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Notice

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

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Information about this document as published in the Federal Register.

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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 15, 2005 to April 28, 2005. The last biweekly notice was published on April 26, 2005 (70 FR 21449).

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 Start Printed Page 24646Atomic 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, 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 (OCNGS), Ocean County, New Jersey

Date of amendment request: March 28, 2005.

Description of amendment request: The licensee proposed to revise the licensing bases of OCNGS in the area of radiological dose analyses for the design-basis accidents (DBAs). Specifically, the licensee proposed to use the alternative source terms (AST) depicted in Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” instead of the source terms used in the current licensing basis and depicted in Technical Information Document 14844, “Calculation of Distance Factors for Power and Test Reactor Sites.” The acceptance criteria for the postulated consequences using AST are set forth in 10 CFR 50.67 and General Design Criterion 19, “Control Room.” The licensee has performed radiological consequence analysis for the most limiting DBAs that result in offsite and control room operator exposure to support a full-scope implementation of the AST. If approved, the amendment would: (1) Revise Section 3.2.A, “Standby Liquid Control System,” of the Technical Specifications (TSs) to add a specification to require that the subject system is operable when the reactor is at or greater than 212 degrees Fahrenheit; (2) revise various pages of the TS Bases to reflect use of the AST methodology. The issuance of the requested amendment would also signify the NRC staff's approval to revise the OCNGS Updated Final Safety Analysis Report to reflect implementation of the AST in the OCNGS licensing basis.

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 Start Printed Page 24647consideration. The NRC staff's analysis is presented below:

The first standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The AST is an input to calculations used to evaluate the consequences of an accident, and does not by itself affect the plant response, or the actual pathway of the radiation release. It does, however, better represent the physical characteristics of the release, so that appropriate mitigation techniques may be applied. The proposed amendment does not affect the design of plant systems, structures, or components (SSCs), or their operational characteristics or function. As a result, implementing the AST would not have any increase on the frequency of occurrence for previously analyzed accidents. It may be argued that the calculated radiological consequences are different because a different set of assumptions, with accompanying acceptance criteria, are used. However, since there is no design or operational change associated with the proposed amendment, the actual consequences of the same accident would not be changed regardless of what methodology was used before the accident to arrive at postulated consequences. As a result, implementing the AST would not increase the consequences of any previously evaluated accident.

The second standard requires that operation of the unit in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment does not alter the design, configuration, or method of operation of any SSC. Therefore, no new initiators or precursors of a new or different kind of accident are created that could result in a new or different kind of accident.

The third standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the establishment of setpoints to initiate alarms or actions. These are principally documented in the OCNGS licensing basis documents such as the Updated Final Safety Analysis Report, and none of these would be changed by the amendment. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the NRC staff's analysis, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the proposed amendment 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 Section Chief: Richard J. Laufer.

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: March 4, 2005.

Description of amendments request: The proposed amendments would delete Section 2.F (2.G in Unit 3) of the Operating License which requires reporting violations of the requirements in Section 2.C of the Operating License. The amendments will also make administrative and editorial changes to the Technical Specifications (TSs). Changes to TS 1.4, “Frequency,” and TS 3.4.3, “RCS Pressure and Temperature (P/T) Limits,” will correct editorial errors. The changes to TS 2.1.1, “Reactor Core SLs,” and TS 3.3.1, “Reactor Protective System (RPS) Instrumentation—Operating,” will remove the reference to departure from nucleate boiling ratios (DNBR) based on operating cycle, since only one of the listed DNBR values is now valid. TS 3.1.10, “Special Test Exceptions (STE)—MODES 1 and 2,” will be changed to correct an inconsistency between the limiting condition for operation and the TS Bases. The changes to TS 3.7.2, “Main Steam Isolation Valves (MSIVs)” and TS 3.7.3, “Main Feedwater Isolation Valves (MFIVs)” will correct the applicability for these specifications. The change to TS 3.8.1, “AC Sources—Operating” will add a note to a surveillance requirement. Changes to TS 3.8.4, “DC Sources—Operating” and TS 3.8.6, “Battery Cell Parameter” will remove the reference to AT&T batteries. The changes to TS 5.5.9, “Steam Generator (SG) Tube Surveillance Program” will correct the reference for NRC notification.

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 includes [the] following changes that are considered to be administrative and/or editorial changes:

The reporting requirement in License Condition 2.F (2.G in Unit 3) is adequately addressed by the requirements identified in 10 CFR 50.72, “Immediate notification requirements for operating nuclear power reactors” and 10 CFR 50.73, “Licensee event report system.” Since Condition 2.F (2.G in Unit 3) is adequately addressed by the requirements in 10 CFR 50.72 and 10 CFR 50.73, the Condition is not required. Therefore, this is considered an administrative change that eliminates regulatory requirements that are adequately addressed by the requirements in 10 CFR 50.72 and 10 CFR 50.73.

The changes to Technical Specifications (TS) 1.4 and 3.4.3 are editorial changes only. These changes maintain the format of the Technical Specifications and correct editorial errors in the Technical Specifications.

The changes to Technical Specifications 2.1.1 and 3.3.1 remove requirements that are no longer applicable to the Palo Verde Nuclear Generating Station (PVNGS) units. As part of Amendment 133 to the PVNGS Operating License, the minimum DNBR was revised based on Unit operating cycle, ≥1.30 (through operating cycle 10)” and ≥1.34 (operating cycle 11 and later).” All three PVNGS units have completed operating cycle 10. Therefore, the reference to the minimum d[e]parture from nucleate boiling ratio (DNBR) through operating cycle 10 (≥1.30) is no longer required.

The changes to Technical Specification 3.1.10 correct an inconsistency between the Technical Specification limiting condition for operation (LCO) and Bases. The Bases for this specification states that “Even if an accident occurs during PHYSICS TESTS with one or more LCOs suspended, fuel damage criteria are preserved because the limits on power distribution and shutdown capability are maintained during PHYSICS TESTS.” The limits on power distribution are maintained by TSs 3.2.1, “Linear Heat Rate (LHR)” and 3.2.4 “Departure from Nucleate Boiling Ratio (DNBR).” These changes ensure that shutdown capability is maintained during physics tests.

The changes to Technical Specifications Section 3.7.2, “Main Steam Isolation Valves (MSIVs)” and Section 3.7.3, “Main Feedwater Isolation Valves (MFIVs)” correct an inconsistency between the applicability and the required actions. The changes are consistent with the guidance in NUREG-1432, “Standard Technical Specifications, Combustion Engineering Plants.” Therefore, this is considered an administrative change that corrects an inconsistency in the Technical Specifications.

The changes to Technical Specifications Section 3.8.1, “AC Sources—Operating,” correct an inconsistency in the surveillance requirements that were revised in Amendment 129 to the PVNGS Operating License. A note was not included with the change to one of the surveillance requirements. This change adds the note to Start Printed Page 24648the surveillance requirement. Therefore, this is considered an administrative change that corrects an inconsistency in the Technical Specifications.

The changes to Technical Specifications Section 3.8.4, “DC Sources—Operating” and Section 3.8.6, “Battery Cell Parameters” removes the requirements and references to the AT&T batteries. APS has replaced the AT&T batteries with low specific gravity batteries in all three units. Therefore, this is considered an administrative change that removes unnecessary requirements and references.

The changes to Technical Specifications Section 5.5.9, “Steam Generator (SG) Tube Surveillance Program,” updates the requirement to notify the NRC based on the January 23, 2001 rule change to 10 CFR 50.72. Therefore, this change corrects NRC notification requirements in Technical Specifications, based on the January 23, 2001 rule change to 10 CFR 50.72 (65 FR 63786, 10/25/00).

As discussed above the proposed amendment involves administrative and/or editorial changes only. The proposed amendment does not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events. The proposed changes do not involve the addition or removal of any equipment or any design changes to the facility. The proposed changes do not affect plant operations, any design function or an analysis that verifies the capability of structures, systems, and components (SSCs) of the plant. The proposed changes do not change any of the previously evaluated accidents in the updated final safety analysis report (UFSAR). The proposed changes do not affect SSCs, operating procedures, and administrative controls that have the function of preventing or mitigating any of these accidents.

Therefore, the proposed changes do not represent 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.

As discussed in standard 1, the proposed amendment only involves administrative and/or editorial changes. No actual plant equipment or accident analysis will be affected by the proposed changes. The proposed changes will not change the design function or operation of any SSCs. The proposed changes will not result in any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases. The proposed amendment does not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events.

Therefore, this proposed change does not create the possibility of an accident of a different kind than previously evaluated.

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

Response: No.

As discussed in standard 1, the proposed amendment only involves administrative and/or editorial changes. Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel and fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. This request involves administrative and/or editorial changes only. No actual plant equipment or accident analysis will be affected by the proposed changes. Additionally, the proposed changes will not relax any criteria used to establish safety limits, will not relax any safety system settings, or will not relax the bases for any limiting conditions for operation.

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

NRC Section Chief: Robert A. Gramm.

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: January 27, 2005.

Description of amendments request: The proposed amendment would allow entry into a mode or other specified condition in the applicability of a Technical Specification (TS), while in a condition statement and the associated required actions of the TSs, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would be eliminated, several notes or specific exceptions would be revised to reflect the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 3.0.4 would be revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's TS Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated January 27, 2005.

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 entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by 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). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change 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 [a] Margin of Safety.

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the Start Printed Page 24649plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Carey Fleming, Esquire, Counsel, Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, Baltimore, MD 21202.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: March 14, 2005.

Description of amendment request: The proposed amendments would delete Technical Specification (TS) Section 5.5.4, “Post Accident Sampling,” requirements to maintain a Post Accident Sampling System (PASS). Licensees were generally required to implement PASS upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide 1.97, Revision 3, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Access Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the NRC's lessons learned from the accident that occurred at TMI Unit 2. Requirements related to PASS were imposed by Order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. Lessons learned and improvements implemented over the last 20 years have shown that the information obtained from PASS can be readily obtained through other means or is of little use in the assessment and mitigation of accident conditions.

The NRC staff issued a notice of opportunity for comment in the Federal Register on March 3, 2003 (68 FR 10052) on possible amendments to eliminate PASS, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in a license amendment application in the Federal Register on May 13, 2003 (68 FR 25664). The licensee affirmed the applicability of the following NSHC determination in its application dated March 14, 2005.

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 PASS was originally designed to perform many sampling and analysis functions. These functions were designed and intended to be used in post accident situations and were put into place as a result of the TMI-2 accident. The specific intent of the PASS was to provide a system that has the capability to obtain and analyze samples of plant fluids containing potentially high levels of radioactivity, without exceeding plant personnel radiation exposure limits. Analytical results of these samples would be used largely for verification purposes in aiding the plant staff in assessing the extent of core damage and subsequent offsite radiological dose projections. The system was not intended to and does not serve a function for preventing accidents and its elimination would not affect the probability of accidents previously evaluated.

In the 20 years since the TMI-2 accident and the consequential promulgation of post accident sampling requirements, operating experience has demonstrated that a PASS provides little actual benefit to post accident mitigation. Past experience has indicated that there exists in-plant instrumentation and methodologies available in lieu of a PASS for collecting and assimilating information needed to assess core damage following an accident. Furthermore, the implementation of Severe Accident Management Guidance (SAMG) emphasizes accident management strategies based on in-plant instruments. These strategies provide guidance to the plant staff for mitigation and recovery from a severe accident. Based on current severe accident management strategies and guidelines, it is determined that the PASS provides little benefit to the plant staff in coping with an accident.

The regulatory requirements for the PASS can be eliminated without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. The elimination of the PASS will not prevent an accident management strategy that meets the initial intent of the post-TMI-2 accident guidance through the use of the SAMGs, the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

Therefore, the elimination of PASS requirements from Technical Specifications (TS) (and other elements of the licensing bases) does not involve a significant increase in the consequences of any 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 elimination of PASS related requirements will not result in any failure mode not previously analyzed. The PASS was intended to allow for verification of the extent of reactor core damage and also to provide an input to offsite dose projection calculations. The PASS is not considered an accident precursor, nor does its existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radioisotopes within the containment building.

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

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

The elimination of the PASS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. Methodologies that are not reliant on PASS are designed to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The use of a PASS is redundant and does not provide quick recognition of core events or rapid response to events in progress. The intent of the requirements established as a result of the TMI-2 accident can be adequately met without reliance on a PASS.

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

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

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

Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP, 1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: John A. Nakoski.Start Printed Page 24650

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: September 23, 2004, as supplemented by letter dated April 19, 2005.

Description of amendment request: The amendment would revise the reactor operational limits, as specified in the River Bend Station Core Operating Limits Report (COLR), to compensate for the inoperability of the End of Cycle Recirculation Pump Trip (EOC-RPT) instrumentation. This will provide an alternative to the existing Limiting Condition for Operation for the EOC-RPT instrumentation. The revised Technical Specification will require that either the EOC-RPT instrumentation be operable or that Minimum Critical Power Ratio and Linear Heat Generation Rate limits for the inoperable EOC-RPT be placed in effect as specified in the COLR.

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

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

Response: No.

The End of Cycle Recirculation Pump Trip (EOC-RPT) functions to insert negative reactivity in response to certain anticipated transients. The EOC-RPT is a mitigation function and not the initiator of any evaluated accident or transient. Operation with inoperable EOC-RPT instrumentation and compliance with new restrictive Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) operating limits establish sufficient margin to the core thermal MCPR safety limit (SL) and the thermal mechanical design limits as would be the case with operable EOC-RPT instrumentation and existing MCPR and LHGR limits.

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 change will not create any new modes of plant or equipment operation. The proposed change allows the option to apply an additional penalty factor to the MCPR and LHGR when the EOC-RPT is inoperable. With the addition of the penalty factor, the margin to the MCPR SL and the thermal mechanical design limits are maintained. Therefore, operating the plant with the proposed change will not create the possibility of a new or different kind of accident from any previously analyzed.

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.

By establishing a new restrictive MCPR and LHGR operating limit, there are no changes to the plant design and safety analysis. There are no changes to the reactor core design instrument setpoints. The margin of safety assumed in the safety analysis is not affected. Applicable regulatory requirements will continue to be met and adequate defense-in[-]depth will be maintained. Sufficient safety margins will be maintained.

The analytical methods used to determine the revised core operating limits were reviewed and approved by the NRC, and are described in Technical Specification 5.6.5. Specific analyses were prepared by the RBS fuel vendor to develop core operating limits without crediting the EOC-RPT. Therefore, implementation of the proposed changes will not involve a significant reduction in the margin 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: Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Allen G. Howe.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: February 15, 2005.

Description of amendment request: The proposed amendment would approve application of an alternative source term methodology with the exception that Technical Information Document 14844, “Calculation of Distance Factors for Power Test Reactor Sites,” will continue to be used as the radiation dose basis for equipment qualification.

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

The implementation of AST assumptions has been evaluated in revisions to the analyses of the following limiting DBAs at the Byron Station and Braidwood Station.

Loss-of-Coolant Accident

Fuel Handling Accident

Control Rod Ejection Accident

Locked Rotor Accident

Main Steam Line Break Accident

Steam Generator Tube Rupture Accident

Based upon the results of these analyses, it has been demonstrated that, with the requested changes, the dose consequences of these limiting events are within the regulatory guidance provided by the NRC for use with the AST methodology. This guidance is presented in RG 1.183, and Standard Review Plan Section 15.0.1. The AST is an input to calculations used to evaluate the consequences of an accident and does not by itself affect the plant response or the actual pathway of the activity released from the fuel. It does, however, better represent the physical characteristics of the release such that appropriate mitigation techniques may be applied.

The AST methodology follows the guidance provided in RG 1.183 and satisfies the dose limits in 10 CFR 50.67. Even though these limits are not directly comparable to the previously specified whole body and thyroid requirements of 10 CFR 50, Appendix A, General Design Criteria (GDC) 19, “Control room,” and 10 CFR 100.11, “Determination of exclusion area, low population zone, and population center distance,” the results of the AST analyses have demonstrated that the 10 CFR 50.67 limits are satisfied. Therefore, it is concluded that AST does not involve a significant increase in the consequences of an accident previously evaluated.

Implementation of AST provides increased operating margins for the control room ventilation system filtration efficiencies. It also relaxes containment integrity requirements while handling irradiated fuel that has decayed for greater than 48 hours and during core alterations. Automatic initiation of the radiation isolation mode for the control room is not credited in the accident analysis which allows relaxation of certain Technical Specification surveillance requirements.

The equipment affected by the proposed changes is mitigative in nature and relied upon after an accident has been initiated. Application of the AST does result in changes to the functions and operation of various filtration systems as described in the Updated Final Safety Analysis Report (UFSAR). These effects have been considered in the evaluations for these proposed changes. While the operation of various systems does change with the implementation of AST, the affected systems are not accident initiators; and application of the AST methodology, itself, is not an initiator of a design basis accident. The proposed changes to the TS revise certain equipment performance requirements but do not require any physical changes to the plant.

As a result, the proposed changes do not affect any of the parameters or conditions Start Printed Page 24651that could contribute to the initiation of any accidents. Relaxation of operability requirements during the specified conditions will not significantly increase the probability of occurrence of an accident previously analyzed. Since design basis accident initiators are not being altered by adoption of the AST, the probability of an accident previously evaluated is not affected.

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

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

The proposed changes do not involve a physical change to the plant. Implementation of AST provides increased operating margins for filtration system efficiencies. Application of AST also allows for the relaxation of containment integrity requirements while handling irradiated fuel that has decayed for greater than 48 hours and during core alterations. Automatic initiation of the radiation isolation mode for the control room is no longer credited in the accident analysis.

Similarly, the proposed changes do not require any physical changes to any structures, systems or components involved in the mitigation of any accidents. Therefore, no new initiators or precursors of a new or different kind of accident are created. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed changes.

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

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

Approval of a change from the original source term methodology (i.e., TID 14844) to an AST methodology, consistent with the guidance in RG 1.183, will not result in a significant reduction in the margin of safety. The safety margins and analytical conservatisms associated with the AST methodology have been evaluated and were found acceptable. The results of the revised DBA analyses, performed in support of the proposed changes, are subject to specific acceptance criteria as specified in RG 1.183. The dose consequences of these DBAs remain within the acceptance criteria presented in 10 CFR 50.67 and RG 1.183.

The proposed changes continue to ensure that the doses at the exclusion area boundary (EAB) and low population zone boundary (LPZ), as well as the control room, are within the specified regulatory limits.

Therefore, based on the above discussion, 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 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, et al., Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio; Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania; Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: February 22, 2005.

Description of amendment request: The requested change will delete Technical Specification requirements related to Occupational Radiation Exposure Reports and Monthly Operating Reports.

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated February 22, 2005.

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:

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

Response: No.

The proposed change eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. 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 change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above, the requested change does not involve a significant hazards consideration.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chiefs: Gene Y. Suh, Richard J. Laufer.

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

Date of amendment request: March 22, 2005.

Description of amendment request: The proposed amendments revise the Technical Specifications (TS) for several Reactor Protection System functional units. The steam/feedwater flow mismatch coincident with steam generator water level—low reactor trip is being deleted, the reactor trip on turbine trip interlock is being changed from P-7 to P-8, the value of the P-8 interlock setpoint is being changed from 45 percent rated thermal power (RTP) to 40 percent RTP, and the value of the P-8 interlock allowable value is being changed from 48 percent RTP to 43 percent RTP.

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 the operability requirements, surveillance requirements and the interlock setpoint for two Reactor Trip System functional units. The affected trip functional units are not initiators of any accident previously evaluated. The proposed changes to the affected trip functional units do not adversely affect the initiators of any accident previously evaluated. A best estimate Start Printed Page 24652analysis has shown that a turbine trip without a reactor trip below 40% power does not challenge the pressurizer PORVs [power operated relief valves] or the steam generator safety valves; thereby, not adversely affecting the probability of a small break LOCA [loss of coolant accident] due to a stuck open PORV, or an excessive cooldown event due to a stuck open steam generator safety valve. As a result, the probability of any accident previously evaluated is not significantly increased by the proposed changes.

The steam/feedwater flow mismatch coincident with steam generator water level—low reactor trip is not credited as a primary trip in any previously evaluated accidents. The reactor trip on turbine trip below the P-8 interlock is not credited as a primary trip in any previously evaluated accidents. Therefore, the mitigation functions that have been assumed in the accident analyses will continue to be performed by the systems and components currently credited in the analyses; and the accident analysis results are not affected by the changes to the affected trip functional units. The P-8 setpoint is not an initial condition of any accident previously evaluated. Therefore, the accident analysis results are not affected by changes to the P-8 setpoint. No safety analyses previously performed in the Turkey Point Units 3 and 4 UFSAR [Updated Final Safety Analysis Report] required reanalysis for these proposed changes. All accident analyses acceptance criteria continue to be met. The proposed changes do not create any new credible limiting single failure. As a result, the consequences of any accident previously evaluated are not significantly increased by the proposed changes.

In conclusion, operation of the facility in accordance with the proposed amendments does 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.

No changes are being made to the plant that would introduce any new accident causal mechanisms. The proposed changes do not adversely affect previously identified accident initiators and do not create any new accident initiators. No new limiting single failures or accident scenarios are created by the proposed changes. No new challenges to any installed safety system are created by these proposed changes. The proposed changes do not result in any event previously deemed incredible being made credible.

The steam/feedwater flow mismatch coincident with steam generator water level—low reactor trip is not credited as an inhibitor of any potential or actual accident initiators. So, deletion of this reactor trip functional unit will not create the possibility of a new or different kind of accident from any previously evaluated.

Changing the interlock for the reactor trip on turbine trip from P-7 to P-8 changes the power level associated with enabling and disabling the reactor trip on turbine trip function. The turbine pressure input to the reactor protection system permissives is not an accident initiator and is not credited in the accident analyses. Changing the P-8 allowable and trip setpoint values changes the power level associated with enabling and disabling the reactor trip functions currently associated with P-8. The change does not affect how the associated trip functional units operate or function. Since these interlock changes do not affect the way that the associated trip functional units operate or function, the changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

No UFSAR safety analyses were changed or modified as a result of these proposed changes. Therefore, all margins associated with the current UFSAR safety analyses acceptance criteria are unaffected. The current UFSAR safety analyses remain bounding. No UFSAR Chapter 14 events explicitly credit the steam/feedwater flow mismatch reactor trip function and the reactor trip on turbine trip function below the P-8 setpoint value. The safety systems credited in the safety analyses will continue to be available to perform their mitigation functions. Changing the P-8 setpoint from 45% to 40% is in the conservative direction for the Reactor Coolant Flow—Low Reactor Trip and the Reactor Coolant Pump Breaker Position Reactor Trip. Therefore, the proposed changes do not result in a significant reduction in a margin of safety; and operation of the facility in accordance with the proposed amendments would 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 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.

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

Date of amendment request: March 28, 2005.

Description of amendment request: The proposed amendment would revise the Technical Specifications to allow the option of not measuring the moderator temperature coefficient within 7 effective full-power days after reaching an equilibrium boron concentration of 300 parts per million. This option would be available if the benchmark criteria in WCAP-13749-P-A and the revised prediction specified in the core operating limits report are satisfied.

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

1. The proposed change[s] do[es] not involve a significant increase in the probability or consequences of an accident previously evaluated.

The probability or consequences of accidents previously evaluated in the UFSAR [updated final safety analysis report] are unaffected by this proposed change. There is no change to any equipment response or accident mitigation scenario, and this change results in no additional challenges to fission product barrier integrity. The proposed change does not alter the design, configuration, operation, or function of any plant system, structure, or component. Further, the existing limits on moderator temperature coefficient (MTC) established by the Technical Specifications (TS), based on assumptions in the safety analyses, remain unchanged and continue to be satisfied. As a result, the outcomes of previously evaluated accidents are unaffected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed change[s] do[es] not create the possibility of a new or different kind of accident from any previously evaluated.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. The proposed change neither installs or removes any plant equipment, nor alters the design, physical configuration, or mode of operation of any plant structure, system, or component. The MTC is a variable that must remain within prescribed limits, but it is not an accident initiator. No physical changes are being made to the plant, so no new accident causal mechanisms are being introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposed change[s] do[es] not involve a significant reduction in the margin of safety.

The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect on the availability, operability, or performance of the safety-related systems and components. The proposed change does not alter the design, configuration, operation, or function of any plant system, structure, or component. The ability of any operable structure, system, or component to perform its designated safety function is unaffected by Start Printed Page 24653this change. A change to a surveillance requirement is proposed based on an alternative method of confirming that the surveillance is met. The Technical Specifications establish limits for the moderator temperature coefficient (MTC) based on assumptions in the accident analyses. Applying the conditional exemption from the MTC measurement changes the method of meeting the surveillance requirement; however, this change does not modify the TS values and ensures adherence to the current TS limits. The basis for the derivation of the MTC limits from the moderator density coefficient (MDC) assumed in the accident analysis is unchanged. Further, the safety analysis assumption of a constant MDC and its assumed value will not change. Therefore, the margin of safety as defined in the TS is not reduced and 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: M. S. Ross, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Section Chief: Darrell J. Roberts.

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

Date of amendment request: March 28, 2005.

Description of amendment request: The proposed amendment would revise Seabrook Station, Unit No. 1 (Seabrook) Technical Specification (TS) 3/4.9.13, “Spent Fuel Assembly Storage.” This revision would reflect a revised criticality safety analysis supporting a two-zone spent fuel pool consisting of BORAFLEX® and BORAL® fuel assembly storage racks. Additionally, the proposed change would create TS 3/4.9.15, “Spent Fuel Pool Boron Concentration.”

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

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

The proposed license amendment incorporates the results of a revised criticality analysis for the spent fuel pool without making any physical changes to the facility. The revised criticality analysis for the spent fuel pool (1) credits boron during movement of fuel in the spent fuel pool, (2) assumes no neutron-absorbing material in the BORAFLEX ® storage racks, and (3) applies a conservative penalty in the analysis of BORAL ® racks. These changes do not increase the probability of a fuel assembly being misplaced within the spent fuel pool. The movement of fuel assemblies will continue to be controlled by approved procedures, and the placement of spent fuel will be controlled by the revised Technical Specifications. The proposed changes do not alter or prevent the ability of structures, systems, or components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR).

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 in the Seabrook Station UFSAR. The consequences of a misplaced fuel assembly are not increased because the analysis demonstrates that the fuel will remain sub-critical with a minimum of 872 ppm [part per million] boron in the spent fuel pool. The new technical specification included in this proposed change will ensure that the minimum boron concentration is established during the movement of fuel in the spent fuel pool. Further, the proposed changes neither increase the types and amounts of radioactivity released offsite nor increase occupational or public radiation exposures.

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

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

The proposed changes to the TS do not alter the operation of the spent fuel storage system or its ability to perform its design function. The proposed changes do not include any physical changes to the plant and do not introduce a new or different accident from any type previously evaluated. A misplaced fuel assembly does not represent a new or different type [of] accident, and the analysis shows that the fuel remains sub-critical for the limiting case of a misplaced fuel assembly. Similarly, continuing to take credit for boron in the spent fuel under accident conditions does not create the possibility of a new or different kind of accident. The previous criticality analyses took credit for soluble boron in the spent fuel pool water to show acceptable results in the analyses of fuel misloading events.

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

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

The changes proposed by this license amendment ensure that the spent fuel will remain sub-critical under normal and accident conditions. The controlled placement of fuel assemblies within the spent fuel pool will maintain Keff less than or equal to 0.95 as required by TS 5.6.1.1 for spent fuel storage. The proposed amendment maintains the 0.95 limit on Keff by restricting the placement of spent fuel and by crediting soluble boron in the fuel pool water.

To assure that the true reactivity will be less than the calculated reactivity, the analyses contain conservative assumptions for calculating the safety limits for the spent fuel rack. With this proposed change, Keff will be less than or equal to 0.95 with a 95% probability at a 95% confidence level.

Therefore, the proposed amendment does not result in 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. S. Ross, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.

NRC Section Chief: Darrell J. Roberts.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of amendment request: April 1, 2005.

Description of amendment request: The licensee proposed to revise Section 3.8.7, “Inverters—Operating,” of the Technical Specifications (TSs), extending the time allowed to fix inoperable emergency uninterruptible power supply (UPS) inverters from the current 24 hours to 7 days.

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's analysis is presented below:

The first standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment does not affect the design of the emergency UPS inverters, the operational characteristics or function of the inverters, the interfaces between the inverters and other plant systems, or the reliability of the inverters. An inoperable emergency UPS inverter was not considered an initiator of a previously analyzed event. In addition, the required actions and the associated completion times specified by the TSs are not initiators of previously evaluated accidents. As a result, extending the completion time Start Printed Page 24654for an inoperable emergency UPS inverter would not have a significant impact on the frequency of occurrence for a previously analyzed accident. Furthermore, the proposed amendment will not result in modifications to plant activities associated with inverter maintenance, but rather, provides operational flexibility by allowing additional time to perform inverter corrective maintenance and post-maintenance testing on-line. The proposed extension of inoperable time will not significantly affect the capability of inverters to perform their safety function, which is to ensure an uninterruptible supply of 120-volt alternating current (ac) electrical power to the associated power distribution subsystems. The licensee performed a probabilistic risk assessment which concluded that the increase in plant risk is small. Therefore, the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The second standard requires that operation of the unit in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment does not alter the design, configuration, or method of operation of the emergency UPS inverters or their associated 120-volt ac uninterruptible power distribution subsystems, nor does the amendment alter any safety analyses inputs and assumptions. The proposed extended emergency UPS inverter completion time does not reduce the number of emergency UPS inverters below the minimum required for safe shutdown or accident mitigation, and does not affect the parameters within which NMP2 is operated or the setpoints at which protective or mitigative actions are initiated. The use of the alternate safety-related maintenance supply to power the 120-volt ac uninterruptible power distribution subsystem is consistent with the NMP2 design. If a station blackout event were to occur while an emergency UPS inverter is out of service, a dedicated portable power supply would be connected to provide a continuous source of power to the connected systems. Accordingly, no new failure modes, system interactions, or accident responses will be created that could result in a new or different kind of accident.

The third standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the establishment of setpoints to initiate alarms or actions. The proposed amendment will not affect any margin of safety as defined in the NMP2 Updated Safety Analysis Report. The amendment does not change the design or operational parameters of the UPS inverters as compared to original plant design. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: April 1, 2005.

Description of amendment request: The proposed amendment would provide one-time extension to the completion time for restoration of a service water train to operable status in Technical Specification (TS) 3.7.8, “Service Water System (SWS).”

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 amendment does not involve a significant increase in the probability of an accident previously evaluated because the extended Technical Specification action completion time is not an accident initiator. Therefore the probability is not increased significantly.

The proposed amendment does not involve a significant increase in the consequences of an accident previously evaluated. With service water pump P-7C inoperable, 100% of the required post-accident SWS cooling capability remains available with the redundant train maintained operable. A risk analysis was performed to show that the consequences are not significantly increased. The compensatory measures provide additional assurance that there is no significant increase in the consequences of an accident associated with extending the Technical Specification action completion time for the service water system for an additional 96 hours.

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

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

Response: No.

The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment only extends the Technical Specification action completion time and does not involve a physical alteration of any system, structure or component (SSC), or change in the way any SSC is operated. The proposed amendment does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the changes being requested.

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 amendment does not involve a significant reduction in a margin of safety. With service water pump P-7C inoperable, 100% of the required post-accident service water system cooling capability remains available with the redundant train maintained operable. Therefore, there is no significant reduction in the margin of safety.

Based on the availability of redundant systems, the compensatory measures that will be taken, and the low probability of an accident that could not be mitigated by the available systems, the proposed amendment would 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 Part 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

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

NRC Section Chief: L. Raghavan.

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

Date of amendment request: January 28, 2005.

Description of amendment request: The proposed amendment would revise Start Printed Page 24655the SSES 2 Technical Specification (TS) Table 3.3.5.1-1 “Emergency Core Cooling System Instrumentation,” to change Function 3.e “HPCI [High-Pressure Coolant Injection] System,” conditions referenced from Required Action A.1 from “D” to “C.” This is an editorial revision to correct a typographical error that has been present since PPL converted to the Improved Technical Specifications in 1998.

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 of occurrence or consequences of an accident previously evaluated?

Response: No.

The proposed change to the Unit 2 TS Table 3.3.5.1 provides a correction to a typographical error that occurred when preparing a change to Unit 2 Technical Specification Table 3.3.5.1-1 in the response to an NRC Request for Additional Information (RAI). The request was initiated during NRC review of documents submitted by PPL for the conversion to the Improved Technical Specifications. This proposed change is considered to be administrative in nature because it was originally submitted correctly and was inadvertently changed in response to the RAI.

Therefore, this 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.

As stated above, the proposed change to the Unit 2 TS Table 3.3.5.1 provides a correction to a typographical error that occurred when preparing the response to an NRC Request for Additional Information. The request was initiated by the NRC during its review of documents submitted by PPL for the conversion to the Improved Technical Specifications. This proposed change is administrative in nature.

Therefore, these proposed changes do 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.

Again, the proposed change to the Unit 2 TS Table 3.3.5.1 provides a correction to a typographical error that occurred when preparing the response to an NRC Request for Additional Information. The request was initiated by the NRC during its review of documents submitted by PPL for the conversion to the Improved Technical Specifications. This proposed change is administrative in nature.

Therefore, these 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: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Section Chief: Richard J. Laufer.

PSEG Nuclear LLC, Docket Nos. 50-272, Salem Nuclear Generating Station, Unit No. 1 Salem County, New Jersey

Date of amendment request: February 23, 2005.

Description of amendment request: The proposed changes will revise Technical Specification (TS) Steam Generator (SG) requirements for Salem Nuclear Generating Station, Unit No. 1. The proposed changes would replace TS 3/4.4.5 “Steam Generator (SG)” with “Steam Generator Tube Integrity;” add a new TS 6.8.4.i, “Steam Generator Program;” and add a new reporting requirement TS 6.9.1.10 “Steam Generator Tube Inspection Report.” Additionally, the proposed changes would revise TS 3/4.4.6.2, “Reactor Coolant System Operational Leakage.” Specifically, the Limiting Condition for Operation and ACTION and Surveillance Requirements of TS 3/4.4.6.2 would be revised to clarify the requirements related to primary-to-secondary leakage. These changes would facilitate implementation of industry initiative Nuclear Energy Institute (NEI) 97-08, “Steam Generator Program Guidelines,” to allow a comprehensive, performance-based approach to managing SG performance at Salem Nuclear Generating Station, Unit No. 1.

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[s] require[s] a Steam Generator Program that includes performance criteria that will provide reasonable assurance that the steam generator (SG) tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cool down and all anticipated transients included in the design specification). The SG performance criteria are based on tube structural integrity, accident induced leakage, and operational leakage.

The structural integrity performance criterion is:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

The accident induced leakage performance criterion is:

The primary-to-secondary accident induced leakage rate for any design basis accidents, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG.

The operational leakage performance criterion is:

The reactor coolant system operational primary-to-secondary leakage through any one SG shall be limited to 150 gallons per day.

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

For other design basis accidents such as main steam line break (MSLB), rod ejection, and reactor coolant pump locked rotor the tubes are assumed to retain their structural integrity (i.e., they are assumed not to rupture). These analyses assume that primary-to-secondary leakage for all SGs is 1 gallon per minute or increases to 1 gallon per minute as a result of accident-induced stresses. The accident induced leakage criterion retained by the proposed changes accounts for tubes that may leak during design basis accidents. The accident induced leakage criterion limits this leakage to no more than the value assumed in the accident analysis. Start Printed Page 24656

The SG performance criteria proposed as part of these TS changes identify the standards against which tube integrity is to be measured. Meeting the performance criteria provides reasonable assurance that the SG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident. The performance criteria are only a part of the Steam Generator Program required by the proposed addition of TS 6.8.4.i. The program defined by NEI 97-06 includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring.

The consequences of design basis accidents are, in part, functions of the DOSE EQUIVALENT I-131 in the primary coolant and the primary-to-secondary leakage rates resulting from an accident. Therefore, limits are included in the Salem TS for operational leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure the plant is operated within its analyzed condition. The Salem analysis of the limiting design basis accident assumes that primary-to-secondary leak rate after the accident is 1 gallon per minute with no more than 500 gallons per day through any one SG, and that the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the TS values before the accident.

The proposed change[s] do[es] not affect the design of the SGs, their method of operation, or primary coolant chemistry controls. The proposed approach updates the current TS and enhances the requirements for SG inspections.

The proposed change[s] do[es] not adversely impact any other previously evaluated design basis accident and [are] an improvement over the current TS.

Therefore, the proposed changes do not affect the consequences of a[n] SGTR accident and the probability of such an accident is reduced. In addition, the proposed changes do not affect the probabilities or consequences of an MSLB, rod ejection, or a reactor coolant pump locked rotor event.

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 changes do not affect the design of the SGs, their method of operation, or primary or secondary coolant chemistry controls. In addition, the proposed change[s] do[es] not impact any other plant system or component. The change[s] enhance[s] SG inspection requirements.

Therefore, the proposed changes do 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 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[s] do[es] not affect tube design or operating environment. The proposed change[s] [are] expected to result in an improvement in the tube integrity by implementing the Steam Generator Program to manage SG tube inspection, assessment, 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.

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

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: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: Darrell J. Roberts.

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

Date of amendment requests: March 24, 2005.

Description of amendment requests: The proposed change would revise the following Technical Specifications (TSs):

  • TS 1.1, Definitions, correct the definition of SHUTDOWN MARGIN (SDM).
  • TS 3.1.1, SHUTDOWN MARGIN (SDM)—Tavg > 2000F, and TS 3.1.2, SHUTDOWN MARGIN (SDM)—Tavg < 2000F, relocate the numerical shutdown margin requirements to the Core Operating Limits Report (COLR).
  • TS 3.1.3, Reactivity Balance, increase the required action time from 72 hours to 7 days when the “Core reactivity balance not within limit.”
  • TS 3.1.5, Control Element Assembly (CEA) Alignment, TS 3.1.6, Shutdown Control Element Assembly (CEA) Insertion Limits, and TS 3.1.7, Regulating CEA Insertion Limits, remove the requirement to verify SDM.
  • TS 3.2.4, Departure From Nucleate Boiling Ratio (DNBR), relocate to the COLR the power margin that must be accommodated when the Core Operating Limit Supervisory System (COLSS) is in service and neither CEA calculator is OPERABLE.
  • TS 5.7.1.5, CORE OPERATING LIMITS REPORT (COLR), identify that the limits for TSs 3.1.1 and 3.1.2 shall be in the COLR.

The proposed changes are consistent with the Standard Technical Specifications for Combustion Engineering Plants, NUREG-1432, Revision 3.

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 Limiting Conditions of Operation (LCOs) and Core Operating Limits Report (COLR) will continue to restrict operation to within the regions that provide acceptable results. The safety analysis will continue to be performed in accordance with the Nuclear Regulatory Commission (NRC) approved San Onofre Units 2 and 3 reload analysis methodology.

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 change does not add any new equipment, modify any interfaces with any existing equipment, alter the equipment's function, or change the method of operating the equipment. The proposed change does not alter plant conditions in a manner that could affect other plant components. The proposed change does not cause any existing equipment to become an accident initiator. Start Printed Page 24657

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

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

Response: No.

Safety Limits ensure that Specified Acceptable Fuel Design Limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences. All fuel limits and design criteria will continue to be met, based on the NRC approved San Onofre Units 2 and 3 reload analysis methodology. Therefore, the proposed change will have no impact on the margins as defined in the Technical Specification bases.

Therefore, this 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 requests involve no significant hazards consideration.

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

NRC Section Chief: Robert A. Gramm.

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

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

Date of application for amendments: December 16, 2004.

Brief description of amendments: The amendments delete TS 5.6.1, “Occupational Radiation Exposure Report” and TS 5.6.4, “Monthly Operating Reports,” as described in the Notice of Availability published in the Federal Register on June 23, 2004 (69 FR 35067).

Date of issuance: April 27, 2005.

Effective date: April 27, 2005, and shall be implemented within 90 days of the date of issuance.

Amendment Nos.: Unit 1-154, Unit 2-154, Unit 3-154.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: February 1, 2005 (70 FR 5236).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 27, 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: November 17, 2004.

Brief Description of amendments: The amendments eliminate the requirements to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: April 19, 2005.

Effective date: April 19, 2005.

Amendment Nos.: 235 and 263.

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 7763).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 19, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: November 17, 2004.

Brief description of amendment: The amendment eliminates the requirements to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: April 19, 2005.

Effective date: April 19, 2005.

Amendment No.: 204.

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

Date of initial notice in Federal Register: February 15, 2005 (70 FR 7763)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 19, 2005.

No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of application for amendment: November 17, 2004.

Brief description of amendment: This amendment revises Technical Specifications by eliminating the requirements to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: April 19, 2005.

Effective date: April 19, 2005.

Amendment No.: 118.

Facility Operating License No. NPF-63. Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: February 15, 2004 (70 FR 7763).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 19, 2005. Start Printed Page 24658

No significant hazards consideration comments received: No.

Entergy Northwest, Docket No. 50-397, Columbia Generating Station, Benton County, Washington

Date of application for amendment: September 23, 2004, as supplemented by letter dated January 13, 2005.

Brief description of amendment: The change revises Columbia Generating Station's licensing basis by replacing the current plant-specific reactor pressure vessel material surveillance program with the boiling water reactor vessels and internals project (BWRVIP) integrated surveillance program (ISP). Specifically, the amendment revises Columbia's final safety analysis report to include participation in the ISP as described in the program document BWRVIP-86-A, “BWR [Boiling Water Reactor] Vessel and Internals Project Updated BWR Integrated Surveillance Program (ISP) Implementation Plan,” dated October 2002.

Date of issuance: April 28, 2005.

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

Amendment No.: 192.

Facility Operating License No. NPF-21: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 26, 2004 (69 FR 62471).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 28, 2005.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas

Date of amendment request: December 20, 2004.

Brief description of amendment: The amendment deletes TS 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports,” as described in the Notice of Availability published in the Federal Register on June 23, 2004 (69 FR 35067).

Date of issuance: April 14, 2005.

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

Amendment No.: 223.

Renewed Facility Operating License No. DPR-51: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 18, 2005 (70 FR 2890).

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

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

Date of application for amendment: October 22, 2004.

Brief description of amendment: These amendments revise the Technical Specifications by eliminating the requirements associated with hydrogen recombiners and hydrogen monitors.

Date of issuance: April 14, 2005.

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

Amendment No.: 243 and 228.

Facility Operating License Nos. DPR-26 and DPR-64: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 1, 2005 (70 FR 5240).

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

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York

Date of application for amendment: October 25, 2004.

Brief description of amendment: These amendments revise the Technical Specifications by eliminating the requirements to submit monthly operating reports and occupational radiation exposure reports.

Date of issuance: April 14, 2005.

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

Amendment No.: 242 and 227.

Facility Operating License Nos. DPR-26 and DPR-64: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 1, 2005 (70 FR 5241).

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

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: April 30, 2004.

Brief description of amendments: The amendments modify the technical specification (TS) requirements to adopt the provisions of the industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints.”

Date of issuance: April 5, 2005.

Effective date: As of the date of issuance and shall be implemented within 180 days.

Amendment Nos.: 141, 141, 134, 134.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 26, 2004.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 5, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 15, 2004.

Brief description of amendments: The proposed amendment will delete the Technical Specification (TS) requirements related to hydrogen/oxygen monitors. The proposed TS changes support implementation of the revisions to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled 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: April 28, 2005.

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

Amendment Nos.: 213/205.

Facility Operating License Nos. DPR-19, DPR-25: The amendments revised the Technical Specifications. Start Printed Page 24659

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 April 28, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 15, 2004.

Brief description of amendments: The amendments delete the Technical Specification requirements to maintain hydrogen recombiners and hydrogen/oxygen monitors and related Surveillance Requirements. The revised Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, “Combustible Gas Control for Nuclear Power Plants,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design qualification criteria for hydrogen and oxygen monitors.

Date of issuance: April 22, 2005.

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

Amendment Nos.: 172/158.

Facility Operating License Nos. NPF-11 and NPF-18: 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 April 22, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 25, 2003.

Brief description of amendments: The amendment revised the Technical Specifications (TSs) associated with Reactor Coolant System—CHEMISTRY. Specifically, the amendment relocates Reactor Coolant System—CHEMISTRY, in its entirety from the TSs to the Technical Requirements Manual (TRM). In addition, the amendment deletes the specific activity requirements related to E-Bar, gross beta and gross gamma.

Date of issuance: April 18, 2005.

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

Amendment Nos.: 174 and 136.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments revised the TSs.

Date of initial notice in Federal Register: February 17, 2004 (69 FR 7522).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 18, 2005.

No significant hazards consideration determination comments received: No.

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

Date of application for amendment: November 17, 2004.

Brief description of amendment: The amendment eliminates the requirements to submit monthly operating reports and annual occupational radiation exposure reports.

Date of issuance: April 19, 2005.

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

Amendment No.: 217.

Facility Operating License No. DPR-72: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: February 15, 2005 (70 FR 7768).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 19, 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: June 28, 2004.

Description of amendment request: The amendment revised the Seabrook Station, Unit No. 1 Technical Specifications (TSs) to align the language of Surveillance Requirement 4.9.4 with that of Limiting Condition for Operation 3.9.4, “Containment Building Penetrations.” The amendment changes the requirement from “during core alterations and the movement of irradiated fuel” to “during the movement of recently irradiated fuel.”

Date of issuance: April 21, 2005.

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

Amendment No.: 102.

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

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53110).

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

No significant hazards consideration comments received: No.

National Aeronautics and Space Administration, Docket No. 50-30, the Plum Brook Test Reactor, Sandusky, Ohio

Date of application for amendment: January 14, 2005.

Brief description of amendment: The amendment clarifies the license requirements for confirmation of Final Status Survey results prior to backfilling or covering of excavated areas.

Date of issuance: April 21, 2005.

Effective date: The license amendment is effective as of its date of issuance.

Amendment No.: 12.

Facility License No. TR-3: This amendment consists of changes to the Facility License.

Date of initial notice in Federal Register: March 15, 2005 (70 FR 12743).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation enclosed with the amendment dated April 21, 2005.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410, Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New York

Date of application for amendments: January 24, 2005.

Brief description of amendments: The amendments deleted Sections 6.6.1 and 5.6.1, “Occupational Radiation Exposure Report,” and Sections 6.6.4 and 5.6.4, “Monthly Operating Reports,” from the NMP1 and NMP2 Technical Specifications.

Date of issuance: April 19, 2005.

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

Amendment Nos.: 188 and 115.

Facility Operating License Nos. DPR-63 and NPF-69: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: February 15, 2005 (70 FR 7769).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 19, 2005.

No significant hazards consideration comments received: No

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of application for amendment: September 22, 2004.

Brief description of amendment: The amendment extended the validity of the Start Printed Page 24660reactor pressure vessel pressure-temperature limit curves from May 1, 2005, to May 1, 2006.

Date of issuance: April 25, 2005.

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

Amendment No.: 197.

Facility Operating License No. NPF-22: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: December 7, 2004 (69 FR 70721).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 25, 2005.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

Dates of application for amendments: February 26 and April 28, 2004, as supplemented by letters dated July 8 and October 20, 2004.

Brief description of amendments: The amendments revised Technical Specification (TS) Section 5.6.6, Reactor Coolant System (RCS) Pressure Temperature Limits Report (PTLR), to facilitate future licensee-controlled changes to the PTLR. The changes include a revised PTLR that provides new heatup and cooldown limits and Cold Overpressure Protection System (COPS) setpoints, and to recalculate the minimum size of the pressurizer power operated relief valve orifice of the RCS vent. In addition, the changes relocate the COPS arming temperature to the PTLR, and lower the COPS arming temperature from 350 °F to 220 °F. The licensee also included TS bases changes to support the changes to the TSs.

Date of issuance: March 28, 2005.

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

Amendment Nos.: 136 (Unit 1) and 115 (Unit 2).

Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19575) and April 22, 2004 (69 FR 34707)

The supplements dated July 8 and October 20, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 28, 2005.

No significant hazards consideration comments received: No

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

Date of application for amendments: December 2, 2004.

Brief description of amendments: The amendments modify technical specification (TS) requirements for mode change limitations in Limiting Condition for Operation 3.0.4 and Surveillance Requirement 4.0.4 consistent with Industry/TS Task Force (TSTF) Standard TS Change Traveler, TSTF-359, Revision 9, “Increased Flexibility in Mode Restraints.” A notice of availability for this TS improvement using the Consolidated Line Item Improvement Process was published in the Federal Register (FR) on April 4, 2003 (68 FR 16579).

Date of issuance: April 11, 2005.

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

Amendment Nos.: 301, 290.

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

Date of initial notice in Federal Register: January 18, 2005 (70 FR 2901)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated April 11, 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 23, 2003, as supplemented by letter dated June 9, 2004.

Brief description of amendments: The amendments revise the Technical Specifications (TSs) to extend the interval between local leak rate tests for the containment purge and vent valves with resilient seats (containment purge valves, hydrogen purge valves, and containment pressure relief valves).

Date of issuance: April 13, 2005.

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

Amendment Nos.: 116 and 116.

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

Date of initial notice in Federal Register: November 12, 2003 (68 FR 64140).

The supplement dated June 9, 2004, 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 April 13, 2005.

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 13, 2004.

Brief description of amendment: The amendment revised Surveillance Requirement (SR) 3.8.1.7 (fast-start test), SR 3.8.1.12 (safety injection actuation signal test), SR 3.8.1.15 (hot restart test), and SR 3.8.1.20 (redundant unit test) to clarify what voltage and frequency limits are applicable during the transient and steady state portions of the diesel generator start testing performed by these SRs.

Date of issuance: April 21, 2005.

Effective date: April 21, 2005, and shall be implemented within 90 days from the date of issuance.

Amendment No.: 161.

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

Date of initial notice in Federal Register: January 18, 2005 (70 FR 2904)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 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 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, Start Printed Page 24661which 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 NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or 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 Start Printed Page 24662under 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).

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: April 11, 2005, as supplemented on April 14, 2005.

Description of amendment request: The amendments revise Technical Specification (TS) 5.5.9, “Steam Generator (SG) Tube Surveillance Program,” to incorporate changes in the SG inspection scope for Braidwood Station, Unit 2 only, during refueling outage 11.

Date of issuance: April 25, 2005.

Effective date: April 25, 2005.

Amendment Nos.: 135, 135.

Facility Operating License Nos. NPF-72 and NPF-77: Amendment revises the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Joliet Herald News, April 15 and 18, 2005, and Morris Daily Herald, April 19, 2005. The announcement provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated April 25, 2005.

Attorney for licensee: Thomas S. O'Neil.

NRC Section Chief: Gene Y Suh.

Start Signature

Dated at Rockville, Maryland, this 2nd day of May 2005.

For the Nuclear Regulatory Commission.

Ledyard B. Marsh,

Director, Division of Licensing Project Management, 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 protective order.

Back to Citation

[FR Doc. E5-2207 Filed 5-9-05; 8:45 am]

BILLING CODE 7590-01-P