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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 January 20, 2005, through February 3, 2005. The last biweekly notice was published on February 1, 2005 (70 FR 5233).

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

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

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

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

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

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

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible Start Printed Page 7763effect 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 email 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 email to pdr@nrc.gov.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina; Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina; Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendments request: November 17, 2004.

Description of amendments request: The requested change would delete Technical Specification (TS) 5.6.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “Monthly Operating Reports,” for the Brunswick and H. B. Robinson plants. The equivalent change is being requested for the Shearon Harris facility by deleting TS 6.9.1.2.a and TS 6.9.1.2.c under “Annual Reports” and TS 6.9.1.5, “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 November 17, 2004.

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 (TS) 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 Start Printed Page 7764equipment, 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: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Michael L. Marshall.

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: January 5, 2005.

Description of amendment request: The proposed amendments would revise the Technical Specification (TS) 5.5.19 associated with the Lee Combustion Turbine (LCT) testing program. TS 5.5.19.b currently requires verification that an LCT can supply the equivalent of one Unit's maximum safeguard loads, plus two Units' Mode 3 loads, when connected to the system grid every 12 months. In the proposed amendments, this requirement would be more clearly specified as “plus two Units’ safe shutdown loads.”

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) Involve a significant increase in the probability or consequences of an accident previously evaluated:

Duke proposes to revise TS 5.5.19.b to clarify the Lee Combustion Turbine (LCT) testing requirements. The proposed change makes the wording of the test requirement consistent with the UFSAR [Updated Final Safety Analysis Report] and the original wording of the TS requirement before administrative changes were made in Amendment 232, 232, 231, and Amendment 300, 300, and 300. LCT testing has no impact on the probability of an accident analyzed in the UFSAR. The LCT can be credited to mitigate the consequences of an accident analyzed in the UFSAR. However, this clarification of LCT testing requirements has no impact on its ability to mitigate the consequences of an accident. As such, the proposed LAR [license amendment request] does not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from any kind of accident previously evaluated:

Duke proposes the revise TS 5.5.19.b to clarify the Lee Combustion Turbine (LCT) testing requirements. The proposed change makes wording of the test requirement consistent with the UFSAR and the original wording of the TS requirement before administrative changes were made in Amendment 232, 232, 231, and changes were made in Amendment 300, 300, and 300. These changes do not alter the nature of events postulated in the Safety Analysis Report nor do they introduce any unique precursor mechanisms. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Involve a significant reduction in a margin of safety.

The proposed TS change does not unfavorably affect any plant safety limits, set points, or design parameters. The changes also do not unfavorably affect the fuel, fuel cladding, RCS [reactor coolant system], or containment integrity. Therefore, the proposed TS change, which clarifies TS requirements associated with the LCT testing program, does not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: July 29, 2004.

Description of amendment request: The proposed amendment would delete the requirements from the technical specifications (TS) to maintain a hydrogen dilution system, a hydrogen purge system, and hydrogen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. The revised Title 10 of the Code of Federal Regulations (10 CFR) section 50.44, “Combustible gas control for nuclear power reactors,” eliminated the requirements for hydrogen recombiners and related vent and purge systems and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors.

The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of availability of a model no significant hazards consideration determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model no significant hazards consideration determination in its application dated July 29, 2004.

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 revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The NRC has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity.

With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in RG 1.97 is intended for key variables that most directly indicate Start Printed Page 7765the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44, the NRC found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents.

The regulatory requirements for the hydrogen monitors can be relaxed 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. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines, the emergency plan, the emergency operating procedures, and site survey monitoring that support modification of emergency plan protective action recommendations.

Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from the TS, does not involve a significant increase in the probability or 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 the hydrogen recombiner [dilution/purge system for Davis Besse] requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner [dilution/purge system for Davis Besse] and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner [dilution/purge system for Davis Besse] and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides 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 hydrogen recombiner [dilution/purge system for Davis Besse] requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, 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.

The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The NRC has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage.

Category 3 hydrogen monitors are adequate 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 intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors.

Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability.

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

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

NRC Section Chief: Gene Y. Suh.

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

Date of amendment request: December 20, 2004.

Description of amendment request: The proposed change would revise Technical Specification (TS) 3/4.9.2, “Refueling Operations—Instrumentation,” concerning source range neutron flux monitors to be consistent with Improved Standard Technical Specifications.

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 changes affect the Limiting Condition for Operation [LCO] for Refueling Operations—Instrumentation, in particular, the LCO sections pertaining to the source range neutron flux detectors will be changed to be more like the corresponding sections in the Improved Standard Technical Specifications. The source range neutron flux detectors have no control functions and are therefore not accident initiators. Consequently, the proposed changes will have no impact on the probability of any accident previously evaluated. The detectors are not credited in mitigating the consequences of any accident; therefore, the proposed changes will have no impact on the consequences of any accident previously evaluated.

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

Response: No.

The changes affect the Limiting Condition for Operation for Refueling Operations—Instrumentation, in particular, the source range neutron flux detectors. The source range neutron flux detectors will continue to operate in the same manner as previously considered. Accident initial conditions and assumptions remain as previously analyzed.

The proposed changes do not introduce any new or different accident initiators. 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 changes affect the Limiting Condition for Operation for Refueling Operations—Instrumentation; in particular, the source range neutron detectors. These detectors have no control functions, and are not credited in mitigating the consequences of any accident. The source range neutron detectors are not associated with a safety limit. In addition, the proposed changes to TS will not result in design changes to the source range neutron detectors or in changes to how the source range detectors are used. Therefore, the proposed changes will not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Gene Y. Suh.

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

Date of amendment request: January 5, 2005.

Description of amendment request: The license amendment would revise Technical Specification 3/4.3.2.1, “Safety Features Actuation System [SFAS] Instrumentation,” to permit a single inoperable SFAS functional unit to be placed in a bypassed condition indefinitely. Start Printed Page 7766

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 would permit a single SFAS instrument string functional unit to be placed in bypass indefinitely. The primary function of SFAS is to monitor station conditions and actuate the engineered safety features when needed in order to prevent or limit fission product and energy release from the core, to isolate the containment vessel, and to initiate the operation of the Engineered Safety Features (ESF) equipment in the event of a loss-of-coolant accident (LOCA).

The SFAS is a possible accident initiator in that an inadvertent system level actuation could result in a transient or accident. The existing Technical Specification requirements for SFAS allow operation indefinitely with a single SFAS functional unit in trip, which results in a 1-out-of-3 channel logic. In this condition, the spurious actuation in one of the three remaining corresponding functional unit would result in an inadvertent system level actuation. Under the proposed change, indefinite operation in a 2-out-of-3, 1-out-of-3, or 1-out-of-2 channel logic would be allowed. The likelihood of a spurious system level actuation for any of the configurations allowed under the proposed change is no greater than the likelihood of spurious actuation under the 1-out-of-3 channel logic allowed under the existing Technical Specification requirements. Therefore, operation of the SFAS actuation from that permitted by the existing Technical Specifications.

Under the proposed change, the SFAS will continue to perform this function with a high level of reliability. The proposed change would allow operation of the SFAS in a condition with reduced redundancy from what is currently required by the Technical Specifications. Operation of the SFAS with reduced redundancy was evaluated against the design criteria to which the system was designed. The design criteria applicable to the SFAS, including the single failure criterion, continue to be met. The proposed change does not prevent the SFAS from mitigating the consequences of previously analyzed accidents.

The proposed change would not increase the likelihood of an inadvertent SFAS actuation. The proposed change would not prevent the SFAS from mitigating the consequences of previously analyzed accidents. 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 alter the SFAS design function or the manner in which that function is performed. Under the proposed change, the SFAS will continue to perform its function with a high degree of reliability. No new failure modes or accident initiators are created by the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed change would allow operation of the SFAS in a condition with reduced redundancy from what is currently required by the Technical Specifications. Operation of the SFAS with reduced redundancy was evaluated against the design criteria to which the system was designed. This evaluation shows that with the SFAS in the conditions permitted by the proposed change, the SFAS still satisfies all the applicable design criteria, including the single failure criterion. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Gene Y. Suh.

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

Date of amendment request: January 11, 2005.

Description of amendment request: The proposed amendment would revise the Updated Safety Analysis Report (USAR) by modifying the design requirements for protection from tornado missiles. Specifically, the proposed amendment would allow certain structures, systems, and components that are not currently provided with physical protection from tornado-induced missiles to be evaluated for acceptability based on the Electrical Power Research Institute “Tornado Missile Risk Evaluation Methodology” (TORMIS).

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 would reflect use of the Electric Power Research Institute (EPRI) Topical Report “Tornado Missile Risk Evaluation Methodology” (EPRI NP-2005), Volumes I and II. As noted in the NRC Safety Evaluation on this report dated October 26, 1983, “The current licensing criteria governing tornado missile protection are contained in Standard Review Plan (SRP) Sections 3.5.1.4 and 3.5.2. These criteria generally specify that safety-related systems be provided positive tornado missile protection (barriers) from the maximum credible tornado threat. However, SRP Section 3.5.1.4 includes acceptance criteria permitting relaxation of the above deterministic guidance, if it can be demonstrated that the probability of damage to unprotected essential safety-related features is sufficiently small.”

“Certain Operating License (OL) applicants and operating reactor licensees have chosen to demonstrate compliance with tornado missile protection criteria for certain portions of the plant * * * by providing a probabilistic analysis which is intended to show a sufficiently low risk associated with tornado missiles. Some* * * have utilized the tornado missile probabilistic risk assessment (PRA) methodology developed by” EPRI in the Topical Report listed above. The NRC noted that this report “can be utilized when assessing the need for positive tornado missile protection for specific safety-related plant features.” This methodology has subsequently been utilized in nuclear power plant licensing actions.

As permitted in NRC Standard Review Plan (NUREG-0800) sections, the total probability will be maintained below an allowable level, i.e., an acceptance criteria threshold, which reflects an extremely low probability of occurrence. The DBNPS [Davis-Besse Nuclear Power Station] approach assumes that if the probability calculation result for the total plant identifies that the cumulative probability of tornado missiles striking an unprotected portion of a safety system or component required for safe shutdown in the event of a tornado exceeds 10−6 per year, then unique missile barriers would need to be installed to lower the total probability below the acceptance criteria of 10−6 per year.

With respect to the probability of occurrence of an accident previously evaluated in the USAR, the possibility of a tornado reaching the DBNPS site and causing damage to plant structures, systems, and components is an event considered in the USAR. The changes being proposed herein do not affect the probability that the natural phenomena (a tornado) will reach the plant, but they do, from a licensing basis perspective, affect the probability that missiles generated by the winds of the tornado might strike certain plant systems or components. As recently determined, there are a limited number of safety-related components that could theoretically be struck by a tornado generated missile. The Start Printed Page 7767total (cumulative) probability of a tornado missile striking an unprotected component will be maintained below an extremely low acceptance criteria to ensure overall plant safety. Due to the extremely low probability of a tornado missile impacting an essential component, the small increase in the probability of accident initiation is not considered significant.

With respect to the consequences of an accident previously evaluated, there is an extremely low probability of a malfunction of an unprotected essential component due to tornado missile impact. Due to (1) the extremely low probability of a tornado missile striking essential equipment as calculated by TORMIS, and (2) the low probability that any tornado missile strikes would cause sufficient damage to prevent essential equipment from performing its accident-mitigating function, a loss of accident mitigation capability is not considered credible. Therefore, the radiological consequences of accidents are not significantly affected.

The proposed change is not considered to constitute a significant increase in the probability of occurrence or the consequences of an accident, due to the extremely low total probability of a tornado missile strike and thus an extremely low probability of a radiological release. Therefore, the proposed change does not involve a significant increase in the probability or consequences of previously evaluated accidents.

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

Response: No.

The possibility of a tornado reaching the DBNPS site is a design basis event considered in the USAR. This change involves recognition of the acceptability of performing tornado missile probability calculations in accordance with established regulatory guidance. The change therefore deals with an established design basis event (the tornado). Therefore, the proposed change would not contribute to the possibility of a new or different kind of accident from those previously analyzed.

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

Response: No.

This request does not involve a significant reduction in a margin of safety. The existing licensing basis for the DBNPS with respect to the design basis event of a tornado reaching the plant is to provide positive missile barriers for all systems and components required for safe shutdown in the event of a tornado. With the change, it will be recognized that there is an extremely low probability, below an established acceptance limit, that a limited subset of these systems and components could be struck. The change to missile protection based on extremely low probability (less than 1 x 10−6 per year cumulative strike probability) of occurrence of tornado generated missile strikes on portions of these systems and components is not considered to constitute a significant decrease in the margin of safety due to that extremely low probability. Therefore, the changes associated with this license amendment do not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: Gene Y. Suh.

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

Date of amendment request: September 10, 2004.

Description of amendment request: The proposed amendment would delete the requirements from the technical specifications (TS) to maintain hydrogen recombiners and hydrogen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. The revised 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,” eliminated the requirements for hydrogen recombiners and related vent and purge systems and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors.

The NRC staff issued a notice of availability of a model no significant hazards consideration determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model no significant hazards consideration determination in its application dated September 10, 2004.

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 revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity.

With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in RG 1.97 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents.

The regulatory requirements for the hydrogen monitors can be relaxed 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. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines, the emergency plan, the emergency operating procedures, and site survey monitoring that support modification of emergency plan protective action recommendations.

Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from the TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated. Start Printed Page 7768

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 the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides 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 hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, 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.

The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage.

Category 3 hydrogen monitors are adequate 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 intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors.

Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability.

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

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Gene Y. Suh.

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

Date of amendment request: November 17, 2004.

Description of amendment request: The requested change would delete Technical Specification (TS) 5.7.1.1.a, “Occupational Radiation Exposure Report,” and TS 5.7.1.2, “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 November 17, 2004.

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 (TS) 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: David T. Conley, Associate General Counsel II—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Michael L. Marshall.

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

Date of amendment request: August 6, 2004.

Description of amendment request: The proposed amendment deletes the requirements from the Technical Specifications (TSs) to maintain hydrogen recombiners and hydrogen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TSs for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors.

The NRC staff issued a notice of availability of a model no significant hazards consideration determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated August 6, 2004.

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:

Start Printed Page 7769

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

The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity.

With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in RG 1.97 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents.

The regulatory requirements for the hydrogen monitors can be relaxed 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. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines (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 the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or 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 the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides 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 [a] Margin of Safety

The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, 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.

The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage.

Category 3 hydrogen monitors are adequate 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 intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors.

Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability.

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

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

NRC Section Chief: Richard J. Laufer.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410, Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP1 and NMP2), Oswego County, New York

Date of amendment request: January 24, 2005.

Description of amendment request: The licensee proposed amendments to delete 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, respectively. 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 January 24, 2005.

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 by referencing the model NSHC analysis published by the NRC staff. The model NSHC analysis is reproduced 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. Start Printed Page 7770

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.

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 J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: October 28, 2004.

Brief description of amendments: The proposed amendment deletes the requirements from the Technical Specifications (TS) to maintain hydrogen recombiners and hydrogen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors.

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 September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated October 28, 2004.

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 revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity.

With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. Category 1 in RG 1.97 is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents.

The regulatory requirements for the hydrogen monitors can be relaxed 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. Classification of the hydrogen monitors as Category 3, and removal of the hydrogen monitors from TS will not prevent an accident management strategy through the use of the severe accident management guidelines (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 the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or 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 the hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides 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 hydrogen recombiner requirements and relaxation of the hydrogen monitor requirements, including removal of these requirements from TS, 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.

The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage.

Category 3 hydrogen monitors are adequate 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 intent of the requirements established as a result of the TMI Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors.

Therefore, this change does not involve a significant reduction in the margin of safety. Removal of hydrogen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Start Printed Page 7771

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

NRC Section Chief: Michael K. Webb (Acting).

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

Date of amendment request: November 4, 2004.

Description of amendment request: The proposed changes would relocate the inservice testing requirements, remove the inservice inspection requirements, and add a Bases Control Program to the Administrative Controls section of the Technical Specifications (TS).

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 Surry Units 1 and 2 in accordance with the proposed Technical Specifications change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change is administrative in nature, and station operations are not being affected. The ASME [American Society of Mechanical Engineers] Code requirements are established, reviewed and approved by ASME, the industry and ultimately endorsed by the NRC for inclusion into 10 CFR 50.55a. Updates to the ASME Code reflect advances in technology and consider information obtained from plant operating experience to provide enhanced inspection and testing. Thus, the proposed change only modifies TS to appropriately reference the recently NRC approved Inservice Testing Program for the fourth interval at Surry Power Station. Consequently, the probability or consequences of an accident previously evaluated are not increased.

2. The proposed Technical Specifications change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

As noted above, the proposed change is administrative in nature, and no new accident precursors are being introduced. Since the inservice testing will continue to be performed in accordance with an NRC approved program, adequate assurance is provided to ensure the safety-related pumps and valves would operate as required. No new testing is required that could create a new or different type of accident. Consequently, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed Technical Specifications change does not involve a significant reduction in a margin of safety.

Performing inservice testing of pumps and valves to the NRC approved program for the fourth interval at Surry Power Station provides adequate assurance that the safety-related pumps and valves will continue to perform their intended safety function. This is an administrative change in nature and as such does not involve a significant reduction in the margin of safety.

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

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

NRC Section Chief: John A. Nakoski.

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

The following notices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. They are repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant (WBN), Unit 1, Rhea County, Tennessee

Date of amendment request: September 15, 2004.

Brief description of amendment request: In accordance with Technical Specification Task Force (TSTF) 285, Charging Pump Swap Low-Temperature Over-Pressurization Allowance, LCO 3.4.12, Cold Overpressure Mitigation System (COMS), is being revised to modify and relocate two notes in the WBN Technical Specifications. The changes are all administrative, except a change which would allow two charging pumps to be made capable of injecting into the Reactor Coolant System to support pump swap operations for a period not to exceed one hour instead of the currently allowed 15 minutes.

Date of publication of individual notice in Federal Register: February 1, 2005 (70 FR 5226).

Expiration date of individual notice: March 3, 2005 (public comments) and April 4, 2005 (hearing requests).

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, Start Printed Page 7772(301) 415-4737 or by e-mail to pdr@nrc.gov.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of application of amendment: September 30, 2003.

Description of amendment request: The amendment modifies Technical Specification (TS) 4.3.1, “Criticality,” adds TS 3.7.16, “Spent Fuel Pool Boron Concentration,” and adds TS 3.7.17, “Spent Fuel Pool Storage.” Specifically, the amendment increases the maximum enrichment limit of the fuel assemblies that can be stored in the Unit 2 spent fuel pool by taking credit for soluble boron, burnup, and configuration control in maintaining acceptable margins of subcriticality.

Date of issuance: January 27, 2005.

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

Amendment No.: 246.

Renewed License No. DPR-69: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 20, 2004 (69 FR 2739).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 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: July 26, 2004, as supplemented January 26, 2005.

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

Date of issuance: February 2, 2005.

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

Amendment Nos.: 234 and 261.

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

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53100). The January 26, 2005, supplement contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 2, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 25, 2003, as supplemented June 9, and July 30, 2003, and September 13, 2004.

Brief description of amendments: The amendments revised the Technical Specifications to incorporate a Steam Generator (SG) program that defines a performance-based approach to maintaining SG tube integrity. The SG program includes performance criteria that define the basis for tube integrity and provides reasonable assurance that SG tubing will remain capable of fulfilling its safety function of maintaining reactor coolant system pressure boundary integrity. The proposed amendments add a new TS for SG tube integrity (3.4.18) and revise the TS for reactor coolant operation leakage (3.4.13), SG tube surveillance program (5.5.9), and SG tube inspection report (5.6.8).

Date of issuance: January 13, 2005.

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

Amendment Nos.: 218 and 212.

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

Date of initial notice in Federal Register: July 8, 2003 (68 FR 40712).

The supplements dated June 9, and July 30, 2003, and September 13, 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 January 13, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 18, 2004.

Brief description of amendments: The amendments revised the Technical Specifications to remove references to Safety Injection Steam Line Pressure-Low.

Date of issuance:

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

Amendment Nos.: 224 and 206.

Renewed Facility Operating License Nos. NPF-9 and NPF-17: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 2004 (69 FR 64987).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 27, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: August 15, 2003, as supplemented on April 9, 2004.

Brief description of amendments: The amendments revise Technical Specification (TS) 3.4.15, “RCS Leakage Detection Instrumentation”, to require one containment sump monitor and one containment atmosphere particulate radioactivity monitor to be operable in Modes 1, 2, 3, and 4.

Date of issuance: January 14, 2005.

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

Amendment Nos.: 140, 133.

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 28, 2003 (68 FR 61477).

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

No significant hazards consideration comments received: No.

Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: December 2, 2003, as supplemented by letters dated September 14 and December 10, 2004, and January 7, 2005.

Brief description of amendment: This amendment revised the Technical Specifications (TSs) to permit operation Start Printed Page 7773with a reduced reactor coolant system flow corresponding to a steam generator (SG) tube plugging level of 30-percent per SG. This amendment also includes the transition to Westinghouse Reload Safety Evaluation Methodology (WCAP-9272).

Date of issuance: January 31, 2005.

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

Amendment No.: 138.

Renewed Facility Operating License No. NPF-16: Amendment revised the TSs.

Date of initial notice in Federal Register: March 18, 2004 (69 FR 12873).

The September 14 and December 10, 2004, and January 7, 2005, supplements did not affect the original proposed no significant hazards determination, or expand the scope of the request as noticed in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 19, 2004, as supplemented on July 16, 2004.

Brief description of amendment: The amendment revised Section 3/4.6.2, “Protective Instrumentation,” to establish a 24-month operating cycle calibration frequency for the intermediate range monitor instrumentation. In addition, the amendment authorized relocation of the limiting conditions for operation and surveillance requirements for certain control rod withdrawal block instruments from Section 3/4.6.2 to the Updated Final Safety Analysis Report.

Date of issuance: January 25, 2005.

Effective date: January 25, 2005.

Amendment No.: 186.

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

Date of initial notice in Federal Register: May 25, 2004 (69 FR 29769).

The July 16, 2004, letter provided clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination.

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

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa

Date of application for amendment: October 5, 2004.

Brief description of amendment: The amendment deletes Technical Specification (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: January 31, 2005.

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

Amendment No.: 256.

Facility Operating License No. DPR-49: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 2004 (69 FR 64989).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: January 30, 2004.

Brief description of amendment: The amendment changes the Technical Specifications (TSs) to (1) clarify the permissive setpoint for the source range monitor detector-not-fully-inserted rod block bypass, (2) correct a typographical error in the surveillance requirement for suppression pool temperature monitoring, (3) clarify the setpoint for the pressure suppression chamber-reactor building vacuum breakers instrumentation, (4) clarify the operating force requirements for the pressure suppression chamber-drywell vacuum breakers surveillance test, and (5) make corrections resulting from license Amendments 130 and 132.

Date of issuance: January 28, 2005.

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

Amendment No.: 141.

Facility Operating License No. DPR-22. Amendment revised the TSs.

Date of initial notice in Federal Register: April 13, 2004 (69 FR 19573).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 5, 2004.

Brief description of amendment: The amendment deletes technical specification (TS) 6.7.A.2, “Requirement to submit an Occupational Radiation Exposure Report,” TS 6.7.A.3, “Requirement to submit a Monthly Operating Report,” and TS 6.7.A.6, “Requirement to report safety/relief valve failures and challenges” as described in the Notice of Availability published in the Federal Register on June 23, 2004 (69 FR 35067).

Date of issuance: February 1, 2005.

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

Amendment No.: 142.

Facility Operating License No. DPR-22. Amendment revised the TSs.

Date of initial notice in Federal Register: November 9, 2004 (69 FR 64989).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 1, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 5, 2004.

Brief description of amendment: The amendment deletes technical specification 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: January 10, 2005.

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

Amendment No.: 220.

Facility Operating License No. DPR-20: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: November 9, 2004 (69 FR 64989).

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

No significant hazards consideration comments received: No. Start Printed Page 7774

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

Date of application for amendment: January 30, 2004.

Brief description of amendment: The amendment eliminates requirements for hydrogen recombiners and relocates the requirements for hydrogen monitors to the licensee's Commitment Management Program.

Date of issuance: January 11, 2005.

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

Amendment No.: 221.

Facility Operating License No. DPR-20: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: March 2, 2004 (69 FR 9862).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 11, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: October 5, 2004.

Brief description of amendments: The amendments delete technical specification (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: January 31, 2005.

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

Amendment Nos.: 168, 158.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 2004 (69 FR 64989).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 31, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: February 13, 2004, as supplemented by letters dated November 5 and December 10, 2004.

Brief description of amendments: The amendments revise Technical Specifications (TSs) 3.3.1, “Reactor Trip System (RTS) Instrumentation,” 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” and 3.3.6, “Containment Ventilation Isolation Instrumentation,” to adopt the completion time, test bypass time, and surveillance frequency time changes approved by the NRC in Topical Reports WCAP-14333-P-A, “Probabilistic Risk Analysis of the RPS [reactor protection system] and ESFAS Test Times and Completion Times,” and WCAP-15376-P-A, “Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times.” The amendments revise the required actions for certain action conditions; increase the completion times for several required actions (including some notes); delete notes in certain required actions; and increase frequency time intervals (including certain notes) in several surveillance requirements.

Date of issuance: January 31, 2005.

Effective date: January 31, 2005, and shall be implemented within 180 days of the date of issuance.

Amendment Nos.: Unit 1—179; Unit 2—181.

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

Date of initial notice in Federal Register: March 30, 2004 (69 FR 16622). The supplemental letters dated November 5 and December 10, 2004, provided clarifying information that did not change the scope of the original application as noticed or the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 31, 2005.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County, Georgia

Date of application for amendments: April 26, 2004, as supplemented by letters dated August 17 and September 7, 2004.

Brief description of amendments: The amendment revised the Technical Specification Section 5.5.12, “Primary Containment Leakage Rate Testing Program” to reflect a one-time deferral of the Type A Containment Integrated Leak Rate Test (ILRT). This change extends the 10-year interval between ILRTs to 15 years from the previous ILRT.

Date of issuance: February 1, 2005.

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

Amendment No.: 187.

Renewed Facility Operating License No. NPF-5: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 3, 2004 (69 FR 46591).

The supplements dated August 17 and September 7, 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 February 1, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: December 17, 2003, as supplemented by letters dated October 28 and November 16, 2004.

Brief description of amendment: The amendment revises TSs 3.3.1, “Reactor Trip System (RTS) Instrumentation,” 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” and 3.3.9, “Boron Dilution Mitigation System (BDMS)” to adopt the completion time, test bypass time, and surveillance time interval changes in NRC-approved WCAP-14333-P-A, “Probabilistic Risk Analysis of the RPS [reactor protection system] and ESFAS Test Times and Completion Times,” and WCAP-15376-P-A, “Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times.” The TS changes revise required actions for certain action conditions; increase the completion times for several required actions (including some notes); delete notes in certain required actions; increase frequency time intervals (including certain notes) in several surveillance requirements (SRs); add an action condition and required actions; revise notes in certain SRs; and revise Table 3.3.2-1. There is also an administrative correction to the format of the TSs. Start Printed Page 7775

Date of issuance: January 31, 2005.

Effective date: January 31, 2005, and shall be implemented within 120 days of its date of issuance.

Amendment No.: 165.

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

Date of initial notice in Federal Register: February 3, 2004 (69 FR 5211).

The supplemental letters dated October 28 and November 16, 2004, provided clarifying information that did not change the scope of the original application as noticed or the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 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 15, 2003, as supplemented by letters dated October 7 and November 12, 2004.

Brief description of amendment: The amendment revises Technical Specifications (TSs) 3.3.1, “Reactor Trip System (RTS) Instrumentation,” and 3.3.2, “Engineered Safety Feature Actuation System (ESFAS) Instrumentation,” to adopt the completion time, test bypass time, and surveillance frequency time changes approved by the NRC in Topical Reports WCAP-14333-P-A, “Probabilistic Risk Analysis of the RPS [reactor protection system] and ESFAS Test Times and Completion Times,” and WCAP-15376-P-A, “Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times.” The amendment revises the required actions for certain action conditions; increase the completion times for several required actions (including some notes); delete notes in certain required actions; and increase frequency time intervals (including certain notes) in several surveillance requirements.

Date of issuance: January 31, 2005.

Effective date: January 31, 2005, and shall be implemented within 180 days of the date of issuance.

Amendment No.: 156.

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

Date of initial notice in Federal Register: February 3, 2004 (69 FR 5212).

The supplemental letters dated October 7 and November 12, 2004, provided clarifying information that did not change the scope of the original application as noticed or the NRC staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 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: October 7, 2004.

Brief description of amendment: The amendment revises Section 5.3, “Unit Staff Qualifications,” of the technical specifications (TSs) to add the qualification requirements for the shift manager and the control room supervisor. In addition, based on a comparison review performed by the NRC and Wolf Creek Nuclear Operating Corporation personnel, editorial corrections are being made to the TSs.

Date of issuance: January 31, 2005.

Effective date: January 31, 2005, and shall be implemented within 90 days from the date of issuance.

Amendment No.: 159.

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

Date of initial notice in Federal Register: November 23, 2004 (68 FR 68188).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 31, 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, which are set forth in the license amendment.

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

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

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

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

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant Start Printed Page 7776hazards 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 email 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 under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

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

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

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

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

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

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

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Start Printed Page 7777Office 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 email 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).

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

Date of amendment request: January 15, 2005.

Description of amendment request: The amendment revises the Operating License to add a license condition to allow a one-time extension of the allowed outage time for the west centrifugal charging pump.

Date of issuance: January 16, 2005.

Effective date: January 16, 2005.

Amendment No.: 285.

Facility Operating License No. DPR-58: Amendment revises the Operating License.

Public comments requested as to proposed no significant hazards consideration (NSHC): No. The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated January 16, 2005.

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: M. Kotzalas, Acting.

Start Signature

Dated at Rockville, Maryland, this 7th day of February, 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. 05-2788 Filed 2-14-05; 8:45 am]

BILLING CODE 7590-01-P