06-744. [No title available]

I. Background

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

This biweekly notice includes all notices of amendments issued, or proposed to be issued from January 6, 2006 to January 19, 2006. The last biweekly notice was published on January 17, 2006 (71 FR 2586).

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

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

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

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

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request 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].

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

Date of amendments request: December 23, 2005.

Description of amendments request: The amendments would increase the emergency diesel generator (EDG) allowed out of service time (AOT) from 72 hours to 10 days, allow EDG starting air receiver pressure to momentarily drop below limits during successful starting of an EDG, and remove from the Technical Specifications the statement that the two groups of pressurizer heaters are capable of being powered from an emergency power supply.

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 Technical Specification (TS) change to increase the emergency diesel generator (EDG) allowed out of service time (AOT) from 72 hours to 10 days will not cause an accident to occur and will not result in any change in the operation of the associated accident mitigation equipment. The EDGs are not accident initiators. The EDGs are designed to mitigate the consequences of previously evaluated accidents including a loss of offsite power. Extending the AOT for a single EDG would not affect the previously evaluated accidents since the remaining EDG supporting the redundant Engineered Safety Features (ESF) systems would continue to be available to perform the accident mitigation functions. The duration of this TS AOT considers that there is a minimal possibility that an accident will occur while a component is removed from service. A risk informed assessment was performed which concluded that the increase in plant risk is small and consistent with the guidance contained in Regulatory Guide 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications.” The design basis accidents will remain the same postulated events described in the PVNGS [Palo Verde Nuclear Generating Station] Updated Final Safety Analysis Report (UFSAR). In addition, extending the EDG AOT will not impact the consequences of an accident previously evaluated. The consequences of previously evaluated accidents will remain the same during the proposed 10 day AOT as during the current 72 hour AOT. The ability of the remaining TS-required EDG to mitigate the consequences of an accident will not be affected since no additional failures are postulated while equipment is inoperable within the TS AOT. The remaining EDG is sufficient to mitigate the consequences of any design basis accident.

The proposed addition of a note to Condition F of TS 3.8.3, would allow EDG starting air receiver pressure to momentarily drop below limits during successful starting of an EDG. The EDG air starting system will not be operated or be configured any differently than that which it is currently required and designed for. This proposed change will only add a note for clarification to Condition F of TS 3.8.3. This note describes entering this Condition is not necessary when the EDG starts normally and is operating per required procedures. Momentary transients outside the air receiver pressure range do not invalidate the successful start and running of the EDG. A successful start of the EDG indicates the starting air system has performed its required safety function. This proposed change will not increase the probability or consequence of an accident previously evaluated.

The proposed TS change associated with the requirements for the pressurizer heaters to be supplied by emergency power will not result in any change in plant design. These components will continue to be powered from Class 1E power sources as described in the proposed TS Bases change associated with this change. As a result, the operation and reliability of the pressurizer heaters will not be affected by the proposed description change. In addition, operation of the pressurizer heaters is not assumed to mitigate any design basis accident. The proposed changes will not cause an accident to occur and will not result in a change in the operation of any accident mitigation equipment. The design basis accidents remain the same postulated events described in the PVNGS UFSAR.

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

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

Response: No.

The proposed changes do not involve a change in the design, configuration, or method of operation of the plant that could create the possibility of a new or different [kind of] accident. Equipment will be operated in the same configuration and manner that is currently allowed and designed for. The proposed changes do not introduce any new failure modes. This license amendment request does not impact any plant systems that are accident initiators or adversely impact any accident mitigating systems.

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

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

Response: No.

The EDG reliability and availability are monitored and evaluated, in accordance with 10 CFR 50.65 (Maintenance Rule) performance criteria, to assure EDG out of service times do not degrade operational safety over time. Extension of the EDG AOT will not erode the reduction in severe accident risk that was achieved with implementation of the Station Blackout (SBO) rule (10 CFR 50.63) or affect any safety analyses assumptions or inputs. The SBO coping analysis is unaffected by the AOT extension since the EDGs are not assumed to be available during the coping period. The assumptions used in the coping analysis regarding EDG reliability are unaffected since preventive maintenance and testing will continue to be performed to maintain the reliability assumptions.

Accident mitigation functions will be maintained by the other TS-required EDG availability to supply power to the safety related Class 1E electrical loads. The availability of the TS-required offsite power, combined with the availability of the PVNGS SBO Gas Turbine Generators (GTGs) and the use of the Configuration Risk Management Program required by 10 CFR 50.65(a)(4), provide adequate compensation for the small incremental increase in plant risk of the proposed EDG AOT extension. This small increase in plant risk while operating is offset by a reduction in shutdown risk resulting from the increased availability and reliability of the EDGs during refueling outages, and avoiding transition risk incurred during unplanned plant shutdowns. In addition, the calculated risk measures associated with the proposed AOT are below the acceptance criteria defined in Regulatory Guide 1.177.

The proposed change to add a note to Condition F of TS 3.8.3 does not involve changes to setpoints or limits established or assumed by the accident analyses. This note only applies to those occasions when after a successful start of an EDG has occurred and the starting air receiver pressure has momentarily dropped below its limit. This change allows for not declaring the EDG inoperable solely due to this momentary drop in pressure during a successful start of the EDG. No safety margin will be impacted by this change.

The proposed TS change associated with the wording description of LCO [Limiting Condition of Operation] 3.4.9, “Pressurizer,” for the requirement of the pressurizer heaters to be supplied by emergency power does not adversely affect equipment design or operation, and there are no changes being made to the TS-required safety limits or system settings that would adversely affect plant safety. The emergency power requirements for the pressurizer heaters, which came from the Three Mile Island (TMI) action item requirement II.E.3.1, “Emergency Power Requirements for Pressurizer Heater,” of NUREG-0737, “Clarification of TMI Action Plan Requirements,” will continue to be met. The pressurizer heaters used to satisfy the NUREG-0737 and LCO 3.4.9 requirements are, by design, permanently connected to Class 1E power supplies as described in the PVNGS Updated Final Safety Analyses Report, Section 18.II.E.3.1.

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 that review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.

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

NRC Branch Chief: David Terao.

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

Date of amendment request: November 15, 2005.

Description of amendment request: The proposed change modifies the technical specifications (TS) to clarify the wording of emergency closed cooling water (ECCW) Surveillance Requirement (SR) 3.7.10.2. The current wording in SR 3.7.10.2 requires that automatic valves on the ECCW system actuate on an actuation signal. However, the TS Bases for the SR identify more than just valves tested to include the automatic start capability of the ECCW pump in each subsystem. Therefore, the wording of this SR would be modified to clarify that its purpose is to verify actuation of the entire subsystem on an actual or simulated signal, rather than just verify valve actuation.

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

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

There are no physical modifications being made to any plant system or component. The only change is to a Surveillance Requirement within the Technical Specifications, in order to improve understanding and avoid misinterpretation of the requirements. The original intent of ECCW SR 3.7.10.2 is maintained by the change being proposed. The revised Technical Specification requirements do not impact initiators of previously evaluated accidents or transients.

The specification being revised is associated with a system used to mitigate the consequences of accidents. The change to the wording of ECCW SR 3.7.10.2 does not impact the capability of the associated system to perform its required function. The reworded ECCW SR more clearly requires that the system[']s total actuation capability be maintained.

The change does not affect how plant systems are controlled or operated or tested. The change continues to provide confirmation of the capability of plant components to respond as required to mitigate the consequences of events. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

There are no physical modifications being made to any plant system or component, and the proposed change introduces no new method of operation of the plant, or its systems or components. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

The change to the ECCW SR continues to ensure the ECCW subsystems are tested on the same periodicity to verify their capability to respond to actuation signals from the Emergency Core Cooling System (ECCS) Instrumentation Functions of Low Water Level and High Drywell Pressure. Therefore, the necessary function of the Technical Specification requirements is maintained, and the proposed change does not involve a significant reduction in a margin of safety.

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

Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy Corporation, Mail Stop A-GHE-107, 76 South Main Street, Akron, OH 44308.

NRC Branch Chief: Mindy Landau, Acting.

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

Date of amendment request: December 13, 2005.

Description of amendment request: The proposed amendments would revise technical specification (TS) requirements for surveillance requirements for containment integrated leakage rate testing in TS 5.5.14.a to allow a one-time extension of the interval between reactor containment vessel integrated leakage rate tests (ILRTs) from 10 to 15 years.

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

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

Response: No.

This license amendment proposes to revise the Technical Specifications to allow for the one time extension of the containment integrated leakage rate test interval from 10 to 15 years. The containment vessel function is purely mitigative. There are no design basis accidents initiated by a failure of the containment leakage mitigation function. The extension of the containment integrated leakage rate test interval will not create any adverse interactions with other systems that could result in initiation of a design basis accident. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased.

The potential consequences of the proposed change have been quantified by analyzing the changes in risk that would result from extending the containment integrated leakage rate test interval from 10 to 15 years. The increase in risk in terms of person-rem per year within 50 miles resulting from design basis accidents was estimated to be of a magnitude that NUREG-1493, “Performance-Based Containment Leak-Test Program”, indicates is imperceptible. The Nuclear Management Company has also analyzed the increase in risk in terms of the frequency of large early releases from accidents. The increase in the large early release frequency resulting from the proposed extension was determined to be within the guidelines published in Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis”. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. The Nuclear Management Company has determined that the increase in conditional containment failure probability from reducing the containment integrated leakage rate test frequency from 1 test per 10 years to 1 test per 15 years would be small.

Continued containment integrity is also assured by the history of successful containment integrated leakage rate tests, and the established programs for local leakage rate testing and in-service inspections which are unaffected by the proposed change. Therefore, the probability of occurrence or the consequences of an accident previously analyzed are not significantly increased.

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 to extend the containment integrated leakage rate test interval from 10 to 15 years does not create any new or different accident initiators or precursors. The length of the containment integrated leakage rate test interval does not affect the manner in which any accident begins. The proposed change does not create any new failure modes for the containment and does not affect the interaction between the containment and any other system. Thus, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The risk-based margins of safety associated with the containment integrated leakage rate test are those associated with the estimated person-rem per year, the large early release frequency, and the conditional containment failure probability. The Nuclear Management Company has quantified the potential effect of the proposed change on these parameters and determined that the effect is not significant. The non-risk-based margins of safety associated with the containment integrated leakage rate test are those involved with its structural integrity and leak tightness. The proposed change to extend the containment integrated leakage rate test interval from 10 to 15 years does not adversely affect either of these attributes. The proposed change only affects the frequency at which these attributes are verified. Therefore, the proposed change does not involve a significant reduction in margin of safety.

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

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

NRC Acting Branch Chief: Timothy Kobetz.

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

Date of amendment request: November 18, 2005.

Description of amendment request: The proposed amendment would change the SSES 1 and 2 Technical Specifications (TSs) to implement the Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). Specifically, the average power range monitor (APRM) flow-biased scram and rod block trip setpoints would be revised to permit operation in the MELLLA region. The current flow-biased rod block monitor (RBM) would also be replaced by a power dependent RBM implemented through the referenced proposed upgrade to a digital power range neutron monitor system (PRNMS). The change from the flow-biased RBM to the power-dependent RBM would also require new trip setpoints. In addition, the flow-biased APRM scram and rod block trip setdown requirement would be replaced by more direct power and flow-dependent thermal limits to reduce the need for APRM gain adjustments, and to allow more direct thermal limits administration during operation other than rated conditions. Finally, the proposed amendment would change the methods used to evaluate the annulus pressurization (AP), mass blowdown, and early release resulting from the postulated recirculation suction line break (RSLB).

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

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

Response: No.

Proposed Change No. 1: The proposed change eliminates the Average Power Range Monitor (APRM) flow-biased scram and rod block trip setpoint setdown requirements and substitutes power and flow dependent adjustments to the Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) thermal limits. Thermal limits will be determined using NRC approved analytical methods. The proposed change will have no effect upon any accident initiating mechanism. The power and flow dependent adjustments will ensure that the MCPR safety limit will not be violated as a result of any Anticipated Operational Occurrence (AOO), and that the fuel thermal and mechanical design bases will be maintained. Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

Proposed Change No. 2: The proposed change expands the power and flow operating domain by relaxing the restrictions imposed by the formulation of the APRM flow-biased scram and rod block trip setpoints and the replacement of the current flow-biased RBM with a new power dependent RBM, which will be implemented using a digital Power Range Neutron Monitoring System (PRNMS). The APRM and RBM are not involved in the initiation of any accident; and the APRM flow-biased scram and rod block functions are not credited in any PPL safety licensing analyses.

The analysis of the instrument line break event resulted in an insignificant change in the radiological consequences. The change for the instrument line break was an insignificant increase of 0.1 Rem.

Since the proposed changes will not affect any accident initiator, or introduce and initial conditions that would result in NRC approved criteria being exceeded, and since the APRM and RBM will remain capable of performing their design functions, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

Proposed Change No. 3: The methods used to evaluate Annulus Pressurization (AP) and mass blowdown and energy releases resulting from the postulated Recirculation Suction Line Break (RSLB) at the MELLLA conditions are changed to use more realistic, but still conservative, methods of analysis to determine an AP mass and energy release profile for AP loads resulting from the postulated RSLB. The releases resulting from the RSLB at off-rated conditions have been demonstrated to be bounded by the current design basis loads. Since the proposed changes do not affect any accident initiator and since the RSLB AP releases remain bounded by the current design basis, the proposed changes do not involve a significant increase in the probability or radiological consequences of an accident previously evaluated. Therefore the proposed changes do not involve a significant increase in the probability or consequences of any accident previously evaluated.

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

Response: No.

Proposed Change No. 1: The proposed change eliminates the Average Power Range Monitor (APRM) flow-biased scram and rod block setpoint setdown requirements and substitutes power and flow dependent adjustments to the Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) thermal limits. Because the thermal limits will continue to be met, no analyzed transient event will escalate into a new or different type of accident due to the initial starting conditions permitted by the adjusted thermal limits. Therefore, the proposed change will not create the possibility of a new or different kind of accident previously evaluated.

Proposed Change No. 2: The proposed change expands the power and flow operating domain by relaxing the restrictions imposed by the formulation of the APRM flow-biased scram and rod block trip setpoints and the replacement of the current flow-biased RBM with a new power dependent RBM, which will be implemented using a digital Power Range Neutron Monitoring System (PRNMS). Changing the formulation for the APRM flow-biased scram and rod block trip setpoints and from a flow-biased RBM to a power dependent RBM does not change their respective functions and manner of operation. The change does not introduce a sequence of events or introduce a new failure mode that would create a new or different type of accident. The APRM flow-biased rod block trip setpoint will continue to block control rod withdrawal when core power significantly exceeds normal limits and approaches the scram level. The APRM flow-biased scram trip setpoint will continue to initiate a scram if the increasing power/flow condition continue beyond the APRM flow-biased rod block setpoint. The power dependent RBM will prevent rod withdrawal when the power dependent RBM rod block setpoint is reached. No new failure mechanisms, malfunctions, or accident initiators are being introduced by the proposed changes. In addition, operating within the expanded power flow map will not require any systems, structures or components to function differently than previously evaluated and will not create initial conditions that would result in a new or different kind of accident from any accident previously evaluated.

Proposed Change No. 3: The methods used to evaluate Annulus Pressurization (AP) and mass blowdown and energy releases resulting from the postulated Recirculation Suction Line Break (RSLB) at the MELLLA conditions are changed to use more realistic, but still conservative, methods of analysis to determine an AP mass and energy release profile for AP loads resulting from the postulated RSLB. The proposed changes to the methods of analysis to determine AP mass and energy releases resulting from the postulated RSLB do not change the design function or operation of any plant equipment. No new failure mechanisms, malfunctions, or accident initiators are being introduced by the proposed changes. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

Proposed Change No. 1: The proposed change eliminates the Average Power Range Monitor (APRM) flow-biased scram and rod block setpoint setdown requirements and substitutes power and flow dependent adjustments to the Minimum Critical Power Ratio (MCPR) and Linear Heat Generation Rate (LHGR) thermal limits. Replacement of the APRM setpoint setdown requirement with power and flow dependent adjustments to the MPR and LHGR thermal limits will ensure that margins to the fuel cladding Safety Limit are preserved during operation at other than rated conditions. Thermal limits will be determined using NRC approved analytical methods. The power and flow dependent adjustments will ensure that the MPR safety limit will not be violated as a result of any Anticipated Operational Occurrence (AOO), and that the fuel thermal and mechanical design bases will be maintained. The 10 CFR 50.46 acceptance criteria for the performance of the Emergency Core Cooling System (ECCS) following postulated Loss-Of-Coolant Accidents (LOCAs) will continue to be met. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

Proposed Change No. 2: The proposed change expands the power and flow operating domain by relaxing the restrictions imposed by the formulation of the APRM flow-biased scram and rod block trip setpoints and the replacement of the current flow-biased RBM with a new power dependent RBM, which will be implemented using a digital Power Range Neutron Monitoring System (PRNMS). The APRM flow-biased rod block trip setpoint will continue to block control rod withdrawal when core power significantly exceeds normal limits and approaches the scram level. The APRM flow-biased scram trip setpoint will continue to initiate a scram if the increasing power/flow condition continues beyond the APRM flow-biased rod block setpoint. The RBM will continue to prevent rod withdrawal when the power dependent RBM rod block setpoint is reached. The MPR and LHGR thermal limits will be developed to ensure that fuel thermal mechanical design bases shall remain within the licensing limits during a rod withdrawal error event and to ensure that the MPR safety limit will not be violated as a result of a rod withdrawal error event. Operation in the expanded operating domain will not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. Anticipated operational occurrences and postulated accident within the expanded operating domain will be evaluated using NRC approved methods. Therefore, the proposed change will not involve a significant reduction in the margin of safety.

Proposed Change No. 3: The methods used to evaluate Annulus Pressurization (AP) and mass blowdown and energy releases resulting from the postulated Recirculation Suction Line Break (RSLB) at the MELLLA conditions are changed to use more realistic, but still conservative, methods of analysis to determine an AP mass and energy release profile for AP loads resulting from the postulated RSLB. Mass and energy releases for AP loads resulting from the postulated RSLB remain bounded by the current design basis releases. 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: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Branch Chief: Richard J. Laufer.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: October 11, 2005.

Description of amendment request: The proposed amendment would remove the Technical Specification (TS) 3.1.5 requirement for the Standby Liquid Control (SLC) system to be operable in Operational Condition 5 (refueling) with any control rod withdrawn. Corresponding changes would also be made to the SLC Initiation sections of Tables 3.3.2-1 and 4.3.2-1.

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

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

Response: No.

The proposed changes to delete the operability requirement for the SLC System in OPERATIONAL CONDITION 5* (OPERATIONAL CONDITION 5 with any control rod withdrawn) does not affect the probability or consequences of an accident previously evaluated. In STARTUP and POWER OPERATION, the SLC System is required to provide shutdown capability. In HOT SHUTDOWN and COLD SHUTDOWN, control rods are not able to be withdrawn since the reactor mode switch is in Shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. Design basis accident mitigation scenarios for OPERATIONAL CONDITION 5 do not depend on, or require, SLC System operability. In REFUELING mode, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate shutdown margin in accordance with TS LIMITING CONDITION FOR OPERATION 3.1.1 ensures that the reactor will not become critical. Since the purpose of the SLC System is to bring the reactor to a cold shutdown condition from normal power operations and maintain it in a cold shutdown condition, there is no design basis for the SLC System to be required to be OPERABLE when only a single control rod can be withdrawn. In addition, the reactor protection system and the control rod system would continue to be able to provide protection in the unlikely event that an inadvertent criticality occurs.

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

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

Response: No.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated in the UFSAR [updated final safety analysis report]. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. Specifically, no new hardware is being added to the plant as part of the proposed change, no existing equipment is being modified, and no significant changes in operations are being introduced.

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

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

Response: No.

The proposed changes will not alter any assumptions, initial conditions, or results of any accident analyses. The purpose of the SLC System is to bring the reactor to and maintain it in a cold shutdown condition following a failure to scram during plant operations. The SLC System is not designed to terminate an inadvertent criticality during REFUELING. Shutdown margin, either demonstrated or analytically determined, in accordance with Technical Specifications and procedural controls, will assure that an inadvertent criticality event will not occur during REFUELING. In addition, the reactor protection system and control rod system provide protection in the unlikely event that an inadvertent criticality occurs. The proposed change does not affect the ability of the SLC System to achieve plant shutdown under analyzed conditions (POWER OPERATION and STARTUP).

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

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

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

NRC Branch Chief: Darrell J. Roberts.

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].

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: October 5, 2004, as supplemented March 22, August 29, and October 31, 2005.

Brief description of amendments: The amendments revised the BVPS-1 and 2 Technical Specifications (TSs) 3/4.3.1, “Reactor Trip System Instrumentation,” and 3/4.3.2, “Engineered Safety Feature Actuation Instrumentation,” to modify steam generator (SG) level allowable value (AV) setpoints. Specifically, the TS changes increased the AVs of the SG water level-low-low setpoints from 14.6 percent and 16 percent to 19.6 percent and 20 percent of the narrow range (NR) instrument span for BVPS-1 and 2, respectively. These are the AVs of setpoints specified in TS Table 3.3-1 to initiate a reactor trip, and the actuation setpoints specified in TS Table 3.3-3 to start the auxiliary feedwater pumps. Also, for BVPS-2, the AV of the SG water level-high-high setpoint increased from 81.1 percent to 92.7 percent of the NR span. This is the AV of a setpoint for actuation of the turbine trip and the feedwater system isolation specified in TS Table 3.3-3.

Date of issuance: January 11, 2006.

Effective date: Upon issuance and shall be implemented within 60 days.

Amendment Nos.: 270 and 152.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: November 23, 2004 (69 FR 68183). The supplements dated March 22, August 29, and October 31, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 11, 2006.

No significant hazards consideration comments received: No.

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

Date of amendment request: January 10, 2005.

Description of amendment request: The amendment revised the Seabrook Station, Unit No. 1, Technical Specifications (TSs) to extend the interval for the performance of Containment Air Lock Interlock Surveillance Requirement 4.6.1.3 from 6 months to 24 months.

Date of issuance: January 6, 2006.

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

Amendment No.: 106.

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

Date of initial notice in Federal Register: May 24, 2005 (70 FR 29796). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 6, 2006.

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: September 1, 2005.

Brief description of amendments: The amendments delete the Technical Specification requirements for Occupational Radiation Exposure Reports and Monthly Operating Reports.

Date of Issuance: January 13, 2006.

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

Amendment Nos.: 198 and 141.

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

Date of initial notice in Federal Register: October 25, 2005 (70 FR 61661). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 13, 2006.

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: July 21, 2005.

Brief description of amendments: The amendments delete the Technical Specification requirements for Occupational Radiation Exposure Reports and Monthly Operating Reports.

Date of issuance: January 13, 2006.

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

Amendment Nos: 228 and 224.

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

Date of initial notice in Federal Register: October 25, 2005 (70 FR 61660).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 13, 2006.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: July 29, 2005.

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

Date of issuance: January 12, 2006.

Effective date: As of the date of issuance and shall be implemented within 45 days.

Amendment Nos.: 292, 274.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 6, 2005 (70 FR 72673). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 12, 2006.

No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: March 8, 2005, as supplemented by letter dated August 18, 2005.

Brief description of amendment: The amendment revised the Technical Specification 2.1.1.2 for the single recirculation loop Safety Limit Minimum Critical Power Ratio value to reflect results of a cycle-specific calculation.

Date of issuance: January 4, 2006.

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

Amendment No.: 215.

Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 29, 2005 (70 FR 15944). The supplement dated August 18, 2005, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 4, 2006.

No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: July 21, 2005.

Brief description of amendment: The amendment revises the technical specifications testing frequency for the surveillance requirement (SR) in TS 3.1.4, “Control Rod Scram Times.” Specifically, the proposed change would revise the frequency for SR 3.1.4.2, control rod scram time testing, from “120 days cumulative operation in MODE 1” to “200 days cumulative operation in MODE 1.”

Date of issuance: January 5, 2006.

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

Amendment No.: 216.

Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 25, 2005 (70 FR 61661). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 5, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: June 30, 2004.

Brief description of amendment: The amendment revised Table 4.2.1, “Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation,” of the Technical Specifications to shorten the test interval between surveillance tests for the scram discharge volume high level rod block, and the safety/relief valve low-low set logic inhibit timer.

Date of issuance: January 12, 2006.

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

Amendment No.: 144.

Facility Operating License No. DPR-22. Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: January 18, 2005 (70 FR 2892). The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 12, 2006.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of application for amendment: January 11, 2005.

Brief description of amendment: The amendment deletes requirements from the Technical Specifications for annual Occupational Radiation Exposure Reports and Monthly Operating Reports.

Date of issuance: January 11, 2006.

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

Amendment No.: 161.

Facility Operating License No. NPF-57: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: March 29, 2005 (70 FR 15946). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 11, 2006.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of application for amendment: February 25, 2005.

Brief description of amendment: The amendment revised Technical Specification 3.1.3.1, “Control Rod Operability,” for the condition of having one or more scram discharge volume vents or drain lines with inoperable valves.

Date of issuance: January 13, 2006.

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

Amendment No.: 162.

Facility Operating License No. NPF-57: This amendment revised the Technical Specifications.

Date of initial notice in Federal Register: June 7, 2005 (70 FR 33217). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 13, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendments: January 11, 2005.

Brief description of amendments: The amendments deleted requirements from the Technical Specifications (TSs) for annual Occupational Radiation Exposure Reports and Monthly Operating Reports.

Date of issuance: January 11, 2006.

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

Amendment Nos.: 270 and 251.

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

Date of initial notice in Federal Register: March 29, 2005 (70 FR 15946) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 11, 2006.

No significant hazards consideration comments received: No

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

Date of application for amendments: February 15, 2005.

Brief description of amendments: These amendments delete the total water and steam volume of the reactor coolant system from TS 5.4.2.

Date of issuance: January 11, 2006.

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

Amendment Nos.: 269 and 250.

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

Date of initial notice in Federal Register: March 29, 2005 (70 FR 15940). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 11, 2006.

No significant hazards consideration comments received: No.

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

Date of application for amendment: April 4, 2005, as supplemented by letters dated September 30 and November 8, 2005.

Brief description of amendment: The amendment supports the steam generator replacement project by temporarily allowing one of the shield building dome penetrations to be opened up to five hours a day, six days a week while in Modes 1-4 during Cycle 7 operation until entering Mode 5 at the start of the Cycle 7 refueling outage in fall 2006.

Date of issuance: January 6, 2006.

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

Amendment No.: 59.

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

Date of initial notice in Federal Register: July 19, 2005 (70 FR 41446). The supplemental letters provided clarifying information that was within the scope of the initial notice and did not change the initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 6, 2006.

No significant hazards consideration comments received: No.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, [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 or the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

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

Date of amendment request: January 3, 2006, as supplemented by letters dated January 6 and 10, 2006.

Description of amendment request: Entergy Operations, Inc. (Entergy) requests an emergency Technical Specification (TS) change to the Steam Generator Level—Low allowable value of Limiting Condition for Operation 3.3.11, “Emergency Feedwater [EFW] Initiation and Control (EFIC) System Instrumentation.” Operation at 100 percent power with the current allowable value involves an increased risk of spurious EFW initiation. Therefore, Entergy requests a revised TS allowable value of ≥ 9.34 inches and a limiting trip setpoint value of ≥ 10.42 inches in order to achieve and maintain 100 percent power operation. An actuation time delay of ≤ 10.4 seconds is also proposed to minimize the possibility of inadvertent actuations during anticipated transients such as main feedwater transients or main turbine trips.

Date of issuance: January 13, 2006.

Effective date: As of the date of issuance and shall be implemented within 7 days from the date of issuance.

Amendment No.: 227.

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

Public comments requested as to proposed no significant hazards consideration (NSHC): No. The Commission's related evaluation of the amendment, finding of emergency circumstances, state consultation, and final NSHC determination are contained in a safety evaluation dated January 13, 2006.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Stawn, 1700 K Street, NW., Washington, DC 20006-3817.

NRC Branch Chief: David Terao.

Dated at Rockville, Maryland, this 20th day of January 2006.

For the Nuclear Regulatory Commission.

Catherine Haney,

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

Footnotes

1.  To the extent that the applications contain attachments and supporting documents that are not publicly available because they are asserted to contain safeguards or proprietary information, petitioners desiring access to this information should contact the applicant or applicant's counsel and discuss the need for a protective order.