05-12987. Biweekly Notice; Application 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 June 10, 2005 to June 23, 2005. The last biweekly notice was published on June 21, 2005 (70 FR 35735).

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].

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

Date of amendments request: May 26, 2005.

Description of amendments request: The amendments would revise the Technical Specification (TS) requirements related to steam generator (SG) tube integrity, consistent with those in NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, “Steam Generator Tube Integrity.” The proposed amendment also includes changes to the revised SG program in TS Section 5.5.9 to specify the SG tube inspection length through the SG tubesheet and establish plugging criteria in the inspected tubesheet region for the remaining original SGs containing Alloy 600 mill annealed (MA) tubes. This change is being proposed to establish conformance with the NRC position identified in Generic Letter (GL) 2004-01, “Requirements for Steam Generator Tube Inspections.”

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 analysis that established the inspection length through the SG tube sheet for the PVNGS Alloy 600 MA-tube SGs took into account the reinforcing effect the tubesheet has on the external surface of an expanded SG tube. Tube-bundle integrity will not be adversely affected by the implementation of the revised tube inspection scope. SG tube burst or collapse cannot occur within the confines of the tubesheet; therefore, the tube burst and collapse criteria of draft Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR Steam Generator Tubes,” are inherently met. Any degradation below the inspection length is shown by analyses and test results to be acceptable, thereby precluding an event with consequences similar to a postulated tube rupture event.

Tube burst is precluded for cracks within the tubesheet by the constraint provided by the tubesheet. Thus, structural integrity is maintained by the tubesheet constraint. However, a 360-degree circumferential crack or many axially oriented cracks could permit severing of the tube and tube pullout from the tubesheet under the axial forces on the tube from primary to secondary pressure differentials. Analysis and testing was performed to define the length of non-degraded tubing that is sufficient to compensate for the axial forces on the tube and thus prevent pullout. That length is bounded by the inspection length proposed in this change.

In conclusion, incorporation of the revised inspection scope into PVNGS TS maintains existing design limits and 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 performance based requirements are an improvement over the requirements imposed by the current TS.

Implementation of the proposed Steam Generator 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 Steam Generator 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.

Tube-bundle integrity is expected to be maintained during all plant conditions upon implementation of the proposed tube inspection scope. Use of this scope does not introduce a new mechanism that would result in a different kind of accident from those previously analyzed. Even with the limiting circumstances of a complete circumferential separation of a tube occurring below the inspection length into the tubesheet, SG tube pullout is precluded and leakage is predicted to be maintained within the Updated Final Safety Analysis Report limits during all plant conditions.

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

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

Response: No.

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 also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a 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 Steam Generator Program to manage SG tube inspection, assessment, repair, and plugging. The requirements established by the Steam Generator Program are consistent with those in the applicable design codes and standards and are an improvement over the requirements in the current TS.

Upon implementation of the revised inspection scope, operation with potential cracking below the Inspection Extent length in the expansion region of the SG tubing will meet the margin of safety as defined by Regulatory Guide (RG) 1.83 [Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes], draft RG 1.121 [Bases for Plugging Degraded PWRSteam Generator Tubes], and the requirements of General Design Criteria 14, 15, 31, and 32 of Appendix A to 10 CFR 50.

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

Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, Arizona 85072-2034. NRC Acting Section Chief: Daniel S. Collins.

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

Date of amendments request: June 3, 2005.

Description of amendments request: The proposed amendments would revise the Updated Final Safety Analysis Report (UFSAR) for Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2 and 3. The proposed amendments would reflect a modification performed by the licensee that replaced the automatic water makeup function for the emergency diesel generator jacket water cooling system with that of manual operator actions.

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 emergency diesel generator (EDG) is a system that must function in response to an accident that has been evaluated in either Chapter 6 or 15 of the PVNGS UFSAR. It is designed to respond to certain described accident scenarios. None of the accidents evaluated are initiated within the EDG system. Therefore, this request to allow the replacement of the automatic makeup feature(s) with a manual feature can not increase the probability of an accident previously postulated in the UFSAR.

None of the accidents evaluated which credit operation of the EDG system require automatic fill of the DGCWS [Diesel Generator Cooling Water System] in order to mitigate the consequences of the accident. The fill system, whether automatic in nature as originally designed or manual, simply maintains the EDG in the ready state.

Therefore, the proposed change does not involve a significant increase in the 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 EDG is a piece of equipment important to safety. This modification replaces the automatic water makeup function for the EDG jacket water cooling system with that of manual operator actions. The jacket water makeup is needed for normal leakage and possible evaporation. Area walkdowns occur twice daily when the diesel generator is in a standby mode (not running) and more frequently (thirty minutes after initial loading and every two hours while loaded) when the EDG is being tested or has responded to an emergency event. The area operator walkdown procedures instruct the operators to log the standpipe level and ensure it is in the normal operating range. If the level is not, operators are required to restore level and conduct further investigation of the condition and notify appropriate personnel. This ensures that enough water remains in the jacket water system to allow the diesel to remain operational and evaluations are performed in order to detect any abnormal leakrates. Therefore, the normal area operator walkdowns and frequencies are adequate to ensure that sufficient jacket water standpipe inventory is maintained.

With this modification, the EDG is still maintained and monitored for proper conditions in a standby status to ensure that it will respond to emergencies when called upon. Once the EDG responds to an emergency signal and is loaded, its jacket water system is required to be monitored every two hours to help ensure that all parameters are observed and maintained for proper operation, including its jacket water standpipe level.

So, with these measures in place it can be expected that the EDG will be maintained capable of performing as designed to any emergency safety signal. The [E]DG safety system and its support jacket water cooling system do not initiate any accident events.

Therefore, the modification of this non-safety support system cannot 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 PVNGS UFSAR states that the design basis function for the emergency diesel generators is to provide a standby source of onsite Class 1E AC power for the two trains of engineered safety features equipment for safe plant shutdown and decay heat removal in the event of loss of preferred (off-site) power. Supporting this design basis function of supplying emergency power is the function of the emergency diesel generator jacket cooling water system, which is to remove rejected heat from each diesel engine at the rated design load of the emergency diesel generator. The UFSAR further describes the emergency diesel generator jacket cooling water surge tank (standpipe), stating that the surge tank is sized to provide an adequate reservoir to compensate for any minor leaks. The UFSAR also described makeup to the jacket cooling water system as being automatically actuated and provided from the safety-grade condensate transfer system or manually from the demineralized water systems. The subject modification replaced the automatic features with manual operator action—the sources of the makeup water have not changed.

The PVNGS engineering analyses and the safety analyses that demonstrate the functional goals and the design basis of the emergency diesel generator system do not credit any makeup water supply to the jacket cooling water system of the emergency diesel generator for an initial 25 hours into an event. Operator monitoring and manual makeup provides adequate control for maintaining the DGCWS standpipe level, both for standby and loaded conditions. An automatically actuated makeup water supply is not essential to the safe and continued operation of the emergency diesel generator. Makeup water is provided as a convenient source of water to compensate for anticipated normal system losses and evaporation. It is not provided to serve as an emergency source of makeup water to the jacket cooling water system in the event of a major failure or leak occurring within the jacket cooling water system.

Makeup to the system is required to compensate for normal expected system losses, minor leaks, and evaporation. In addition, an engineering calculation has been performed to address 10 CFR 50, Appendix R concerns, which demonstrates that no operator action is required or credited during the first twenty-five hours of emergency diesel generator loaded operation provided that the initial water level is at the specified minimum level. This twenty-five hour period before operator intervention, which is assumed to occur, sufficiently bounds the thirty minutes of no operator action that is normally assumed in most of the accident analyses.

In addition, the area operator walkdown procedures instruct the operators to log the standpipe level and ensure it is in the normal operating range. If the level is not, operators are required to restore level and conduct further investigation of the condition and notify appropriate personnel. This ensures that enough water remains in the jacket water system to allow the diesel to remain operational and evaluations are performed in order to detect any abnormal leakrates.

Therefore, APS has concluded that the proposed license amendment request does not involve a significant reduction in a margin of safety.

Based on the above, Arizona Public Service Company (APS) concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of “no significant hazards consideration” is justified.

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

Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, Arizona 85072-2034.

NRC Section Chief: Daniel S. Collins.

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

Date of amendments request: June 7, 2005.

Description of amendments request: The proposed amendment would revise Technical Specification (TS) 3.1.1, “Shutdown Margin,” to modify Required Action B.1 restricting a positive reactivity addition. The proposed amendment would also correct an administrative error regarding an incorrect TS reference in TS 3.4.17, “Special Test Exception RCS [reactor coolant system] Loops—Modes 4 and 5.”

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

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

The intent of this change is to clarify a Technical Specification involving positive reactivity additions to the shutdown reactor so that small, controlled, safe insertions of positive reactivity will be allowed where they are now categorically prohibited, posing a potential conflict between two required actions. These controlled activities could result in a slight change in the probability of an event occurring as a RCS manipulation that is currently prohibited would now be allowed. However, RCS manipulations are rigidly controlled to minimize the possibility of a significant reactivity increase.

In addition, there is sufficient shutdown margin available in this condition to allow for slight reactivity changes without significantly increasing the probability of an accident previously evaluated.

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

The administrative error was in the marked up Technical Specification pages submitted with a proposed change. The correct Technical Specification number was provided in the proposal letter and was used by the staff in the discussion for accepting the proposed change. Correcting this administrative error does not change the significant hazards discussion previously submitted.

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

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

This proposed change involving positive reactivity addition allows for a minor plant operational adjustment without adversely impacting the safety analysis required shutdown margin. It does not involve any change to plant equipment or the shutdown margin requirements in the Technical Specifications.

The administrative error was in the marked up Technical Specification pages submitted with a proposed change. The correct Technical Specification number was provided in the proposal letter and was used by the staff in the discussion for accepting the proposed change. Correcting this administrative error does not change the significant hazards discussion previously submitted.

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

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

The margin of safety in Modes 3, 4 and 5 is preserved by the calculated shutdown margin which prevents an inadvertent criticality. The proposed change involving positive reactivity addition will permit reductions in discretionary shutdown margin that is beyond Technical Specification requirements. However, the shutdown margin required by the Technical Specifications is not changed. By not impacting the shutdown margin, the margin of safety is not affected.

The administrative error was in the marked up Technical Specification pages submitted with a proposed change. The correct Technical Specification number was provided in the proposal letter and was used by the staff in the discussion for accepting the proposed change. Correcting this administrative error does not change the significant hazards discussion previously submitted.

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

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

Attorney for licensee: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.

NRC Section Chief: Richard J. Laufer.

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

Date of amendments request: June 7, 2005.

Description of amendments request: The proposed amendment would revise the Technical Specifications (TSs) to eliminate the use of the defined term Core Alterations. The proposed amendment would incorporate the changes reflected in TS Task Force (TSTF) Travelers 471-T (TSTF-471-T) and TSTF-51-A. In addition, the proposed amendment would revise TS 3.9.2, “Nuclear Instrumentation,” by replacing “Core Alterations” with “positive reactivity additions” in the required action for an inoperable source range monitor during refueling operations. The limiting conditions for operation in TS 3.9.4, “Shutdown Cooling (SDC) and Coolant Recirculation—High Water Level,” would also be revised by replacing “core alterations” with “movement of fuel assemblies within containment.”

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

The proposed change eliminates the use of the defined term CORE ALTERATIONS from the Technical Specifications. Core alterations are not an initiator of any accident previously evaluated except a fuel handling accident. Those revised Technical Specifications that protect the initial conditions of a fuel handling accident also require the suspension of movement of irradiated fuel assemblies, which protects the initial condition of a fuel handling accident. Therefore, suspension of CORE ALTERATIONS do not affect the initiators of the accidents previously evaluated and suspension of CORE ALTERATIONS does not affect the mitigation of the accidents previously evaluated.

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

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

No new or different accidents result from utilizing the proposed change. The changes do not involve a physical modification of the plant (i.e., no new or different type of equipment will be installed) or a significant change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions.

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

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

Only two accidents are postulated to occur during plant conditions where CORE ALTERATIONS may be made: A fuel handling accident and a boron dilution accident. Suspending movement of irradiated fuel assemblies prevents a fuel handling accident. Also requiring the suspension of CORE ALTERATIONS is redundant to suspending movement of irradiated fuel assemblies and does not increase the margin of safety. CORE ALTERATIONS have no effect on a boron dilution accident. Core components are not involved in the initiation or mitigation of a boron dilution accident. Therefore, CORE ALTERATIONS have no effect on the margin of safety related to a boron dilution accident.

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 amendments request involves no significant hazards consideration.

Attorney for licensee: Carey Fleming, Sr. Counsel—Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202.

NRC Section Chief: Richard J. Laufer.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina and Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: October 27, 2004.

Description of amendment request: The amendments would revise the facility operating licenses (FOLs) to remove a license condition that limits the maximum rod average burnup for any rod to 60 GWd/mtU. This deletion would allow the 62 GWd/mtU limit, approved by the NRC, as documented in Duke Topical Report DPC-2009-P-A, to become the burnup limit. The amendments would also revise both of the station's Updated Final Safety Analysis Reports (Section 4.0) to include a new discussion of the fuel burnup limit. Additionally, approval would allow Duke to make an administrative revision to Duke Topical Report DPC-NE-2009-P-A, Revision 2, to reference the approval of these amendments and to reflect removal of the current license condition. Furthermore, the amendments would remove the McGuire FOL Section 2.E, that lists reporting requirements with regard to Maximum Power Level, Fire Protection, Protection of the Environment (Unit 2 FOL only), and Physical Protection. It would also remove the Catawba FOL Section 2.F, that lists reporting requirements with regard to Maximum Power Level, Updated Final Safety Analysis Report, Antitrust Conditions, Fire Protection, and Additional Conditions.

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

1. Would implementation of the changes proposed in this LAR [License Amendment Request] involve a significant increase in the probability or consequences of an accident previously evaluated?

No, deletion of the fuel burnup limit currently stated as an additional license condition in the McGuire and Catawba Facility Operating Licenses has no impact on accident probabilities. Further, as determined in the NRC's environmental assessment which supports the increased burnup limit (NUREG/CR-6703, Environmental Effects of Extending Fuel Burnup Above 60 GWd/mtU), the potential environmental consequences of postulated accidents are not expected to increase significantly with increased burnup. Duke concurs with this assessment conclusion for the burnup range in this LAR.

The deletion of the reporting requirements from the FOLs is solely administrative. No plant equipment or accident analyses will be affected by this deletion.

2. Would implementation of the changes proposed in this LAR create the possibility of a new or different kind of accident from any accident previously evaluated?

No, implementation of this amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms will be created as a result of the NRC approval of this LAR. No changes are being made to the plant which will introduce any new accident causal mechanisms. This amendment does not otherwise impact any plant structures, systems, or components that are accident initiators; therefore, no new accident types are being created.

3. Would implementation of the changes proposed in this LAR involve a significant reduction in a margin of safety?

No, margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. These barriers are not significantly affected by the changes proposed in this LAR. The effect of the increased burnup on fuel cladding was considered in the NRC's environmental assessment supporting the increase in the fuel burnup limit. Further, the proposed limit is equal to that approved for the fuel rod cladding at McGuire and Catawba.

The deletion of the reporting requirements from the FOLs is solely administrative in nature. No plant equipment or accident analyses will be affected by this deletion.

The margin of safety is established through the design of the plant structures, systems, 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 event, and thereby protect the fission product barriers. The proposed changes have no significant impact on any of these considerations in regard to the physical plant or the manner in which it is operated.

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

NRC Section Chief: Evangelos C. Marinos

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: October 11, 2004.

Description of amendment request: The proposed amendments apply to Technical Specifications 3.8.1, “AC Sources—Operating,” and 3.8.9, “Distribution Systems—Operating.” They would extend several completion times and would modify several Surveillance Requirement (SR) Notes. Additionally, they would correct a recently identified non-conservative situation that currently exists with SR 3.8.1.4.

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:

First Standard

Will implementation of the changes proposed in this license amendment request involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The changes proposed in this license amendment request increase the Technical Specifications Completion Times for the emergency diesel generators and electrical power and distribution systems. Increasing these Completion Times will not cause a significant increase in the probability or consequences of an accident which has been previously evaluated. This license amendment request is supported by an extensive risk-informed study performed by the nuclear industry and documented in a topical report and Technical Specifications Task Force travelers that have been submitted for NRC review and approval. Within this study, the risk impacts of increasing the Completion Times were calculated and compared against the acceptability guidelines contained in the applicable regulatory guides and found to be acceptable. The emergency diesel generators and electrical power and distribution systems and equipment affected by this license amendment request will remain highly reliable. Thus there will be no significant increase in the probability or consequences of an accident which has been previously evaluated.

The proposed changes that modify Surveillance Requirement notes are consistent with an NRC [Nuclear Regulatory Commission]-approved industry initiative. Implementation of these changes will require that the plant's risk be managed. Thus there will be no significant increase in the probability or consequences of an accident which has been previously evaluated.

The proposed change that corrects the non-conservative Surveillance Requirement only increases a Technical Specifications parameter value in the conservative direction. Thus this change will not contribute to any increase in the probability or consequences of an accident which has been previously evaluated.

Second Standard

Will implementation of the changes proposed in this license amendment request create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The proposed changes would create no new accidents since no changes are being made that introduce any new accident casual mechanisms. The deterministic evaluation that supports this license amendment request consisted of a review of plant systems and safety functions impacted by entry into the expanded Completion Times, the performance of testing in previously prohibited operating modes, or increasing a Technical Specification mandated parameter in the conservative direction. The emergency diesel generators and electrical power and distribution systems were quantitatively and qualitatively assessed. It was determined that no new accidents or transients would be introduced by the proposed changes.

Third Standard

Will implementation of the changes proposed in this license amendment request involve a significant reduction in a margin of safety?

No. The impact of the proposed changes on the safety margins was considered in the deterministic evaluations that support this license amendment request. Extending the Completion Times, performing testing activities to confirm operability, or conservatively increasing a Technical Specification controlled parameter does not adversely impact any assumptions or inputs in the transient analyses contained in the McGuire Updated Final Safety Analysis Report (UFSAR). The proposed changes have no negative impact upon the ability of the fission product barriers (fuel cladding, the reactor coolant system, and the containment system) to perform their design functions during and following an accident situation. Additionally, the proposed changes have no adverse impact on setpoints or limits established or assumed within the UFSAR.

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: Ms. Lisa F. Vaughn, Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

NRC Section Chief: Evangelos C. Marinos.

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

Date of amendment request: May 24, 2005.

Description of amendment request: The proposed amendment would revise the steam generator (SG) tube inspection scope for Byron Station, Unit 2 for Refueling Outage 12 and the subsequent operating cycle. The proposed changes modify the inspection requirements for portions of SG tubes within the hot leg tubesheet region of the SGs.

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 previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The proposed changes that alter the SG inspection criteria do not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed changes will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

Of the applicable accidents previously evaluated, the limiting transients with consideration to the proposed changes to the SG tube inspection criteria, are the SG tube rupture (SGTR) event and the steam line break (SLB) accident.

During the SGTR event, the required structural integrity margins of the SG tubes will be maintained by the presence of the SG tubesheet. SG tubes are hydraulically expanded in the tubesheet area. Tube rupture in tubes with cracks in the tubesheet is precluded by the constraint provided by the tubesheet. This constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet and from the differential pressure between the primary and secondary side. Based on this design, the structural margins against burst, discussed in Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR [pressurized water reactor] SG Tubes,” are maintained for both normal and postulated accident conditions.

The proposed changes do not affect other systems, structures, components or operational features. Therefore, the proposed changes result in no significant increase in the probability of the occurrence of a SGTR accident.

At normal operating pressures, leakage from primary water stress corrosion cracking (PWSCC) below the proposed limited inspection depth is limited by both the tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint. Consequently, negligible normal operating leakage is expected from cracks within the tubesheet region. The consequences of an SGTR event are affected by the primary-to-secondary leakage flow during the event. Primary-to-secondary leakage flow through a postulated broken tube is not affected by the proposed change since the tubesheet enhances the tube integrity in the region of the hydraulic expansion by precluding tube deformation beyond its initial hydraulically expanded outside diameter.

The probability of a SLB is unaffected by the potential failure of a SG tube as this failure is not an initiator for a SLB.

The consequences of a SLB are also not significantly affected by the proposed changes. During a SLB accident, the reduction in pressure above the tubesheet on the shell side of the SG creates an axially uniformly distributed load on the tubesheet due to the reactor coolant system pressure on the underside of the tubesheet. The resulting bending action constrains the tubes in the tubesheet thereby restricting primary-to-secondary leakage below the midplane.

Primary-to-secondary leakage from tube degradation in the tubesheet area during the limiting accident (i.e., SLB) is limited by flow restrictions resulting from the crack and tube-to-tubesheet contact pressures that provide a restricted leakage path above the indications and also limit the degree of potential crack face opening as compared to free span indications. The primary-to-secondary leak rate during postulated SLB accident conditions would be expected to be less than that during normal operation for indications near the bottom of the tubesheet (i.e., including indications in the tube end welds). This conclusion is based on the observation that while the driving pressure causing leakage increases by approximately a factor of two, the flow resistance associated with an increase in the tube-to-tubesheet contact pressure, during a SLB, increases by up to approximately a factor of three. While such a leakage decrease is logically expected, the postulated accident leak rate could be conservatively bounded by twice the normal operating leak rate if the increase in contact pressure is ignored. Since normal operating leakage is limited to less than 0.104 gpm (150 gpd) per TS 3.4.13, “RCS Operational Leakage,” the associated accident condition leak rate, assuming all leakage to be from lower tubesheet indications, would be bounded by approximately 0.2 gpm. This value is well within the assumed accident leakage rate of 0.5 gpm discussed in Updated Final Safety Analysis Table 15.1-3, “Parameters Used in Steam Line Break Analyses.” Hence it is reasonable to omit any consideration of inspection of the tube, tube end weld, bulges/overexpansions or other anomalies below 17 inches from the top of the hot leg tubesheet. Therefore, the consequences of a SLB accident remain unaffected.

Based on the above discussion, the proposed changes do not involve an increase in the 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 changes do not involve the use or installation of new equipment and the currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed changes will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases.

Based on this evaluation, 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 significant reduction in a margin of safety?

Response: No.

The proposed changes maintain the required structural margins of the SG tubes for both normal and accident conditions. Nuclear Energy Institute (NEI) 97-06, “Steam Generator Program Guidelines,” Revision 1 and Regulatory Guide (RG) 1.121, “Bases for Plugging Degraded PWR Steam Generator Tubes,” are used as the bases in the development of the limited hot leg tubesheet inspection depth methodology for determining that SG tube integrity considerations are maintained within acceptable limits. RG 1.121 describes a method acceptable to the NRC for meeting General Design Criteria (GDC) 14, “Reactor coolant pressure boundary,” GDC 15, “Reactor coolant system design,” GDC 31, “Fracture prevention of reactor coolant pressure boundary,” and GDC 32, “Inspection of reactor coolant pressure boundary,” by reducing the probability and consequences of a SGTR. RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation the probability and consequences of a SGTR are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.

For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially oriented cracking, Westinghouse letter LTR-CDME-05-32-P, “Limited Inspection of the Steam Generator Tube Portion Within the Tubesheet at Byron Unit 2 and Braidwood Unit 2,” Revision 1, dated May 2005, defines a length of degradation free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot leg tubesheet inspection depth criteria will preclude unacceptable primary-to-secondary leakage during all plant conditions. The methodology for determining leakage provides for large margins between calculated and actual leakage values in the proposed limited hot leg tubesheet inspection depth criteria.

Therefore, the proposed changes do not involve a significant hazards consideration under the criteria set forth in 10 CFR 50.92(c).

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. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: May 11, 2005. The proposed amendment supercedes, in its entirety, a previous amendment request dated April 29, 2004, published in the Federal Register on May 25, 2004 (69 FR 29766).

Description of amendment request: The proposed amendment would revise technical specification (TS) 3/4.4.10, “Reactor Coolant System—Structural Integrity, ASME Code Class 1, 2, and 3 Components,” to allow a one-time extension of the surveillance interval for the reactor vessel internals vent valves from September 2005 to March 2006.

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

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

Response: No.

The proposed one-time surveillance interval exception does not alter the design, operation, or testing method of any structure, system, or component. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated. In addition, no accident initiators are affected and no previously analyzed accident scenario is changed. Initiating conditions and assumptions remain as previously analyzed. 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 one-time surveillance interval exception does not alter the design, operation, or testing method of any structure, system, or component. The proposed change does not introduce any new or different accident initiators. 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 proposed one-time surveillance interval exception does not affect the capabilities of the Reactor Vessel Internals Vent Valves. Therefore, the proposed change will not involve 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: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Gene Y. Suh.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: May 22, 2005.

Description of amendment request: The proposed amendment would adopt a qualified alternate repair criteria (ARC) for axial tube end cracking (TEC) indications in the Davis-Besse Nuclear Power Station, Unit 1 once-through steam generator tubes. Specifically, the proposed amendment would revise the technical specification surveillance requirements for steam generator tube inservice inspection to include the TEC ARC. The technical basis for the ARC is provided in Babcock & Wilcox Owners Group Topical Report BAW-2346P, “Alternate Repair Criteria for Tube End Cracking in the Tube-to-Tubesheet Roll Joint of Once-Through Steam Generators,” dated April 1999.

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

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

Response: No.

The proposed amendment does not increase the probability of any accident. Steam generator tube failure is an initiating condition for the steam generator tube rupture (SGTR) accident. The proposed TEC ARC does not affect the probability of an SGTR because the TEC ARC is limited to crack indications that are precluded from burst due to the presence of the tubesheet. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

The proposed amendment does not increase the consequences of any previously evaluated accident. Primary-to-secondary leakage affects the radiological consequences of accidents evaluated in the Updated Safety Analysis Report. The proposed amendment may result in an increase in post-accident primary-to-secondary leakage. Analyses have been performed to determine the expected post-accident leakage from each TEC left in service. The proposed amendment would impose inservice inspection and leakage assessment requirements that would ensure that the expected post-accident primary-to-secondary leakage through TECs and all other sources is maintained below the value assumed in the accident analyses. Therefore, the proposed change does not involve a significant increase in the 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 TEC ARC does not introduce any new failure modes or accident scenarios. Analyses have demonstrated that structural and leakage integrity is maintained for normal operating and accident conditions. Any failure of a tube from a TEC would be bounded by the SGTR analysis. 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 proposed amendment does not reduce the structural margin of the steam generator tubes. Structural integrity of the tube is maintained since the TEC ARC is limited to crack indications that are precluded from burst due to the presence of the tubesheet. The proposed amendment would impose inservice inspection and leakage assessment requirements that will ensure that the expected post-accident primary-to-secondary leakage through TECs and all other sources is maintained below the value assumed in the accident analyses. 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: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308

NRC Section Chief: Gene Y. Suh.

Florida Power and Light Company, Docket Nos. 50-335 and 50-389, St. Lucie Nuclear Plant, Units 1 and 2, St. Lucie County, Florida

Date of amendment request: April 21, 2005.

Description of amendment request: The submittal requests revision to several Technical Specifications (TSs) using seven TS Task Force (TSTF) generic changes. The seven TSTFs (nos. 5, 65, 101, 258, 299, 308, and 361) delete redundant safety limit violation notification requirements; adopt use of generic titles for utility positions; change the auxiliary feedwater pump test frequency to be consistent with the inservice test program frequency; remove redundant requirements and add other requirements to Section 5.0, Administrative Controls; clarify the meaning of “refueling cycle” for system integrated leak test intervals in the Primary Coolant Sources Outside Containment program; 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. In addition, the proposed amendments revise the TSs to adopt the Improved Standard Technical Specification (ISTS) requirements for remote shutdown instrumentation and the ISTS actions and action times for accident monitoring instrumentation.

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

The proposed changes revise administrative requirements, actions, action times, surveillance requirements, and surveillance frequencies. The revised requirements are not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased by the proposed changes. The Technical Specifications continue to require the systems, structures, and components associated with the revised requirements to be operable. Therefore, any mitigation functions assumed in the accident analyses will continue to be performed. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed amendments do not involve a significant increase in the probability or consequences of any accident previously evaluated.

(2) Operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes do not alter the design or physical configuration of the plant. No changes are being made to the plant that would introduce any new accident causal mechanisms. Therefore, operation of the facility in accordance with the proposed amendments do not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

The proposed changes do not change the design or function of plant equipment. The proposed changes do not significantly reduce the level of assurance that any associated plant equipment will be available to perform its function. The proposed changes provide operating flexibility without significantly affecting plant operation. Therefore, operation of the facility in accordance with the proposed amendments would 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Section Chief: Michael L. Marshall, Jr.

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

Date of amendment request: May 25, 2005.

Description of amendment request: The proposed amendment would delete from the Cooper Nuclear Station (CNS) Technical Specifications (TSs) temporary notes that have expired and are no longer in effect.

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.

Deleting temporary notes that have expired from the CNS TS does not impact the plant design or how the plant is operated, nor does it affect any of the conditions that could cause an accident. Thus, this change does not result in a significant increase in the probability of an accident previously evaluated. Removing the expired temporary notes does not reduce the requirements for maintaining systems needed to mitigate postulated accidents as described in the CNS Updated Safety Analysis Report. Thus, this change does not result in a significant increase in the consequences of an accident previously evaluated. 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.

Deleting temporary notes that have expired does not involve a change to the plant design or to how the plant is operated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

Deleting temporary notes that have expired does not result in a relaxation of any limit associated with the performance of systems required to mitigate postulated accidents, nor does it reduce any of the requirements for maintaining those systems. 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: Mr. John C. McClure, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: David Terao.

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

Date of amendment request: September 30, 2004, as supplemented on May 28, 2005.

Description of amendment request: The proposed amendment would revise the information in the Updated Final Safety Analysis Report regarding the application of leak-before-break methodology to the accumulator A and B lines and the pressurizer surge line. The application of leak-before-break methodology would permit the exclusion of these lines from the evaluation of dynamic effects associated with postulated high energy line breaks.

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 previous[ly] evaluated?

Response: No.

The proposed changes use an approved fracture mechanics methodology, in accordance with 10 CFR [Part] 50, Appendix A, GDC [General Design Criterion] -4 to demonstrate that the probability of fluid system rupture for these lines attached to the Reactor Coolant System is extremely low under conditions associated with the design basis for the piping.

The proposed changes do not adversely affect accident initiators or precursors nor significantly alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely alter or prevent the ability of structures, systems, and components (SSCs) 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 an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupation/public radiation exposures. The proposed changes do not affect the probability of an accident occurring since they reflect a change in plant design basis that is consistent with current Regulations. The proposed changes cannot increase the consequences of postulated accidents since LOCA [loss-of-coolant accident] and methods containment analysis will not be changed. Therefore, the 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 does not create the possibility of a new or different kind of accident, since it simply provides an analytical justification for demonstrating that the probability of a fluid system rupture is extremely small. Leak-before-break justifications per GDC-4 still require that ECCS [emergency core cooling system], containment, and EQ [environmental qualification] requirements be maintained consistent with the original postulated accident assumptions—only protection from dynamic effects is modified.

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

Response: No.

The proposed changes apply very conservative approved analytical methods to demonstrate that the probability of a fluid system rupture is very low. This analysis justifies differences in protection from dynamic effect [and] is associated with these extremely low probability ruptures. For overall ECCS, containment, and EQ requirements, there will be no changes to the licensing basis.

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: Daniel F. Stenger, Ballard Spahr Andrews & Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 20005.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: March 8, 2005.

Description of amendment request: The amendments proposed by Southern Nuclear Operating Company (SNC) would revise the Technical Specifications (TS) to delete Function 11, Reactor Coolant Pump (RCP) Breaker Position, in TS 3.3.1, “Reactor Trip System (RTS) Instrumentation.”

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?

No. The proposed changes do not significantly increase the probability or consequences of an accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR). All of the safety analyses have been evaluated for impact. The elimination of RCP Breaker Position reactor trip will not initiate any accident; therefore, the probability of an accident has not been increased. An evaluation of dose consequences, with respect to the proposed changes, indicates there is no impact due to the proposed changes and all acceptance criteria continue to be met. Therefore, these 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 previously evaluated?

No. The proposed changes do not create the possibility of a new or different kind of accident than any accident already evaluated in the UFSAR. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed changes. The changes have no adverse effects on any safety-related system. Therefore, all accident analyses criteria continue to be met and these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

No. The proposed changes do not involve a significant reduction in a margin of safety. All analyses that credit the RCS Low Flow reactor trip function have been reviewed and no changes to any inputs are required. The evaluation demonstrated that all applicable acceptance criteria are met. 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: M. Stanford Blanton, Esq., Balch and Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201.

NRC Section Chief: Evangelos C. Marinos.

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

Date of amendment request: June 2, 2005.

Description of amendment request: The proposed amendment would change Technical Specification (TS) 3.4.6.1, “Reactor Coolant System Leakage Detection Systems,” to specifically require only one containment radioactivity monitor (particulate channel) to be operable in Modes 1, 2, 3 and 4. Additionally, corresponding changes to the Surveillance Requirement (SR) 4.4.6.1 and 4.4.6.2.1, “Reactor Coolant System Operational Leakage,” are also requested.

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

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

Response: No.

The proposed change has been evaluated and determined to not increase the probability or consequences of an accident previously evaluated. The proposed change does not make any hardware changes and does not alter the configuration of any plant system, structure or component (SSC). The proposed change only removes the containment atmosphere gaseous radioactivity monitor as an option for meeting the operability requirement for TS 3.4.6.1, and correspondingly from the requirements of SR 4.4.6.1 and 4.4.6.2.1.a. Therefore, the probability of occurrence of an accident is not increased. The TS will continue to require diverse means of leakage detection equipment, thus ensuring that leakage due to cracks would continue to be identified prior to breakage and the plant shutdown accordingly. Additionally, the proposed change is not modeled in the South Texas Project probabilistic risk assessment and has no impact on core damage frequency or large early release frequency. Therefore, the consequences of an accident are not increased.

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

Response: No.

The proposed change does not involve the use or installation of new equipment and the currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed changes will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases. The proposed change does not affect any SSC associated with an accident initiator. Based on this evaluation, 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 change does not make any alteration to any RCS leakage detection components. The proposed change only removes the gaseous channel of the containment atmosphere radioactivity monitor as an option for meeting the operability requirement for TS 3.4.6.1, and correspondingly from the requirements of SR 4.4.6.1 and 4.4.6.2.1.a. The proposed amendment continues to require diverse means of leakage detection equipment with capability to promptly detect RCS leakage. Although not required by TS, additional diverse means of leakage detection capability are available. Based on this evaluation, the proposed change does not involve a significant reduction in a margin of safety.

Based upon the NRC staff's review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.

NRC Section Chief: David Terao.

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

Date of amendment request: April 27, 2005.

Description of amendment request: The proposed amendment would revise the applicability for Items 18.A and 18.B of Technical Specification (TS) Table 3.3-1, “Reactor Trip System Instrumentation,” and TS Table 4.3-1, “Reactor Trip System Instrumentation Surveillance Requirements.” This change will add a footnote that indicates that the Mode 1 applicability is limited to operation above the P-9 (50-percent rated thermal power) value. Additionally, the action for an inoperable turbine stop valve closure channel is being revised to be consistent with the design of this function. Finally, an option consistent with the latest standard TSs (NUREG-1431, Revision 3) is added to permit a reduction in thermal power to below the P-9 interlock within 10 hours for an inoperable turbine stop valve closure channel.

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

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

Response: No.

The proposed changes revise the applicability and actions for inoperable reactor trip functions from a turbine trip event. These changes do not alter these functions physically or how they are maintained. By clarifying the proper applicability and enhancing the actions for these functions the availability of these trips and compensatory measures for inoperable conditions are improved. The availability change implements the required conditions for turbine trip operability that are consistent with their ability to perform the reactor trip functions. The action changes correct inappropriate requirements for minimum channels to be operable and the allowance to bypass channels in consideration of the logic design for the turbine stop valve closure channels. The change to allow power reduction as an alternative to tripping an inoperable channel for the turbine stop valve closure channels, provides a more conservative response than currently allowed.

Since these changes will not affect the ability of these trips to perform the initiation of reactor trips when appropriate, the offsite dose consequences for an accident will not be impacted. Equally, the potential to cause an accident is not affected because no plant system or component has been altered by the proposed changes. 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 changes only affect the applicability and action requirements for the turbine trip functions. This does not affect any physical features of the plant or the manner in which these functions are utilized. The proposed applicability will require the functions to be operable when they are able to perform their trip functions. The actions will handle inoperable channels such that their safety function will be satisfied or the unit will be placed in a condition that does not require these trip functions. 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 proposed changes do not alter any plant setpoints or functions that are assumed to actuate in the event of postulated accidents. In fact, the proposed changes do not alter any plant feature and only alters the requirements for when the function must be operable and the actions to take should a channel become inoperable during these conditions. The proposed changes ensure the functionality of the turbine trips when assumed in the analysis and provides actions for inoperable channels that preserve the safety functions for accident mitigation. 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 Section 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: April 27, 2005.

Description of amendment request: The proposed amendment would relocate a number of technical specification (TS) requirements to the Technical Requirements Manual (TRM).

The proposed amendment would relocate the provisions for TS 3.1.3.4 (Rod Drop Time), TS 3.3.2 (Movable Incore Detectors), TS 3.3.3.4 (Meteorological Instrumentation), TS 3.4.7 (Reactor Coolant System Chemistry), TS 3.4.11 (Reactor Coolant System Head Vents), TS 3.7.2 (Steam Generator Pressure and Temperature Limitations), TS 3.7.10 (Sealed Source Contamination), TS 3.9.5 (Refueling Operations Communications), and TS 3.9.6 (Manipulator Crane).

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

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

Response: No.

The proposed change only relocates requirements to TRM that are not required to be included in the TSs in accordance with 10 CFR 50.36. Changes to the TRM require evaluations and reviews in accordance with 10 CFR 50.59 to ensure that the health and safety of the public is not adversely affected. The proposed relocation retains the current TS requirements and only alters the location of these provisions. This relocation cannot affect the probability or consequences of an accident as this is only an administrative revision that will not alter any plant equipment or processes. 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.

Since the proposed change only relocates the current TS requirements without change, there is not a potential for a change in the accident generation potential. This change will not alter plant components, systems, or operating practices. 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 proposed change relocates specifications that do not meet the threshold for inclusion in the TSs as defined in 10 CFR 50.36. This change will not alter the requirements for these functions or plant setpoints or functions that maintain the margins of safety. 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 Section Chief: Michael L. Marshall, Jr.

Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consderation 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.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Brunswick County, North Carolina

Date of amendment request: May 17, 2005.

Brief description of amendment request: The amendments replace the existing requirement of Technical Specification 3.4.5, “RCS [Reactor Coolant System] Leakage Detection Instrumentation,” Required Action D.1, to enter Limiting Condition for Operation (LCO) 3.0.3 if required leakage detection systems are inoperable with the requirement to be in Mode 3 within 12 hours and Mode 4 within 36 hours.

Date of publication of individual notice in Federal Register: June 13, 2005 (70 FR 34161).

Expiration date of individual notice: June 27, 2005 (for comments); August 12, 2005 (for hearing requests).

Notice of Issuance of Amendments to Facility Operating Licenses

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

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

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

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

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

Date of application for amendment: February 24, 2005.

Brief description of amendment: The amendment revised the Technical Specifications, Section 3.1.1, “Protective Instrumentation Requirements,” notes aa and bb, correcting missed wording which led to incorrect statements of the as-designed service water pump and reactor building closed cooling water system pump trip conditions. The amendment also made an editorial correction to pages 3.6-1 and 3.6-2.

Date of Issuance: June 23, 2005.

Effective date: June 23, 2005 and shall be implemented within 60 days of issuance.

Amendment No.: 255.

Facility Operating License No. DPR-16: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 29, 2005 (70 FR 15941). The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated June 23, 2005.

No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

Date of application for amendment: October 21, 2004, as supplemented January 4, 2005.

Brief description of amendment: The amendment deletes the Technical Specification (TS) requirements to submit monthly operating reports and annual occupational radiation exposure reports. The change is consistent with Revision 1 of the Nuclear Regulatory Commission (NRC) approved Industry/Technical Specifications Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-369, “Removal of Monthly Operating Report and Occupational Radiation Exposure Report.” This TS improvement was published in the Federal Register on June 23, 2004 (69 FR 35067), as part of the Consolidated Line Item Improvement Process.

Date of issuance: June 17, 2005.

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

Amendment No.: 254.

Facility Operating License No. DPR-50. Amendment revised the TSs.

Date of initial notice in Federal Register: April 12, 2005 (70 FR 19114). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 17, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 13, 2004, as supplemented on April 21, 2005.

Brief description of amendments: The amendments revised License Condition 2.E of each unit's operating license by replacing the current wording with wording from Generic Letter (GL) 86-10, “Implementation of Fire Protection Requirements.”

Date of issuance: June 15, 2005.

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

Amendment Nos.: 273 and 250.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the operating licenses.

Date of initial notice in Federal Register: December 7, 2004 (69 FR 70715). The supplement dated April 21, 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. The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated June 15, 2005.

No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of application for amendments: January 27, 2005.

Brief Description of amendments: The amendments revised respective Technical Specifications (TS) testing frequency for the surveillance requirement (SR) in TS 3.1.4, “Control Rod Scram Times.” The change revises the test frequency of SR 3.1.4.2, control rod scram time testing, from “120 days cumulative operation in MODE 1” to “200 days cumulative operation in MODE 1.”

Date of issuance: May 31, 2005.

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

Amendment Nos.: 236 and 264.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: February 15, 2005 (70 FR 12745). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 31, 2005.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-423, Millstone Power Station, Unit Nos. 1, 2, and 3, New London County, Connecticut

Date of application for amendments: December 21, 2004.

Brief description of amendments: The amendments eliminate requirements for annual Occupational Radiation Exposure Reports, annual reports regarding challenges to pressurizer relief and safety valves, and Monthly Operating Reports.

Date of issuance: June 13, 2005.

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

Amendment Nos.: 114, 286, and 223.

Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: The Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 12, 2005 (70 FR 19114). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 13, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 22, 2004, as supplemented by letters dated February 8 and April 7, 2005.

Brief description of amendments: The amendments revised the Technical Specifications (TS) 3.3.2, “Engineered Safety Features Actuation System Instrumentation,” and TS 3.3.6, “Containment Air Release and Addition Isolation Instrumentation,” to permit an 18-month surveillance interval for certain Westinghouse Type AR slave relays and for certain Potter and Brumfield MDR-Series slave relays.

Date of issuance: May 24, 2005.

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

Amendment Nos.: 224 and 219.

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

Date of initial notice in Federal Register: September 14, 2004 (69 FR 55468). The supplements dated February 8 and April 7, 2005, provided additional information that clarified the application, did not expand the scope of the March 22, 2004, 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 24, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 10, 2004, as supplemented by letter dated January 31, 2005.

Brief description of amendments: The amendments revised the Technical Specifications to extend the interval between local leakage rate tests of the containment purge and vent valves with resilient seals (that is, in the containment purge system, hydrogen purge system, and containment air release and addition system).

Date of issuance: June 10, 2005.

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

Amendment Nos.: 225 and 222.

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

Date of initial notice in Federal Register: December 21, 2004 (69 FR 76487).

The supplement dated January 31, 2005, provided additional information that clarified the application, did not expand the scope of the June 10, 2004, 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 June 10, 2005.

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: May 12, 2004, as completely superseded by application dated July 8, 2004, and supplemented by letters dated October 14, 2004, and January 19, March 7, and April 7, 2005.

Brief description of amendment: The Index is deleted from the Technical Specifications.

Date of issuance: June 22, 2005.

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

Amendment No.: 260.

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

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53106). The supplements dated October 14, 2004, and January 19, March 7, and April 7, 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 June 22, 2005.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of application for amendments: October 21, 2004.

Brief description of amendments: The amendment deletes the Technical Specification (TS) requirements to submit monthly operating reports and annual occupational radiation exposure reports. The change is consistent with Revision 1 of NRC-approved Technical Specifications Task Force (TSTF) 369, “Elimination of Requirements for Monthly Operating Reports and Occupational Radiation Exposure Reports.” This TS improvement was published in the Federal Register (69 FR 35067) on June 23, 2004, as part of the Consolidated Line Item Improvement Process.

Date of issuance: June 14, 2005.

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

Amendments Nos.: 254 and 257.

Renewed Facility Operating License Nos. DPR-44 and DPR-56: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 12, 2004 (70 FR 19116). The Commission's related evaluation of the amendments are contained in a Safety Evaluation dated June 14, 2005.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: September 15, 2004.

Brief description of amendments: The amendments deleted the technical specification (TS) requirements related to hydrogen and oxygen monitors. The TS changes support implementation of the revisions to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, “Combustible Gas Control for Nuclear Power Reactors,” that became effective on October 16, 2003. The changes are consistent with Revision 1 of the NRC-approved Industry/Technical Specifications Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-447, “Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen Monitors.”

Date of issuance: June 14, 2005.

Effective date: June 14, 2005.

Amendment Nos.: 226/221.

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

Date of initial notice in Federal Register: February 1, 2005 (70 FR 5243).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 14, 2005.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: August 2, 2004.

Brief description of amendment: This amendment deleted Technical Specification 6.8.4.c, “Post-Accident Sampling,” and the related requirements to maintain a Post-Accident Sampling System.

Date of issuance: June 10, 2005.

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

Amendment No.: 264.

Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.

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

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

No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: October 22, 2004, as supplemented by letter dated December 16, 2004.

Description of amendment request: The amendment revised the Seabrook Station, Unit No. 1 Technical Specifications (TSs) to allow for individual entry into the limiting condition for operation (LCO) for each instrument, and extends the allowed outage times for LCOs 3.3.3.6.a and 3.3.3.6.b.

Date of issuance: June 15, 2005.

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

Amendment No.: 103.

Facility Operating License No. NPF-86: The amendment revised the TSs.

Date of initial notice in Federal Register: November 2, 2004 (69 FR 63560). The December 16, 2004 supplement provided clarifying information that did not change the scope of the proposed amendment as described in the original notice of proposed action published in the Federal Register, and 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 June 15, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: May 5, 2005, as supplemented June 9, 2005.

Brief description of amendment: The amendment revises the Facility Operating License and Technical Specifications to modify the auxiliary feed water (AFW) pump suction protection requirements and change the design basis as described in the Updated Safety Analysis Report to revise the functionality of the discharge pressure switches to provide pump runout protection, which requires operator actions to restore the AFW pumps for specific post-accident recovery activities.

Date of issuance: June 20, 2005.

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

Amendment No.: 183.

Facility Operating License No. DPR-43: Amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: May 13, 2005 (70 FR 25619). The supplement dated June 9, 2005, provided clarifying information that did not change the scope of the May 5, 2005 application, nor the initial 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 June 20, 2005.

No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: September 30, 2004.

Brief description of amendments: The amendments revise Technical Specifications related to the reactor coolant pump flywheel inspection program by increasing the inspection interval from current 10 years to 20 years.

Date of issuance: June 10, 2005.

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

Amendment Nos.: 118/118.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 1, 2005 (70 FR 9998).

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

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: September 12, 2003, as supplemented by letters dated November 20, 2003, March 30, April 20, May 7, May 27, August 18, and November 3, 2004, and February 17, 2005.

Brief description of amendment: These amendments revise the Technical Specifications to incorporate a full-scope application of an alternate source term methodology in accordance with Title 10 of the Code of Federal Regulations, Section 50.67.

Date of issuance: June 15, 2005.

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

Amendment Nos.: 240 and 221.

Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: December 9, 2003 (68 FR 68672). The supplements dated November 20, 2003, March 30, April 20, May 7, May 27, August 18, and November 3, 2004, and February 17, 2005, contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 15, 2005.

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: August 30, 2004.

Brief description of amendment: These amendments revise the Technical Specifications by extending the inspection interval for reactor coolant pump flywheels to 20 years.

Date of issuance: June 15, 2005.

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

Amendment Nos.: 241 and 222.

Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: March 15, 2005 (70 FR 12751).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 15, 2005.

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: August 30, 2004.

Brief Description of amendments: These amendments revise the Technical Specifications by extending the inspection interval for reactor coolant pump flywheels to 20 years.

Date of issuance: June 21, 2005.

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

Amendment Nos.: 242 and 241.

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

Date of initial notice in Federal Register: March 15, 2005 (70 FR 12751).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 21, 2005.

No significant hazards consideration comments received: No.

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

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

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

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

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

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

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

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

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

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1 [phone], [phone], or by e-mail to [e-mail]. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact.^[1] Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

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

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

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

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

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

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

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

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

Date of amendment request: June 16, 2005, as supplemented June 19, 2005.

Description of amendment request: The amendment revises the Technical Specifications to remove the requirement to have an operable containment spray flow path capable of taking suction from the containment sump.

Date of issuance: June 21, 2005.

Effective date: June 21, 2005.

Amendment No.: 184.

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

Public comments requested as to proposed no significant hazards consideration (NSHC): No. The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated June 21, 2005.

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

NRC Section Chief: L. Raghavan.

Dated in Rockville, Maryland, this 27th day of June 2005.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

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

Footnotes

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