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Notice

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

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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 April 13, 2007 to April 26, 2007. The last biweekly notice was published on April 24, 2007 (72 FR 20375).

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

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

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rulemaking, Directives and Editing Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The Start Printed Page 26174petition 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, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. 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 (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely 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 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: November 27, 2006.

Description of amendment request: The amendment would revise the Oyster Creek Technical Specification (TS) 6.9.1.d, “Annual Radioactive Effluent Release Report,” by changing the requirement to submit the report within 60 days of January 1. Specifically, the revised requirement would be to submit the report prior to May 1 of each year.

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

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

Response: No.

The proposed change involves a revision to the required submittal date for the Radioactive Effluent Release Report, and is administrative in nature. The change will not alter the physical design or operation of any plant structure, system, or component.

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

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

Response: No.

The proposed change is administrative in nature. The proposed change has no impact on the design, function or operation of any plant structure, system or component and does not affect any accident analyses. Accordingly, the change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure, system, or component to perform their safety function.

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

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

Response: No.

The proposed change is administrative in nature, does not negate any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there is no change being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant 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: Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief: Harold K. Chernoff. Start Printed Page 26175

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

Date of amendment request: March 28, 2007.

Description of amendment request: The proposed change would revise the required wattage specified in the River Bend Station, Unit 1 (RBS), Technical Specification 5.5.7.e, Ventilation Filter Testing Program, for the Control Room Fresh Air System (CRFAS) heater for testing. The proposed required wattage for testing the CRFAS heater would be revised from 23 ± 2.3 kilowatt (kW), to “≥ 15 kW.”

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

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

Response: No.

This change specifies the required power (in kW) for the Control Room ventilation electric heaters to decrease relative humidity of the air to less than 70% relative humidity as required for proper operation of the charcoal absorber components based on calculated requirements. The heater will continue to perform its intended design function as designed. The heater is not an accident precursor.

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 heater will continue to perform its function as designed. The heater provides humidity control for the Control Room filter unit during a design basis accident. Changing the test acceptance criteria to a calculated value has no influence on, nor does it contribute in any way to, the possibility of a new or different kind of accident or malfunction from those previously analyzed. No change has been made to the design, function or method of performing testing. No safety-related equipment or safety functions are altered as a result of this change.

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.

No margin of safety is changed as a result of this change. The heater will continue to perform its design function. Testing methodology has not changed. The function of the heater is unchanged. The acceptance criterion has been changed to a calculated value rather than the name plate rating to make testing more realistic. The heater will continue to operate to perform its intended design function.

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

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: March 30, 2007.

Description of amendment request: The proposed amendment would revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification (TS) to support a partial re-rack of the storage racks in the ANO-2 spent fuel pool (SFP). The proposed amendment would revise TS 3.9.12, “Fuel Storage,” and its associated tables, figures, and surveillance requirements, TS 5.3, “Fuel Storage,” and add TS 6.5.17, “Metamic Coupon Sampling Program.” The ANO-2 TS 3.9.12 would be changed to: (1) Support higher fuel assembly U-235 enrichment; (2) apply the appropriate loading restrictions; and (3) delete the dry cask loading restrictions. ANO-2 TS 5.3.1b would be changed to reflect a different SFP boron concentration that is needed to assure K-effective (Keff) remains less than or equal to 0.95. ANO-2 TS 5.3.2a would be modified to reflect a higher fuel assembly U-235 enrichment. A new coupon sampling program would be added as TS 6.5.17. In addition, TS Surveillance Requirement 4.9.12.d would be added to direct performance of the coupon sampling program.

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.

Fuel Handling Accidents

The current licensing bases for the dose consequences associated with a fuel handling accident (FHA), which was performed considering a maximum U-235 enrichment of 5.0 wt% and a maximum burnup of 65 megawatt-days/kilograms of uranium, does not exceed 25% of 10 CFR [Title 10 of the Code of Federal Regulations] 100 limits. The proposed change is bounded by the current analysis and therefore, there is no increase in the dose consequences associated with a[n] FHA.

During rack removal and installation, safe load paths will be determined and written procedures followed to ensure that the racks are not carried over any fuel assemblies. With the proposed limitations on rack and cask movement, there should be no impact to spent fuel and no radiological consequences due to fuel rack installation. The racks will be moved with a single failure proof crane. Therefore, a postulated drop of a rack is not a credible accident.

The probability of having a[n] FHA has not increased.

Criticality Accidents Associated With a Dropped Fuel Assembly

The three fuel assembly drop accidents described below can be postulated to increase reactivity. However, for these accident conditions, the double contingency principle of ANS [American National Standard] N16.1-1975 is applied. This states that it is unnecessary to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since its absence would be a second unlikely event.

Three types of drop accidents have been considered: a vertical drop accident, a horizontal drop accident, and an inadvertent drop of an assembly between the outside periphery of the rack and the pool wall. The structural damage to the pool liner, the racks, and fuel assembly resulting from a dropped fuel assembly striking the rack, the pool floor, or another assembly located in the racks is primarily dependent on the mass of the falling object, drop height, and structural configuration of the rack. The two parameters related to the fuel assembly (mass and drop height) are not changed by the proposed rack modification. The new rack design was evaluated for all postulated structural drops and the structural damage to these items remains within acceptable limits. In all cases the proposed TS limit for boron concentration ensures that a five percent subcriticality margin is met for the postulated accidents.

Criticality Accidents Associated With a Misplaced Fuel Assembly

The fuel assembly misplacement accident was considered for all storage configurations. An assembly with high reactivity is assumed to be placed in a storage location which requires restricted storage based on initial U-235 loading, cooling time, and burnup. The presence of boron in the pool water assumed in the analysis has been shown to offset the worst case reactivity effect of a misplaced Start Printed Page 26176fuel assembly for any configuration. This boron requirement is less than the boron concentration required by the ANO-2 TS. Thus, a five percent subcriticality margin is met for postulated accidents, since any reactivity increase will be much less than the negative worth of the dissolved boron.

Optimum Moderation Accident

For fuel storage applications in the SFP, water is usually present. An “optimum moderation” accident is not a concern in SFP storage racks because the rack design prevents the preferential reduction of water density between the cells of a rack (e.g., boiling between cells). In addition, the criticality analysis has demonstrated that the effective neutron multiplication factor (Keff) will remain less than 1.0 when the SFP is fully flooded with unborated water.

An “optimum moderation” accident in the new fuel vault was evaluated and the conclusions of that evaluation confirmed that the reactivity effect is less than the regulatory limit of 0.98 for Keff.

Loss of SFP Cooling

The proposed modification to the ANO-2 SFP racks does not result in a change to the SFP cooling system and therefore the probability of a loss of SFP cooling is not increased.

The consequences of a loss of spent fuel pool cooling were evaluated and found to not involve a significant increase as a result of the proposed changes. A thermal-hydraulic evaluation for the loss of SFP cooling was performed. The analysis determined that the minimum time to boil is about two hours following a complete core off load and a complete loss of forced cooling. This provides sufficient time for the operators to restore cooling or establish an alternate means of cooling before the water shielding above the top of the racks falls below 10 feet. Therefore, the proposed change represents no increase in the consequences of loss of pool cooling.

Seismic Event

The proposed rack modification does not result in an increase in the probability or consequences of a design basis seismic event. The new racks were analyzed and all structural acceptance criteria are shown to be met during seismic events. The structural capability of the SFP and liner will not be exceeded as a result of the new rack design.

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 presence of soluble boron in the pool water assumed in the criticality analysis is less than the boron concentration required by the ANO-2 TSs. Thus, a five percent subcriticality margin is met for postulated accidents, since any reactivity increase will be much less than the negative worth of the dissolved boron.

No new or different types of fuel assembly drop scenarios are created by the proposed change. During the installation of the new racks, the possibility of dropping a rack is not a credible accident since a single failure proof crane and safe load paths will be used for rack movements. No new or different fuel assembly misplacement accidents will be created. Administrative controls currently exist to assist in assuring fuel misplacement does not occur.

No changes are proposed to the spent fuel pool cooling system or makeup systems and therefore no new accidents are considered related to the loss of cooling or makeup capability.

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.

With the presence of a nominal boron concentration, the SFP storage racks will be designed to assure a subcritical array with a five percent subcritical margin (95% probability at the 95 % confidence level). This has been verified by criticality analyses.

Credit for soluble boron in the SFP water is permitted under accident conditions. The proposed modification that will allow installation of the new racks does not result in the potential of any new misplacement scenarios. Criticality analyses have been performed to determine the required boron concentration that would ensure the maximum Keff does not exceed 0.95. The ANO-2 TS for the minimum SFP boron concentration is greater than that required to ensure Keff remains below 0.95. Therefore, the margin of safety defined by taking credit for soluble boron will be maintained.

The structural analysis of the new spent fuel racks along with the evaluation of the SFP structure indicated that the integrity of these structures will be maintained. The structural requirements were shown to be satisfied, thus the safety margins were maintained.

In addition the proposed change includes a coupon sampling program that will monitor the physical properties of the MetamicTM absorber material. The monitoring program provides a method of verifying that the assumptions used in the SFP criticality analyses remain valid.

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

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: March 1, 2007.

Description of amendment request: The proposed change would revise Grand Gulf Nuclear Station, Unit 1, Technical Specification (TS) Tables 3.3.5.1-1 and 3.3.5.2-1 to modify the allowable values of the low Condensate Storage Tank (CST) level setpoints for the High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) suction swap from the CST to the Suppression Pool. The change is necessary to correct an error in the original plant design. The error, under certain conditions, could prevent a swap of the HPCS and RCIC suction flow paths to the Suppression Pool. Currently, the erroneous setpoints have been corrected to a higher level, and are administratively controlled in accordance with the Administrative Letter 98-10, “Dispositioning of Technical Specifications That Are Insufficient To Assure Plant Safety.”

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

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

Response: No.

This change will adjust the setpoint for an automatic swap of the suction for the HPCS and RClC systems from the Condensate Storage Tank (CST) to the Suppression Pool. The Suppression Pool is the source of water credited in the accident analyses. This transfer is not the initiator of any analyzed accident. The setpoint adjustment will allow a transfer of the suction to an assured safety-related water source earlier in the event and will have no effect on the probability or consequences of an accident previously evaluated.

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

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

Response: No.

Transfer of the suction source for HPCS and RClC will occur sooner as a result of this change. No new operational conditions beyond those currently allowed are introduced. This change is consistent with the safety analyses assumptions and current Start Printed Page 26177plant operating practices. This simply corrects the setpoint consistent with the accident analyses and therefore cannot create the possibility of a new or different kind of accident from any previously evaluated accident.

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 does not reduce safety, but rather allows the transfer from the CST to the Suppression Pool sooner. The Suppression Pool is the source of water credited in the accident analyses. This change is consistent with the safety analyses assumptions and current plant operating practices. No new operational conditions beyond those currently allowed are created by these changes.

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

NRC Branch Chief: Thomas G. Hiltz.

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

Date of amendment request: October 18, 2006.

Description of amendment request: A change is proposed to the technical specifications (TSs) of LaSalle County Station, Units 1 and 2 (LaSalle), consistent with TS Task Force Traveler No. 432 (TSTF-423), “Technical Specification End States, NEDC-32988-A,” to the standard TSs for boiling-water reactor plants, to allow for some systems entry into hot shutdown rather than cold shutdown, to repair equipment if risk is assessed and managed consistent with the program in place for complying with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.65(a)(4). The proposed amendment would modify the TS to risk-informed requirements regarding selected required action end states provided in TSTF-423, Revision 0.

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

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

The proposed change allows a change to certain required end states when the TS Completion Times for remaining in power operation will be exceeded. Most of the requested technical specification (TS) changes are to permit an end state of hot shutdown (Mode 3) rather than an end state of cold shutdown (Mode 4) contained in the current TS. The request was limited to: (1) Those end states where entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable technical specification, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical. Risk insights from both the qualitative and quantitative risk assessments were used in specific TS assessments. Such assessments are documented in Section 6 of GE NEDC-32988, Revision 2, “Technical Justification to Support Risk Informed Modification to Selected Required Action End States for BWR Plants.” They [risk assessments] provide an integrated discussion of deterministic and probabilistic issues, focusing on specific technical specifications, which are used to support the proposed TS end state and associated restrictions. The [NRC] staff finds that the risk insights support the conclusions of the specific TS assessments. Therefore, the probability of an accident previously evaluated is not significantly increased, if at all. The consequences of an accident after adopting proposed TSTF-423, are no different than the consequences of an accident prior to adopting TSTF-423. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). If risk is assessed and managed, allowing a change to certain required end states when the TS Completion Times for remaining in power operation are exceeded, i.e., entry into hot shutdown rather than cold shutdown to repair equipment, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change and the commitment by the licensee to adhere to the guidance in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0, “Technical Specifications End States, NEDC-32988-A,” will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

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

The proposed change allows, for some systems, entry into hot shutdown rather than cold shutdown to repair equipment, if risk is assessed and managed. The [Boiling Water Reactor Owners Group] BWROG's risk assessment approach is comprehensive and follows [NRC] staff guidance as documented in [Regulatory Guides] RGs 1.174 and 1.177. In addition, the analyses shows that the criteria of the three-tiered approach for allowing TS changes are met. The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A risk assessment was performed to justify the proposed TS changes. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

LaSalle has reviewed the proposed no significant hazards consideration determination published on March 23, 2006, (71 FR 14743) as part of the consolidated line item improvement and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration.

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

NRC Branch Chief: Russell Gibbs.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of amendment request: February 28, 2007.

Description of amendment request: The proposed amendments would revise Technical Specification (TS) 5.2.2, “Plant Staff”, and TS 5.3, “Plant Staff Qualifications”, requirements for shift technical advisor (STA) qualifications. The proposed changes will specify that personnel who perform the function of STA shall meet the qualification requirements of the Commission Policy Statement on Engineering Expertise on Shift, published in the Federal Register on October 28, 1985 (50 FR 43621). This change will allow qualified personnel to perform the function of STA without Start Printed Page 26178also holding a senior reactor operator (SRO) license.

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.

This license amendment request proposes to add a new sentence to Technical Specification 5.2.2 specifying that personnel who perform the function of shift technical advisor shall meet the qualification requirements of the Commission Policy Statement on Engineering Expertise on Shift and remove shift technical advisor qualification requirements from Technical Specification 5.3.1. This change will allow qualified personnel to perform the function of shift technical advisor without also holding a senior reactor operator license.

The proposed changes are administrative changes to Technical Specifications Chapter 5, the administrative chapter of the Technical Specifications. Shift technical advisors perform the function of on-shift technical advisor to the shift supervisor and do not operate the plant. Therefore, the changes proposed in this license amendment request 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.

This license amendment request proposes to add a new sentence to Technical Specification 5.2.2 specifying that personnel who perform the function of shift technical advisor shall meet the qualification requirements of the Commission Policy Statement on Engineering Expertise on Shift and remove shift technical advisor qualification requirements from Technical Specification 5.3.1. This change will allow qualified personnel to perform the function of shift technical advisor without also holding a senior reactor operator license.

The Technical Specification changes proposed in this license amendment are administrative, do not change the manner in which the plant is operated, and 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.

This license amendment request proposes to add a new sentence to Technical Specification 5.2.2 specifying that personnel who perform the function of shift technical advisor shall meet the qualification requirements of the Commission Policy Statement on Engineering Expertise on Shift and remove shift technical advisor qualification requirements from Technical Specification 5.3.1. This change will allow qualified personnel to perform the function of shift technical advisor without also holding a senior reactor operator license.

The proposed changes are administrative changes to Technical Specifications Chapter 5, the administrative chapter of the Technical Specifications. Shift technical advisors perform the function of on-shift technical advisor to the shift supervisor and do not operate the plant. Thus, the Technical Specification changes proposed in this license amendment request 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 requests involve no significant hazards consideration.

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

NRC Branch Chief: L. Raghavan.

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

Date of amendment request: January 18, 2007.

Brief description of amendments: The amendments requested would revise Technical Specifications (TS) requirement 3.8.1, “AC Sources—Operating,” Extension of Completion Times for Offsite Circuits.

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

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

Response: No.

The proposed Technical Specification (TS) Completion Time (CT) extension does not significantly increase the probability of occurrence of a previously evaluated accident because the startup transformers (STs) are not initiators of previously evaluated accidents involving a loss of offsite power (LOOP). The proposed changes to the TS Required Actions CTs do not affect any of the assumptions used in the deterministic or the PSA [probabilistic safety assessment] analysis relative to LOOP initiating event frequency. Implementation of the proposed changes does not result in a risk significant impact. The onsite AC [alternating current] power sources will remain highly reliable and the proposed changes will not result in a significant increase in the risk of plant operation. This is demonstrated by showing that the impact on plant safety as measured by the increase in core damage frequency (CDF) is less than 1E-06 per year and the increase in large early release frequency (LERF) is less than 1E-07 per year. In addition, for the CT changes, the incremental conditional core damage probabilities (ICCDP) and incremental conditional large early release probabilities (ICLERP) are less than 5E-07 and 5E-08, respectively. These changes meet the acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore, since the onsite AC power sources will continue to perform their functions with high reliability as originally assumed and the increase in risk as measured by ΔCDF, ΔLERF, ICCDP, and ICLERP risk metrics is within the acceptance criteria of existing regulatory guidance, there will not be a significant increase in the consequences of any accidents.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, and components (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. The proposed changes are consistent with safety analysis assumptions and resultant consequences.

The proposed TS CT extension will continue to provide assurance that the sources of power to 6.9 kV [kilovolts] AC buses perform their function when called upon. Extending the TS CT to 30 days does not affect the design of the STs, the operational characteristics of the STs, the interfaces between the STs and other plant systems, the function, or the reliability of the STs. Thus, the STs will be capable of performing their accident mitigation functions and there is no impact to the radiological consequences of any accident analysis.

The Configuration Risk Management Program (CRMP) in TS 5.5.18 is an administrative program that assesses risk based on plant status. The risk-informed CT will be implemented consistent with the CRMP and approved plant procedures. When utilizing the 30-day extension, requirements of the CRMP per TS 5.5.18 call for the consideration of other measures to mitigate the consequences of an accident occurring while a[n] ST is inoperable. Furthermore, administrative controls will be applied when exercising the 30-day CT extension and are adequate to maintain defense-in-depth and sufficient safety margins.

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? Start Printed Page 26179

Response: No.

The proposed changes do not result in a change in the manner in which the electrical distribution subsystems provide plant protection. There [are] no design changes associated with the proposed changes. The changes to the CT do not change any existing accident scenarios, nor create any new or different accident scenarios.

The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter any of the assumptions made in the safety analysis. The changes to the CT do not affect the accident analysis directly; the CT is strictly tied to the PRA [probabilistic risk assessment] and the risk associated with the occurrence of a low-probability event during the limited time the component is unavailable.

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

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. Neither the safety analyses nor the safety analysis acceptance criteria are impacted by these changes. The proposed changes will not result in plant operation in a configuration outside the current design basis. The proposed activities only involve changes to certain TS CTs.

The proposed change does not involve a change to the plant design or operation and thus does not affect the design of the STs, the operation characteristics of the STs, the interfaces between the STs and other plant systems, or the function or reliability of the STs. Because the STs' performance and reliability will continue to be ensured by the proposed TS change, the proposed changes do not result in a reduction in the margin of safety.

Therefore the proposed change does not involve a 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: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Branch Chief: David Terao.

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

Date of amendment request: December 19, 2006.

Brief description of amendments: The amendments requested would revise Technical Specification (TS) requirement 5.5.16, “Containment Leakage Rate Testing Program,” for consistency with the requirements of paragraph 50.55a(g)(4) of Title 10 of the Code of Federal Regulations (10 CFR) for components classified as Code Class CC.

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

Response: No.

The proposed change revises the TS administrative controls programs for consistency with the requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components classified as Code Class CC.

The proposed change affects the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program. In addition, the proposed change allows those examinations to be performed during power operation as opposed to during a refueling outage. The frequency of visual examinations of the concrete surfaces of the containment and the mode of operation during which those examinations are performed has no relationship to or adverse impact on the probability of any of the initiating events assumed in the accident analyses. The proposed change would allow visual examinations that are performed pursuant to NRC approved [American Society of Mechanical Engineers] (ASME) Section XI Code requirements (except where relief has been granted by the NRC) to meet the intent of visual examinations required by Regulatory Guide 1.163, without requiring additional visual examinations pursuant to the Regulatory Guide. The intent of early detection of deterioration will continue to be met by the more rigorous requirements of the Code required visual examinations. As such, the safety function of the containment as a fission product barrier is maintained.

The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. It does not involve the addition or removal of any equipment, or any design changes to the facility.

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

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

Response: No.

The proposed change revises the TS administrative controls programs for consistency with the requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components classified as Code Class CC.

The change affects the frequency of visual examinations that will be performed for the concrete surfaces containments. In addition, the proposed change allows those examinations to be performed during power operation as opposed to during a refueling outage. The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism. Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure.

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

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

Response: No.

The proposed change revises the Improved Standard Technical Specification Administrative Controls program requirements for consistency with the requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components classified as Code Class CC.

The change affects the frequency of visual examinations that will be performed for the concrete surfaces containments. In addition, the proposed change allows those examinations to be performed during power operation as opposed to during a refueling outage. The safety function of the containment as a fission product barrier will be maintained.

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: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Branch Chief: Thomas Hiltz.

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 Start Printed Page 2618010 CFR Chapter I, which are set forth in the license amendment.

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

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

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

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

Date of application for amendment: March 28, 2005, as supplemented by letters dated November 2, 2005, January 24, February 2, March 16, March 23, and March 28, 2007.

Brief description of amendment: The amendment revises the Oyster Creek Licensing Basis in the area of radiological dose analyses for design-basis accidents using the alternative source terms depicted in Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.” Additionally, the amendment revises the Oyster Creek Technical Specifications (TSs) consistent with the amended design-basis.

Date of Issuance: April 26, 2007.

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

Amendment No.: 262.

Facility Operating License No. DPR-16: The amendment revised the TSs.

Date of initial notice in Federal Register: May 10, 2005 (70 FR 24646). The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed to significant hazards consideration determination.

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 26, 2007.

No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

Date of application for amendment: June 1, 2006, as supplemented by letters dated November 20, 2006, and February 22, 2007.

Brief description of amendment: The amendment revises Surveillance Requirement 3.5.2 in the HBRSEP2 Technical Specifications.

Date of issuance: April 4, 2007.

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

Amendment No.: 213.

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

Date of initial notice in Federal Register: December 19, 2006 (71 FR 75992). The supplemental letters provided additional information that was within the scope of the original notice 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 April 4, 2007.

No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of application for amendment: November 27, 2006.

Brief description of amendment: The amendment revised Technical Specification (TS) 5.5.9 to relocate the specific American Society of Testing and Materials (ASTM) Standard from the Administrative Controls Section of TS to a licensee-controlled document. Also, the revision to TS 5.5.9 allows the performance of an alternate water and sediment content test to establish the acceptability of new fuel oil prior to addition to the storage tank has been added to the clear and bright test.

Date of issuance: April 12, 2007.

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

Amendment No.: 174.

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

Date of initial notice in Federal Register: January 3, 2007 (72 FR 149).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 12, 2007.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: October 13, 2006.

Brief description of amendment: The amendment revised Facility Operating License No. NPF-58 by deleting License Condition 2.F, which specifies reporting of violations of Operating License Section 2.C, and eliminates Technical Specification 5.6.6, which contains a reporting condition similar to Operating License Section 2.C.(6).

Date of issuance: April 19, 2007.

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

Amendment No.: 140.

Facility Operating License No. NPF-58: This amendment revised the Technical Specifications and License.

Date of initial notice in Federal Register: November 21, 2006 (71 FR 67394).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 19, 2007.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, Van Buren County, Michigan

Date of application for amendment: August 31, 2006, as supplemented on December 15, 2006, and March 1 and April 4, 2007.

Brief description of amendment: The amendment conforms the license to reflect the transfer of Renewed Facility Operating License No. DPR-20 to Entergy Nuclear Palisades, LLC, as Start Printed Page 26181owner, and Entergy Nuclear Operations, Inc., as operator, as approved by Order of the Commission dated April 6, 2007, and as revised on April 10, 2007.

Date of issuance: April 11, 2007.

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

Amendment No.: 224.

Facility Operating License No. DPR-20: Amendment revised the Renewed Operating License and Technical Specifications.

Date of initial notice in Federal Register: November 16, 2006 (71 FR 66805).

The December 15, 2006, and March 1 and April 4, 2007, supplemental letters contained clarifying information and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 6, 2007, as revised on April 10, 2007.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power Plant, Unit 3, Humboldt County, California

Date of application for amendment: December 20, 2006.

Brief description of amendment: The amendment revises License Condition 2.B.3(c) to allow the receipt, possession, and use of byproduct, source, or special nuclear material without restriction to amount or atomic number, for sample analysis or instrument calibration or associated with radioactive apparatus or components.

Date of issuance: April 17, 2007.

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

Amendment No.: 39.

Facility Operating License No. DPR-7: This amendment revises the license.

Date of initial notice in Federal Register: February 13, 2007 (72 FR 6788).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 17, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 29, 2006.

Brief description of amendments: The amendments revised Technical Specification (TS) 3.4.1, “Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” and TS 5.6.5, “Core Operating Limits Report (COLR).” This amendment relocated the RCS DNB parameters for pressurizer pressure and RCS average temperature to the COLR. In addition, TS 5.6.5 was revised to add topical reports WCAP-8567-P-A, “Improved Thermal Design Procedure,” and WCAP-11596-P-A, “Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores.”

Date of issuance: April 17, 2007.

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

Amendment Nos.: Unit 1—195; Unit 2—196.

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

Date of initial notice in Federal Register: February 13, 2007 (72 FR 6786).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 17, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: May 1, 2006, as supplemented October 9, 2006, and February 21, 2007.

Brief description of amendments: The amendments relocate the main steamline discharge radiation monitors (R46) from Technical Specification (TS) 3/4.3.3.1, “Radiation Monitoring Instrumentation” to TS 3/4.3.3.7, “Accident Monitoring Instrumentation.” In addition, the amendments modify TS definition 1.31, “Source Check.”

Date of issuance: April 19, 2007.

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

Amendment Nos.: 280 and 263.

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

Date of initial notice in Federal Register: July 18, 2006 (71 FR 40753). The supplements dated October 9, 2006, and February 21, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on July 18, 2006 (71 FR 40753).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 19, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: June 7, 2006.

Brief description of amendments: The amendments delete the Technical Specification (TS) requirements related to hydrogen recombiners and hydrogen analyzers. The changes support the implementation of a revision to Title 10 of the Code of Federal Regulations, Section 50.44, “Combustible gas control for nuclear power reactors” that became effective on October 16, 2003. A notice of availability for this TS improvement using the consolidated line item improvement process was published in the Federal Register on September 25, 2003 (68 FR 55416).

Date of issuance: April 19, 2007.

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

Amendment Nos.: 281 and 264.

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

Date of initial notice in Federal Register: August 29, 2006 (71 FR 51231).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 19, 2007.

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 21, 2006.

Description of amendment request: The amendments revised Technical Specification (TS) Limiting Condition for Operation 3.10.1, and the associated Bases, to expand its scope to include provisions for temperature excursions greater than 212 °F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with inservice leak or hydrostatic testing, while considering operational conditions to be in Mode 4.

Date of issuance: April 16, 2007.

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

Amendment Nos.: 270, 299 & 258. Start Printed Page 26182

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

Date of initial notice in Federal Register: February 13, 2007 (72 FR 6791).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 16, 2007.

No significant hazards consideration comments received: No.

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

Date of amendment request: December 15, 2006.

Brief description of amendment: The amendment revised the Technical Specifications to adopt NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-372, “Addition of LCO [Limiting Condition for Operation] 3.0.8, Inoperability of Snubbers.” The amendment added (1) a new LCO 3.0.8 addressing situations where one or more required snubbers are unable to perform their associated support function(s) and (2) a reference to LCO 3.0.8 in LCO 3.0.1, which describes when LCOs shall be met.

Date of issuance: April 17, 2007.

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

Amendment No.: 173.

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

Date of initial notice in Federal Register: January 3, 2007 (72 FR 154).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 17, 2007.

No significant hazards consideration comments received: No.

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

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

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

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

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

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

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

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

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

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 Start Printed Page 26183NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or 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, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. 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 (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely 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).

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear Plant, Unit 3, Limestone County, Alabama

Date of application for amendment: April 6, 2007 (TS-460-T).

Brief description of amendment: This amendment approves a one-time extension of the Completion Time for emergency diesel generator (EDG) ‘3D’ from 7 days to 14 days. The extension allows continued operation while repairs, post-maintenance testing, and surveillance testing of the subject EDG are completed.

Date of issuance: April 6, 2007.

Effective date: April 6, 2007, to be implemented within 30 days.

Amendment No.: 257.

Renewed Facility Operating License No. DPR-68: Amendment revises the Technical Specifications.

The Commission's related evaluation of the amendment, finding of emergency circumstances, and final determination of no significant hazards consideration, are contained in a Safety Evaluation dated April 6, 2007.

Public comments requested as to proposed no significant hazards consideration: No.

Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902. Start Printed Page 26184

NRC Section Chief: Thomas H. Boyce.

Start Signature

Dated at Rockville, Maryland, this 1st day of May 2007.

For the Nuclear Regulatory Commission.

Harold K. Chernoff,

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

End Signature End Preamble

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 protection order.

Back to Citation

[FR Doc. E7-8679 Filed 5-7-07; 8:45 am]

BILLING CODE 7590-01-P