04-27614. Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from November 25, 2004, through December 9, 2004. The last biweekly notice was published on December 7, 2004 (69 FR 70712).

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Date of amendment request: June 10, 2004.

Description of amendment request: The amendments would revise Technical Specification (TS) 3.6.3, “Containment Isolation Valves,” to allow the surveillance frequencies for leakage rate testing to be specified in the Catawba Nuclear Station Containment Leak Rate Testing 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:

First Standard

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

No.

This amendment will not change any previously evaluated accidents such as the postulated “Fuel Handling Accident (FHA) in Containment”. No credit is assumed for VP containment isolation in the FHA within containment. The Containment Purge (VP) System and Hydrogen Purge (VY) System containment isolation valves are sealed closed during modes 1 through 4. The Containment Air Release and Addition (VQ) System containment isolation valves are designed to close within 5 seconds of a containment phase “A” isolation signal. The prevention and mitigation of these accidents is not affected by this change.

Test data demonstrates that the likelihood of a malfunction of a resilient seal in one of the VP, VY, or VQ valves is not increased by this change in the surveillances. The systems will continue to be able to perform their design functions of isolating containment during the evaluated accidents. Test procedures will continue to monitor the leakage of these valves to ensure the design function will continue to be met. There is no impact on previously evaluated accidents since the valves will continue to close and seal or remain closed as originally assumed in the accident scenarios.

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

Second Standard

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

No.

This change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing any normal plant operation. The change does not alter assumptions made in the safety analyses or licensing basis. This change will not affect or degrade the ability of the Containment Purge System, Hydrogen Purge System, or Containment Air Release and Addition System valves to perform their specified safety functions. Therefore, the change does not create the possibility of a new or different kind of credible accident from any accident previously evaluated.

Third Standard

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

No.

SR 3.6.3.6 currently states: “The measured leakage rate for Containment Purge System and Hydrogen Purge System valves must be < 0.05 L_a (Design Leakage Rate) when pressurized to Pa (Design Containment Pressure). The measured leakage rate for Containment Air Release and Addition valves must be < 0.01 L_a when pressurized to P_a. These required maximum leak rates will not be changed by this amendment. Testing of these valves to measure leakage through the valve seats will continue, only at a different frequency based on past test results. This will be a nominal frequency of 18 months for the VP System and in accordance with 10 CFR 50, Appendix J, Option B for the VQ and VY Systems. Therefore, the proposed changes listed above do not involve a significant reduction in a margin of safety.

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

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

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: July 19, 2004.

Description of amendment request: The amendments would revise Technical Specification (TS) 3.8.4, “DC Sources—Operating” and TS 3.8.6, “Battery Cell Parameters” to allow for the replacement of the existing nickel cadmium diesel generator batteries with conventional lead acid batteries.

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

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

The DG batteries are not accident initiating equipment; they are accident mitigating equipment. As such, they cannot affect the probability of any accident being initiated. The performance of the replacement batteries will exceed that of the existing batteries. Therefore, no accident consequences will be adversely impacted.

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

The DG batteries are not capable by themselves of initiating any accident. Other than the replacement of the batteries themselves and the associated modification work (e.g., installation of the battery HVAC system), no physical changes to the overall plant are being proposed. No changes to the overall manner in which the plant is operated are being proposed. Therefore, no potential for new accident types is generated.

(3) The proposed license amendments do not involve a significant reduction in a margin of safety.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their intended functions. These barriers include the fuel cladding, the reactor coolant system, and the containment. The modification to replace the DG batteries will not have any impact on these barriers. In addition, no accident mitigating equipment will be adversely impacted as a result of the battery replacement. The replacement batteries will have overall performance capabilities equal to or greater than those for the existing batteries. Therefore, existing safety margins will be preserved.

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

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

NRC Section Chief: John A. Nakoski.

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

Date of amendment request: September 21, 2004.

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

The NRC staff issued a notice of availability of a model no significant hazards consideration determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model no significant hazards consideration determination in its application dated September 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 revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity.

With the elimination of the design-basis LOCA hydrogen release, hydrogen and oxygen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG 1.97, Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen and oxygen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. Also, as part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment.

The regulatory requirements for the hydrogen and oxygen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, classification of the oxygen monitors as Category 2 and removal of the hydrogen and oxygen monitors from TS will not prevent an accident management strategy through the use of the SAMGs [severe accident management guidelines], the emergency plan (EP), the emergency operating procedures (EOPs), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

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

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

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

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

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

The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety.

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

Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI Unit 2, accident can be adequately met without reliance on safety-related hydrogen monitors. Category 2 oxygen monitors are adequate to verify the status of an inerted containment.

Therefore, this change does not involve a significant reduction in the margin of safety. The intent of the requirements established as a result of the TMI Unit 2, accident can be adequately met without reliance on safety-related oxygen monitors. Removal of hydrogen and oxygen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability.

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

Attorney for licensee: Thomas C. Poindexter, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Robert A. Gramm.

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

Date of amendment request: April 14, 2004.

Description of amendment request: The proposed amendment would add a new section to the Technical Specifications (TSs) and two new Limiting Conditions for Operations (LCOs) to allow certain reactor coolant system (RCS) hydrostatic and system leakage pressure tests to be performed with the reactor pressure vessel temperature above 212° Fahrenheit (F). The first LCO would allow specified TS requirements to be changed to permit performance of special tests and operations, which otherwise could not be performed if required to comply with the requirements of the TSs. The second LCO would require reactor low water level instrumentation, standby gas treatment system, and secondary containment to be OPERABLE to allow certain RCS pressure tests to be performed with the reactor pressure vessel temperature above 212° F, and provides for an exemption from the requirements for OPERABILITY for other systems that currently go into effect when in Hot Shutdown or when RCS temperature is greater than 212° F. It will also update the Table of Contents to reflect the proposed changes.

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

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

The proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated. The probability of an accident previously evaluated is not significantly increased because the proposed change will not alter the method by which RCS hydrostatic pressure and leak testing is performed. Under this proposed change the secondary containment, standby gas treatment system and associated initiation instrumentation are required to be operable during the performance of RCS hydrostatic pressure and leak testing and would be capable of handling any airborne radioactivity or steam leaks that could occur. The required pressure testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of a main steamline break (MSLB) outside the primary containment. Accordingly, the consequences of previously evaluated accidents are not increased significantly.

The proposed update to the Table of Contents is editorial in nature. Since this update is administrative in nature, it cannot increase the probability or consequences of a previously analyzed accident.

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 previously evaluated?

The proposed amendment change will not alter the way that hydrostatic pressure and leak testing is performed. Therefore, the proposed change will 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?

The proposed amendment will not involve a significant reduction in a margin of safety for a postulated MSLB outside of primary containment. The proposed changes and additions result in increased system operability requirements above those that currently exist during the performance of RCS hydrostatic pressure and leak testing. The incremental increase in stored energy in the vessel during testing will be conservatively bounded by the consequences of the postulated MSLB outside of primary containment. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The proposed update to the Table of Contents is editorial in nature. Since this update is administrative in nature, the proposed change does not involve a significant reduction in a margin of safety

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: 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: September 2, 2004.

Description of amendment request: The proposed amendment would revise Technical Specification 4.5.B.2.2 (TS) to change the surveillance requirement frequency for air testing the drywell and suppression pool (torus) spray headers and nozzles from “once every 5 years” to “following maintenance that could result in nozzle blockage.”

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 drywell and torus headers and spray nozzles are not assumed to be initiators of any accidents previously evaluated. Maintenance practices and normal environmental conditions to which the system is subjected are adequate to ensure operability of the systems. Since the system will be able to perform its accident mitigation function, the consequences of accident previously evaluated are not increased. 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 revised surveillance does not introduce any new mode of plant operation, does not involve physical modification of the plant, or any new operating modes, and cannot introduce new accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

Maintenance practices and normal environmental conditions to which the system is subjected are adequate to ensure operability of the systems. As the spray nozzles are expected to remain fully capable of performing their post-accident mitigation function, margin of safety is not reduced. 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 J. Roberts.

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

Date of amendment request: September 2, 2004.

Description of amendment request: The proposed amendment would remove a license condition that currently requires the reactor not to be operated for more than 24 hours if one recirculation loop is out of service. It would revise Technical Specifications (TSs) to allow the minimum critical power ratio (MCPR) safety limit to be changed for single loop operations (SLOs). It would also revise the current jet pump limiting condition for operation and surveillance requirements to allow for the conduct of a TS required surveillance during SLOs. The proposed amendment would modify the TSs to address SLO operating conditions and restrictions, and delete a TS condition related to thermal-hydraulic stability. It would update the TSs for average planar linear heat generation rate for SLOs, and update the thermal power applicability restrictions to be consistent with NUREG-1433, Revision 3, “Standard Technical Specifications for General Electric Boiling-Water Reactors.” It would also revise the TSs for linear heat generation rate and MCPR for thermal power applicability restrictions. The proposed amendment makes an administrative change to have MCPR recalculated when reactor power is equal to or greater than 25 percent. Lastly, it would update the TSs' table of contents and TS pages to administratively reflect all of these proposed changes.

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 license and technical specification changes will allow the plant to be operated with one recirculation pump for longer than 24 hours provided that appropriate limits are instituted. Extended single recirculation loop operation has been evaluated and methodologies have been established for determining appropriate operating limits. Implementation of the single recirculation loop operating limits ensures that system operation is in conformance with the conditions established to minimize the probability of accidents and the associated consequences. Required completion times for implementing the system operating limits and restoring out of specification limits minimize the probability that an accident occurs when out of specification conditions exist while allowing for deliberate operator action.

Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence 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 license and technical specification changes will allow plant operation with a single recirculation loop for longer than 24 hours. The proposed changes introduce an additional recirculation system-operating mode, however, existing system component operating equipment or operating characteristics will not change. The Pilgrim Station Single Loop Analysis Report identifies required operating limits that apply when the system will be operated in the single loop operation mode. Implementation of these operating limits will ensure that the system is operated in accordance with design. Additionally, revised jet pump surveillance ensures that loop specific surveillance is performed as required to validate the bounding assumptions of existing accident analyses. As such, no new failure mechanisms are created and existing design evaluations bound system operation.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident.

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

Response: No.

The proposed license and technical specification changes identify the operating limits that apply to single recirculation loop operation. These proposed recirculation system limits were identified to ensure that system operation would be in conformance to the conditions evaluated in applicable accident and transient analyses. Implementation of the proposed limits for single recirculation loop operation ensures that safety margins are maintained. Required completion times for implementing the system operating limits minimizes the possibility that an accident occurs when out of specification conditions exist.

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

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

Attorney for licensee: 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 Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (licensee), Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: September 16, 2003 as supplemented by letter dated March 15, 2004.

Description of amendment request: The proposed amendment would relocate the current definition of surveillance frequency to new Technical Specification (TS) Sections 4.0.2 and 4.0.3, and revise the requirements for missed surveillance in Section 4.0.3. This change is consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) change TSTF-358, Revision 5. The proposed change would allow a longer period of time to perform a missed surveillance. The time is 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 conjunction with the proposed change, the proposed amendment would add the requirements for a Bases Control Program which is consistent with Section 5.5 of NUREG 1433. In addition, the current definition of surveillance interval (definition “Z”) would be re-worded and relocated to new Section 4.0.1 consistent with Surveillance Requirement 3.0.1 of NUREG 1433. Appropriate Bases, also consistent with NUREG 1433 would be adopted for the new sections. An editorial change would be made to TS 6.7.C which references the current definition of surveillance frequency to now reference the new Section 4.0.2.

The NRC staff issued a notice of opportunity 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 model NSHC determination in its application dated September 16, 2003. The model NSHC determination analysis for changes to the TS associated with missed surveillances, and the NSHC determination analysis provided by the licensee for the remaining TS changes, is provided herein.

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

With regard to the proposed change to the TS associated with missed surveillances, the proposed change relaxes the time allowed to perform a missed surveillance. 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. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

With regard to the remaining proposed changes to the TSs, the proposed changes do not involve physical changes to the plant or introduce any new modes of operation. Accordingly, continued assurance is provided that the process variables, structures, systems, and components are maintained such that there will be no degradation of any fission product barrier which could increase the radiological consequences of an accident. Therefore, the proposed changes do 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

With regard to the proposed changes to the TSs associated with missed surveillances, 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. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

With regard to the remaining proposed changes to the TSs, 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. Thus, the changes do 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

With regard to the proposed changes to the TSs associated with missed surveillances, 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. Therefore, these changes do not involve a significant reduction in a margin of safety.

With regard to the remaining proposed changes to the TSs, the administrative changes do not alter the basic operation of process variables, systems, or components as described in the safety analysis. No new equipment is introduced. Accordingly, the proposed changes do not involve a significant reduction in a margin of safety.

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

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

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: Allen G. Howe.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: October 5, 2004.

Description of amendment request: The proposed amendment would revise Technical Specification Section 6.7.C “Primary Containment Leak Rate Testing Program,” to allow a one-time extension to the 10-year interval for performing the next Type A containment integrated leak rate test (ILRT). Specifically, the change would allow the test to be performed within 15 years from the last ILRT, which was performed in April 1995.

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

1. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed revision to Technical Specifications adds a one-time extension to the current interval for Type A testing. The current test interval of 10.6 years, based on past performance, is extended on a one-time basis to fifteen years from the last Type A test. The proposed extension to Type A testing cannot increase the probability of an accident previously evaluated since the containment Type A testing extension is not a modification and the test extension is not of a type that could lead to equipment failure or accident initiation.

The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident since research documented in NUREG-1493 has found that, generically, very few potential containment leakage paths are not identified by Type B and C tests. The NUREG concluded that reducing the Type A (ILRT) testing frequency to once per twenty years was found to lead to an imperceptible increase in risk. These generic conclusions were confirmed by a plant specific risk analysis performed using the current Vermont Yankee Probabilistic Safety Assessment (PSA) internal events model that concluded the consequences are low to negligible.

Testing and inspection programs in place also provide a high degree of assurance that the containment will not degrade in a manner detectable only by Type A testing. The last two successful Type A tests indicate a very leak tight containment. Type B and C testing required by Technical Specifications will identify any containment opening such as valves that would otherwise be detected by the Type A tests. Inspections, including those required by the ASME [American Society of Mechanical Engineers] code and the Maintenance Rule are performed in order to identify indications of containment degradation that could affect that leak tightness.

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

2. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing. The current test interval of 10.6 years, based on past performance, would be extended on a one time basis to fifteen years from the last Type A test. The proposed extension to Type A testing cannot create the possibility of a new or different type of accident since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident.

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

3. The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.

The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing. The current test interval of 10.6 years, based on past performance, would be extended on a one time basis to fifteen years from the last Type A test. The proposed extension to Type A testing will not significantly reduce the margin of safety. The NUREG-1493 generic study of the effects of extending containment leakage testing found that a 20-year extension in Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 found that, generically, the design containment leakage rate contributes about 0.1 percent to the individual risk and that the decrease in Type A testing frequency would have a minimal affect on this risk since 95% of the potential leakage paths are detected by Type C testing. This was further confirmed by a plant specific risk assessment using the current Vermont Yankee PSA internal events model that concluded the risk associated with this change is negligibly small and/or non-risk significant.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: Allen G. Howe.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: October 6, 2004.

Description of amendment request: The proposed amendment would revise Technical Specification Surveillance Requirement 4.5.B.1 related to air testing of the drywell spray headers and nozzles. Specifically, the amendment would change the test frequency from once every 5 years to following maintenance that could result in nozzle blockage.

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

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

Response: No.

The proposed amendment would revise the Technical Specification surveillance requirements associated with the air test of the drywell spray headers and nozzles. The frequency of the air test would be changed from a fixed 5-year frequency to following maintenance that could result in nozzle blockage.

This surveillance test is performed while the plant is in a cold shutdown condition and the equipment is not required to be operable. The testing is to verify that the spray headers and nozzles are not obstructed. The proposed change in the surveillance test frequency will not result in any design changes to systems, structures, or components, or their method of operation. The drywell spray headers and nozzles are not initiators of any accidents previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability of any accident previously evaluated.

The drywell spray headers provide a means to control both temperature and pressure inside the primary containment, within design limits, under post-accident conditions. Due to the system design and operation considerations discussed in the licensee's application, the potential for corrosion product formation is minimized. In addition, the Vermont Yankee foreign material exclusion program has been judged to be sufficient to ensure that foreign material is not inadvertently introduced into the system. The proposed testing requirements are considered sufficient to provide a high degree of confidence that containment spray will function when required. Therefore, the proposed change does not involve a significant increase in the consequences of any accident previously evaluated.

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

Response: No.

The proposed change in the surveillance test frequency does not create the possibility of a new or different kind of accident, since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode, creating an accident or affecting the mitigation of an accident.

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

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

Response: No.

The proposed change revises the surveillance requirement to verify that the drywell spray headers and nozzles are unobstructed. Industry experience, Vermont Yankee surveillance history and the environmental conditions the system is subjected to are adequate to ensure continued system availability. As the spray nozzles are expected to remain unobstructed and be able to perform their post-accident function, plant safety is not affected.

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

Based on this review, it appears that the three standards of 10 CFR 50.92 are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: Allen G. Howe.

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

Date of amendment request: July 9, 2004.

Description of amendment request: The Humboldt Bay Power Plant (HBPP), Unit 3, is a decommissioning nuclear power plant that was permanently shutdown in July 1976. In December of 2003, Pacific Gas and Electric (PG&E or the licensee) applied for a license to store its spent fuel in an onsite dry cask independent spent fuel storage installation (ISFSI). Moving the spent fuel to an ISFSI would permit the licensee to begin significant decommissioning activities. The licensee has chosen to use a Holtec HI-STAR HB spent fuel cask handling system involving a spent fuel multipurpose canister and overpack. To facilitate spent fuel transfer from the HBPP spent fuel pool to the ISFSI, the licensee will also need to install a new crane that can be used to lift the cask handling system loaded with spent fuel assemblies. The licensee states it will be able to satisfy the applicable guidance of NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants,” and NUREG-0554. “Single-Failure Proof Cranes for Nuclear Power Plants,” in performing the necessary movement of the HBPP spent fuel to dry cask storage. The licensee has requested a license amendment that approves the use of the crane and associated changes to the HBPP Defueled Safety Analysis Report (DSAR) along with analyses, design, and procedural changes required to implement transfer of the spent fuel from the spent fuel pool to the ISFSI.

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

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

No. With the HI-STAR HB System and the associated design and handling procedures, all cask drops and other events, which could damage other spent fuel, have been precluded through the robust handling systems, and mechanical arrangement that preclude crane movement over spent fuel, meeting the guidelines of NUREG-0612. Revisions of the HBPP procedures implementing the control of heavy loads ensures that PG&E will meet the NUREG-0612 guidelines and will protect the fuel storage locations and the new HI-STAR HB System loading/unloading activities. As a result of this design approach, a cask-handling accident that results in a significant offsite radiological release is not considered credible as demonstrated by the probabilistic evaluation that was performed using the guidelines of NUREG-0612 Appendix B and updated information from NUREG-1774 [“A Survey of Crane Operating Experience at U.S. Nuclear Power Plants from 1968 through 2002.”]

Other HBPP licensing-basis events, such as the drop of a spent fuel assembly, have not been affected by these changes and remain bounding events for potential radiological consequences.

The proposed design of the dry cask system, the handling system, and associated procedural controls provide assurance that: (1) operational errors and mishandling events, and (2) support system malfunctions will not result in an increase in the probability or consequence of an accident previously analyzed.

The proposed changes to use the Holtec HI-STAR HB system have been evaluated for seismic events and tornado missile impacts and it has been determined that these changes will not result in an increase in the probability or consequences of an accident previously evaluated. The Fire Protection Program will ensure that the combustible materials are properly controlled such that the total combustibles meet the current program commitments. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident.

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

No. The engineering design measures and the handling procedures preclude the possibility of new or different kinds of accidents. Damage to 10 CFR 50 structures, systems, and components from the cask handling and associated activities, and events resulting from possible damage to contained fuel have been considered. Both the types of accidents and the results remain within the envelope of existing HBPP DSAR licensing basis analyses, as demonstrated by the PG&E and Holtec analyses.

The rupture of multipurpose canister (MPC) dewatering, forced helium dehydration or related closure system lines or the malfunction of equipment during cask handling operations resulting in radiological consequences are bounded by the HBPP DSAR fuel-handling accident analysis.

Other design considerations, such as spent fuel pool (SFP) thermal, water chemistry and clarity, criticality, and structural, were evaluated and determined not to introduce the possibility of a new or different kind of accident from any previously evaluated.

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

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

No. With the Holtec HI-STAR HB System, and the associated design and handling procedures, cask drops and other events have been precluded through robust load handling systems, providing defense-in-depth as described in NUREG-0612. Cask tipovers, while not considered credible, are shown to be below the 60g limit, preventing damage to the contained fuel assemblies (and associated structures), and meeting the analysis guidelines of NUREG-0612. As the existing licensing basis assumes a nonmechanistic drop damaging the SFP and all fuel, the result of this design approach with the minimization of drops and the associated structural challenges assure the margin of safety has been maintained.

Other HBPP licensing-basis events, such as the drop of a spent fuel assembly, have not been affected by these changes and remain bounding events. Revision of HBPP procedures implementing the control of heavy loads to incorporate the additional restrictions on heavy loads movement will not affect the procedures or methodology used and will, therefore, not affect margins.

Adverse effects from seismic events and/or cask drops or tipovers have been evaluated, assuring that the fuel, MPC, and overpack remain within their design bases. Since design basis criteria are fully satisfied, there is no impact on the margin of safety.

The Fire Protection Program will continue to ensure that the combustible materials are properly controlled such that the total combustibles meet the current program commitments. Thus, there are no significant reductions in margin of safety associated with these changes.

Other design considerations, such as SFP thermal, water chemistry, criticality, and structural, were evaluated and determined to not involve a reduction in a margin of safety.

Based on the above evaluations, the licensee concludes that the activities associated with the above changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92 and accordingly, a finding by the NRC of no significant hazards consideration is justified.

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: Richard F. Locke, Esquire, Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Section Chief: Claudia Craig.

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

Date of amendment request: October 6, 2004

Brief description of amendments: The proposed change will revise the Technical Specification (TS) 3.8.1, “AC Sources—Operating,” to allow surveillance testing of the onsite diesel generators (DGs) during power operation.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The licensee's analysis 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 design of plant equipment is not being modified by the proposed changes. In addition, the DGs and their associated emergency loads are accident mitigating features. As such, testing of the diesel generators (DGs) themselves is not associated with any potential accident-initiating mechanism. Therefore, there will be no significant impact on any accident probabilities by the approval of the requested changes.

The changes include an increase in the online time that a DG under test will be paralleled to the grid (for SRs [Surveillance Requirements] 3.8.1.10 and 3.8.1.14) or unavailable due to testing (per SR 3.8.1.13). However, the overall time that the DG is paralleled in all modes (outage/non-outage) should remain unchanged. As such, the ability of the tested DG to respond to a design basis accident [DBA] could be adversely impacted by the proposed changes. However, the impacts are not considered significant based, in part, on the ability of the remaining DG to mitigate a DBA or provide safe shutdown. With regard to SR 3.8.1.10 and SR 3.8.1.14, experience shows that testing per these SRs typically does not perturb the electrical distribution system and share the same electrical configuration alignment as the current monthly surveillance. In addition, operating experience and qualitative evaluation of the probability of the DG or bus loads being adversely affected concurrent with or due to a significant grid disturbance, while the DG is being tested, support the conclusion that the proposed changes do not involve any significant increase in the likelihood of a safety-related bus blackout or damage to plant loads.

The SR changes that are consistent with TSTF [Technical Specification Task Force]-283 have been approved generically and for individual Licensees. The on-line tests allowed by the TSTF are only to be performed for the purpose of establishing OPERABILITY. Performance of these SRs during restricted MODES will require an assessment to assure plant safety is maintained or enhanced.

Deletion of expired TS LCO [Limiting Condition for Operation] 3.8.1, Required Action A.3, one-time 21-day Completion Time allowance for Startup Transformer XST2 preventive maintenance is an administrative change only. 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 changes would not create any new accidents since no changes are being made to the plant that would introduce any new accident causal mechanisms. Equipment will be operated in the same configuration as currently allowed for other DG SRs that allow testing during at-power operation. Deletion of expired TS LCO 3.8.1, Required Action A.3, one-time 21-day Completion Time allowance for Startup Transformer XST2 preventive maintenance is an administrative change only. This license amendment request does not impact any plant systems that are accident initiators; neither does it adversely impact any accident mitigating systems.

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 changes do not involve a significant reduction in the margin of safety. The margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed changes do not directly affect these barriers, nor do they involve any significant adverse impact on the DGs which serve to support these barriers in the event of an accident concurrent with a loss of offsite power. The proposed changes to the testing requirements for the plant DGs do not affect the OPERABILITY requirements for the DGs, as verification of such OPERABILITY will continue to be performed as required (except during different allowed MODES). The changes have an insignificant impact on DG availability, as continued verification of OPERABILITY supports the capability of the DGs to perform their required function of providing emergency power to plant equipment that supports or constitutes the fission product barriers. Only one DG is to be tested at a time, so that the remaining DG will be available to safely shut down the plant if required. Consequently, performance of the fission product barriers will not be impacted by implementation of the proposed amendment.

In addition, the proposed changes involve no changes to setpoints or limits established or assumed by the accident analysis. On this and the above basis, no safety margins will be impacted.

Deletion of expired TS LCO 3.8.1, Required Action A.3, one-time 21-day Completion Time allowance for Startup Transformer XST2 preventive maintenance is an administrative change only.

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

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

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

NRC Section Chief: Mike Webb, Acting Chief.

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

Date of amendment request: October 13, 2004.

Brief description of amendments: The proposed changes will revise the Technical Specifications (TSs) to incorporate two topical reports used to determine the core operating limits of Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, and delete reference to four topical reports and a reference to NUREG-0800 that are no longer required to support CPSES, Units 1 and 2, 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. The licensee's analysis is presented below:

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

Response: No.

The proposed change is administrative in nature and as such does not impact the condition or performance of any plant structure, system or component. The core operating limits are established to support Technical Specifications 3.1, 3.2, 3.3, and 3.4. The core operating limits ensure that fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO). The methods used to determine the core operating limits for each operating cycle are based on methods previously found acceptable by the NRC and listed in TS section 5.6.5.b. Application of these approved methods will continue to ensure that acceptable operating limits are established to protect the fuel cladding integrity during normal operation and AOOs. The requested Technical Specification changes do not involve any plant modifications or operational changes that could affect system reliability, performance, or possibility of operator error. The requested changes do not affect any postulated accident precursors, do not affect any accident mitigation systems, and do not introduce any new accident initiation mechanisms.

As a result, the proposed change to the CPSES Technical Specifications does not involve any increase in the probability or the consequences of any accident or malfunction of equipment important to safety previously evaluated since neither accident probabilities nor consequences are being affected by this proposed administrative change.

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

Response: No.

The proposed change is administrative in nature, and therefore does not involve any changes in station operation or physical modifications to the plant. In addition, no changes are being made in the methods used to respond to plant transients that have been previously analyzed. No changes are being made to plant parameters within which the plant is normally operated or in the setpoints, which initiate protective or mitigative actions, and no new failure modes are being introduced.

Therefore, the proposed administrative change to the CPSES Technical Specifications does not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any previously evaluated.

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

Response: No.

The proposed change is administrative in nature and does not impact station operation or any plant structure, system or component that is relied upon for accident mitigation. Furthermore, the margin of safety assumed in the plant safety analysis is not affected in any way by the proposed administrative change.

Therefore, the proposed change to the CPSES Technical Specifications does not involve any reduction in a margin of safety.

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

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

NRC Section Chief: Michael Webb, Acting Chief.

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.

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

Date of amendment request: November 8, 2004.

Description of amendment request: To revise Technical Specification Section 4.4.5.4 to modify the definitions of steam generator (SG) tube “Plugging Limit” and “Tube Inspection.”

Date of publication of individual notice in the Federal Register: November 24, 2004 (69 FR 68408).

Expiration date of individual notice: January 24, 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 [phone], [phone] or by e-mail to [e-mail].

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

Date of application for amendment: April 1, 2004.

Brief description of amendment: The amendment revises Technical Specification (TS) requirements to adopt the provisions of the TS Task Force (TSTF) change TSTF-359, regarding increased flexibility in mode changes. The availability of TSTF-359 for adoption by licensees was announced in the Federal Register on April 4, 2003 (68 FR 16579).

Date of issuance: November 29, 2004.

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

Amendment No.: 163.

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

Date of initial notice in Federal Register: August 24, 2004 (69 FR 52037).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated November 29, 2004.

No significant hazards consideration comments received: No.

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

Date of amendment request: October 7, 2004, as supplemented by letters dated November 12 and 18, 2004.

Description of amendment request: The amendment revised the Safety Limit Minimum Critical Power Ratio in Technical Specification 2.1.1.2 to reflect the results of cycle-specific calculations performed for Fermi 2 operating Cycles 10 and 11.

Date of issuance: November 30, 2004.

Effective date: As of the date of issuance and shall be implemented prior to startup for Fermi 2 Cycle 11 operation.

Amendment No.: 164.

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

Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. (November 9, 2005; 69 FR 64986) The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided an opportunity to request a hearing by January 10, 2005, but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.

The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated November 30, 2004.

Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.

NRC Section Chief: L. Raghavan.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania

Date of application for amendments: January 26, 2004, as supplemented September 13, 2004.

Brief description of amendments: These amendments authorized changes to the BVPS-1 and 2 Updated Final Safety Analysis Reports (UFSARs) to revise the level of the Ohio River that is assumed at the onset of an accident during power operation to be 654.0′ mean sea level (msl) instead of 649.0′ msl for BVPS-1 and 2. The proposed change is consistent with current Technical Specification 3.7.5.1, which requires the plant to shut down when the Ohio River reaches a level below 654.0′ msl.

Date of issuance: November 29, 2004.

Effective date: As of the date of issuance and shall submit the changes authorized by these amendments with the next update of the UFSARs in accordance with 10 CFR 50.71(e).

Amendment Nos.: 264 and 145.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments authorize changes to the UFSARs.

Date of initial notice in Federal Register: March 16, 2004 (69 FR 12369).

The supplement dated September 13, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 29, 2004 .

No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of application for amendments: November 21, 2003, as supplemented by letters dated May 18, and August 23, 2004.

Brief description of amendments: These amendments revised the St. Lucie Unit 1 and 2 Technical Specifications (TSs) to eliminate certain pressure sensor response time testing requirements. Elimination of these tests is discussed in the Combustion Engineering Owners Group Topical Report CE NPSD-1167, Revision 2, “Elimination of Pressure Sensor Response Time Testing Requirements,” which was approved by the NRC staff in letters dated July 24, 2000, and December 5, 2000. Specifically, these amendments revise the St. Lucie Units 1 and 2 TS Definitions 1.12, “Engineered Safety Features Response Time,” and 1.26, “Reactor Protection System Response Time.”

Date of Issuance: November 30, 2004.

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

Amendment Nos.: 195 and 137

Renewed Facility Operating License Nos. DPR-67 and NPF-16: Amendments revised the TSs.

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57675).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 30, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: November 26, 2002, as supplemented by letters dated September 8, 2003, October 30, 2003, June 21, 2004, and October 8, 2004.

Brief description of amendments: The amendments increased the total spent fuel wet storage capacity for each unit, by adding a spent fuel storage rack in the cask area in each unit's spent fuel pool. Each rack increased both units' storage capacity by 131 fuel assemblies. The amendments also included the addition of the design of the racks in Section 5.6.1.1.c of the Technical Specifications (TSs), and revised the stated spent fuel capacity in TS Section 5.6.3 and the location called out in the Design Features Sections 5.6.1.1a and b of the TSs referring to Updated Final Safety Analysis Report Appendix 14D rather the Westinghouse Report WCAP-14416-P.

Date of issuance: November 24, 2004.

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

Amendment Nos: 226 and 222.

Renewed Facility Operating License Nos. DPR-31 and DPR-41: Amendments revised the TSs.

Date of initial notice in Federal Register: January 28, 2003 (69 FR 4246). The supplemental letters provided clarifying information that did not expand the scope of the original application or change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in an Environmental Assessment dated October 17, 2003, and in a Safety Evaluation dated November 24, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendment: August 17, 2004.

Brief description of amendment: The amendment revised Section 3.3.1, “Oxygen Concentration,” of the Technical Specifications to add a new action, allowing 24 hours to restore the oxygen concentration within the limit of <4% by volume if the limit is exceeded when the reactor is operating in the power operating condition.

Date of Issuance: November 29, 2004.

Effective date: November 29, 2004 and shall be implemented within 15 days of issuance.

Amendment No.: 185.

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

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

The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated November 29, 2004.

No significant hazards consideration comments received: No.

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

Date of application for amendments: September 12, 2003, and its supplements dated April 23, June 4, and August 30, 2004.

Brief description of amendments: The amendments increase the current steam generator narrow range water level-low low setpoints from greater or equal to 7.0 percent allowable value and 7.2 percent nominal trip setpoint to greater than or equal to 14.8 percent allowable value and 15.0 percent nominal trip setpoint. The reactor trip setpoint is specified in TS Table 3.3.1-1, “Reactor Trip System Instrumentation,” and the actuation setpoint to start the auxiliary feedwater pumps is specified in TS Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation.”

Date of issuance: December 2, 2004.

Effective date: December 2, 2004, and shall be implemented within 90 days from the date of issuance.

Amendment Nos.: Unit 1-178; Unit 2-180.

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

Date of initial notice in Federal Register: November 25, 2003 (68 FR 66138) The April 23, June 4, and August 30, 2004, supplemental letters provided additional clarifying information, 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 the amendments is contained in a Safety Evaluation dated December 2, 2004.

No significant hazards consideration comments received: No.

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 application for amendments: December 5, 2003, as supplemented by letter dated June 4, 2004.

Brief description of amendments: The amendments revised SSES 1 and 2 Technical Specifications (TSs) by adding a requirement to apply linear heat generation (LHGR) limits if the main turbine bypass system becomes inoperable. The proposed changes clarify TS 3.7.6 to state that both minimum critical power ratio and LHGR limits for an inoperable main turbine bypass system are required if the system becomes inoperable.

Date of issuance: December 3, 2004.

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

Amendment Nos.: 218 and 193.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 6, 2004 (69 FR 698). The supplement dated June 6, 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 the amendments is contained in a Safety Evaluation dated December 3, 2004.

No significant hazards consideration comments received: No.

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: August 12, 2004, as superseded by letter dated October 5, 2004, as supplemented by letter dated October 11, 2004.

Brief description of amendments: The amendments revise Technical Specification (TS) 3/4.4.5, in conjunction with the new administrative control TS 6.8.3.o and reporting requirement TS 6.9.1.7, to establish a new programmatic, largely performance-based framework for ensuring SG tube integrity. The reactor coolant system leakage requirements of TS 3.4.6.2 are also revised.

Date of issuance: November 24, 2004.

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

Amendment Nos.: Unit 1—164; Unit 2—154.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 31, 2004 (69 FR 53113). The October 5, 2004, letter which superseded the August 12, 2004, letter and the supplement dated October 11, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not significantly change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 24, 2004.

No significant hazards consideration comments received: No.

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: August 26, 2004.

Brief description of amendments: The amendments eliminate the requirements in the TS associated with hydrogen recombiners and hydrogen monitors. A notice of availability for this TS improvement using the consolidated line item improvement process was published in the Federal Register on September 25, 2003 (68 FR 55416).

Date of issuance: November 30, 2004.

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

Amendment Nos.: Unit 1—165; Unit 2—155.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 28, 2004 (69 FR 57996).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated November 30, 2004.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 13th day of December 2004.

For the Nuclear Regulatory Commission.

James E. Lyons,

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