Source: https://www.federalregister.gov/documents/2002/10/29/02-27243/biweekly-notice-applications-and-amendments-to-facility-operating-licenses-involving-no-significant
Timestamp: 2018-03-20 10:21:30
Document Index: 609887405

Matched Legal Cases: ['art 2', 'art] 100', 'art] 100', 'art] 50', 'art] 50', 'art 50', 'art 72', 'art 73', 'art 2']

October 10, 2002, and shall be implemented within 30 days of issuance.
66005-66019 (15 pages)
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. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear (SQN) Plant, Unit 1, Hamilton County, Tennessee
https://www.federalregister.gov/d/02-27243 https://www.federalregister.gov/d/02-27243
This biweekly notice includes all notices of amendments issued, or proposed to be issued from, October 4, 2002, through October 17, 2002. The last biweekly notice was published on October 15, 2002 (67 FR 63687).
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 Rgister 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 Start Printed Page 66006Pike, 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, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.
By November 29, 2002, 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,[1] which is available at the Commission's PDR, located at One White Flint North, 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.
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 Start Printed Page 66007granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Date of amendments request: September 6, 2002.
Description of amendments request: The amendments would replace the peak linear heat safety limit, in Technical Specification (TS) 2.1.1.2, “Reactor Core SLs [Safety Limits],” by a peak fuel centerline temperature safety limit to have a safety limit in the TSs that would not be exceeded during normal operation or anticipated operational occurrences (AOOs), in accordance with Section 50.36(c)(1)(ii)(A) of Title 10 of the Code of Federal Regulations (10 CFR).
The proposed change does not require any physical change to plant systems, structures, or components nor does it require any change in systems or plant operations. The proposed change does not result in any change to safety analysis methods or results. The change to establish peak fuel centerline temperature as the Safety Limit is consistent with the PVNGS Units 1, 2 and 3 licensing bases for ensuring that the fuel design limits are met. Operations and analysis will continue to be in accordance with the PVNGS Units 1, 2 and 3 licensing bases. The peak fuel centerline temperature is the basis for protecting the fuel and is consistent with the safety analysis. [The peak linear heat rate and peak fuel centerline temperature safety limits are not initiators of accidents.]
The PVNGS Units 1, 2 and 3 Updated Final Safety Analysis Report (UFSAR) Chapter 15 accident analyses for AOOs where the peak linear heat rate may exceed the existing Safety Limit of 21 kW/ft are the control element assembly (CEA) Withdrawal events at Subcritical and Low Power conditions. The analyses for these AOOs indicate that the peak fuel centerline temperature is not exceeded. The existing safety analyses, which remain unchanged, do not affect any accident initiators that would create a new accident. [The peak linear heat rate and peak fuel centerline temperature safety limits are not initiators of accidents.]
The proposed change does not result in any change to safety analysis methods or results. Therefore, by changing the Safety Limit from peak linear heat rate to peak fuel centerline temperature the margins as established in the PVNGS Units 1, 2 and 3 Technical Specifications and UFSAR are unchanged.
Based on the above, APS [Arizona Public Service Company] concludes that the activities associated with the proposed amendment[s] 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.
Date of amendment request: September 20, 2002.
Description of amendment request: The proposed amendment revises Technical Specification (TS) 3.9.5, Shutdown Cooling (SDC) and Coolant Circulation—Low Water Level, for Unit Nos. 1 and 2 to add two notes to allow operational changes in the Shutdown Cooling System to support operations and testing. The changes would allow the SDC pumps to be deenergized for less than or equal to 15 minutes when switching from one train to another. The second change would allow one SDC loop to be inoperable for up to 2 hours for surveillance testing, provided that the other loop was operable and in operation.
The system affected by this proposed amendment is the Shutdown Cooling (SDC) System. This system mitigates the consequences of a boron dilution event and removes decay heat from the Reactor Coolant System when the unit is in Mode 6. This proposed amendment revises the Technical Specification to allow the SDC pumps to be deenergized for less than or equal to 15 minutes to allow swapping from one operating train to another, and would allow one SDC loop to be inoperable for up to two hours for surveillance testing. Because this system is used for the mitigation of an accident, it is not an accident initiator. Therefore, the probability of an accident previously evaluated is not increased.
The only design basis accident considered in this Mode is a boron dilution event. Consideration is also given to a loss of decay heat removal in this Mode as well. Both of these conditions are evaluated in the Updated Final Safety Analysis Report (UFSAR). The evaluations consider operation of the SDC system to mitigate these conditions. Removing this system from service for a limited amount of time, with other operational restrictions, limits the consequences to those already assumed in the UFSAR. Thus, no increase in offsite dose occurs under this conditions. Therefore, the consequences of an accident previously evaluated have not increased.
The proposed changes do not involve a significant change in the operation of the plant and no new accident initiation mechanism is created by the proposed changes. The SDC System is not being altered by this amendment request. No substantial changes are made in the way in which the SDC System is operated. The only change made would allow both SDC pumps to be Start Printed Page 66008deenergized to swap operating trains, and one SDC inoperable for less than two hours to allow for surveillance testing. Since the SDC System is an accident mitigating system only, changes in when this system is needed to operate cannot create a new [kind] of accident.
The margin of safety provided by the SDC System is to provide boration control and to remove decay and sensible heat from the Reactor Coolant System as described in the UFSAR. Removal of system components from service as described above, and with limitations in place to prevent boron dilution and loss of decay and sensible heat removal, does not significantly impact the margin of safety. The SDC System will continue to be able to provide its safety function under this conditions. Operators will continue to have adequate time to respond to any off-normal events. Removing the system from service, for a limited period of time, with other operational restrictions limits the consequences to those already assumed in the UFSAR. Therefore, no reduction in [a] margin of safety has occurred because the event results in the UFSAR are not changed by operation in the proposed conditions.
Date of amendment request: August 26, 2002.
Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) for diesel fuel oil for the plant's onsite diesel-generator power sources. The proposed changes would allow the use of an optional water and sediment content test, would relocate the specific version of certain American Society for Testing and Materials (ASTM) references to licensee controlled documents, would add several new ASTM references, and would relocate the requirement for a 10-year diesel fuel oil tank inspection and cleaning to licensee controlled documents. The licensee stated that the changes are consistent with the Standard Technical Specification Travelers (TSTF) 374, Revision 0 and TSTF 2, Revision 1. Associated changes are also proposed for the TS Bases.
The following discussion is a summary of the evaluation of the change contained in this proposed amendment against the 10 CFR 50.92 (c) requirements to demonstrate that all three standards are satisfied. A no significant hazards consideration is indicated if operation of the facility in accordance with the proposed amendments would not:
The proposed changes relocate the specific American Society for Testing and Materials (ASTM) Standard references from the Administrative Controls Section of Technical Specifications (TS) to a licensee-controlled document. Since any changes of the licensee-controlled document will be evaluated to the requirements of 10 CFR 50.59, “Changes, tests, and experiments,” no increase in the probability or consequences of an accident previously evaluated is involved. In addition, the “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to the storage tanks has expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The Bases for SR 3.8.3.3 (CNS) and 3.8.3.2 (MNS) are revised to indicate that the API gravity is tested in accordance with ASTM D1298 or D287.
Relocating the specific ASTM Standard references from the TS to a licensee-controlled document, allowing a water and sediment test to be performed to establish the acceptability of new fuel oil, and revising the TS Bases will not affect or degrade the ability of the emergency diesel generators (DGs) to perform their specified safety function. Fuel oil quality will continue to meet ASTM requirements.
In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire and 3.8.3.6 for Catawba are revised to remove the requirement for a 10-year tank inspection and cleaning. This requirement will be moved to a licensee-controlled document. Any changes of the licensee-controlled document will be evaluated to the requirements of 10CFR 50.59 “Changes, tests, and experiments,”.
This change will not affect or degrade the ability of the emergency diesel generators (DGs) to perform their specified safety function. Fuel oil quality will continue to meet ASTM requirements.
The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained.
The proposed changes do not alter or prevent the ability of structures, systems, or components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated.
Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.
The proposed changes relocate the specific ASTM Standard references from the Administrative Controls Section of TS to a licensee-controlled document. In addition, the “clear and bright” test used to establish the acceptability of new fuel oil for use prior to addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The proposed changes revise Bases B 3.8.3 to reference the current specific ASTM standards. The Bases for SRs 3.8.3.3 (CNS) and 3.8.3.2 (MNS) are revised to indicate that the API gravity is tested in accordance with ASTM D1298 or D287.
In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire and 3.8.3.6 for Catawba are revised to remove the requirement for a 10-year tank inspection and cleaning. This requirement will be moved to a licensee-controlled document. Any changes of the licensee-controlled document will be evaluated to the requirements of 10CFR50.59 “Changes, tests, and experiments,”.
The changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis or licensing basis. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes relocate the specific ASTM Standard references from the Administrative Control Section of TS to a licensee-controlled document. Instituting the proposed changes will continue to ensure the use of the current applicable ASTM Start Printed Page 66009Standards to evaluate the quality of both new and stored fuel oil designated for use in the emergency diesels. The detail associated with the specific ASTM Standard references is not required to be in the TS to provide adequate protection of the public health and safety, since the TS still retains the requirement for compliance with the applicable ASTM standard. Changes to the licensee-controlled document are performed in accordance with the provisions of 10 CFR 50.59. Should it be determined that future changes involve a potential reduction in a margin of safety, NRC review and approval would be necessary prior to the implementation of the changes. This approach provides an effective level of regulatory control and provides for a more appropriate change control process.
The “clear and bright” test used to establish the acceptability of new fuel oil for use prior to the addition to storage tanks has been expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil. The proposed changes revise Bases B 3.8.3 to allow reference to the current ASTM standard. The Bases for SR 3.8.3.3 is revised to indicate that the API gravity is tested in accordance with ASTM D1298 or D287. The level of safety of facility operation is unaffected by the proposed changes since there is no change in the intent of the TS requirements of assuring fuel oil is of the appropriate quality for emergency DG use.
In addition Surveillance Requirements (SR) 3.8.3.5 for McGuire and 3.8.3.6 for Catawba are revised to remove the requirement for a 10-year tank inspection and cleaning. This requirement will be moved to a licensee-controlled document. Any changes of the licensee-controlled document will be evaluated to the requirements of 10CFR50.59 “Changes, tests, and experiments”. The level of safety of the facility operation is unaffected by the proposed changes since there is no change in the intent of the SR to clean and inspect the fuel tanks.
Therefore, the proposed changes listed above do not involve a significant reduction in a margin of safety.
Date of amendment request: October 15, 2001, as supplemented by letter dated August 27, 2002.
Description of amendment request: The proposed amendment request provides additional information to support a modification to Technical Specification 3.4.7 and limits Reactor Coolant System activity permitted by the ACTION statement to 60 microcuries per gram (μCi/gm) at all power levels. The letdown line break accident analysis in the Final Safety Analysis Report is also changed to reflect revised dose consequences. This notice supercedes the biweekly Federal Register notice dated November 28, 2001 (66 FR 59504), based on the original application dated October 15, 2001.
Response: The proposed change to the Technical Specifications (TS) conservatively limits Reactor Coolant System (RCS) activity permitted by Action Statement 3.4.7.a to 60 μCi/gm at all reactor power levels. The proposed change to the Final Safety Analysis Report (FSAR) Section 15.6.3.1 revises the letdown line break accident analyses.
The probability of a previously evaluated accident is not affected by this change because the pre-existing iodine spike is not an accident initiator and the new letdown line break accident analysis does not affect any plant Structure, Systems, or Component (SSC) but merely determines the consequences of the previously evaluated accident.
This TS change is conservative in that it will reduce the accident consequences for events occurring at lower power levels. The new letdown line break accident analysis meets the original Safety Evaluation Report (SER) and the current Standard Review Plan (SRP) acceptance criteria of a small fraction of the 10 CFR [Part] 100 limits.
Response: The probability of a new or different accident is not affected by this change because the new letdown line break analysis does not affect any plant Structure, Systems, or Component but merely determines the consequences of the previously evaluated accident.
Response: The TS change is more limiting in that it will reduce the accident consequences for events occurring at lower power levels.
The new letdown line break accident analysis, assuming one operating charging pump, meets the original SER and current SRP acceptance criteria of a small fraction of the 10 CFR [Part] 100 limits. This single pump analysis provides a suitable licensing basis analysis and has sufficient conservatism to accommodate two and three pump operating scenarios that may exist during the operating cycle.
Date of application for amendment request: September 27, 2002.
Description of amendment request: The proposed amendment changes Appendix B, “Environmental Protection Plan (Non-Radiological),” of the license by removing a parenthetical reference to a superseded section of 10 CFR 51.
The proposed change deletes a reference to a superseded section of 10 CFR 51, “Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions,” found in the non-radiological Environmental Protection Plans (EPPs) for Byron Station, LaSalle County Station and Quad Cities Nuclear Power Station, Units 1 and 2. The EPP (Non-Radiological) is Appendix B to the Facility Operating License. The change is administrative in nature. No physical changes to the facilities will result from the proposed change. The initial conditions and methodologies used in accident analyses remain unchanged. The Start Printed Page 66010proposed change does not revise or alter the design assumptions for systems or components used to mitigate the consequences of accidents. Thus, accident analyses results are not impacted by this proposed change.
The proposed change deletes a reference to a superseded section of 10 CFR 51.5. The change is administrative in nature. No physical or operational changes to the facilities will result from the proposed change.
The proposed change does not affect the design or operation of any system, structure, or component (SSC) in the plant. The safety functions of the related SSCs are not changed in any manner, nor is the reliability of any SSC reduced. The change does not affect the manner by which the facility is operated and does not change any facility, structure, system, or component. No new or different type of equipment will be installed by this proposed change.
The proposed change is administrative in nature and has no impact on the margin of safety of any Technical Specification. There is no impact on safety limits or limiting safety system settings. The change does not affect any plant safety parameters or setpoints. The proposed change deletes an inaccurate reference to a section of 10 CFR 51 that has been superseded. No physical or operational changes to the facility will result from the proposed changes.
Date of amendment request: January 16, 2002.
Description of amendment request: The amendment would make administrative, editorial, and format (including repagination) changes to the technical specification (TS) Bases index and the Administrative Control section of TSs. Specifically, the amendments would relocate the TS Bases page listings from the TS index to a TS Bases index, and remove certain duplicative administrative requirements from Section 6, “Administrative Controls,” of the TSs.
No. The proposed administrative changes to the TS index and to Section 6 of the TSs do not result in changes being made to structures, systems, or components (SSCs), or to event initiators or precursors. Also, the proposed changes do not impact the design of plant systems such that previously analyzed SSCs would now be more likely to fail. The initiating conditions and assumptions for accidents described in the Updated Final Safety Analysis Report (UFSAR) remain as previously analyzed. Thus, the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.
The previously analyzed SSCs are unaffected by the proposed changes and continue to provide assurance that they are capable of performing their intended design function in mitigating the effects of design basis accidents (DBAs). As such, the consequences of accidents previously evaluated in the UFSAR will not be increased and no additional radiological source terms are generated. Therefore, there will be no reduction in the capability of those SSCs in limiting the radiological consequences of previously evaluated accidents and reasonable assurance that there is no undue risk to the health and safety of the public will continue to be provided. Thus, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.
Therefore, the proposed administrative changes do not significantly increase the probability or consequences of any accident previously evaluated.
No. The proposed administrative changes do not involve physical changes to analyzed SSCs or changes to the modes of plant operation defined in the technical specification. The proposed changes do not involve the addition or modification of plant equipment (no new or different type of equipment will be installed) nor do they alter the design or adversely affect operation of any plant systems. No new accident scenarios, accident or transient initiators or precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes.
The proposed administrative changes do not cause the malfunction of safety-related equipment assumed to be operable in accident analyses. No new or different mode of failure has been created and no new or different equipment performance requirements are imposed for accident mitigation. As such, the proposed changes have no effect on previously evaluated accidents.
Therefore, the proposed administrative changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
No. The proposed administrative changes do not affect any previously evaluated accident. The proposed changes do not adversely affect the TS requirements and will continue to ensure that the necessary plant equipment is operable in the plant conditions where these systems are required to operate to mitigate a DBA as described in the analyses presented in the UFSAR. Thus, the proposed administrative, editorial, and format changes do not affect plant safety.
Therefore, the proposed administrative changes do not involve a significant reduction in a margin of safety.
Description of amendment request: The proposed amendment would revise the Unit 2 reactor coolant system (RCS) pressure-temperature curves in Technical Specification (TS) Figures 3.4-2 and 3.4-3 and associated TS Bases. The revised curves will bound operation of the unit for the remainder of its current license duration and bound operation with planned license amendments to increase the power level at which the unit is allowed to operate.
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 Start Printed Page 66011consideration, which is presented below:
The proposed change will revise the RCS pressure-temperature curves to bound operation of the reactor for up to 32 EFPY at a power level of up to 3800 MW for the current fuel cycle and beyond, to reflect new fluence analysis methodology, to reflect the use of ASME [American Society of Mechanical Engineers] Code Case N-641, to include boltup limits, and to no longer include instrument uncertainty margins.
The proposed change will not result in physical changes to structures, systems, or components (SSCs), or to event initiators or precursors. The proposed change will not affect the ability of personnel to control RCS [Reactor Coolant System] pressure at low temperatures and, thereby, ensure the integrity of the RCPB [Reactor Coolant Pressure Boundary]. Use of ASME Code Case N-641 will be approved by the NRC through approval of a Donald C. Cook Nuclear Plant-specific exemption to requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G. Therefore, the proposed revision to the RCS pressure-temperature curve changes will have been determined in accordance with NRC accepted methodologies. These methodologies provide adequate assurance that the reactor vessel will withstand the effects of normal cyclic loads due to temperature and pressure changes, and provide an acceptable level of protection against brittle failure. Additionally, the proposed changes will not impact the design or operation of plant systems such that previously analyzed SSCs will be more likely to fail. The initiating conditions and assumptions for accidents described in the UFSAR will remain as previously analyzed. Therefore, the proposed changes will not involve a significant increase in the probability of an accident previously evaluated.
The proposed RCS pressure-temperature curves will continue to provide adequate margins of protection for the RCPB. The proposed changes have been determined, through supporting analyses, to be in accordance with the methodologies and criteria set forth in the applicable regulations, or in accordance with technically adequate alternatives. Compliance with these methodologies provides adequate margins of safety and ensures that the RCPB will withstand the effects of normal cyclic loads due to temperature and pressure changes as well as the loads associated with postulated faulted events as described in the UFSAR. The format changes will improve the appearance of the affected pages but will not affect any requirements. Therefore, the proposed change will not significantly reduce the margin of safety.
Date of amendment request: September 30, 2002.
Description of amendment request: The proposed amendment requests permission to change Kewaunee Nuclear Power Plant (KNPP) Facility Operating License DRP-43 to use an upgraded computer code for design basis accident containment integrity analyses. KNPP is currently licensed to use code for Generation of Thermal-Hydraulic Information for Containment (GOTHIC) version 6.0a. The proposed amendment requests to use GOTHIC 7.0p2 (GOTHIC 7).
Accident analyses affected by GOTHIC have each been evaluated and found to show good agreement between the GOTHIC 7 analysis and the current analysis of record (AOR). Safety analysis results using GOTHIC 7 are shown to satisfy all applicable design and safety analysis acceptance criteria. Since GOTHIC 7 conforms to design bases and its results are bounded by the existing safety analyses, its use within limits of the bounding accident analyses will not cause an increase in the probability or consequences of an accident previously evaluated. Adherence to safety analysis acceptance criteria prevents use of GOTHIC 7 from creating new challenges to components and systems that could adversely affect their ability to mitigate accident consequence or diminish integrity of any fission product barrier.
Thus, the requested upgrade to GOTHIC 7 with [mist diffusion layer] MDL modeling option will not increase probability or the consequences of an accident previously evaluated.
Upgrade to GOTHIC 7 is a change in analysis methods applied to Kewaunee [design basis accident] DBA. Analysis methods are not accident initiators. GOTHIC 7 will be applied in the same manner currently licensed and it is consistent with current plant design bases and licensed accident analysis methodologies. It does not adversely affect any fission product barrier, nor does it alter the safety function of safety related systems, structures, and components depended upon for accident prevention or mitigation. Equipment important to safety will continue to function within design. As demonstrated by the [Numerical Start Printed Page 66012Applications Inc.] NAI report, GOTHIC 7 yields a representation of expected plant response for affected design basis accidents that is more accurate but remains conservative. GOTHIC 7 predicted results for affected DBA remain bounded by the limiting analyses of record.
Thus, the requested upgrade to GOTHIC 7 does not create the possibility of a new or different kind of accident from any previously evaluated.
Upgrade to GOTHIC 7 affects Kewaunee design basis [loss of coolant accident] LOCA and [main steamline break] MSLB DBA containment analyses. The results predicted by GOTHIC 7 for these DBA analyses remain within limiting design basis accidents of record. GOTHIC 7 accuracy and conservatism in this application has been verified through benchmark analyses against the current analyses of record, validated against recognized standard data, and found to be appropriate for application to Kewaunee DBA. Safety analysis acceptance criteria are satisfied and adherence to safety analysis acceptance criteria using GOTHIC 7 assures that Technical Specification limits will not be exceeded during normal operation.
Thus, upgrade to GOTHIC 7 does not involve a significant reduction in the margin of safety.
Description of amendment request: The proposed amendment would delete Surveillance Requirement (SR) 4.6.B.2, “Primary System Boundary—Reactor Vessel Temperature and Pressure,” from the Monticello Technical Specifications (TSs) on the basis of the licensee's commitment to (1) relocate the current requirements to the Updated Safety Analysis Report (USAR) and (2) implement the Boiling Water Reactor Vessel and Internals Program Integrated Surveillance Program as approved by the Nuclear Regulatory Commission (NRC) in a letter dated February 1, 2002. SR 4.6.B.2 currently states: “Test specimens representing the reactor vessel, base weld, and weld heat affected zone metal shall be installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The material sample program shall conform to ASTM [American Society for Testing and Materials] E 185-66.” The licensee would also make related changes to the TS Bases 3.6/4.6.
The proposed change relocates the requirement of the TS Surveillance Requirement to a Licensee controlled document and implements an integrated surveillance program that has been evaluated by the NRC staff as meeting the requirements of paragraph III.C of Appendix H to 10 CFR [Part] 50. The proposed change of relocating a TS Surveillance Requirement to the Monticello USAR and implementing an integrated surveillance program is not considered a precursor or initiator of an accident previously evaluated. The proposed change does not impact current plant operations or the design function of any structure, system or component. Consequently, the proposed change does not significantly increase the probability of any accident previously evaluated.
The proposed change provides the same assurance of Reactor Pressure Vessel integrity as has always been assured. The relocation of the TS Surveillance Requirement provides an acceptable method for implementing the integrated surveillance program which was evaluated by the NRC staff as meeting the requirements of 10 CFR [Part] 50, Appendix H, paragraph III.C. The relocation of the TS Surveillance or the implementation of an integrated surveillance program is not an input or consideration in any accident previously evaluated, thus the proposed change will not increase the probability of any such accident occurring. The proposed amendment does not involve any change to the configuration or method of operation of any plant equipment that is used to mitigate the consequences of an accident, nor does it affect any assumptions or conditions in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.
Therefore, operation of the facility in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed amendment does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. No equipment interfaces are modified and no changes to any equipment function or the method of operating the equipment are being made. The proposed change, to relocate the TS Surveillance and implement an integrated surveillance program, maintains an equivalent level of RPV [reactor pressure vessel] material surveillance and does not introduce any new accident initiators. The proposed change will not change the design, configuration or operation of the plant.
Therefore, operation of the facility in accordance with the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed amendment has been evaluated as providing an acceptable alternative to the plant-specific RPV material surveillance program that meets the requirements of the regulations for RPV material surveillance. The proposed change does not exceed or alter a design basis or safety limit. The change relocates a TS Surveillance Requirement and implements an integrated surveillance program and as such does not significantly reduce the margin of safety.
Therefore, operation of the facility in accordance with the proposed change does not involve a significant reduction in a margin of safety.
Date of amendment request: September 23, 2002.
Description of amendment request: The proposed amendments would change the SSES 1 and 2 Technical Specifications (TSs) by revising limiting condition for operation (LCO) 3.6.2.3 to add a new Condition B, which permits both residual heat removal (RHR) suppression pool cooling subsystems to be inoperable for 8 hours, rather than immediately initiating a unit shutdown. By making this change, the licensee is incorporating Technical Specifications Task Force change traveler number 230 into its TSs.
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 Start Printed Page 66013consideration, which is presented below:
The proposed change relaxes the Required Actions of [LCO 3.6.2.3] by allowing 8 hours to restore one RHR suppression pool cooling subsystem to OPERABLE status when both subsystems have been determined to be inoperable. Required Actions and their associated Completion Times are not initiating conditions for any accident previously evaluated. The proposed 8 hour Completion Time provides some time to restore required subsystem(s) to OPERABLE status, yet is short enough that operating an additional 8 hours is not a significant risk. Consequently, this change in Required Actions does not significantly increase the probability of occurrence of any accident previously evaluated. The Required Actions in the proposed change have been developed to provide assurance that appropriate remedial actions are taken in response to the degraded condition, considering the operability status of the RHR Suppression Pool Cooling System and the capability of minimizing the risk associated with continued operation. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve a physical modification or alteration of plant equipment (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The Required Actions and associated Completion Times in the proposed change have been evaluated to ensure that no new accident initiators are introduced. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The relaxed Required Actions do not involve a significant reduction in a margin of safety. The proposed change has been evaluated to minimize the risk of continued operation with both RHR suppression pool cooling subsystems inoperable. The operability status of the RHR Suppression Pool Cooling System, a reasonable time for repair or replacement of required features, and the low probability of a design basis accident occurring during the repair period have been considered in the evaluation. Therefore, this change does not involve a significant reduction in a margin of safety.
3. Does not involve a significant reduction in [a] margin of safety?
Accordingly, based on the above, the proposed change does not involve a significant reduction in [a] margin of safety.
Description of amendment request: The proposed amendments would remove license condition 2.C.3.f from the Unit 1 operating license and license condition 2.C.4 from the Unit 2 operating license, and replace them with a commitment in Section 9.1.4.2.2.5 of the Updated Final Safety Analysis Report (UFSAR). Specifically, license conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively, require NRC approval of the lifting devices which attach the spent fuel cask to the crane prior to use of the spent fuel cask crane for the purpose of moving Start Printed Page 66014spent fuel casks. Subsequent to issuance of FOLs NPF-2 and NPF-8, the NRC issued NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants,” which endorsed the use of ANSI N14.6 for the design and inspection of special lift devises thereby eliminating the need for license conditions 2.C.3.f and 2.C.4. Accordingly, SNC proposes that license conditions 2.C.3.f and 2.C.4 be removed from FOLs NPF-2 and NPF-8, respectively, and replaced with a commitment in the FNP UFSAR to ANSI N14.6 for the design, fabrication, testing, and quality assurance requirements associated with the spent fuel cask lift device.
The proposed change replaces license conditions 2.C.3.f and2.C.4 to FOLs NPF-2 and NPF-8, respectively, with a commitment in the FNP Updated Final Safety Analysis Report (UFSAR) to the requirements of ANSI N14.6, as clarified by NUREG-0612, for the design, fabrication, testing, maintenance, and quality assurance requirements applicable to the spent fuel cask special lift device. The proposed change does not involve a physical change to or require new or different operability requirements for plant systems, structures, or components. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, provides methods acceptable to the NRC for assuring the safe handling of heavy loads. NUREG-0612 endorses the use of ANSI N14.6 for the design, fabrication, testing, maintenance, and quality assurance requirements applicable to special lifting devices used to handle heavy loads in the proximity of safe shutdown equipment and irradiated spent fuel, thereby eliminating the need for license conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively. Accordingly, removal of license conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NFP-8, respectively, does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change replaces license conditions 2.C.3.f and 2.C.4 from FOLs NPF-2 and NPF-8, respectively, with a commitment in the FNP UFSAR to the requirements of ANSI N14.6, as clarified by NUREG-0612, for the design, fabrication, testing, maintenance, and quality assurance requirements applicable to the spent fuel cask special lift device. The proposed change does not involve: (1) A physical change to plant systems, structures or components; or (2) require new or different operability requirements for plant systems, structures, or components. SNC's commitment to the guidance provided in ANSI N14.6, as clarified by NUREG-0612, provides assurance that the spent fuel cask special lift device, in conjunction with the use of the single-failure proof spent fuel cask crane, will preclude the possibility of a cask drop accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in a margin of safety?
The proposed change does not involve a physical change to the plant or impact the operability requirements of systems, structures, or components considered important to safety. As stated above, the use of ANSI N14.6, as clarified by NUREG-0612, has been endorsed by the NRC in NUREG-0612. The proposed change replaces license conditions 2.C.3.f and 2.C.4 to FOLs NPF-2 and NPF-8, respectively, with a commitment in the FNP UFSAR to the requirements of ANSI N14.6, as clarified by NUREG-0612, for the design, fabrication, testing, maintenance, and quality assurance requirements for the spent fuel cask crane special lift device. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Date of amendment request: March 29, 2002 (TSC 02-02) as supplemented by a letter dated October 10, 2002.
Description of amendment request: The proposed amendment deletes several of the Unit 1 Technical Specification (TS) surveillance requirements (SR) contained in TS 3/4.4.5, “Steam Generators” (SGs), associated with the voltage-based SG alternative repair criteria (ARC). In addition the proposed changes would delete License Condition 2.C.9.d which references commitment letters associated with SG inspection activities.
Tennessee Valley Authority's [TVA's] proposed TS amendment does not compromise limits associated with SG tube integrity. TVA's proposed change removes existing SG tube plugging criteria (i.e., ARC) from the TS and reestablishes the standard TS criteria (40 percent through-wall criteria). This change is inherently more conservative.
The proposed revision does not alter plant equipment, test methods or operating practices. The proposed change continues to provide controls for safe operation of SQN SGs within the required limits. The proposed change does not contribute to events or assumptions associated with postulated design basis accidents (i.e., SG tube rupture). The proposed change does not affect operator indicators or actions required to diagnose or mitigate a SG tube rupture accident. The proposed revisions continue to maintain the required safety functions. Accordingly, the probability of an accident or the consequences of an accident previously evaluated is not increased.
TVA's proposed amendment removes existing repair criteria and incorporates the more conservative TS limit for SG tube plugging (i.e., plug tubes with degradation depths equal to or greater than 40 percent through-wall). This change will not give rise to new failure modes. The failure of a SG tube to maintain leakage integrity during operation is an analyzed event in the SQN Updated Final Safety Analysis Report. Accordingly, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
TVA's proposed TS amendment is conservative with respect to the margin of safety. The margin of safety is preserved through ensuring structural integrity and leakage integrity of the SG tubes.
TVA's proposed change to remove ARC from the TS does not compromise structural integrity or leakage integrity of SG tubes. The proposed change invokes the standard TS tube plugging criteria limit (40 percent through-wall criteria) which is inherently more conservative.
The proposed change does not affect the plant conditions, setpoints, or safety limits that could result in precursors to accidents or degrade accident mitigation systems. Plant system safety functions are not altered by the proposed change. Consequently, the proposed TS revisions does not reduce 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 Start Printed Page 66015amendment request involves no significant hazards consideration.
Date of application for amendments: September 6, 2002 (TS 00-14).
Brief description of amendments: The proposed amendments would change the Sequoyah (SQN) Units 1 and 2 Technical Specification (TS) 3/4.4.9.1, “Pressure/Temperature [P-T] Limits, Reactor Coolant System” and TS 3/4.4.12, “Low Temperature Overpressure Protection [LTOP] Systems.” The proposed amendment provides two changes to the these specifications as described below:
1. The proposed change relocates the information provided in these TSs into a pressure temperature limit report (PTLR) format in accordance with U.S. Nuclear Regulatory Commission (NRC) Generic Letter (GL) 96-03, “Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits.”
2. The proposed change also upgrades these TSs to the standard TS requirements for Westinghouse plants (NUREG-1431, Revision 2). In addition, the Tennessee Valley Authority (TVA) proposed a change to SQN TS 3/4.4.9.2, “Pressurizer,” to relocate the requirements of this TS into the SQN Technical Requirements Manual (TRM).
The proposed revision does not affect plant equipment, test methods or operating practices. The modification to SQN TSs is consistent with the Standard Technical Specifications for Westinghouse Plants and continues to provide controls for safe operation within the required limits. The revised specifications provide appropriate administrative controls for the RCS [reactor coolant system] P-T limits and LTOP setpoints within the PTLR for future revisions as needed. The proposed changes do not contribute to events or assumptions associated with postulated design basis accidents (DBA). The proposed revisions continue to maintain the required safety functions. Accordingly, the probability of an accident or the consequences of an accident previously evaluated is not increased.
The proposed revisions are not the result of changes to plant equipment, test methods, or operating practices. The proposed revision to the SQN RCS P-T limits, and LTOP setpoints continues to ensure that conservative fracture toughness margins are maintained to protect against reactor pressure vessel failure and overpressure conditions. The modified P-T limits and LTOP setpoints are based on NRC approved methodology in conjunction with alternative methods provided in ASME Code Case N-640, “Alternative Requirement Fracture Toughness for Development of P-T Limit Curves for ASME [American Society of Mechanical Engineers] Section XI, Division 1” and WCAP-15315, “Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Operating PWR [pressurized water reactor] and BWR [boiling water reactor] Plants.”
The proposed changes to incorporate the PTLR format is administrative in nature and provide controls for maintaining RCS P-T limits and LTOP setpoints for future revisions as needed.
The reactor vessel P-T limits and LTOP setpoints are operational limits and are not considered to be contributors to the generation of postulated accidents. The safety functions of the associated systems remain unchanged and do not affect the assumptions of DBAs. The operational limits and setpoints continue to be governed within the TSs/PTLR. Accordingly, the proposed changes do not create the possibility of a new or different kind of accident.
TVA's proposed TS amendment provides revised reactor pressure vessel P-T limits and LTOP setpoints that are within the design capabilities of the RCS Safety Structures, Systems and Components (SSC) and pressure control systems. The limits are based on conservative design margins that ensure that plant operation is within the design capacity of the reactor vessel materials. Accordingly, the function of the RCS to provide a fission product barrier is not compromised.
TVA's proposed change to include revised P-T and LTOP limits does not result in a change to system design features. The proposed change does not affect plant conditions that result in precursors to accidents or cause degradation of accident mitigation systems. The plant system safety functions are not altered by the proposed change.
The proposed changes to the P-T limits and LTOP setpoints change the calculations and method from that described in the current TS Bases to one based on ASME Code Case N-640 and WCAP-15315. The effect of this change is to allow plant operation with different limits while continuing to retain conservative margins for assuring integrity of the reactor vessel and the RCS. Consequently, the proposed TS revisions do not significantly reduce the margin of safety.
Date of application for amendment: July 9, 2002. Start Printed Page 66016
Brief description of amendment: The proposed amendment would revise the Technical Specifications to remove the cycle-specific allowances on (1) Rod insertion limits during individual rod position indicator channel calibrations and (2) rod position indicator channel accuracy for operation at or below 50 percent power. The proposed amendment also would revise the control rod indicated misalignment limits.
Date of publication of individual notice in Federal Register: October 7, 2002 (67 FR 62500).
Expiration date of individual notice: November 6, 2002.
Date of application for amendments: October 1, 2002.
Brief description of amendments: The amendments revise the licensing basis as described in the Updated Final Safety Analysis Report (UFSAR) to allow lifting heavier loads with the reactor building crane during the Unit 1 refueling outage beginning in November 2002.
Date of publication of individual notice in Federal Register: October 4, 2002 (67 FR 62270).
Expiration date of individual notice: November 4, 2002.
Date of application for amendment: September 11, 2001, as supplemented on June 27 and September 19, 2002.
Brief description of amendment: The amendment revised the Technical Specifications, Section 3.9, “Refueling,” and its corresponding bases to permit the continuation of core alterations during refueling operations with the refueling interlocks inoperable by providing alternate actions which will preserve the intended design function of the inoperable interlocks.
Date of Issuance: October 10, 2002.
Effective date: October 10, 2002, and shall be implemented within 30 days of issuance.
Date of initial notice in Federal Register: March 5, 2002 (67 FR 10008). The June 27 and September 19, 2002, letters provided clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated October 10, 2002.
Date of application for amendment: March 13, 2002, as supplemented May 10, August 14, September 5, September 23, and October 4, 2002.
Brief description of amendment: The amendment revises the Technical Specifications (TS) for HBRSEP2 to permit selective implementation of alternative radiological source term and modify the TS requirement for movement of irradiated fuel and performing core alterations.
Date of issuance: October 4, 2002.
Date of initial notice in Federal Register: April 30, 2002 (67 FR 21285). The May 10, August 14, September 5, September 23, and October 4, 2002, supplements contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 4, 2002.
Date of application for amendment: January 31, 2002, as supplemented by letters dated June 12, June 25, July 22, September 16, and October 2, 2002.
Brief description of amendment: This amendment increases the licensed power level by approximately 1.7%, from 3,833 megawatts thermal (MWt) to 3,898 MWt. These changes result from increased feedwater flow measurement accuracy to be achieved by utilizing high accuracy ultrasonic flow measurement instrumentation.
Date of initial notice in Federal Register: April 2, 2002 (67 FR 15622). Start Printed Page 66017The June 12, June 25, July 22, September 16, and October 2, 2002, supplemental letters 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 October 10, 2002.
Description of amendment request: The amendment revises the Cooper Nuclear Station's Technical Specifications (TS) 5.5.7, “Ventilation Filter Testing Program (VFTP),” reflecting a correction of an erroneous reference to American Society of Mechanical Engineers N510-1980.
Effective date: The amendment is effective on the date of issuance, to be implemented within 30 days from the date of issuance.
Facility Operating License No. DPR-46: Amendment revises the Technical Specifications.
Date of initial notice in Federal Register: September 5, 2001 (66 FR 46480). The Commission related evaluation of the amendment is contained in a Safety Evaluation dated September 30, 2002.
Date of application for amendments: June 28, 2002, as supplemented on August 15, August 16, and October 2, 2002.
Brief description of amendments: The amendments change the Salem Technical Specifications (TS) requirements for Fuel Decay Time prior to commencing movement of irradiated fuel. TS Limiting Condition for Operation 3/4.9.3, “Decay Time,” is revised to allow fuel movement in the containment to commence 100 hours after the reactor has become subcritical between October 15th through May 15th. Should refueling occur between May 16th and October 14th, the current 168 hours decay time limit will remain in place. These requirements are valid through the year 2010.
Amendment Nos.: 251 and 232.
Date of initial notice in Federal Register: August 30, 2002 (67 FR 55887). The August 15, August 16, and October 2, 2002, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 10, 2002.
Brief description of amendments: These amendments change the Salem Technical Specifications (TSs) requirements associated with its containment spray nozzles. The frequency of TS Surveillance Requirement (SR) 4.6.2.1.d for verifying that the containment spray nozzles are unobstructed is changed from a fixed 10-year frequency to after activities that could result in nozzle blockage. In this case, PSEG will be required to evaluate the work performed to determine the impact to the containment spray system, or perform an air or smoke flow test. The applicable Bases pages are also revised to reflect this change.
Amendment Nos.: 252 & 233.
Date of initial notice in Federal Register: August 20, 2002 (67 FR 53989). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 10, 2002.
Date of application for amendments: November 1, 2001, as supplemented on October 1, 2002.
Brief description of amendments: The changes modify the provisions under which equipment may be considered operable when either its normal or emergency power source is inoperable. Technical Specifications (TS) Section 3.0.5 was deleted and additional limiting conditions for operation were incorporated into electrical power systems TS 3.8.1.1, “A.C. Sources—Operating.” The corresponding TS Bases were modified accordingly. The proposed changes are consistent with the recommendations contained in NUREG-1431, Rev. 2, “Standard Technical Specifications for Westinghouse Plants.”
Date of issuance: October 11, 2002.
Amendment Nos.: 253 and 234.
Date of initial notice in Federal Register: February 5, 2002 (67 FR 5331). The October 1, 2002 supplement was within the scope of the original application and did not change the staff's proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated October 11, 2002.
Brief description of amendment: The amendment eliminates the security plan requirements from the 10 CFR Part 50 licensed site after the spent nuclear fuel has been transferred to the 10 CFR Part 72 licensed Independent Spent Fuel Storage Installation and is based in part on exemptions from specific requirements set forth in 10 CFR Part 73 and 10 CFR 50.54(p).
Effective date: October 10, 2002, to be implemented within 30 days.
Facility Operating License No. DPR-54: The amendment revised the Operating License and the Technical Specifications.
Date of initial notice in Federal Register: March 21, 2001 (66 FR 15930). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 10, 2002.
No significant hazards consideration comments received: No. Start Printed Page 66018
Brief description of amendment: The amendment revises Surveillance Requirements (SRs) 3.3.1.2 and 3.3.1.3 of the technical specifications on the reactor trip system (RTS) instrumentation. The change to SR 3.3.1.2 replaces the reference to the nuclear instrumentation system channel output by a reference to the power range channel output and deletes Note 1 to the SR. The change to SR 3.3.1.3 is editorial.
Effective date: October 2, 2002, and shall be implemented within 6 months of the date of issuance, including the incorporation of changes to the Technical Specification Bases as described in the licensee's application dated July 25, 2002.
Date of initial notice in Federal Register: August 20, 2002 (67 FR 53992). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated October 2, 2002.
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, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Assess 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 Public Document Room (PDR) Reference staff at 1-800-397-4209, 304-415-4737 or by email to pdr@nrc.gov.
The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. By November 29, 2002, 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 Start Printed Page 66019petition 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,[2] which is available at the Commission's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, and electronically on the Internet at the NRC web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If there are problems in accessing the document, contact the PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to
pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or 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.
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: Rulemakings and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of the 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 petition for leave to intervene and request for hearing 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.
Description of amendment request: The amendments consist of a one-time change to the Dresden Updated Final Safety Analysis Report (UFSAR) to state that lifting heavy loads up to and including 116 tons is allowed prior to and during the upcoming Dresden Unit 3 refueling outage number 17.
Amendment No.: 196 and 189.
Facility Operating License Nos. DPR-19 and DPR-25: Amendment revises the UFSAR.
Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Joliet Herald News, dated October 1, 2002. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received.
The Commission's related evaluation of the amendment, finding of exigent circumstances, state consultation, and final NSHC determination are contained in a Safety Evaluation dated October 4, 2002.
Dated at Rockville, Maryland, this 18th day of October 2002.
2. “The most recent version of Title 10 of the Code of Federal Regulations, published January 1, 2002, inadvertently omitted the last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), regarding petitions to intervene and contentions. Those provisions are extant and still applicable to petitions to intervene. Those provisions are as follows: “In all other circumstances, such ruling body or officer shall, in ruling on—
(iii) The possible effect of any order that may be entered in the proceeding on the petitoner's interest.
[FR Doc. 02-27243 Filed 10-28-02; 8:45 am]