to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Being in a TS condition and the associated required actions are not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by these changes. The addition of a requirement to assess and manage the risk introduced by these changes will further minimize possible concerns. Therefore, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2—The Proposed Changes Do Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated

No new or different accidents result from utilizing the proposed changes. The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The proposed changes do not alter assumptions made in the safety analysis and are consistent with the safety analysis assumptions and current plant operating practice. Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by these changes will further minimize possible concerns. Thus, these changes do not create the possibility of a new or different kind of accident from an accident previously evaluated.

Criterion 3—The Proposed Changes Do Not Involve a Significant Reduction in the Margin of Safety

The proposed changes in TS Section 1.4, Example 1.4-1, are considered administrative and will have no effect on the application of the TS requirements. Therefore, the margin of safety provided by the TS requirements is unchanged. The proposed changes in TS LCO 3.0.4 and SR 3.0.4 allow entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The GGNS TS allows operation of the plant without the full complement of equipment through the TS conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS LCO condition on the margin of safety is not considered significant. The proposed changes do not alter the required actions or completion times of the TS. The proposed changes allow TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The changes also eliminate current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, these changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Terence A. Burke, Associate General Council—Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, Mississippi 39213.

NRC Branch Chief: David Terao.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County, Illinois.

Date of amendment request: January 8, 2007.

Description of amendment request: The proposed amendment would revise the technical specification (TS) requirements for selected reactor trip system (RTS) instrumentation, engineered safety feature actuation system (ESFAS) instrumentation, and containment ventilation isolation instrumentation to adopt completion times, test bypass time, and surveillance test interval changes. The changes are based on Westinghouse Electric Company, LLC, topical reports WCAP-14333-P-A, Revision 1, “Probabilistic Risk Analysis of the [Reactor Protection System] RPS and ESFAS Test Times and Completion Times,” and WCAP-15376-P-A, Revision 1, “Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times.”

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

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

Overall protection system performance will remain within the bounds of the previously performed accident analyses since no hardware changes are proposed. The same RTS and ESFAS instrumentation will continue to be used. The protection systems will continue to function in a manner consistent with the plant design basis. These changes to the TS do not result in a condition where the design, material, and construction standards that were applicable prior to the change are altered.

The proposed changes will not modify any system interface. The proposed changes will not affect the probability of any event initiators. There will be no degradation in the performance of or an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report.

The determination that the results of the proposed changes are acceptable was established in the NRC Safety Evaluations prepared for WCAP-14333-P-A, (issued by letter dated July 15, 1998) and for WCAP-15376-P-A, (issued by letter dated December 20, 2002). Implementation of the proposed changes will result in an insignificant risk impact.

Applicability of these conclusions has been verified through plant-specific reviews and implementation of the generic analysis results in accordance with the respective NRC Safety Evaluation conditions.

The proposed changes to the CTs [completion times], test bypass times, and Surveillance Frequencies reduce the potential for inadvertent reactor trips and spurious engineered safeguard features actuations, and therefore do not increase the probability of any accident previously evaluated. The proposed changes do not change the response of the plant to any accidents and have an insignificant impact on the reliability of the RTS and ESFAS signals. The RTS and ESFAS will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This is demonstrated by showing that the impact on plant safety, as measured by the increase in core damage frequency (CDF) is less than 1.0E-06 per year and the increase in large early release frequency (LERF) is less than 1.0E-07 per year. In addition, for the CT changes, the incremental conditional core damage probabilities (ICCDP) and incremental conditional large early release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, respectively. These changes meet the acceptance criteria in Regulatory Guides (RGs) 1.174 and 1.177. Therefore, since the RTS and ESFAS will continue to perform their functions with high reliability, as originally assumed, and the increase in risk, as measured by ΔCDF, ΔLERF, ICCDP, ICLERP risk metrics, is within the acceptance criteria of existing regulatory guidance, there will not be a significant increase in the consequences of any accidents.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed changes are consistent with safety analysis assumptions and resultant consequences.

Therefore, this change does not increase the probability or consequences of any accident previously evaluated.

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no hardware changes nor are there any changes in the method by which any safety-related plant system performs its safety function. The proposed changes will not affect the normal method of plant operation. No performance requirements will be affected or eliminated. The proposed changes will not result in physical alteration to any plant system nor will there be any change in the method by which any safety-related plant system performs its safety function. There will be no setpoint changes or changes to accident analysis assumptions.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of these changes. There will be no adverse effect or challenges imposed on any safety-related system as a result of these changes.

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

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

The proposed changes do not affect the acceptance criteria for any analyzed event nor is there a change to any Safety Analysis Limit. There will be no effect on the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the departure from nucleate boiling limits, fuel centerline temperature, or any other margin of safety. The radiological dose consequence acceptance criteria listed in the NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” will continue to be met.

Redundant RTS and ESFAS trains are maintained, and diversity with regard of the signals that provide reactor trip and engineered safety features actuation is also maintained. All signals credited as primary or secondary, and all operator actions credited in the accident analyses will remain the same. The proposed changes will not result in plant operation in a configuration outside the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RGs 1.174 and 1.177. Although there was no attempt to quantify any positive human factors benefit due to increased CTs and bypass test times, it is expected that there would be a net benefit due to a reduced potential for spurious reactor trips and actuations associated with testing.

Implementation of the proposed changes is expected to result in an overall improvement in safety, as follows:

• Reduced testing will result in fewer inadvertent reactor trips, less frequent actuation of ESFAS components, less frequent distraction of operations personnel without significantly affecting RTS and ESFAS reliability.
• Improvements in the effectiveness of the operating staff in monitoring and controlling plant operation will be realized. This is due to less frequent distraction of the operators and shift supervisor to attend to instrumentation Required Actions with short CTs.
• Longer repair times associated with increased CTs will lead to higher quality repairs and improved reliability.
• The CT extensions for the reactor trip breakers will provide additional time to complete test and maintenance activities while at power, potentially reducing the number of forced outages related to compliance with reactor trip breaker CT, and provide consistency with the CT for the logic trains.

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 requested amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Bradley J. Fewell, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Russell A. Gibbs.

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

Date of amendment request: January 30, 2007.

Description of amendment request: The proposed amendment would revise Technical Specifications (TSs) Surveillance Requirement (SR) 3.5.1.3.b to correctly state that the required pressure at which the Alternate Nitrogen System is determined to be operable should be greater than or equal to 410 psig, not the currently stated pressure of greater than or equal to 220 psig. The safety-related Alternate Nitrogen System provides an alternate pressure source to equipment required during or following an accident. The licensee has determined that the current acceptance value specified by SR 3.5.1.3.b is non-conservative and needs to be corrected to the higher value.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (10 CFR) Part 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration (NSHC). The NRC staff reviewed the licensee's analysis, and has performed its own analysis as follows:

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

No. The proposed amendment would only correct the acceptance value specified by SR 3.5.1.3.b. The acceptance value of the nitrogen supply was not considered to be a precursor to, and does not affect the probability of, an accident. In addition, there is no design or operation change associated with the proposed amendment.

Therefore, the proposed amendment does not increase the probability of an accident previously evaluated.

The corrected, higher pressure of the Alternate Nitrogen System will ensure that nitrogen is available to operate equipment after an accident, as designed. The increased acceptance value will not decrease the functionality of the Alternate Nitrogen System, or the functionality of the plant equipment it supports. Therefore, the plant systems required to mitigate accidents will remain capable of performing their design functions. As a result, the proposed amendment will not lead to a significant change in the consequences of any accident.

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

No. The proposed amendment does not involve a physical alteration of any system, structure, or component (SSC) or a change in the way any SSC is operated. The proposed amendment does not involve operation of any SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the revised acceptance value.

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

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

No. The proposed amendment only changes the acceptance value of the Alternate Nitrogen System. There will be no modification of any TSs limiting condition for operation, no change to any limit on previously analyzed accidents, no change to how previously analyzed accidents or transients would be mitigated, no change in any methodology used to evaluate consequences of accidents, and no change in any operating procedure or process. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Branch Chief: L. Raghavan.

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 amendment requests: February 8, 2007.

Description of amendment requests: This license amendment request will (1) revise Technical Specification (TS) Surveillance Requirement (SR) 3.3.7.3.a to lower the allowable value for dropout and raise the allowable value for pickup of the degraded voltage function, and (2) revise TS SR 3.8.1 to lower the diesel generator minimum output voltage due to lower settings for the degraded voltage function.

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.

This proposed change revises the Technical Specification (TS) Surveillance Requirement (SR) 3.3.7.3.a allowable values of the Degraded Voltage Function and SRs 3.8.1.2, .7, .9, .11, .12, .15, .16, .17, .19, and .20 for Diesel Generator (DG) minimum operable voltage. This proposed change will allow Southern California Edison (SCE) to widen the operating band while maintaining adequate conservatism for the degraded relay settings and overall loop uncertainties while keeping 218 kV as the minimum voltage on the offsite transmission grid necessary to support operability of the immediate access offsite power source (also referred to as the normal preferred power source). This will be accomplished by lowering the dropout and increasing the pickup settings of the degraded voltage protection relays. Following approval of this proposed change, the 4.16 kV Class 1E buses would remain on the normal preferred power source at or above a grid voltage of 218 kV while protecting all Class 1E equipment from degraded grid conditions.

The degraded voltage protection circuits are designed to protect electrical equipment against the effects of degraded voltage on the offsite transmission networks. Therefore, these circuits are generally not considered to be accident initiators. However, spurious actuation of the degraded voltage protection relays could result in the loss of the preferred power source (offsite source of alternating current (AC) power). The proposed change lowers the allowable value for dropout and raises the allowable value for pickup for the degraded voltage protection relays. This results in an increase in operating band and a lower probability of spurious actuation of these degraded voltage signals. Therefore, there is no increase in the probability of a Loss of Offsite Power (preferred power source) as a result of this proposed change.

The safety function of the degraded voltage protection circuits is to ensure the operability of Class 1E equipment. SCE has performed calculations that demonstrate that operation in accordance with this proposed change will not result in operation of plant equipment at degraded voltages. Therefore, there is no increase in the consequences of any accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any 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 allowable values of the degraded voltage relays and the DG minimum operating voltage will provide an acceptable level of protection for plant equipment.

This proposed change affects only the voltage settings of the degraded voltage protection relays and voltage regulator setting of the DG for lowering the required bus voltage. There is no other change to the degraded voltage function. There are no physical modifications necessary to the degraded voltage protection relays or the DG. There are no changes to the actions performed by the relays or the DG following actuation. Therefore, there are no new failure modes or effects introduced by this proposed change.

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

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

Response: No.

The proposed degraded voltage protection schemes are designed to ensure that plant equipment will not operate at a degraded voltage and the DG Automatic Voltage Regulator (AVR) is set to provide adequate voltage for resetting of the relays and satisfactory operation of the Safety Related equipment. The proposed degraded voltage allowable values will not affect the existing protection criterion for plant equipment. This maintains the existing margin of safety for plant equipment.

Therefore, there is no significant reduction in a margin of safety as a result of the proposed amendment.

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

Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.

NRC Branch Chief: David Terao.

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 amendment request: February 13, 2007.

Description of amendment request: The proposed amendment would modify the licensee's Technical Specification (TS) Section 3.9.1, “Refueling Equipment Interlocks,” to add required actions to allow insertion of a control rod withdrawal block and verification that all control rods are fully inserted as alternate actions to suspending in-vessel fuel movement in the event that one or more required refueling equipment interlocks are inoperable. These changes are based on Technical Specification Task Force (TSTF) change TSTF-225, Revision 2, “Fuel movement with inoperable refueling equipment interlocks” and are consistent with the current Boiling Water Reactor (BWR)/4 Standard Technical Specifications (STS), NUREG-1433, Volume 1, Revision 3.0.

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?

The proposed change provides additional actions for an inoperable refueling equipment interlock. The proposed actions will allow fuel movement with inoperable refueling interlocks, however, those actions will require the insertion of a continuous control rod withdrawal block, as well as verification that all control rods are fully inserted, before the commencement of fuel movement. Since fuel movement with the refueling interlocks operable allows control rod withdrawal under some circumstances, complete prevention of control rod withdrawal with the refueling interlocks inoperable does not increase the likelihood of a reactivity event, and may in fact decrease its probability of occurrence.

The refueling interlocks are not designed or otherwise intended to prevent or mitigate the consequences of the fuel handling accident. This proposed change does not involve those structures that could have an effect on the fuel handling accident and its consequences, such as the fuel design, the integrity of the refueling platform, and the integrity of the refueling mast and grapple. Furthermore, the consequences of the refueling accident are not increased since, should that accident occur while operating under the provisions of the alternate actions, all control rods will be fully inserted. The consequences of the fuel assembly insertion error event during refueling are not increased since this proposed change preserves the initial conditions of that transient event, i.e., all control rods inserted.

Implementing these changes will not increase the likelihood of an equipment failure resulting from the use of the refueling cranes and hoists. Such protection is afforded by other plant (owner controlled) specifications and procedures. These documents require testing and maintenance of these components separate from the requirements of [Limiting Condition for Operation] LCO 3.9.1.

This submittal does not affect any other system, structure or component that is important with respect to the prevention and mitigation of other accidents or transients.

For the above reasons, this proposed Technical Specifications change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change provides additional actions (the insertion of a control rod block and verification that all control rods are fully inserted) for inoperable refueling interlocks. This change does not involve any permanent alterations to plant systems or components. Nor does it involve changes to operational configurations or to the maintenance and testing of systems or components. Consequently, no new modes of operation are being introduced. Therefore, the 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 the margin of safety?

The proposed change provides additional actions for an inoperable required refueling equipment interlock. The new actions will require that all control rods be fully inserted and that a control rod block be in effect. Under the current specifications, control rod withdrawal is allowed during fuel movement under certain conditions.

The alternate actions of the proposed specifications will not allow rod withdrawal under any circumstances during fuel movement operations, therefore, this proposed change provides a level of safety at least equivalent to the existing actions.

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

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

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Branch Chief: Evangelos C. Marinos.

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 26, 2007.

Description of amendment request: The proposed change adds an operating license condition and revises the Technical Specifications to permit the replacement of main control room (MCR) and emergency switchgear room (ESGR) air-conditioning system (ACS) chilled water piping by using temporary 45-day and 14-day allowed outage times (AOTs) four times in a 24-month time span.

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

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

The proposed change has been evaluated using the risk-informed processes described in Regulatory Guide (RG) 1.174, “An Approach for Using Probabilistic Risk assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” and RG 1.177, “An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications.”

The risk associated with the proposed change was found to be acceptably “small” and therefore not a significant increase in the probability or consequences of an accident previously evaluated.

In addition, the proposed change does not affect the initiators of analyzed events or the assumed mitigation of accident or transient events. During the temporary 45-day and 14-day AOT entries, equipment availability restrictions will restrict or limit the out-of-service time of risk significant plant equipment due to surveillance testing, preventive maintenance, and elective maintenance. In addition, during the replacement activities, compensatory actions will be in place to ensure the availability of chilled water or to provide backup cooling. Therefore, the ACS will continue to perform its required function. As a result, the proposed change to the Surry TS does not involve any significant increase in the probability or the consequences of any accident or malfunction of equipment important to safety previously evaluated since neither accident probabilities nor consequences are being affected by this proposed change.

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

The proposed change does not involve a change in the methods used to respond to plant transients. There is no alteration to the parameters within which the plant is normally operated or in the setpoints, which initiate protective or mitigative actions. The MCR and ESGR ACS will continue to perform its required function. This is assured by the planned implementation of compensatory actions, including provisions for backup cooling. Consequently, no new failure modes are introduced by the proposed change. Therefore, the proposed Surry TS change does not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any accident previously evaluated.

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

Margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated, and the establishment of the setpoints for the actuation of equipment relied upon to respond to an accident or transient event. The proposed change does not affect the ability of the MCR and ESGR ACS to perform its required function. This is assured by the planned implementation of compensatory actions, including provisions for backup cooling. Furthermore, the proposed change has been evaluated using the risk-informed processes described in Regulatory Guide (RG) 1.174, “An approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” and RG 1.177, “An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications.”

The risk associated with the proposed change was found to be acceptably small. Therefore, the proposed change to the Surry TS 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: Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Branch Chief: Evangelos C. Marinos.

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: March 6, 2007.

Description of amendment request: The proposed amendments would revise the licensing basis (Updated Final Safety Analysis Report (UFSAR)) to permit irradiation of the fuel assemblies beginning with Surry Power Station, Unit Nos. 1 and 2, improved fuel assemblies with ZIRLO (Westinghouse trademark) cladding to a lead rod average burnup of 62,000 MWD/MTU.

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

1. The probability of occurrence or the consequence of an accident previously evaluated is not significantly increased.

The activity being evaluated is a slight increase in the lead rod average burnup limit for the fuel assemblies. No change in fuel design or fuel enrichment will be required to increase the lead rod average burnup. The fuel rods at the extended lead rod average burnup will continue to meet the design limits with respect to fuel rod growth, clad fatigue, rod internal pressure and corrosion. There will be no impact on the capability to engage the fuel assemblies with the handling tools. Therefore, it is concluded that the change will not result in an increase in the probability of occurrence of any accident previously evaluated in the UFSAR. The impact of extending the lead rod average burnup to 62,000 MWD/MTU from 60,000 MWD/MTU on the core kinetics parameter, core thermal-hydraulics/[departure from nucleate boiling ratio]DNBR, specific accident considerations, and radiological consequences was considered. Based on the evaluation of these considerations, it is concluded that increasing the lead rod average burnup limit to 62,000 MWD/MTU will not result in a significant increase in the consequences of the accidents previously evaluated in the Surry UFSAR.

2. The possibility for a new or different type of accident from any accident previously evaluated is not created.

The fuel is the only component affected by the change in the burnup limit. The change does not affect the thermal hydraulic response to any transient or accident. The existing fuel rod design criteria continue to be met at the higher burnup limit. Thus, the change does not create the possibility of an accident of a different type.

3. The margin of safety as defined in the Bases to the Surry Technical Specifications is not significantly reduced.

The operation of the Surry cores with a limited number of fuel assemblies with some fuel rods irradiated to a lead rod average burnup of 62,000 MWD/MTU will not change the performance requirements of any system or component such that any design criteria will be exceeded. The normal limits on core operation defined in the Surry Technical Specifications will remain applicable for the irradiation of the fuel to a lead rod average burnup of 62,000 MWD/MTU. Therefore, the margin of safety as defined in the Bases to the Surry Technical Specifications is not significantly reduced.

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

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Branch Chief: Evangelos C. Marinos.

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.

Florida Power Corporation, et al.

Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida.

Date of amendment request: February 8, 2007.

Description of amendment request: To change the basis for protection of spent fuel stored in the spent fuel pool (SFP) in order to eliminate the Final Safety Analysis Report commitment for maintaining the SFP missile shields.

Date of publication of individual notice in the Federal Register : March 13, 2007. (72 FR 11381).

Expiration date of individual notice: May 14, 2007.

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 51.22(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].

Carolina Power & Light Company

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

Date of application for amendment: May 30, 2006, as supplemented by letter dated November 20, 2006.

Brief description of amendment: The amendment revises the existing steam generator tube surveillance program at H. B. Robinson Steam Electric Plant, Unit No. 2.

Date of issuance: March 12, 2007.

Effective date: This license amendment is effective as of the date of issuance and shall be implemented within 60 days.

Amendment No. 212.

Renewed Facility Operating License No. DPR-23. Amendment revises the Technical Specifications.

Date of initial notice in Federal Register : December 19, 2007 (71 FR 75990). The November 20, 2006, supplemental letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated: March 12, 2007.

No significant hazards consideration comments received: No.

Dominion Energy Kewaunee, Inc.

Docket No. 50-305, Kewaunee Power Station, Kewaunee County, Wisconsin.

Date of application for amendment: January 30, 2006, as supplemented by letter dated January 23, 2007.

Brief description of amendment: The amendment modifies the radiological accident analyses and associated technical specifications.

Date of issuance: March 8, 2007.

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

Amendment No.: 190.

Facility Operating License No. DPR-43: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register : March 14, 2006 (71 FR 13172).

The supplemental letter contained clarifying information and did not change the initial no significant hazards consideration determination, and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 8, 2007.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc.

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

Date of application for amendment: February 7, 2006, as supplemented by letters dated August 14, 2006, and January 2, 2007.

Brief description of amendments: The amendment revised the Millstone Power Station, Unit No. 3 Technical Specifications to permit an increase in the allowed outage time from 72 hours to 7 days for the inoperablity of the steam supply to the turbine-driven auxiliary feedwater pump (AFW) or the inoperability of the turbine-driven AFW pump under certain operating mode restrictions.

Date of issuance: February 28, 2007.

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

Amendment No.: 235.

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

Date of initial notice in Federal Register : April 11, 2006 (70 FR 18372).

The supplements dated August 14, 2006, and January 2, 2007, 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 February 28, 2007.

No significant hazards consideration comments received: No.

Entergy Operations, Inc.

Docket Nos. 50-313 and 50-368, Arkansas Nuclear One, Units 1 and 2, Pope County, Arkansas.

Date of amendment request: October 25, 2005, as supplemented by letter dated March 20, 2006.

Brief description of amendments: The changes addressed inventory and inspection requirements associated with the emergency cooling pond, which is a common cooling water source for both units during conditions that may render the normal cooling water source (Dardanelle Reservoir) unavailable.

Date of issuance: March 9, 2007.

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

Amendment Nos.: Unit 1-229, Unit 2-271.

Renewed Facility Operating License Nos. DPR-51 and NPF-6: Amendments revised the Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register : October 24, 2006 (71 FR 62309). The supplemental letter dated March 20, 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 as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 9, 2007.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC

Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1, Oswego County, New York.

Date of application for amendment: October 19, 2006, as supplemented by letter dated January 5, 2007.

Brief description of amendment: The amendment revises the Surveillance Requirement (SR) in Technical Specification (TS) 4.1.1.c, “Scram Insertion Times,” to modify the conditions under which scram time testing (STT) of control rods is required, and to add a requirement to perform STT on a defined portion of control rods, at a specified frequency, during the operating cycle. The amendment also revises the SR in TS 4.1.7.c, “Minimum Critical Power Ratio (MCPR),” to add a requirement to determine the MCPR operating limits following completion of control rod STT per TS 4.1.1.c.

Date of issuance: March 15, 2007.

Effective date: March 15, 2007.

Amendment No.: 193.

Facility Operating License No. DPR-63: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register : December 5, 2006 (71 FR 70562) The supplemental letter dated January 5, 2007, 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.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 15, 2007.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC

Docket No. 50-311, Salem Nuclear Generating Station, Unit No. 2, Salem County, New Jersey.

Date of application for amendment: April 6, 2006.

Brief description of amendment: The amendment changed the Technical Specifications (TSs) to reduce the maximum allowable reactor power level when two main steam safety valves are inoperable.

Date of issuance: March 7, 2007.

Effective date: As of the date of issuance and shall be implemented prior to restart from the steam generator replacement outage.

Amendment No.: 259.

Facility Operating License No. DPR-75: The amendment revised the TSs and the License.

Date of initial notice in Federal Register : November 7, 2006 (71 FR 65144).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 7, 2007.

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: May 1, 2006, as supplemented by letter dated November 3, 2006.

Brief description of amendment: The amendment revises the steam generator tube integrity Technical Specifications consistent with the Nuclear Regulatory Commission's approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, “Steam Generator Tube Integrity,” Revision 4.

Date of issuance: March 1, 2007.

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

Amendment No.: 100.

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

Date of initial notice in Federal Register : June 6, 2006 (71 FR 32605).

The supplemental letter dated November 3, 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 as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 1, 2007.

No significant hazards consideration comments received: No.

Tennessee Valley Authority (TVA)

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

Date of application for amendment: September 22, 2006.

Brief description of amendment: The amendment supplements a June 28, 2004, request to increase the licensed thermal power from 3293 megawatt thermal (MWt) to 3952 MWt, an approximate 20% increase in thermal power. This supplement requests interim approval of an increase in licensed thermal power from 3293 MWt to 3458 MWt with an attendant 30-psi increase in reactor pressure. This represents an approximate 5% increase above the original licensed thermal power of 3293 MWt. An interim approval would provide for operation at 105% power until such time as certain steam dryer analyses can be completed. The NRC staff's review of the remainder of the June 2004 application would resume upon receipt of the satisfactorily completed steam dryer analyses.

Date of issuance: March 6, 2007.

Effective date: Date of issuance, to be implemented prior to restart.

Amendment No.: 269.

Renewed Facility Operating License No. DPR-33: Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register : October 10, 2006 (71 FR 59532). The Commissions related evaluation of the amendment is contained in an Environmental Assessment dated February 12, 2007 (72 FR 6612), and in a Safety Evaluation dated March 6, 2007.

No significant hazards consideration comments received: No.

Union Electric Company

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

Date of application for amendment: March 28, 2006, as supplemented by letter dated November 17, 2006.

Brief description of amendment: The amendment deleted references to specific isolation valves in the chemical and volume control system (CVCS) and modified to allow (1) an exception for decontamination activities and (2) an exception for CVCS resin vessel operation. These are changes to TS 3.3.9, “Boron Dilution Mitigation System (BDMS),” and TS 3.9.2, “Unborated Water Source Isolation Valves.”

Date of issuance: March 8, 2007.

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

Amendment No.: 181.

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

Date of initial notice in Federal Register : May 9, 2006 (71 FR 27004). The supplemental letter dated November 17, 2006, 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 published in the Federal Register. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 8, 2007.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company

Docket Nos. 50-338 and 50-339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia.

Date of application for amendment: October 3, 2006, as supplemented by letter dated January 24, 2007.

Brief description of amendment: The proposed amendments revised the Technical Specifications (TSs) and licensing basis to support the resolution of the Nuclear Regulatory Commission's (NRC's) Generic Safety Issue (GSI) 191, assessment of debris accumulation on containment sump performance and its impact on emergency recirculation during an accident, and NRC Generic Letter (GL) 2004-02.

Date of issuance: March 13, 2007.

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

Amendment Nos.: 250 and 230.

Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the licenses and the technical specifications.

Date of initial notice in Federal Register : December 5, 2006 (71 FR 70563). The supplement dated January 24, 2007, 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 March 13, 2007.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 19th day of March 2007.

For the Nuclear Regulatory Commission.

John W. Lubinski,

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