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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 February 4, 2005, through February 17, 2005. The last biweekly notice was published on February 15, 2005 (70 FR 7762).

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 01F21, 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 Start Printed Page 9987for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

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

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

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

Date of amendment request: June 24, 2004.

Description of amendment request: The proposed amendment would revise Surveillance Requirement (SR) 4.0.2 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period would be extended from the current limit of “ * * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “ * * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement would be added to SR 4.0.2: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.” In addition, a Technical Specifications (TSs) Bases Control Program would be adopted as new TS 6.18.

Basis for proposed no significant hazards consideration determination: The NRC staff issued a notice of Start Printed Page 9988opportunity for comment in the Federal Register on June 14, 2001 (66 FR 32400), on possible amendments concerning missed surveillances, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on September 28, 2001 (66 FR 49714). The licensee affirmed the applicability of the following NSHC determination in its application dated June 24, 2004.

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

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

The proposed change relaxes the time allowed to perform a missed surveillance and adds a Bases Control Program. The time between surveillances is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The equipment being tested is still required to be operable and capable of performing the accident mitigation functions assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly affected. Any reduction in confidence that a standby system might fail to perform its safety function due to a missed surveillance is small and would not, in the absence of other unrelated failures, lead to an increase in consequences beyond those estimated by existing analyses. The addition of a requirement to assess and manage the risk introduced by the missed surveillance will further minimize possible concerns. The addition of a new Section 6.18 to add a Bases Control Program has no effect on the operation or testing of any plant equipment and would not affect any accident initiator. The addition of a Bases Control Program is administrative in nature, and would not affect the probability or consequences of an accident. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. A missed surveillance will not, in and of itself, introduce new failure modes or effects and any increased chance that a standby system might fail to perform its safety function due to a missed surveillance would not, in the absence of other unrelated failures, lead to an accident beyond those previously evaluated. The addition of a requirement to assess and manage the risk introduced by the missed surveillance will further minimize possible concerns. The addition of a Bases Control Program is administrative in nature, and will not create any new accident initiators. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The extended time allowed to perform a missed surveillance does not result in a significant reduction in the margin of safety. As supported by the historical data, the likely outcome of any surveillance is verification that the LCO [Limiting Condition for Operation] is met. Failure to perform a surveillance within the prescribed frequency does not cause equipment to become inoperable. The only effect of the additional time allowed to perform a missed surveillance on the margin of safety is the extension of the time until inoperable equipment is discovered to be inoperable by the missed surveillance. However, given the rare occurrence of inoperable equipment, and the rare occurrence of a missed surveillance, a missed surveillance on inoperable equipment would be very unlikely. This must be balanced against the real risk of manipulating the plant equipment or condition to perform the missed surveillance. In addition, parallel trains and alternate equipment are typically available to perform the safety function of the equipment not tested. Thus, there is confidence that the equipment can perform its assumed safety function. The addition of a Bases Control Program is administrative in nature, serves to ensure that changes to the Bases are made in accordance with approved criteria, and will not have a significant affect on the margin of safety.

Therefore, this change does not involve a significant reduction in a margin of safety. Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

Attorney for licensee: Thomas S. O'Neill, Associate General Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Richard J. Laufer.

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

Date of amendment request: October 15, 2004.

Description of amendment request: The proposed amendment revises surveillance requirements related to the reactor coolant pump flywheel inspections to extend the allowable inspection interval to 20 years.

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated October 15, 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) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's acceptance guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits; or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2) The proposed change does not create the possibility of a new or different kind of Start Printed Page 9989accident from any accident previously evaluated.

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) The proposed change does not involve a significant reduction in a margin of safety.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: 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, Jr.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck Plant, Middlesex County, Connecticut

Date of amendment request: December 1, 2004.

Description of amendment requests: The Haddam Neck Plant (HNP) is currently undergoing active decommissioning. The proposed amendment would revise the License Termination Plan (LTP) to revise the buried debris dose model and surface contamination release limits for various piping sizes. Specifically CYAPCO proposes to:

1. Modify the dose model for volumetrically contaminated concrete, rebar (hereafter referred to as simply “concrete”), the containment liner and embedded piping in basements that are to remain in place at the HNP site. The revised approach results in the offsite disposal of a larger percentage of the concrete structures (approximately 75% of that which would remain under the current approach). The overall effect results in a smaller amount of radioactivity contained in concrete to remain on-site than is allowed by the current LTP. The method of calculating the future groundwater pathway dose using the concrete debris model is being revised to an inventory based approach which will include activity inventories from the containment liner, embedded piping inside surfaces and radioactivity released from volumetrically contaminated concrete (which is controlled by diffusion rate through basement walls and flowable fill). The concrete that will remain is in the containment lower walls and floor mat, the in-core instrumentation sump, and the lower walls and floor of the spent fuel pool in the fuel building. The Basement Fill Model will also be used for other basements and footings that will remain on site using the results of future characterization surveys.

2. Additionally, CYAPCO proposes to include surface contamination release levels for other pipe diameters that may be encountered during the decommissioning beyond that currently included in the LTP for 4 inch piping.

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:

In accordance with 10 CFR 50.92, CYAPCO has reviewed the amendment request and concluded that the amendment request does not involve a Significant Hazards Consideration (SHC). The basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised. The amendment request does not involve an SHC because the amendment request would not:

A. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The activities included in the amendment request are within the bounds of those contained in the HNP Updated Final Safety Analysis Report (UFSAR). The HNP UFSAR Chapter 15 provides a discussion of the radiological events postulated to occur as a result of decommissioning activities with bounding consequences resulting from a resin container accident. This accident is expected to contain more potential airborne activity than can be released from other decommissioning events. The radionuclide distribution assumed for the resin container has a greater inventory of transuranics radionuclides (major dose contributor) than the distribution of plant derived radionuclides in the components involved in other decommissioning activities. The HNP UFSAR also discusses a fuel handling accident in the fuel building, involving the drop of a spent fuel assembly onto the fuel racks. The postulated drop assumes the rupture of all fuel rods in the associated assembly. The probability or consequences of this accident would not be increased during any future fuel operations in the spent fuel building related to decommissioning. Transfer of the spent fuel to canisters for dry cask storage involves additional restrictions contained in the cask certificate of compliance in order to maintain decommissioning activities within the assumptions of and consequences of the fuel handling accident. No systems, structures, or components that could initiate or be required to mitigate consequences of an accident are affected by the amendment request in any way not previously evaluated in the HNP UFSAR. Therefore, the amendment request does not involve any increase in the probability or consequences of any accident previously evaluated.

B. Create the possibility of a new or different kind of accident from any accident previously evaluated.

Accident analyses related to decommissioning activities are addressed in the HNP UFSAR. The activities included in the amendment request are within the bounds of those considered in the HNP UFSAR. Thus, the amendment request does not affect plant systems, structures, or components in any way previously evaluated in the HNP UFSAR. The amendment request does not introduce any new failure modes. Therefore, the amendment request will not create the possibility of a new or different kind of accident from any previously evaluated.

C. Involve a significant reduction in a margin of safety.

The HNP LTP is a plan for demonstrating compliance with radiological criteria for license termination as provided in 10 CFR 20.1402. The margin of safety defined in the statements of consideration for the final rule on the Radiological Criteria for License Termination is described as the margin between 100 mrem/yr public dose limit established in 10 CFR 20.1301 for licensed operation and the 25 mrem/yr dose limit to the average member of the critical group at a site considered acceptable for unrestricted use (one of the criteria of 10 CFR 20.1402). This margin of safety accounts for the potential effects of multiple sources of radiation exposure to the critical group. Since the HNP LTP was designed to comply with the radiological criteria for license termination for unrestricted use, this license amendment request supports this margin of safety. Also, as previously discussed, the bounding accident for decommissioning is the resin container accident. Since the bounding decommissioning accident results in more airborne radioactivity than can be released from the other decommissioning events, the margin of safety associated with consequences of decommissioning accidents is not reduced by this amendment request. Thus, the amendment request does not involve a significant reduction in the margin of safety.

Start Printed Page 9990

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.

NRC Section Chief: Claudia Craig.

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

Date of amendment request: January 19, 2005.

Description of amendment request: The proposed amendments would revise the McGuire, Units 1 and 2, Technical Specification (TS) 5.6.5.b to add an NRC-approved Topical Report to the list of analytical methods used to determine core operating limits.

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:

Criterion 1—Does this LAR Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated?

No. This LAR makes an administrative change to Technical Specification (TS) 5.6.5.b, “Core Operating Limits Report (COLR).” This TS contains a listing of documents (analytical methods) that are used to determine core operating limits. These documents are separately and individually reviewed and approved by the NRC. The current LAR adds a new document, DPC-NE-1005P-A, “Duke Power Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX,” (DPC Proprietary), to the list in TS 5.6.5.b. Topical Report “DPC-NE-1005P-A” has been previously reviewed by the NRC and determined to be appropriate for use at McGuire. The NRC's determination was documented in a safety evaluation report dated August 20, 2004. Based on these considerations, it has been determined that the proposed administrative change has no impact on any accident probabilities or consequences.

Criterion 2—Does This LAR Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated?

No. This LAR is solely administrative in nature since it only adds an NRC-approved licensing basis document to the TS. No new accident causal mechanisms will be created as a result of the NRC approval of this LAR.

Criterion 3—Does This LAR Involve a Significant Reduction in a Margin of Safety?

No. This LAR is solely administrative in nature. The analytical methodologies used to generate the core operating limits are separately and individually reviewed and approved by the NRC, and are unchanged by this LAR. The change contained in this LAR merely revises the McGuire TS in an administrative manner in order to conform with a Duke licensing action that has been previously approved by the NRC. Therefore the change proposed in this amendment request has no impact on 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. Lisa F. Vaughn, Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 28201-1006.

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: December 20, 2004.

Description of amendment request: The requested change will delete Technical Specification (TS) 5.5.1, “Occupational Radiation Exposure Report,” and TS 5.6.4, “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 December 20, 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 (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

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

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

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Allen G. Howe.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: December 14, 2004.

Description of amendment request: The proposed amendment would eliminate certain administrative requirements for safety limit violations that are adequately addressed in 10 CFR 50.36(c)(1)(i)(A), 10 CFR 50.72, 10 CFR 50.73, and by procedures; replace plant-specific titles with generic titles; remove the remaining responsibilities of the Operations Review Committee; replace descriptive details specified in Technical Specification (TS) 3.13.A.1 associated with 10 CFR 50.55a(f), “Inservice Testing Requirements,” with reference to the “Inservice Code Testing Program”; make administrative changes to TS 5.5.4, “Radioactive Effluent Controls Program”; and make editorial corrections and clarifications.

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:

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing Start Printed Page 9991on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed 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 is administrative in nature and does not involve the modification of any plant equipment or affect basic plant operation. There is no impact to any accident previously evaluated.

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

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

Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

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

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

Response: No. The proposed change represents the relocation of specific Technical Specification requirements, based on regulatory guidance and previously approved changes for other stations or deletion of detail redundant to regulations or no longer applicable (i.e., expired one-time exceptions). The proposed change is administrative in nature, does not negate or revise any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the Technical Specifications. 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: J.M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 02360-5599.

NRC Section Chief: Darrell Roberts.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: December 14, 2004.

Description of amendment request: The proposed amendment would remove the additional requirement to perform functional testing of the Average Power Range Monitor (APRM) and Anticipated Transient Without Scram Recirculation Pump Trip Alternate Rod Insertion instrumentation on each startup, even when the nominally required quarterly testing is current. Additionally, performance of the APRM High Flux heat balance calibration is modified to apply only after 12 hours at >25% power. Additional editorial clarifications related to Table 4.2.A through 4.2.G, Note 2 and associated Table references are also proposed.

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

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

Response: No. The proposed changes to eliminate startup-related functional testing, even when the nominally required quarterly testing is current, will not result in a significant increase in the probability or consequences of an accident previously evaluated because there is no change to the requirement that the instrument channels remain operable and are periodically tested throughout the time that the associated function is required. The surveillance continues to be performed at the normal frequency and the normal surveillance frequency has been shown, based on operating experience, to be adequate for assuring that required conditions are established and maintained.

Delaying the APRM [Average Power Range Monitor] heat balance calibration until conditions allow for accurate results will not result in a significant increase in the probability or consequences of an accident previously evaluated because there is no change to the requirement that the instrument channels remain operable. The ability of the APRMs to adequately respond to power excursions from < 25% that assume an APRM trip at 120% is not significantly impacted by deferring the APRM-to-heat balance calibration from the currently required 15% power, until the proposed 12 hours after ≥ 25% power. Additional editorial changes have no technical or operational impact.

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 any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

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

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

Response: No. The proposed changes do not negate any existing equipment or system performance requirements, and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change.

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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 02360-5599.

NRC Section Chief: Darrell Roberts.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: December 14, 2004.

Description of amendment request: The proposed amendment would relocate various requirements from the Technical Specification (TS) to the Final Safety Analysis Report (FSAR) or TS Bases.

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 Start Printed Page 9992consequences of an accident previously evaluated?

Response: No. The proposed relocations are administrative in nature and do not involve the modification of any plant equipment or affect basic plant operation. The associated instrumentation and inspections are not assumed to be an initiator of any analyzed event, nor are these limits assumed in the mitigation of consequences of 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 involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No. The proposed changes to relocate current TS requirements to the FSAR, consistent with regulatory guidance and previously approved changes for other stations, are administrative in nature. These changes do not negate any existing requirement, and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the Technical Specifications to the FSAR. 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: J.M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 02360-5599.

NRC Section Chief: Darrell Roberts.

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

Date of amendment request: December 17, 2004.

Description of amendment request: The proposed amendment would delete the Technical Specification (TS) requirements to submit monthly operating reports and occupational radiation exposure reports.

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in licensing 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 December 17, 2004.

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 NSHC, 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 eliminates the Technical Specifications (TSs) reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No

The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No

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

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

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: October 15, 2004.

Description of amendment request: The proposed amendment revises surveillance requirements related to the reactor coolant pump (RCP) flywheel inspections to extend the allowable inspection interval to 20 years.

The NRC staff issued a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 24, 2003 (68 FR 37590). The notice of availability of the model application was issued on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated October 15, 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 proposed change to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's acceptance guidelines [contained] in RG [Regulatory Guide] 1.174 (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel Start Printed Page 9993is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits; or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: 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, Jr.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, Unit No. 1, St. Lucie County, Florida

Date of amendment request: December 20, 2004.

Description of amendment request: The proposed license amendment would extend the effectiveness of the current Technical Specification pressure/temperature (P/T) limit curves, also called the heatup and cooldown curves, from 23.6 to 35 effective full power years (EFPY). The low temperature overpressure protection requirements, which are based on the P/T limits, would also be extended to 35 EFPY. The proposed amendment would revise Technical Specification Figures 3.1-1b, 3.4-2a, 3.4-2b, and 3.4-3.

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

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

The pressure/temperature (P/T) limit curves in the Technical Specifications are conservatively generated in accordance with the fracture toughness requirements of 10 CFR 50, Appendix G, as supplemented by the ASME [American Society of Mechanical Engineers] Code [Boiler and Pressure Vessel Code], Section Xl, Appendix G recommendations. The adjusted reference temperature (ART) values are based on the Regulatory Guide 1.99, Revision 2, shift prediction and attenuation formula and have been validated by a credible reactor vessel surveillance program. There are no changes to the limit curve, only a change in the period of applicability based on more recent fluence predictions and new best estimate chemistry information. Based on the current fluence projections, analysis has demonstrated that the current P/T limit curves will remain conservative for up to 35 EFPY.

In conjunction with extending the effectiveness of the existing P/T limit curves, the low temperature overpressure protection (LTOP) analysis for 23.6 EFPY is also extended to 35 EFPY. The LTOP analysis confirms that the current setpoints for the power operated relief valves (PORVs) will provide the appropriate overpressure protection at low reactor coolant system (RCS) temperatures. Because the P/T limit curves have not changed, the existing LTOP values have not changed, which include the PORV setpoints.

The P/T limit curves and LTOP analysis have not changed; therefore, the proposed amendment does not represent a change in the configuration or operation of the plant. The results of the existing LTOP analysis have not changed, and the limiting pressures for given temperatures will not be exceeded for the postulated transients. Therefore, assurance is provided that reactor vessel integrity will be maintained. Thus, the proposed amendment does not involve an increase in the probability or consequences of accidents previously evaluated.

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

The requirements for P/T limit curves and LTOP have been in place since the beginning of plant operation. The only changes in these curves are the extension of the period of applicability (EFPY), which is based on new fluence data and the operating time (EFPY) required to reach the same limiting adjusted reference temperature projection used for the current 23.6 EFPY P/T limit curves. Since there is no change in the configuration or operation of the facility as a result of the proposed amendment, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Analysis has demonstrated that the fracture toughness requirements of 10 CFR 50, Appendix G, are satisfied and that conservative operating restrictions are maintained for the purpose of low temperature overpressure protection. The P/T limit curves will provide assurance that the RCS pressure boundary will behave in ductile manner and that the probability of a rapidly propagating fracture is acceptably low. Therefore, operation in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Michael L. Marshall, Jr.

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

Date of amendment request: January 6, 2005.

Description of amendment request: The proposed amendment revises Start Printed Page 9994Technical Specification Section 3/4.4.5, Steam Generators, to allow repair of steam generator tubes by installing Westinghouse Electric LLC Alloy 800 leak limiting sleeves.

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?

No, the leak limiting Alloy 800 tube sleeves are designed using the applicable ASME [American Society of Mechanical Engineers] Boiler and Pressure Vessel Code and meet the design objectives of the original steam generator tubing. The applied stresses and fatigue usage factors for the sleeves are bounded by the limits established in the ASME Code. Mechanical testing has shown that the structural strength of leak limiting sleeves under normal, upset, emergency, and faulted conditions provides margin to the acceptance limits. These acceptance limits bound the most limiting burst margin of three times the normal operating pressure differential as recommended by NRC [U.S. Nuclear Regulatory Commission] Regulatory Guide 1.121. Burst testing of sleeved-tube assemblies has confirmed the analytical results and demonstrated that levels of primary-to-secondary leakage are not expected to exceed acceptable levels during any anticipated plant operating condition.

The leak limiting Alloy 800 sleeve depth-based structural limit is determined using NRC guidance and the pressure-stress equation of the ASME Code, Section III with margin added to account for the configuration of long axial cracks. An Alloy 800 sleeved tube will be plugged on detection of an imperfection in the sleeve or in the pressure boundary portion of the original tube wall.

An evaluation of repaired steam generator tubes, plus testing, and analysis indicates that unacceptable detrimental effects on the leak limiting Alloy 800 sleeve or of a sleeved tube are not expected from the reactor coolant system flow, primary or secondary coolant chemistries, thermal conditions or transients, or pressure conditions as may be experienced at St. Lucie Unit 2. Corrosion testing and historical performance of sleeved steam generator tubes indicates no evidence of sleeve or tube corrosion considered detrimental under anticipated service conditions. The implementation of the proposed tube sleeving has no significant effect on either the configuration of the plant or the manner in which it is operated.

The consequences of a hypothetical failure of a leak limiting Alloy 800 sleeved tube is bounded by the current steam generator tube rupture analysis described in the St. Lucie Unit 2 Updated Final Safety Analysis Report. Due to the slight reduction in the inside diameter caused by the sleeve wall thickness, primary coolant release rates through the parent tube during a tube rupture event would be slightly less than that assumed for the steam generator tube rupture analysis and therefore, would result in lower total primary fluid mass release to the secondary system. A main steam line break or feedwater line break will not cause a steam generator tube rupture since the sleeves are analyzed for a maximum accident differential pressure greater than that predicted in the St. Lucie Unit 2 safety analysis.

Fluid leakage from a sleeved tube during plant operation would be minimal and is well within the allowable Technical Specification leakage limits. Therefore, the proposed tube sleeving 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?

No, the leak limiting Alloy 800 sleeves are designed using the applicable ASME Code as guidance, and therefore, meet the objectives of the original steam generator tubing. As a result, the function of the steam generator will not be significantly affected by the installation of the proposed sleeves. The proposed sleeves do not interact with any other plant systems. Any accident that would result from potential tube or sleeve degradation in the repaired portion of the tube is bounded by the existing steam generator tube rupture accident analysis, thus the potential for a new type of accident is not created. The continued integrity of the sleeved tube is periodically verified by surveillance inspections performed in compliance with Technical Specification requirements. A sleeved tube will be plugged on detection of any service induced imperfection, degradation, or defect in the sleeve and/or pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint).

Implementation of the proposed change has no significant effect on either the configuration of the plant or the manner in which it is operated. Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

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

No, the repair of degraded steam generator tubes with leak limiting Alloy 800 sleeves restores the structural integrity of the degraded tube under normal operating and postulated accident conditions. The reduction in core cooling margin due to the addition of Alloy 800 sleeves is not significant because the cumulative effect of all sleeved and plugged tubes will continue to be less than the currently-allowed core cooling margin threshold established by the total steam generator tube plugging level. Design safety factors utilized for the sleeves are consistent with the safety factors in the ASME Boiler and Pressure Vessel Code used in the original steam generator design. Each tube and portions of the tube with an installed sleeve that constitute the reactor coolant pressure boundary will be monitored; a sleeved tube will be plugged on detection of any service induced imperfection, degradation, or defect in the sleeve and/or pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint). Use of the previously-identified design criteria and design verification testing assures that the margin to safety is not significantly different from that of the original steam generator tubes. Therefore, the proposed repairs employing leak limiting Alloy 800 tube sleeves do not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Michael L. Marshall, Jr.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: September 8, 2004.

Description of amendment request: The proposed amendments would change SSES 1 and 2 Technical Specifications 3.6.4.1, “Secondary Containment,” and 3.6.4.3, “Standby Gas Treatment System (SGTS),” to extend, on a one-time basis, the allowable completion time for required actions for secondary containment inoperable and two SGTS subsystems inoperable, in mode 1, 2, or 3, from 4 hours to 48 hours.

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 does not involve a significant increase in the probability of an accident previously evaluated because neither Secondary Containment nor the Standby Gas Treatment System is an initiator of an accident. Both mitigate accident consequences.

The consequences of a Design Basis Analysis-Loss of Coolant Accident (DBA-LOCA) have been evaluated in the FSAR [Final Safety Analysis Report]. Increasing the completion time for Secondary Containment and two SGTS subsystems inoperable from 4 Start Printed Page 9995hours to 48 does not result in a significant increase in the consequences of a DBA-LOCA event nor change the evaluation of DBA-LOCA events as stated in the FSAR evaluation. The radiological evaluation of DBA-LOCA doses, including doses offsite, Control Room habitability, and exposures for personnel access demonstrates that there would be no significant impact. Movement of irradiated fuel within Secondary Containment will be prohibited during the extended LCO period, to preclude a fuel handling accident, which might lead to a radiological consequence.

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

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

Response: No.

The proposed changes do not involve a physical alteration of the plant. No new or different type of equipment will be installed (damper motors will be replaced) nor will there be changes in methods governing normal plant operation.

The accident analyses affected by this extension are the radiological events that are discussed in the FSAR. The potential for the loss of other plant systems or equipment to mitigate the effects of an accident is not altered.

The proposed changes do not require any new operator response or introduce any new opportunities for operator error not previously considered.

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

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

Response: No.

The increase in completion time for Standby Gas Treatment does not result in any effect on the margin of safety. There is no increase in Core Damage Frequency (CDF) or Large Early Release Frequency (LERF). A recovery plan will be in place to restore the SGTS and Secondary Containment to functional, if a DBA-LOCA accident should occur. Implementation of the compensatory measures minimizes the probability that an accident will be initiated, maximizes the probability that accident mitigation equipment will be available and ensures that SGTS and Secondary Containment will be able to be restored in a timely manner. Thus the potential impact of extending the Completion Time is small. Therefore, this one-time extension will not involve a significant reduction in safety margin.

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

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

NRC Section Chief: Richard J. Laufer.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, Pennsylvania

Date of amendment request: September 8, 2004.

Description of amendment request: The proposed amendment would revise the SSES 1 and 2 Technical Specifications Surveillance Requirement 3.6.1.3.6 to reduce the frequency of performing leakage rate testing for each primary containment purge valve with resilient seals from 184 days to 24 months.

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 proposal would change the Technical Specification Surveillance Requirement for containment purge valves with resilient seals. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the extensive industry operating experience derived from test results has demonstrated that the resilient seal material does not degrade and cause containment isolation valves to leak. Further, these valves are not accident initiators. Thus, the valves will perform as assumed in the accident analyses. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seals. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed nor changes in methods governing normal plant operation). In particular, it does not require the valves to function in any manner other than that which is currently required. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposal would change the Technical Specifications Surveillance Requirement for containment purge valves with resilient seals. The proposed change does not involve a significant reduction in margin of safety because it has no effect on any safety analysis bases or assumptions. It does not change the leakage acceptance criteria. Sufficient data has been collected to demonstrate that resilient seals do not degrade. Testing at the same frequency as other containment isolation valves will not reduce the margin of safety provide by Technical Specifications.

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

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

NRC Section Chief: Richard J. Laufer.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: May 21, 2004.

Description of amendment request: The proposed amendment revises the Reactor Coolant Pump (RCP) Flywheel Inspection Program to extend the allowable inspection interval to 20 years.

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated May 21, 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 proposed change to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete Start Printed Page 9996failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's acceptance guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits; or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Thomas G. Eppink, South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Section Chief: John A. Nakoski.

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

Date of amendment requests: January 28, 2005.

Description of amendment requests: The proposed change would revise Technical Specifications (TSs) 1.1 “Definitions,” 3.4 “Reactor Coolant System [RCS],” and 5.7 “Reporting Requirements” to relocate the RCS pressure-temperature curves and limits from the TSs to a licensee-controlled document identified as the Pressure and Temperature Limit Report.

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

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

Response: No.

This proposed change revises the Technical Specifications by relocating the reactor coolant system (RCS) Pressure and Temperature Limits, Heatup and Cooldown Curves and Low Temperature Overpressure Protection (LTOP) enable temperatures from the Technical Specifications to a RCS Pressure and Temperature Limits Report (PTLR). Relocation of this information will not impact the activity to update the RCS pressure and temperature curves and limits in accordance with the requirements of 10 CFR 50 Appendix G and H to ensure the reactor coolant system's pressure boundary integrity will be protected until end of life (EOL). Consequently, this proposed change is determined to not contribute to the probability of or the initiation of accidents. There is no change to the safety analysis.

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.

This proposed change revises the Technical Specifications by relocating the RCS Pressure and Temperature Limits, Heatup and Cooldown Curves and LTOP enable temperatures from the Technical Specifications to a RCS PTLR to document removal, testing and analyzing the surveillance capsule. This document will be updated by SCE to reflect the testing and analysis of specimens. Removal, testing and analyzing the surveillance capsule resulted in changes to the RCS pressure and temperature limits. These changes are required to maintain the RCS pressure boundary integrity until EOL. Changes to the RCS pressure and temperature curves and limits will not create a new or different kind of accident. There is no change to the safety analysis.

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.

Pressure and temperature curves and limits are provided as limits to plant operation for ensuring RCS pressure boundary integrity and maintained until EOL. Changes to the RCS pressure and temperature curves and limits, resulting from the removal, testing and analyzing of a surveillance capsule, are only made within the acceptable margin limits maintaining the required margin of safety. There is no change to the safety analysis.

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 requests involve no significant hazards consideration.

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment requests: February 3, 2005.

Description of amendment requests: The proposed change would revise Technical Specification 3.6.3, “Containment Isolation Valves,” Surveillance Requirements 3.6.3.3 and 3.6.3.4 for Containment Isolation Valves and Blind Flanges (ClVs) by adding a provision to exempt CIVs that are locked, sealed, or otherwise secured from the position verification surveillance requirements.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the Start Printed Page 9997licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

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

The proposed change does not affect the CIV design or function. In addition, mis-positioned or failed ClVs are not the initiator of any event. The position of a locked, sealed, or otherwise secured valve and blind flange is verified at the time it is locked, sealed, or secured, and these ClVs are administratively controlled to remain in the required position. Further, since the change impacts only the re-verification of the blind flange and valve position as a Technical Specification Surveillance, it does not result in any change in the response of the equipment to an accident.

Based on the above, SCE concludes that deleting the re-verification of the position of a locked, sealed, or secured CIV as a Technical Specification Surveillance does not affect the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new kind of accident from any accident previously evaluated?

This change does not add any new equipment or result in any changes to equipment design or capabilities. This change also does not result in any changes to the operation of the plant. The position of a locked, sealed, or otherwise secured blind flange and valve is verified at the time it is locked, sealed, or secured, and these ClVs are administratively controlled to remain in the required position. Further, since the change impacts only the re-verification of the blind flange and valve position as a Technical Specification Surveillance, it does not result in any change in the response of the equipment to an accident.

Based on the above, SCE concludes that deleting the re-verification of the position of a locked, sealed, or secured CIV as a Technical Specification Surveillance does not create the possibility of a new or different kind of accident from any previously evaluated.

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

The CIVs are administratively controlled and their operation is a nonroutine event. The position of a locked, sealed, or otherwise secured blind flange and valve is verified at the time it is locked, sealed, or secured. Also, no CIVs were found to be out of position from a review of all the San Onofre Units 2 and 3 surveillance data from January 2000 through December 2004. Since the change only deletes the re-verification of the blind flange and valve position as a Technical Specification Surveillance and the administrative controls are in place, the proposed change will provide a similar level of assurance of correct CIV position as the current verifications.

Based on the above, SCE concludes that deleting the re-verification of the position of a locked, sealed, or secured CIV as a Technical Specification Surveillance does not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: January 20, 2005.

Description of amendment request: The proposed amendments would change Technical Specification (TS) 3/4.8.2.1, “DC Sources—Operating,” and TS 3/4.8.2.2, “DC Sources—Shutdown,” with addition of a new TS 3/4.8.2.3, “Battery Parameters”, to incorporate actions for responding to “out-of-limit” conditions, and surveillances for verification of battery parameters. The proposed changes would revise allowed outage times for battery chargers as well as battery charger testing criteria. The proposed changes would also relocate a number of battery surveillance requirements to a licensee-controlled Battery Monitoring and Maintenance Program.

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

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

Response: No.

The proposed change rearranges the Technical Specifications for the direct current [DC] electrical power system, and adds new Conditions and required actions with revised completion times to allow for battery charger inoperability. Neither the direct current electrical power subsystem nor associated battery chargers are initiators of an accident sequence previously evaluated. Performance of plant operational activities in accordance with the proposed Technical Specification changes ensures that the direct current electrical power subsystem is capable of performing its function as previously described. Therefore, the mitigating functions supported by the subject subsystem will continue to provide the protection assumed by the safety analysis.

Relocation of preventive maintenance surveillances and certain operating limits and actions to a “Battery Monitoring and Maintenance Program” will not challenge the ability of the subject subsystem to perform its design function. Maintenance and monitoring currently required will continue to be performed. In addition, the direct current electrical power subsystem is within the scope of 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants,” which will ensure continued control of maintenance activities associated with the subject subsystem.

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 accident from any accident previously evaluated?

Response: No.

The proposed change does not involve any physical alteration of the units. No new equipment is introduced, and installed equipment is not operated in a new or different manner. The proposed changes do not affect setpoints for initiation of protective or mitigating actions.

Operability of the DC electrical power subsystems in accordance with the proposed technical specifications is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the plant.

The proposed changes will not alter the manner in which equipment operation is initiated, nor will the functional demands on credited equipment be changed. No alteration in the operating procedures is proposed, and no change is being made to procedures relied upon in response to an off-normal event. No new failure modes are being introduced, and the proposed change does not alter assumptions made in the safety analyses.

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

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

Response: No.

The proposed change will not adversely affect operation of plant equipment and will not result in a change to the setpoints at which protective actions are initiated. Sufficient DC capacity to support operation of mitigation equipment is ensured. The provisions of the Battery Monitoring and Maintenance Program will ensure that the station batteries are maintained in a highly reliable manner. The equipment fed by the DC electrical system will continue to provide adequate power to safety-related loads in accordance with analysis assumptions.

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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.Start Printed Page 9998

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

NRC Section Chief: Allen G. Howe.

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: September 30, 2004.

Brief description of amendments: The proposed amendment revises TS 5.5.7, “Reactor Coolant Pump [RCP] Flywheel Inspection Program,” to extend the allowable inspection interval to 20 years.

The NRC staff issued a notice of availability of a model safety evaluation and model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on October 22, 2003 (68 FR 60422). The licensee affirmed the applicability of the model NSHC determination in its application dated September 30, 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 proposed change to the RCP flywheel examination frequency does not change the response of the plant to any accidents. The RCP will remain highly reliable and the proposed change will not result in a significant increase in the risk of plant operation. Given the extremely low failure probabilities for the RCP motor flywheel during normal and accident conditions, the extremely low probability of a loss-of-coolant accident (LOCA) with loss of offsite power (LOOP), and assuming a conditional core damage probability (CCDP) of 1.0 (complete failure of safety systems), the core damage frequency (CDF) and change in risk would still not exceed the NRC's acceptance guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per year). Moreover, considering the uncertainties involved in this evaluation, the risk associated with the postulated failure of an RCP motor flywheel is significantly low. Even if all four RCP motor flywheels are considered in the bounding plant configuration case, the risk is still acceptably low.

The proposed change does not adversely affect accident initiators or precursors, nor alter the design assumptions, conditions, or configuration of the facility, or the manner in which the plant is operated and maintained; alter or prevent the ability of structures, systems, components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits; or affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed change does not increase the type or amount of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposure. The proposed change is consistent with the safety analysis assumptions and resultant consequences. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed change in flywheel inspection frequency does not involve any change in the design or operation of the RCP. Nor does the change to examination frequency affect any existing accident scenarios, or create any new or different accident scenarios. Further, the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or alter the methods governing normal plant operation. In addition, the change does not impose any new or different requirements or eliminate any existing requirements, and does not alter any assumptions made in the safety analysis. The proposed change is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside of the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174. There are no significant mechanisms for inservice degradation of the RCP flywheel. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: Allen G. Howe.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, Virginia

Date of amendment request: December 17, 2004.

Description of amendment request: The proposed changes to the Technical Specifications would increase the completion times from 72 hours to 7 days for the following systems: Low-Head Safety Injection (LHSI) Emergency Core Cooling System (ECCS), Auxiliary Feedwater (AFW) System, Quench Spray (QS) System, and Chemical Addition System (CAS).

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

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

The proposed changes do not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. The proposed changes will not alter assumptions relative to the mitigation of an accident or transient event.

The CDF [core damage frequency] impact and the LERF [large early release frequency] impact, as well as the ICCDP [incremental conditional core damage probability] and ICLERP [incremental conditional large early release probability], associated with the proposed completion time changes meet the acceptance criteria in RG [Regulatory Guide] 1.174 and RG 1.177 for the proposed changes. The cumulative CDF and LERF impact for the proposed completion time changes also meet the acceptance criteria in RG 1.174 for the proposed changes.

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

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

The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The overall margin of safety is not decreased due to the increased completion times for the LHSI ECCS, QS including the CAS, and AFW since the systems design and operation are not altered by the proposed increase in completion times. The risk impacts of the changes are also consistent with the acceptance criteria in RG 1.174 and RG 1.177.

For the Chemical Addition System, which is not modeled in the PRA [probabilistic risk assessment] due to its limited capability to mitigate severe accidents, the proposed completion time change takes into account the ability of the spray systems to remove iodine at a reduced capability and the low probability of the worst case DBA [design-basis accident] occurring during this period.Start Printed Page 9999

The codes and standards or their alternatives approved for use by the NRC continue to be met. In addition, the safety analysis acceptance criteria in the licensing basis (e.g., FSAR [final safety analysis report], supporting analyses) continue to be met.

Therefore, 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 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.

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: December 17, 2004.

Description of amendment request: The proposed Technical Specifications (TS) change would revise the reactor coolant system (RCS) pressure temperature (P/T) operating limits, the Low-Temperature Overpressure Protection System (LTOPS) setpoint, and the LTOPS enable temperature (Tenable) basis for cumulative core burnups up to 47.6 effective full-power years (EFPY) and 48.1 EFPY for Surry Power Station, Units 1 and 2, respectively.

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

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

The proposed change modifies the Surry Units 1 and 2 RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value and extends the cumulative core burnup applicability limits for these parameters. The allowable operating pressures and temperatures under the proposed RCS P/T limit curves are not significantly different from those allowed under the existing Technical Specification P/T limits. The revisions in the values for the LTOPS setpoint and LTOPS Tenable do not significantly change the plant operating space. No changes to plant systems, structures or components are proposed, and no new operating modes are established. The P/T limits, LTOPS setpoint, and Tenable value do not contribute to the probability of occurrence or consequences of accidents previously analyzed. The revised licensing basis analyses utilize acceptable analytical methods, and continue to demonstrate that established accident analysis acceptance criteria are met. Therefore, there is no increase in the probability or consequences of any accident previously evaluated.

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

The proposed change modifies the Surry Units 1 and 2 RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value and extends the cumulative core burnup applicability limits for these parameters. The allowable operating pressures and temperatures under the proposed RCS P/T limit curves are not significantly different from those allowed under the existing Technical Specification P/T limits. No changes to plant systems, structures or components are proposed, and no new operating modes are established. Therefore, the proposed changes do not create the possibility of any accident or malfunction of a different type previously evaluated.

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

The proposed revised RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value analysis bases do not involve a significant reduction in the margin of safety for these parameters. The proposed revised RCS P/T limit curves are valid to cumulative core burnups of 47.6 EFPY and 48.1 EFPY for Surry Units 1 and 2, respectively. The proposed revised LTOPS setpoint and Tenable analyses support these same cumulative core burnup limits. The proposed revised RCS P/T limit curves utilize ASME [American Society of Mechanical Engineers] Code Section XI, which supports use of a conservative but less restrictive stress intensity formulation (K1c). The proposed extension of the cumulative core burnup applicability limits along with a small increase in the LTOPS PORV [power-operated relief valve] setpoint is accommodated by the margin provided by ASME Code Section XI. The analyses demonstrate that established analysis acceptance criteria continue to be met. Specifically, the proposed P/T limit curves, LTOPS setpoint and LTOPS Tenable value provide acceptable margin to vessel fracture under both normal operation and LTOPS design basis (mass addition and heat addition) accident conditions. Therefore, the proposed change does not result in a significant reduction in 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.

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

Date of application for amendment: July 15, 2004.

Brief description of amendment: The amendment added references to the list of approved core operating limits analytical methods in Technical Specification 5.6.5.b for Calvert Cliffs, Unit Nos. 1 and 2.

Date of publication of individual notice in Federal Register: December 29, 2004 (69 FR 78056).

Expiration date of individual notice: February 28, 2005.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of amendment request: November 5, 2003, as supplemented by letter dated April 22, 2004.

Brief description of amendment request: The proposed amendment would revise the Point Beach Nuclear Plant (PBNP), Units 1 and 2, Updated Final Safety Analysis Report to reflect the Commission staff's approval of the WCAP-14439-P, Revision 2 analysis entitled, “Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Point Beach Nuclear Plant Units 1 and 2 for the Power Uprate and License Renewal Program.”Start Printed Page 10000

Date of publication of individual notice in Federal Register: February 7, 2005 (70 FR 6466).

Expiration date of individual notice: April 8, 2005.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

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

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

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Date of application for amendment: June 22, 2004.

Brief description of amendment: The proposed amendment revises Technical Specification 3.1.8, “Scram Discharge Volume (SDV) Vent and Drain Valves,” to allow a vent or drain line with one inoperable valve to be isolated instead of requiring the valve to be restored to operable status within 7 days.

Date of issuance: February 10, 2005.

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

Amendment No.: 162.

Facility Operating License No. NPF-62: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: August 31, 2004 (68 FR 53099).

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

No significant hazards consideration comments received: No.

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

Date of application for amendments: December 9, 2003, as supplemented May 19 and August 3, 2004.

Brief description of amendments: The amendments revise Technical Specification 3.7.1, “Main Steam Safety Valves (MSSVs),” to increase the maximum allowable lift setting on two MSSVs on each unit. In addition, the amendments increase the completion time for reducing the Power Level-High Trip setpoint.

Date of issuance: February 10, 2005.

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

Amendment Nos.: 270 and 247.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 26, 2004 (69 FR 62470).

The supplemental letters dated May 19 and August 3, 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.

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: August 19, 2004, as supplemented December 2, 2004.

Brief description of amendment: The amendment revises the reactor coolant system pressure and temperature limits by replacing Technical Specification Section 3.4.3, “RCS Pressure and Temperature (P/T) Limits,” Figures 3.4.3-1 and 3.4.3-2, with figures that are applicable up to 35 effective full-power years.

Date of issuance: February 7, 2005.

Effective date: February 7, 2005.

Amendment No.: 202.

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

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57981). The December 2, 2004, supplement contained clarifying information only that 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 amendment is contained in a Safety Evaluation dated February 7, 2005.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: June 4, 2004, as supplemented on July 27, September 27, and December 14, 2004.

Brief description of amendment: The amendment revises the safety limit values in Technical Specifications 2.1.1.2 for the minimum critical power ratio for both single and two recirculation loop operation.

Date of issuance: February 3, 2005.

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

Amendment No.: 281.

Facility Operating License No. DPR-59: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: July 20, 2004 (69 FR 43459).

The July 27, September 27, and December 14, 2004, letters provided information that clarified the Start Printed Page 10001application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 3, 2005.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: September 1, 2004.

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

Date of issuance: February 3, 2005.

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

Amendment No.: 282.

Facility Operating License No. DPR-59: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57984).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 3, 2005.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: August 19, 2003, as supplemented on March 12, 2004.

Brief description of amendment: The amendment revised Pilgrim Nuclear Power Station (Pilgrim) Technical Specification (TS) Table 3.2.C-1 by changing the rod block monitor (RBM) low power setpoint (LPSP) allowable value from 29% to 25.9%. The amendment corrected the RBM LPSP (currently ≤29%) that was incorrectly inserted into Note 5 for TS Table 3.2.C-1 under License Amendment No. 138, dated July 1, 1991. Pilgrim plant procedures and the Core Operating Limits Report have enforced the correct setpoint value of ≤25.9% since issuance of License Amendment No. 138.

Date of issuance: February 2, 2005.

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

Amendment No.: 210.

Facility Operating License No. DPR-35: The amendment revised the TSs.

Date of initial notice in Federal Register: February 17, 2004 (69 FR 7521).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 5, 2004, as supplemented by letters dated June 22 and December 6, 2004.

Brief description of amendment: This amendment modifies the existing minimum critical power ratio (MCPR) safety limit contained in Technical Specification 2.1.1.2. Specifically, the change modifies the MCPR safety limit values, as calculated by Global Nuclear Fuel (GNF), by decreasing the limit for two recirculation loop operation from 1.10 to 1.08, and decreasing the limit for single recirculation loop operation from 1.11 to 1.10.

Date of issuance: February 3, 2005.

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

Amendment No.: 132.

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

Date of initial notice in Federal Register: May 11, 2004 (69 FR 26189).

The supplements dated June 22 and December 6, 2004, provided clarifying information that did not change the scope of the April 5, 2004, application nor the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 3, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendment: March 31, 2004.

Brief description of amendment: This amendment modified the technical specification (TS) surveillance requirements (SRs) for manual actuation of certain main steam safety/relief valves (S/RVs), including those valves that provide an automatic depressurization system (ADS) and low-low set (LLS) valve function. The specific TS changes revised SR 3.4.4.3 for S/RVs, SR 3.5.1.7 for ADS valves, and SR 3.6.1.6.1 for LLS valves. The changes removed the requirement for the S/RV disks to be lifted from their seats when manually actuated.

The revised SRs specify that the actuator is to stroke when manually actuated, without physically lifting the disks off their seats at power.

Date of issuance: February 10, 2005.

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

Amendment No.: 133.

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

Date of initial notice in Federal Register: May 11, 2004 (69 FR 26188).

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

No significant hazards consideration comments received: No.

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

Date of application for amendment: October 14, 2004.

Brief description of amendment: The amendment corrects errors in Technical Specifications 3.10.i and 6.9.a.4.A.

Date of issuance: February 15, 2005.

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

Amendment No.: 180.

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

Date of initial notice in Federal Register: December 7, 2004 (69 FR 70720).

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

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of application for amendments: October 5, 2004.

Brief description of amendments: The amendments deleted technical specification (TS) 5.6.1, “Occupational Radiation Exposure Reports,” and TS 5.6.3, “Monthly Operating Reports,” as described in the Notice of Availability Start Printed Page 10002published in the Federal Register on June 23, 2004 (69 FR 35067).

Date of issuance: February 7, 2005.

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

Amendment Nos.: 216, 221.

Facility Operating License Nos. DPR-24 and DPR-27: 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 February 7, 2005.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: May 21, 2004.

Brief description of amendment: This amendment deletes the Technical Specification requirements associated with hydrogen recombiners and hydrogen monitors.

Date of issuance: February 3, 2005.

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

Amendment No.: 170.

Renewed Facility Operating License No. NPF-12: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57990).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 3, 2005.

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 2, 2004.

Description of amendment request: The amendments eliminated the requirements for the licensee to submit monthly operating reports and occupational radiation exposure reports.

Date of issuance: January 25, 2005.

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

Amendment Nos.: 252, 291 and 250.

Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 12, 2004 (69 FR 60687).

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

No significant hazards consideration comments received: No.

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

Date of application for amendments: July 8, 2004.

Description of amendment request: The amendments revised Technical Specifications by eliminating the requirements associated with hydrogen monitors.

Date of issuance: February 14, 2005.

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

Amendment Nos.: 253, 292 and 251.

Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 14, 2004 (69 FR 55473).

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

No significant hazards consideration comments received: No.

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

Date of application for amendments: March 3, 2004.

Brief description of amendments: The amendments revised the Updated Final Safety Analysis Report (UFSAR) by modifying the licensing basis for the seismic qualification of round flexible ducting, triangular ducting, and associated air bars installed as part of the suspended ceiling air delivery system in the main control room.

Date of issuance: January 31, 2005.

Effective date: As of the date of issuance and shall be implemented as part of the next UFSAR update made in accordance with 10 CFR 50.71(e).

Amendment Nos.: 298 and 287.

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

Date of initial notice in Federal Register: April 27, 2004 (69 FR 22883).

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.

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

Date of application for amendment: July 8, 2004.

Brief description of amendment: The amendment revises Technical Specification (TS) Section 3.8.4, “DC Sources-Operating.” Specifically, the amendment removes the term “inter-rack” and associated wording from TS Surveillance Requirements 3.8.4.6 and 3.8.4.10 for the 125 Volt Direct Current electrical power subsystems of the emergency diesel generators.

Date of issuance: February 7, 2005.

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

Amendment No.: 54.

Facility Operating License No. NPF-90: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: August 3, 2004 (69 FR 46593).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 7, 2005.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 21, 2004, as supplemented by letters dated November 18 and December 3, 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.Start Printed Page 10003

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

Amendment Nos.: 114, 114.

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

Date of initial notice in Federal Register: March 2, 2004 (69 FR 9866).

The supplemental letters dated November 18 and December 3, 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.

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

Date of amendment request: July 23, 2004.

Brief description of amendment: The amendment eliminates the requirements in the technical specifications associated with hydrogen recombiners and hydrogen monitors.

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.: 157.

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

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53115).

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: July 23, 2004.

Brief description of amendment: The amendment revises the technical specifications by eliminating the requirements to provide the NRC monthly operating reports and annual occupational radiation exposure reports.

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.: 158.

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

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53116).

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 hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated.

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

For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​Start Printed Page 10004reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

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

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

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

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

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

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

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

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

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer or the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).Start Printed Page 10005

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear Station Unit 2, York County, South Carolina

Date of amendment request: February 5, 2005, as supplemented by letter dated February 7, 2005.

Description of amendment request: The amendment revises the system bypass leakage acceptance criterion for the charcoal adsorber in the 2B Auxiliary Building Filtered Ventilation Exhaust System train as listed in Technical Specification 5.5.11, “Ventilation Filter Testing Program.”

Date of issuance: February 7, 2005.

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

Amendment No.: 213.

Renewed Facility Operating License No. NPF-52: Amendments revised the Technical Specifications.

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

The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated February 7, 2005.

Attorney for licensee: Ms. Anne Cottingham, Esquire.

NRC Section Chief: John A. Nakoski.

Start Signature

Dated in Rockville, Maryland, this 17th 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.

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[FR Doc. 05-3627 Filed 2-28-05; 8:45 am]

BILLING CODE 7590-01-P