Operating licenses, amendments; no significant hazards considerations; biweekly notices,
[Federal Register: September 30, 2003 (Volume 68, Number 189)]
[Notices]
[Page 56340-56350]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr30se03-150]
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
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Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any
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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 September 5, 2003, through September 18, 2003. The last biweekly notice was published on September 18, 2003 (68 FR 54747).
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.
Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing 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.
By October 30, 2003, 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 2.714, 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 by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of 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. Petitioner must provide 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 to relief. A petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross- examine witnesses.
If a hearing is requested, 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
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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 with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to hearingdocket@nrc.gov. 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 because of continuing disruptions in delivery of mail to United States Government offices, it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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 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 NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50- 455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois.
Date of amendment request: June 11, 2003.
Description of amendment request: The proposed amendment would add a license condition to increase the completion time (CT) from 72 hours to 144 hours required to restore a unit specific essential service water (SX) train to operable status. The proposed change would be a one time change applicable to Braidwood Station, Unit 1, and both units at Byron Station.
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:
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The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes have been evaluated using the risk informed processes described in RG 1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant- Specific Changes to the Licensing Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, ``dated August 1998. The risk associated with the proposed change was found to be acceptable.
The previously analyzed accidents are initiated by the failure of plant structures, systems, or components. The SX system is not considered an initiator for any of these previously analyzed events. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component that initiated an analyzed event. No active or passive failure mechanisms that could lead to an accident are affected. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.
The unit-specific SX system consists of two separate, electrically independent, 100% capacity, safety related, cooling water trains. Each train consists of a 100% capacity pump, piping, valving, and instrumentation. The pumps and valves are remote and manually aligned, except in the unlikely event of a loss of coolant accident (LOCA). The pumps are automatically started upon receipt of a safety injection signal or an undervoltage on the engineered safety features (ESF) bus, and all essential valves are aligned to their post accident positions. The SX system is also the backup water supply to the auxiliary feedwater system and fire protection system.
The design basis of the SX system is for one SX train, in conjunction with the component cooling water (CC) system and a 100% capacity containment cooling system, to remove core decay heat following a design basis LOCA as discussed in the UFSAR, Section 6.2, ``Containment Systems.'' This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the reactor coolant system by the emergency core cooling system pumps. The SX system is designed to perform its function with a single failure or any active component, assuming the loss of offsite power. The proposed one-time increase in the CT of the operating unit's SX pump is consistent with the philosophy of the current Technical Specification LCO which allows one train of SX to be inoperable for 72 hours. This change only extends the 72 hour perspective; therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
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The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do not involve the use or installation of new equipment and the currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed changes will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases. Based on this evaluation, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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The proposed change does not involve a significant reduction in a margin of safety.
The proposed change does not alter any existing setpoints at which protective actions are initiated and no new setpoints or protective actions are introduced. The design and operations of the SX system remains unchanged. The risk associated with the proposed increase in the time an SX pump is allowed to be inoperable was evaluated using the risk informed processes described in RG 1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk- Informed Decisions on Plant-Specific Changes to the Licensing Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant- Specific, Risk-Informed Decisionmaking: Technical Specifications,'' dated August 1998. The risk was shown to be acceptable. Based on this evaluation, the proposed change does not involve a significant reduction in a margin of safety.
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The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50- 455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois.
Date of amendment request: June 27, 2003.
Description of amendment request: The proposed amendment would revise TS 3.4.10, ``Pressurizer Safety Valves,'' by changing the existing pressurizer safety valve (PSV) lift settings from `` \3\ 2460 psig and [pound] 2510 psig,'' to `` \3\ 2411 psig and [pound] 2509 psig.'' The existing TS represents a +/-1% tolerance band around a lift setting of 2485 psig. The proposed lift setting range of `` \3\ 2411 psig and [pound] 2509 psig'' represents a +/-2% tolerance band around a lift setting of 2460 psig.
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:
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The proposed TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Reanalysis/evaluations were performed to assess all transients that could be potentially impacted by the proposed PSV lift setting and tolerance band change. The proposed change in the PSV tolerance from +/-1% to +/-2% with a reduction in the lift setting from 2485 psig to 2460 psig allows a decrease in the valve minimum opening pressure and therefore, provides earlier pressurizer relief and a reduced RCS pressure. The proposed change does not affect the maximum opening pressure assumed in the non-LOCA analyses since the proposed change in maximum PSV opening pressure is insignificant and in the conservative direction. Therefore, only those transients for which it is conservative to minimize the RCS pressure (i.e., DNB and pressurizer overfill concerns) are potentially impacted by the proposed change. The reanalyses/evaluations of all the affected transients demonstrated acceptable results with no significant increase in the probability or consequences.
Further, any evaluations performed on an overpressure transient conservatively assume the upper limit of the PSV tolerance. The proposed change to the lower tolerance limit of the PSV lift setting means that an overpressure transient may be terminated at a pressure that is lower than assumed in the analysis. It has also been determined that the transient analyses are not adversely affected because the limiting transients are not sensitive to the pressure tolerance decrease. Therefore, the primary system pressure boundary is not challenged by the PSV lower tolerance limit change. The assumed maximum PSV lift setting was not changed, and therefore, does not impact analyses performed for overpressure transients. It has been determined that the design relieving capacity of the PSVs can still be met with the reduction in PSV setpoint. Except for the PSV lower lift setting and increased tolerance, the design and operation of the PSVs remains unchanged.
Based on this analysis, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated in the Byron/Braidwood Stations Updated Final Safety Analysis Report.
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The proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change in the PSV tolerance from +/-1% to +/-2% with a reduction in the lift setting from 2485 psig to 2460 psig allows a decrease in the valve minimum opening pressure and therefore provides earlier pressurizer relief and a reduced RCS pressure. The proposed change does not affect the maximum opening pressure assumed in the accident analyses since the proposed change in maximum PSV opening pressure is insignificant and in the conservative direction. The pressurizer PORVs serve to minimize challenges to the PSVs. An assessment of the impact of reducing the PSV lift setpoint and increasing the tolerance has determined that the resulting margin is sufficient to ensure that the PORVs will actuate prior to the PSVs. Except for the PSV lower lift setting and increased tolerance, the design and operation of the PSVs remain unchanged.
The proposed change does not involve the use or installation of new equipment and all currently installed equipment will not be operated in a new or different manner. No new or different system interactions are created and no new processes are introduced. The proposed change will not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing bases.
Based on this evaluation, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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The proposed TS change does not involve a significant reduction in a margin of safety.
The PSVs provide, in conjunction with the reactor protection system, overpressure protection for the RCS. The PSVs are designed to prevent the system pressure from exceeding the system safety limit, 2735 psig, which is 110% of the design pressure. The change in the upper limit of the PSV tolerance from +1% to +2% with a reduction in the nominal setpoint from 2485 psig to 2460 psig does not challenge the upper limit of the overpressure protection. The change in PSV maximum opening lift setting is insignificant and in the conservative direction with respect to overpressure protection, therefore, the proposed change does not impact analyses performed for overpressure transients. For all non-LOCA events, the analyses/ evaluations support the change in PSV lift setting and tolerance from 2485 psig +/-1% to 2460 psig +/-2%. The LOCA analyses are not impacted because the transient results in a decrease in RCS pressure and therefore, will not challenge the PSV opening pressure lift setting. The change in the PSV lift setting and tolerance also has no effect on the reactor protection or engineered safety features systems trip set points. Thus, the proposed change does not involve a significant reduction in any margin of safety.
Based on the above discussions, it has been determined that the requested TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident from any accident previously evaluated; or involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis- Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio.
Date of amendment request: August 25, 2003.
Description of amendment request: The proposed amendment would revise the Steam and Feedwater Rupture Control System (SFRCS) instrumentation Technical Specifications (TSs) to clearly identify the appropriate actions to be taken if an SFRCS instrumentation channel's output logic becomes inoperable; relocate the SFRCS instrumentation trip setpoints from the TSs to the Updated Safety Analysis Report; and decrease the SFRCS instrument channel functional test frequency from monthly to quarterly and make associated changes to the trip setpoint allowable values.
Basis for proposed no significant hazards consideration determination:
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As required by 10 CFR 50.91(a), the licensees have provided their analysis of the issue of no significant hazards consideration, which is presented below:
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Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not change any accident initiator, initiating condition, or assumption, and do not involve a significant change to plant design or operation. In addition, the proposed changes do not increase the likelihood of a malfunction of any plant structures, systems, or components, do not invalidate assumptions used in evaluating the radiological consequences of an accident, do not alter the source term or containment isolation, and do not provide a new radiation release path or alter radiological consequences. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
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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 introduce a new or different accident initiator or introduce a new or different equipment failure mode or mechanism. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
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Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The SFRCS instrumentation setpoint analyses will continue to adequately preserve the margin of safety. In addition, there are no new or significant changes to the initial conditions contributing to accident severity or consequences. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska.
Date of amendment request: August 25, 2003.
Description of amendment request: The proposed amendment would revise the Technical Specifications (TS) Surveillance Requirement (SR) 3.3.2.1.4 and TS Table 3.3.2.1-1 for mathematical symbols and use of Allowable Values in the place of Analytical Limits.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
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Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change to the Cooper Nuclear Station (CNS) Technical Specifications (TS) corrects the mathematical symbols for the RBM [Rod Block Monitor] LPSP [Low Power Setpoint], IPSP
[Intermediate Power Setpoint] , and the HPSP [High Power Setpoint] to clarify the power ranges at which the RBM upscale trips are in affect. In addition, the change incorporates the use of Allowable Values in the place of Analytical Limits.
Calculation NEC 98-024 Rev. 3, which documents the Analytical Limits and calculates the Allowable Values for the [RBM LPSP, IPSP, and HPSP] have not been altered. The calculation results implemented in procedures 6.1/2RBM.302 remain unchanged. The proposed TS change does not change or invalidate the Analytical Limits.
Based on the above, NPPD [Nebraska Public Power District] concludes that the proposed TS change to modify the mathematical symbols in TS SR 3.3.2.1.4 and TS Table 3.3.2.1-1 footnotes (a), (b), (c), and (e) does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change to the [CNS TS] corrects the mathematical symbols for the RBM LPSP, IPSP, and the HPSP to clarify the power ranges at which the RBM upscale trips are in affect. In addition, the change incorporates the use of Allowable Values in the place of Analytical Limits. The values for the RBM trip setpoints, Analytical Limits, and Allowable Values are not being altered in any way.
Based on the above, NPPD concludes that the proposed TS change to modify the mathematical symbols in TS SR 3.3.2.1.4 and TS Table 3.3.2.1-1 footnotes (a), (b), (c), and (e) does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Do the proposed changes involve a significant reduction in the margin of safety?
The proposed change to the [CNS TS] corrects the mathematical symbols for the RBM LPSP, IPSP, and the HPSP to clarify the power ranges at which the RBM upscale trips are in affect. In addition, the change incorporates the use of Allowable Values in the place of Analytical Limits. This TS change does not change any Analytical Limits or Allowable Value calculations. The methodology by which the RBM Trip Setpoints, Analytical Limits, and Allowable Values are derived has not changed.
Based on the above, NPPD concludes that the proposed TS change to modify the mathematical symbols in TS SR 3.3.2.1.4 and TS Table 3.3.2.1-1 footnotes (a), (b), (c), and (e) does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nine Mile Point Nuclear Station, LLC (NMPNS), Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
Date of amendment request: August 22, 2003.
Description of amendment request: The licensee proposed to revise Section 3.7.1, ``Service Water (SW) System and Ultimate Heat Sink (UHS),'' of the Technical Specifications (TS) to allow continued operation with short-term elevated UHS temperatures. The proposed revision is based on an NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specification change, identified as TSTF-330, ``Allowed Outage Time--Ultimate Heat Sink,'' Revision 3, dated October 16, 2000. Adoption of TSTF-330 would allow continued plant operation with UHS temperatures that temporarily exceed the 82 [deg]F limit.
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:
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Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change allows plant operation to continue if the temperature of the UHS exceeds the TS limit of 82 [deg]F provided that (1) the water temperature, averaged over the previous 24 hour period, is at or below 82 [deg]F, and (2) the UHS temperature is less than or equal to 84 [deg]F. This increase in UHS temperature will not affect the normal operation of the plant to the extent that it would make any accident more likely to occur. The UHS is not an accident initiator. In addition, the proposed change assures adequate margin in the safety systems
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and safety-related heat exchangers to meet the design safety functions at the higher temperature. Thus, the proposed change will have no adverse effect on plant operation, or the availability or operation of any accident mitigation equipment. Furthermore, the proposed change cannot cause an accident, nor will the change significantly affect the plant response to any accidents. Therefore, there will be no increase in the probability or consequences of an accident previously evaluated.
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Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not alter the current plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions. The proposed change will not alter the way any structure, system, or component functions and will not cause an adverse effect on plant operation or accident mitigation equipment. The response of the plant and the operators following a design-basis accident is unaffected by the change. The proposed change does not introduce any new failure modes and the design basis heat removal capability of the affected safety-related components is maintained at the increased UHS temperature limit. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously analyzed.
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Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
NMPNS has performed an evaluation of the safety systems to ensure their safety functions can be met with a UHS water temperature of 84 [deg]F. The higher UHS temperature represents a slight reduction in the margins of safety in terms of these systems' abilities to remove accident heat loads. As part of the evaluation, however, it was verified that these safety systems will still be capable of performing their design-basis functions. The proposed change will have no adverse effect on plant operation or equipment important to safety. The plant responses to accidents will not be significantly affected and the accident mitigation equipment will continue to function as assumed in the accident analysis. Therefore, there will be no significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
Date of amendment request: August 28, 2003.
Description of amendment request: The licensee proposed to change Section 3.1.7, ``Standby Liquid Control (SLC) System,'' of the Technical Specifications (TS) to raise the required average boron concentration in the reactor, resulting from injection of sodium pentaborate solution by the SLC system, to support a transition to the General Electric (GE) 14 fuel design. This design change includes the use of sodium pentaborate solution enriched with the boron-10 isotope. The proposed amendment would add a new surveillance requirement to verify the required boron-10 enrichment of the sodium pentaborate solution prior to addition to the SLC tank. It would also revise the figure that depicts acceptable values of SLC storage tank volume and sodium pentaborate solution concentration by adding a notation regarding the required boron-10 enrichment, and by making a minor adjustment to one of the coordinates that define the Acceptable Operation region on the figure.
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:
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Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The SLC system is designed to provide sufficient negative reactivity to bring the reactor from full power to a subcritical condition at any time in a fuel cycle, without taking credit for control rod movement. The proposed changes to the SLC sodium pentaborate solution requirements maintain the capability of the SLC system to perform this reactivity control function, and assure continued compliance with the requirements of 10 CFR 50.62 for anticipated transients without scram (ATWS). The SLC system is provided to mitigate ATWS events and, as such, is not considered to be an initiator of the ATWS event or any other analyzed accident. The use of sodium pentaborate solution enriched with the Boron-10 isotope, which is chemically and physically similar to the current solution, does not alter the design or operation of the SLC system or increase the likelihood of a system malfunction that could increase the consequences of an accident.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Injection of sodium pentaborate solution into the reactor vessel has been considered in the plant design. The proposed changes revise the SLC boron solution requirements such that the capability of the SLC system to bring the reactor to a subcritical condition without taking credit for control rod movement is maintained, considering operation with an equilibrium core of GE14 fuel. The use of sodium pentaborate solution enriched with the Boron-10 isotope, which is chemically and physically similar to the current solution, does not alter the design, function, or operation of the SLC system. The correct Boron-10 enrichment is assured by the proposed revisions to the TS surveillance requirements. The impact on the solubility limit of enriching the sodium pentaborate solution with the Boron-10 isotope is insignificant; thus, the existing minimum solution and piping temperature specified in the TS will ensure that the boron remains in solution and does not precipitate out in the SLC storage tank or in the SLC pump suction piping. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
-
Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed changes revise the SLC boron solution requirements to maintain the capability of the SLC system to bring the reactor to a subcritical condition without taking credit for control rod movement. These changes support operation with an equilibrium core of GE14 fuel and assure continued compliance with the requirements of 10 CFR 50.62. The minimum required average boron concentration in the reactor core, resulting from the injection of sodium pentaborate solution by the SLC system, has been determined using approved analytical methods. The analysis demonstrates that sufficient shutdown margin is maintained in the reactor such that the reactivity control function of the SLC system is assured. The additional quantity of boron included to allow for imperfect mixing and leakage is being increased from 20 percent to 25 percent. Thus, additional safety margin is provided to bring the reactor subcritical in the event of an ATWS. Therefore, the proposed change does not involve a significant reduction in a margin of safety?
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
Date of amendment request: September 3, 2003.
Brief description of amendments: The proposed change allows entry into a mode or other specified condition in the applicability of a Technical Specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee
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performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TS would be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF-359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated September 3, 2003.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated
The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant Reduction in [a] Margin of Safety
The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina.
Date of amendment request: July 23, 2003.
Description of amendment request: The proposed change will revise the near-end of life (EOL) Moderator Temperature Coefficient (MTC) Surveillance Requirement 4.1.1.3.b by placing a set of conditions on core operation, which if met, would allow exemption from the required MTC measurement. The conditional exemption will be determined on a cycle-specific basis by considering the margin predicted to the surveillance requirement MTC limit and the performance of other core parameters, such as beginning of life MTC measurements and the critical boron concentration as a function of cycle length. The conditional exemption will improve plant availability and minimize disruptions to normal plant operations. Plant safety criteria will not be compromised by the conditional exemption of this one measurement.
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:
-
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
The probability or consequences of accidents previously evaluated in the VCSNS FSAR [Final Safety Analysis Report] are unaffected by this proposed change because there is no change to any equipment response or accident mitigation scenario. There are no additional challenges to fission product barrier integrity. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
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Does this change involve a significant reduction in margin of safety?
The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no effect on the availability, operability, or performance of the safety- related systems and components. A change to the surveillance requirement is proposed, but the limiting conditions for operation required by TS [technical specifications] are not changed.
The TS Bases are founded in part on the ability of the regulatory criteria to be satisfied assuming the limiting conditions for operation are met for the various systems. Conformance to regulatory criteria for operation with the conditional exemption from the near-EOL MTC measurement is demonstrated and the regulatory limits are not exceeded. Therefore, the margin of safety as defined in the TS is not reduced and the proposed change does not involve a significant reduction in a margin of safety.
Pursuant to 10 CFR 50.91, the preceding analyses provide a determination that the
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proposed Technical Specifications change poses no significant hazard as delineated by 10 CFR 50.92.
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: Thomas G. Eppink, South Carolina Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: John A. Nakoski.
Southern California Edison Company, et al., Docket No. 50-361, San Onofre Nuclear Generating Station, Unit 2, San Diego County, California.
Date of amendment requests: August 26, 2003.
Description of amendment requests: The proposed change would revise Technical Specifications (TS) 1.1 ``Definitions,'' 3.4 ``Reactor Coolant System [RCS],'' and 5.7 ``Reporting Requirements. Specifically, the licensee requests to relocate the RCS pressure-temperature curves and limits from the TSs to a licensee-controlled document identified as the PTLR [Pressure and Temperature Limits Report].
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
-
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
Updating the Reactor Coolant System (RCS) pressure and temperature curves and limits in accordance with 10 CFR [Part] 50 Appendices G and H ensures the reactor coolant system's pressure boundary integrity will be protected until End Of Life (EOL) and does not contribute to the probability of or the initiation of accidents. There is no change to the safety analysis.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
-
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
These changes are required to maintain the RCS pressure boundary integrity until EOL. Changes to the RCS pressure and temperature curve and limits will not create a new or different kind of accident. There is no change to the safety analysis.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
-
Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Pressure and temperature curves and limits are provided as limits to plant operation for ensuring RCS pressure boundary integrity is maintained until EOL. No margin of safety is impacted by changes to the RCS pressure and temperature curves and limits. There is no change to the safety analysis.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, SCE concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding 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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.
NRC Section Chief: Stephen Dembek.
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.
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 amendment request: September 3, 2003.
Brief description of amendment request: The proposed amendment would revise Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.5.1.d to replace the phrase ``Each ice basket'' with the phrase ``Ice baskets.'' This change would make the LCO consistent with associated TS Surveillance Requirement (SR) 4.6.5.1.b.2 and would allow the SR to define the detailed requirements for ice basket weight.
Date of publication of individual notice in Federal Register: September 10, 2003 (68 FR 53402).
Expiration date of individual notice: October 10, 2003.
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, 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
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have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina.
Date of application of amendments: February 19, 2003.
Brief description of amendments: The amendments revised the Technical Specifications (TS) 5.5.10, ``Steam Generator (SG) Tube Surveillance Program.'' Specifically, the proposed changes would revise the SG surveillance requirements in the Oconee Units 1, 2, and 3 TSs. Since steam generator replacement outages are respectively scheduled for Fall 2003, Spring 2004, and Fall 2004, the licensee proposes to relocate the program requirements applicable to the original SGs, existing TS 5.5.10 requirements, to TS 5.5.21 and to provide program requirements applicable to the replacement SGs, in TS 5.5.10.
Date of Issuance: September 4, 2003.
Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.
Amendment Nos.: 334, 334, & 335.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR 12949).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 4, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas.
Date of application for amendment: September 19, 2002, as supplemented by letter dated July 18, 2003.
Brief description of amendment: The proposed amendment extends the allowed outage time (AOT) for a single inoperable low pressure safety injection (LPSI) train from 72 hours to 7 days. In addition, an AOT of 72 hours is included for other conditions where the equivalent of a single emergency core cooling system (ECCS) subsystem flow is still available to both the LPSI and high pressure safety injection (HPSI) trains. Also, an action statement is added to restore at least one of each HPSI and LPSI train to operable status within one hour if 100% of ECCS flow is unavailable due to two inoperable HPSI or LPSI trains.
Date of issuance: September 11, 2003.
Effective date: As of the date of issuance to be implemented within 60 days from the date of issuance.
Amendment No.: 251.
Facility Operating License No. NPF-6: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002.
The July 18, 2003, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 11, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois.
Date of application for amendments: December 20, 2002, as supplemented May 30, 2003.
Brief description of amendments: The amendments revise the licensing basis as described in the Updated Final Safety Analysis Report to implement the Boiling-Water Reactor Vessel and Internals Project reactor pressure vessel integrated surveillance program as the basis for demonstrating compliance with the requirements of Appendix H to 10 CFR part 50.
Date of issuance: August 28, 2003.
Effective date: As of the date of issuance and shall be implemented within 60 days.
Amendment Nos.: 217/211.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments revised the licensing basis.
Date of initial notice in Federal Register: February 4, 2003 (68 FR 5669).
The supplement dated May 30, 2003, 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 August 28, 2003.
No significant hazards consideration comments received: No.
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: June 5, 2002, as supplemented August 19 and December 2, 2002, and January 30, February 14, March 19 and 31, June 6 and 24, and September 5, 2003.
Brief description of amendments: The amendments approved selective implementation of an alternative source term methodology for the loss- of-coolant accident (LOCA) and the control rod ejection accident (CREA), incorporation of ARCON96 methodology for release points associated with the LOCA and CREA, elimination of the control room emergency bottled air pressurization system, changes to the control room emergency ventilation system (CREVS), and a change to the BVPS-1 CREVS filter bypass leakage acceptance test criteria.
Date of issuance: September 10, 2003.
Effective date: As of the date of issuance and shall be implemented within 60 days.
Amendment Nos.: 257 and 139.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67 FR 75876). The supplements dated August 19 and December 2, 2002, and January 30, February 14, March 19 and 31, June 6 and 24, and September 5, 2003, provided additional information that clarified the application, did not expand the scope of the application as originally noticed except as noted below, and did not change the staff's original proposed no significant hazards consideration determination. The February 14, 2003, submittal requested the scope of the review be expanded by including in the scope of the review related Updated Final Safety Analysis Report (UFSAR) page changes, but this request was withdrawn in the March 31, 2003, submittal. Additionally, a portion of the requested review was withdrawn in the March 19, 2003, submittal, as these changes were no longer necessary. The portion of the proposed application related to conversion of the BVPS-1 and 2 containments from subatmospheric to atmospheric operating conditions was withdrawn by letter dated September 5, 2003.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 10, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis- Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio.
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Date of application for amendment: May 14, 2003, as supplemented by letters dated June 16, August 2, August 7, and August 20, 2003.
Brief description of amendment: This amendment revised the Technical Specifications to allow a one time exception, only during the Restart Test Plan, to allow entry into Mode 3 of operation without the high-pressure injection pumps being able of taking suction from the low-pressure injection trains when aligned for containment sump recirculation. The exception cannot be used for entry into Mode 2 or Mode 1.
Date of issuance: September 5, 2003.
Effective date: As of the date of issuance and shall be implemented within 90 days.
Amendment No.: 257.
Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR 34668).
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 September 5, 2003.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County, Florida.
Date of application for amendment: February 17, 2003.
Brief description of amendment: The amendment revises Technical Specification (ITS) 3.6.3 ``Containment Isolation Valves,'' to allow verification by administrative means of isolation devices in high radiation areas, and isolation devices that are locked, sealed or otherwise secured.
Date of issuance: September 8, 2003.
Effective date: As of the date of issuance and shall be implemented within 60 days of issuance.
Amendment No.: 209.
Facility Operating License No. DPR-72: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR 18277).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 8, 2003.
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: October 11, 2002, as supplemented by letter dated May 29, 2003.
Description of amendment request: The amendment revises Technical Specification (TS) 3.4.9.1, ``Reactor Coolant System [RCS]--Pressure/ Temperature Limits,'' and TS 3.4.9.3, ``Reactor Coolant System-- Overpressure Protection Systems'' and their associated Bases sections. Specifically, the changes replace TS Figures 3.4-2 ``Reactor Coolant System Heatup Limitations,'' 3.4-3 ``Reactor Coolant System Cooldown Limitations,'' and 3.4-4 ``RCS Cold Overpressure Protection'' to allow operation to 20 Effective Full Power Years.
Date of issuance: September 11, 2003.
Effective date: As of its date of issuance, and shall be implemented within 60 days.
Amendment No.: 89.
Facility Operating License No. NPF-86: Amendment revises the TS.
Date of initial notice in Federal Register: December 10, 2002 (67 FR 75879). The May 29, 2003, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination nor expand the amendment beyond the scope of the initial notice.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 11, 2003.
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: April 24, 2002.
Description of amendment request: The amendment revises surveillance requirements (SRs) in Technical Specification (TS) 4.6.2.1, ``Containment Spray System,'' and TS 4.7.1.2.1b, ``Auxiliary Feedwater System,'' and associated Bases Section 3/4.7.1.2. Specifically, the proposed changes would move SR acceptance criteria for containment spray and auxiliary feedwater pumps from the TSs to the Seabrook Station Technical Requirements Manual.
Date of issuance: September 12, 2003.
Effective date: As of its date of issuance, and shall be implemented within 60 days.
Amendment No.: 90.
Facility Operating License No. NPF-86: The amendment revises the TSs.
Date of initial notice in Federal Register: June 11, 2002 (67 FR 40024).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 12, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, Wisconsin
Date of application for amendments: September 12, 2002, as supplemented March 27 and May 30, 2003.
Brief description of amendments: The amendments add surveillance requirements for Technical Specification (TS) 3.5.2, ``ECCS-- Operating,'' and TS 3.5.3, ``ECCS--Shutdown,'' to verify, every 31 days, that the emergency core cooling system piping is full of water.
Date of issuance: September 5, 2003.
Effective date: As of the date of issuance and shall be implemented within 45 days.
Amendment Nos.: 209 and 214.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments revised the TSs.
Date of initial notice in Federal Register: February 4, 2003 (68 FR 5679).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 5, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, Alabama.
Date of application for amendments: February 13, 2003, as supplemented April 14, 2003.
Description of amendment request: The amendments revised Technical Specification (TS) 4.2.1 ``Fuel assemblies,'' to modify the fuel design description to encompass Framatome Advanced Nuclear Power fuel assemblies, and also to modify TS 4.3 ``Fuel Storage,'' to remove nomenclature specific to Global Nuclear Fuels analysis methods.
Date of issuance: September 5, 2003.
Effective date: September 5, 2003.
Amendment Nos.: 247, 284, 242.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 1, 2003 (68 FR 15763). The April 14, 2003, letter provided clarifying information that did not
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change the scope of the original request or the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 5, 2003.
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: May 1, 2003, as supplemented on July 8, 2003.
Brief description of amendment: The amendment revised Technical Specification (TS) 3.8.7, ``Inverters--Operating.'' The revised TS requires only one inverter for each of the four 120V AC Vital Instrument channels. This amendment is the initial phase of a project that will update the 120V AC Vital Instrument Power System.
Date of issuance: September 8, 2003.
Effective date: As of the date of issuance and shall be implemented prior to Mode 4 entry following the next refueling outage in the fall of 2003.
Amendment No.: 45.
Facility Operating License No. NPF-90: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR 28859). The supplemental letter provided clarifying information that did not expand the scope of the original request 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 September 8, 2003.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri.
Date of application for amendment: June 2, 2003.
Brief description of amendment: The amendment revises the technical specifications (TSs) to increase the specified minimum fuel oil inventories maintained in the fuel oil storage tanks for the diesel generators.
Date of issuance: September 9, 2003.
Effective date: September 9, 2003, and shall be implemented within 60 days from the date of issuance.
Amendment No.: 156.
Facility Operating License No. NPF-30: The amendment revised the Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR 43393).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 9, 2003.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry County, Virginia.
Date of application for amendments: September 5, 2002, as supplemented on April 16, June 9, and July 7, 2003.
Brief Description of amendments: These amendments revise the Technical Specifications to add provisions to permit inspection and related repair of a buried fuel oil storage tank during plant operation by extending the allowed outage time for a buried fuel oil storage tank to 7 days from 24 hours for this purpose.
Date of issuance: September 10, 2003.
Effective date: As of the date of issuance, and shall be implemented within 30 days.
Amendment Nos.: 236 and 235.
Renewed Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the Technical Specifications.
Date of initial notice in Federal Register: August 5, 2003 (68 FR 46247).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated September 10, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 22nd day of September 2003.
For the Nuclear Regulatory Commission. Ledyard B. Marsh, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 03-24477 Filed 9-29-03; 8:45 am]
BILLING CODE 7590-01-P
