Source: https://www.federalregister.gov/documents/2002/07/09/02-16956/biweekly-notice-applications-and-amendments-to-facility-operating-licenses-involving-no-significant
Timestamp: 2018-12-17 02:57:59
Document Index: 117756295

Matched Legal Cases: ['art 2', 'art 72', 'art 20', 'art 20', 'art 50', 'art 50', 'art 50', 'art 50', 'art 100', 'art 20']

45560-45576 (17 pages)
Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear Plant, Charlevoix, County, Michigan
Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-336 and 50-423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London County, Connecticut
Entergy Nuclear Operations, Inc. (ENO), Docket No. 50-003, Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York County, Pennsylvania
Southern California Edison Company, Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California
https://www.federalregister.gov/d/02-16956 https://www.federalregister.gov/d/02-16956
This biweekly notice includes all notices of amendments issued, or proposed to be issued from June 14, 2002 through June 27, 2002. The last biweekly notice was published on June 25, 2002 (67 FR 42814).
By July 25, 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 Start Printed Page 455612.714, [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.
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, 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, 304-415-4737 or by e-mail to pdr@nrc.gov. Start Printed Page 45562
Description of amendment request: The amendment request changes the Defueled Technical Specifications by adding applicability statements to the requirements for storage and inspection of spent fuel and for the program requirements for spent fuel pool water chemistry.
The requested license amendment involves the addition of applicability statements to the program and activity requirements for the storage and inspection of spent fuel activities and requirements and the SFP [spent fuel pool] water chemistry. These applicability statements make requirements applicable whenever irradiated fuel is stored in the SFP. Once irradiated fuel has been completely removed from the SFP and transferred to a certified dry fuel storage container under a general 10 CFR Part 72 license, these program requirements for the SFP are no longer necessary. The program requirements consist of the specification, establishment, implementation, and maintenance of fuel configuration, fuel cooling, and water chemistry for the SFP to minimize the potential effects of decay heat and corrosion.
The corresponding program requirements for fuel storage in dry containers are specified in the container's certificate of conformance and safety analysis report. The corresponding program requirements currently include:
1. Analysis of fuel assemblies to determine maximum temperatures within the fuel assemblies to the temperature at the edge of the assemblies,
2. Design of passive heat removal components to remove heat via convection, conduction, and radiation, and
3. Specifications for canister vacuum drying pressure and helium backfill pressure that would ensure that a sufficiently inert environment is produced within the canister to inhibit corrosion.
The program requirements associated with fuel storage in the SFP do not contribute to accident prevention or mitigation following the complete removal of irradiated fuel. The corresponding program features for fuel storage in dry storage containers are specified and containers are specified and controlled under other applicable license documents. These changes do not significantly increase the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident from any other accident previously evaluated.
The requested amendment involves the addition of applicability statements that will have the effect of making a program requirement associated with the SFP inapplicable when the SFP is no longer used for irradiated fuel storage. The corresponding program requirements are adequately specified in applicable license documents. The elimination of this program requirement following complete removal of irradiated fuel from the SFP does not result in any new or different accident initiators from those already assumed in accidents previously evaluated, nor does it exacerbate any such accidents. Therefore, these changes do not create the possibility of a new or different kind of accident from any previously evaluated.
The safety margins produced as a result of the specification of program requirements for fuel storage in the SFP are adequately maintained in corresponding program requirements associated with fuel storage in dry storage containers. These corresponding program requirements are specified in the dry storage container's certificate of compliance and safety analysis report. Therefore, this change does not involve a significant reduction in a margin of safety.
Description of amendment request: The proposed amendment modifies the Millstone Nuclear Power Station, Unit No. 2 (MP2) and Unit No. 3 (MP3) Technical Specifications (TSs) to change selected MP2 and MP3 radiological-related TSs. These changes are due to the revision to Part 20 of Title 10 of the Code of Federal Regulations.
These changes do not have an impact on the acceptance criteria for any design-basis accident described in the respective MP2 or MP3 Updated Final Safety Analysis Report (UFSAR).
The revisions to the Occupational Radiation Exposure Report, Radioactive Effluent Controls Program, and High Radiation Area Specifications in accordance with TSTF travelers 152, 258, and 308 will have no effect on plant operation. Since the proposed changes are solely administrative or editorial in nature, they do not affect plant operation in any way.
Since the proposed changes are solely administrative or editorial changes to the TSs, they do not affect plant operation in any way. The proposed changes to each unit's TSs will revise them to reflect the requirements of the current 10 CFR Part 20, standardize terminology, provide clearer guidance, clarify inconsistencies, remove extraneous information, and result in minor format changes that will not result in any technical changes to current requirements.
The proposed changes have no effect on any safety analyses assumptions and therefore do not impact any margins of safety. The proposed changes do not impact any acceptance criteria for the design-basis accidents described in the Start Printed Page 45563respective MP2 or MP3 UFSAR and do not impact the consequences of accidents previously evaluated. Therefore, the proposed changes will not result in a reduction in a margin of safety.
Date of amendment request: May 29, 2002.
Description of amendment request: The amendments would revise the Technical Specifications 5.5.2 to allow, on a one-time basis, extension of the interval governing the conduct of containment integrated leak rate test (ILRT) from ten to fifteen years. The amendments represent a one-time exception to the ten-year frequency of the performance-based Type A tests as delineated by Regulatory Guide 1.163, “Performance-Based Containment Leak-Test Program,” September 1995. The amendments will allow conduct of each respective unit's ILRT within fifteen years from the last ILRT performed for each unit.
The following discussion is a summary of the evaluation of the changes contained in these proposed amendments 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 amendments will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed extension to the Type A testing intervals cannot increase the probability of an accident previously evaluated since extension of the intervals is not a physical plant modification that could alter the probability of accident occurrence, nor is it an activity or modification by itself that could lead to equipment failure or accident initiation. The proposed extension to the Type A testing intervals does not result in a significant increase in the consequences of an accident as documented in NUREG-1493. The NUREG notes that very few potential containment leakage paths are not identified by Type B and Type C tests. It concludes that reducing the Type A testing frequency to once per twenty years leads to an imperceptible increase in risk.
Catawba and McGuire provide a high degree of assurance through testing and inspection that the containments will not degrade in a manner detectable only by Type A testing. Recent Type A tests for the Catawba and McGuire units identified containment leakage within acceptance criteria, indicating a very leak tight containment. Inspections required by the ASME Code are also performed in order to identify indications of containment degradation that could affect leak tightness. Separately, Type B and Type C testing, required by TS [Technical Specifications], identify any containment opening from design penetrations, such as valves, that would otherwise be detected by a Type A test. These factors establish that an extension to the Type A test intervals will not represent a significant increase in the consequences of an accident.
The proposed amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed revisions to the Catawba and McGuire TS add a one-time extension to the current interval for Type A testing. The current test interval of ten years, based on past performance, would be extended on a one-time basis to fifteen years from the last Type A test. The proposed extension to Type A test intervals does not create the possibility of a new or different type of accident since there are no physical changes being made to the plants and there are no changes to the operation of the plants that could introduce a new failure mode.
The proposed amendments will not involve a significant reduction in a margin of safety. The proposed revisions to the Catawba and McGuire TS add a one-time extension to the current interval for Type A testing. The current test interval of ten years, based on past performance, would be extended on a one-time basis to fifteen years from the last Type A test. The proposed extension to Type A test intervals will not significantly reduce the margin of safety. The NUREG-1493 generic study of the effects of extending containment leakage testing intervals found that a twenty-year interval resulted in an imperceptable increase in risk to the public. NUREG-1493 found that, generically, the design containment leakage rate contributes about 0.1 percent of the overall risk and that decreasing the Type A testing frequency would have a minimal effect on this risk, since 95 percent of the Type A detectable leakage paths would already be detected by Type B and Type C testing. Similar proposed changes have been previously reviewed and approved by the NRC, and they are applicable to Catawba and McGuire.
Based upon the preceding discussion, Duke Energy Corporation has concluded that the proposed amendments do not involve a significant hazards consideration.
Description of amendment request: The proposed change will revise Appendix 3B and Section 6.2.1.2 of the Updated Safety Analysis Report pertaining to the method of analysis. The proposed change will replace the current vendor THREED code for room pressure-temperature analyses due to High Energy Line Breaks (HELB) with GOTHIC (Generation of Thermal-Hydraulic Information for Containments). The proposed change will allow Entergy Operations, Inc. (EOI) to update the analysis and to evaluate additional changes to the plant.
1. Will the operation of the facility in accordance with these proposed changes involve a significant increase in the probability or consequence of an accident previously evaluated?
Response: The proposed change involves no increase in the probability of the accidents previously evaluated since no physical change to the plant will be made. The change of the High Energy Line Break (HELB) analysis method does not affect the probability of the analyzed event occurring. Start Printed Page 45564The line break locations have not been affected and remain as originally designed.
This submittal is required due to the change of HELB analysis code from the vendor code THREED to the modern industry standard analysis code GOTHIC. This is a change in the methodology for determining the effects of the mass and energy release in the plant as a result of currently postulated events. The change in the evaluation methodology has been benchmarked and reviewed to confirm the results remain consistent with the current analysis. The changes to the model used for the additional analysis allow the use of new, more physically realistic models for Containment and Auxiliary Building pressure/temperature responses and will demonstrate continued qualification of the equipment in these buildings. Mass and energy releases for some cases have also been recalculated to credit pipe friction, which was only credited for certain cases previously.
With these new results the equipment has been reviewed and remains qualified per current programs established at RBS [River Bend Station]. Therefore, the plant will continue to function as designed and thus there will be no impact on consequences.
2. Will the operation of the facility in accordance with these proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No physical change to the plant will be made. The HELB locations were identified by reviewing all the possible break locations in each Auxiliary and Containment Building volume containing high-energy lines. The locations of the breaks remain the same as the previous HELB analyses. The HELB analyses have been evaluated for the current plant configuration. The new HELB analysis has been benchmarked against the previous accepted methods and found to correlate with the previous analysis. Therefore the results can be used to predict plant responses to events. The proposed change uses improved methods for mass and energy release calculation and pressure / temperature responses to determine the EQ [equipment qualification] qualification envelopes. Therefore, no new or different interaction would be created.
3. Will the operation of the facility in accordance with these proposed changes involve a significant reduction in a margin of safety?
Response: The operation of the facility in accordance with the proposed changes will not involve a significant reduction in a margin of safety.
The GOTHIC code has been successfully benchmarked versus the vendor THREED code, which was used in the original design calculations. The HELB analysis results with the benchmarking GOTHIC model are consistent with the THREED results. Therefore, the use of GOTHIC code will not involve a reduction in an identified margin of safety. Given that GOTHIC code is an improved methodology and it has been extensively qualified against the solved analytical problems and testing results, the use of GOTHIC code will produce more accurate pressure/temperature responses for the HELB analyses. The use of the GOTHIC code has been approved for pressure/temperature responses analysis at various other plants including Joseph M. Farley Nuclear Plant, Units 1 and 2, and Waterford [Steam Electric Station, Unit] 3.
The results with the revised methods will be used to show that safety equipment meets the EQ requirements. The peak temperatures and pressures in the HELB GOTHIC benchmark model are within the existing EDC [environmental design criteria] envelopes. Therefore, the pressure/temperature responses from the HELB benchmark analyses have no impact on the equipment qualification.
The methodology in the original design calculations is very conservative. The mass and energy releases without crediting friction introduce excessive amount of high-energy fluid into the break rooms, which is unrealistic. Some HELB calculations have credited both the frictional flows and the additional zone to eliminate excessive conservatism in the pressure/temperature responses. There is no reduction in a margin of safety and the design room differential pressure limits continue to be [met].
The use of this method by EOI RBS is consistent with the guidance given in NRC [U.S. Nuclear Regulatory Commission] Generic Letter 83-11 and Supplement 1, addressing the performance of safety analyses by licensees. EOI has implemented this guidance for the GOTHIC methodology consistent with the intended application. The GOTHIC methodology has been verified and validated by the software vendor. In addition, this methodology is controlled by EOI procedures and under the EOI quality assurance program. This includes EOI and RBS specific verification and validation of this application of GOTHIC and review of the calculations performed.
Date of application for amendment: May 30, 2002.
Description of amendment request: The proposed changes will modify the Indian Point Generating Station, Unit 1 (IP1), Technical Specifications (TSs) and Provisional Operating License No. DPR-5. IP1 is completely enclosed within the protected area for Indian Point Nuclear Generating Station, Unit 2 (IP2). IP1 depends on the IP2 TSs and processes for the implementation of certain regulatory requirements. The requested changes will simplify the IP1 TSs to facilitate the IP2 transition to the Improved TSs. The IP1 TSs will be reformatted, reordered and repaginated for consistency and clarity. ENO also proposes that certain changes supersede requirements of the “Order Approving Decommissioning Plan and Authorizing Decommissioning of Facility” [2] (the Order) to ensure compliance with the current requirements of 10 CFR Part 50.59, “Changes, tests, and experiments.” and 10 CFR Part 50.82, “Termination of license,” for evaluating whether changes can be made to IP1 without NRC approval.
The NSB [Nuclear Services Building] sewage effluent line radiation monitor is not required to function to mitigate any postulated accident. The design or operation of the radiation monitor on the existing sewage effluent discharge line will not be changed by deleting operability and surveillance requirements for the NSB sewage effluent radiation monitor from the IP1 TS. The nuclear services building sewage effluent line is neither an accident initiator nor mitigator.
The other proposed changes do not result in a change to the design or operation of any plant structure, system or component. Therefore any assumptions of the operability or performance of any structure, system or component in accident evaluations are unchanged.
The proposed fire protection TS 2.11 involves deleting requirements from the IP1 TS that are solely applicable to IP2. Any assumptions of the operability or performance of any structure, system or component in IP2 accident evaluations, including the Fire Plan, are unchanged. Therefore, there is no increase in the probability or in the consequences of an accident previously evaluated.
The proposed TS change involves the deletion of operability and surveillance requirements for radioactive effluent monitoring of the NSB sewage effluent from the IP1 TS. The proposed TS changes do not Start Printed Page 45565affect the design or operation of any plant structure, system, or component.
This change to TS 1.0 does not affect a design function for or the operation of any plant structure, system, or component. The change does not affect the method of ENO's compliance with any regulation.
The proposed TS change involving IP1 TS 2.11 statement governs the protection of IP2 safe shutdown systems from fire. Effective protection of IP2 safe shutdown systems from fire is mandated by IP2 License Condition 2.K. The effectiveness of ENO compliance with IP2 License Condition 2.K is not affected by this change. In addition, this change does not affect a design function or the operation of any plant structure, system, or component.
The proposed changes to TS sections 3.1 and 3.2 involve eliminating the duplication of requirements in the IP1 TS and incorporating the requirements by reference to the IP2 TS. A single ENO organization operates both IP1 and IP2. The effective organizational requirements to ensure compliance with all ENO IP1 and IP2 site requirements are mandated by the IP2 TS. The effectiveness of ENO's safety management of the Indian Point site is not affected by this change. In addition, this change does not affect a design function or the operation of any plant structure, system, or component.
The proposed TS change to sections 4.1 and 5.2 involves eliminating the reference in the IP1 TS to the specific applicable section number of the IP2 TS. A single organization operates both IP1 and IP2. The applicable IP2 TS is obvious by the activity title. The effectiveness of ENO's safety management of the Indian Point site is not affected by this change. In addition, this change does not affect a design function or the operation of any plant structure, system, or component.
Effective compliance with the 10CFR20 requirements for radiation protection and monitoring radioactive effluent releases is mandated by other IP1 and IP2 TS and license provisions. The effectiveness of ENO compliance with 10CFR20 requirements is not adversely affected by the elimination of TS requirements for the radiation protection plan and radioactive effluent monitoring on the nuclear services building sewage effluent line.
The proposed TS change involves requirements for the site Meteorological Monitoring and Radiological Environmental Monitoring programs. However, IP2 TS provisions mandate effective compliance for meteorological and radiological environmental monitoring. The effectiveness of ENO compliance with 10CFR50.47, 10CFR100, and 10CFR20 requirements is not adversely affected by this change. In addition, this change does not affect a design function or the operation of any plant structure, system, or component. IP2 TS provisions mandate effective compliance with requirements for radiation protection. The effectiveness of ENO's compliance with 10 CFR 20 is not adversely affected by this change or the change to the section for sealed sources. In addition, this change does not affect a design function or the operation of any plant structure, system, or component.
The proposed TS change involves the location of routine and event reporting requirements. However, other IP2 TS provisions mandate effective compliance with reporting requirements. In addition, this change does not affect a design function or the operation of any plant structure, system, or component.
The effectiveness of ENO's compliance with 10CFR50.59 is not adversely affected by the clarification and relocation of the applicability of the FSAR [Final Safety Analysis Report]. In addition, this change does not affect a design function or the operation of any plant structure, system, or component.
Therefore, the change does not result in a change to any of the safety analyses or any margin of safety.
ENO also requests that the expiration date of IP1 Provisional Operating License No. DPR-5 be changed from “midnight, October 14, 2002,” to “midnight, September 28, 2013,” the current expiration date for Facility Operating License No. DPR-26 for IP2.
In its Safety Evaluation and Environmental Assessment for its January 31, 1996, Order Approving Decommissioning Plan and Authorizing Decommissioning of Facility, the NRC evaluated the acceptability of the possession-only license and safety issues related to SAFSTOR of Indian Point Nuclear Generating Unit No. 1 until September 28, 2013. The requested change does not involve any activity that could change the assumptions of the prior Safety Evaluation and Environmental Assessment.
Therefore, the proposed license amendment does not involve a significant increase in the probability or in the consequences of an accident previously evaluated.
The NRC staff has reviewed the licensee's analyses 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 Fulton, Assistant General Consul, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.
Description of amendment request: The proposed amendment would revise the Facility Operating License and Technical Specifications (TSs) to increase the licensed core thermal power level to 3067.4 megawatts (MWt), which is a 1.4% increase above the currently authorized power level of 3025 MWt. The proposed power uprate involves the improvement in the core power uncertainty allowance originally required for the emergency core cooling system (ECCS) evaluations performed in accordance with Appendix K, “ECCS Evaluation Models,” to Part 50 of Title 10 of the Code of Federal Regulations. In addition, changes would be made in TS Sections 2.2, 3.3, 3.4, 3.7, and the applicable TS Bases would be revised to account for the change in power level.
The evaluations associated with this proposed change to core power level have demonstrated that all applicable acceptance criteria for plant systems, components, and analyses (including the Final Safety Analysis Report Chapter 14 safety analyses) will continue to be met for the proposed 1.4% increase in licensed core thermal power for IP3 [Indian Point Unit 3]. The subject increase in core thermal power will not result in conditions that could adversely affect the integrity (material, design, and construction standards) or the operational performance of any potentially affected system, component or analysis. Therefore, the probability of an accident previously evaluated is not affected by this change. The subject increase in core thermal power will not adversely affect the ability of any safety-related system to meet its intended safety function. Further, the radiological dose evaluations in support of this power uprate effort show that the current FSAR [Final Safety Analysis Report] Chapter 14 radiological analyses are unaffected, and that the current dose analyses of record bound plant operation with the subject increase in licensed core thermal power level.
The evaluations of this proposed amendment show that all applicable acceptance criteria for plant systems, components, and analyses (including FSAR Chapter 14 safety analyses) will continue to be met for the proposed 1.4% power increase in IP3 licensed core thermal power. The subject increase in core thermal power will not result in conditions that could adversely affect the integrity (material, design, and construction standards) or operational performance of any potentially affected system, component, or analyses. The subject increase in core thermal power will not adversely affect the ability of any safety-related system to meet its safety function. Furthermore, the conditions associated with the subject increase in core thermal power will neither cause initiation of any accident, nor create any new credible limiting single failure. The power uprate does not result in changing the status of events previously deemed to be non-credible being made credible. Additionally, no new operating modes are proposed for the plant as a result of this requested change.
Therefore, the subject increase in core thermal power level will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The evaluations associated with this proposed change show that all applicable acceptance criteria for plant systems, components, and analyses (including FSAR Chapter 14 safety analyses) will continue to be met for this proposed 1.4% increase in IP3 licensed core thermal power. The subject increase in core thermal power will not result in conditions that could adversely affect the integrity (material, design, and construction standards) or operational performance of any potentially affected system, component, or analysis. The subject power uprate will not adversely affect the ability of any safety-related system to meet its intended safety function. For example, most IP3 analyses already add a 2% uncertainty allowance to the nominal power level to account solely for power measurement uncertainty. These analyses have not been revised for the 1.4% uprate power level conditions because the sum of increased core power level (1.4%) and the improved power measurement accuracy (uncertainty less than 0.6%) is already bounded by the currently analyzed 2% uncertainty allowance.
Therefore, the subject increase in core thermal power will not involve a reduction in [a] margin of safety.
Date of amendment request: June 3, 2002.
Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.4.9, “Pressurizer,” to increase the pressurizer water level limit when the plant is in Mode 3 (Hot Standby). The current pressurizer water level limit is applicable for Modes 1, 2, and 3, and will remain unchanged for Modes 1 and 2. The proposed amendment would also revise TS 3.8.4, “DC Sources—Operating,” to remove the notes that refer to the one-time amendment allowing the online replacement of station batteries 31 and 32. The notes are no longer applicable since the batteries have been replaced.
Pressurizer water level is an assumed initial condition for certain accident analyses. Plant initial conditions are not accident initiators and do not have an effect on the probability of the accident occurring. The proposed change only revises the specified limit on water level in the pressurizer, so that this change would not affect accident probability.
The specific accidents for which pressurizer water level is an assumed initial condition are a loss of load and a loss of normal feedwater. The limiting accident analysis results occur at full power conditions when the available core thermal power is maximized. The proposed change does not affect the specified pressurizer level limit at any power level from zero to full power. That is, the pressurizer level limit is not being changed in Modes 1 and 2. The proposed change does revise the specified pressurizer water level limit in Mode 3 (Hot Standby) but this does not affect accident analysis results because the limiting analyses will remain those that are postulated to occur in Mode 1 with the plant at full power.
The proposed change does not involve physical changes to existing plant equipment or the installation of any new equipment. The design of the pressurizer, the pressurizer level control system and the pressurizer safety valves is not being changed and the ability of these systems, structures, and components to perform their design or safety functions is not being affected. The proposed change revises the specified limit on pressurizer water level in Mode 3 (Hot Standby) to allow operators greater flexibility in performing a plant cooldown. The method used in performing the plant cooldown is not being changed. This proposed change does not create new failure modes or malfunctions of plant equipment nor is there a new credible failure mechanism.
Pressurizer level is an initial condition assumed in certain accident analyses involving an insurge in the pressurizer and an increasing reactor coolant system (RCS) pressure. These analyses demonstrate that the design pressure for the RCS is not exceeded for the limiting analyses based on the plant at full power. The proposed change does not affect the existing Technical Start Printed Page 45567Specification requirement for Mode 1 (Power Operation) or Mode 2 (Plant Startup) and therefore does not affect the assumptions or results of these accident analyses. The margin for RCS design pressure demonstrated by these analysis results is not being reduced. The proposed change only applies to the pressurizer level limit in Mode 3 (Hot Standby) when there is substantially lower thermal energy available to cause rapid expansion of reactor coolant and an insurge to the pressurizer. Protection of the RCS pressure boundary is still maintained by the pressurizer safety valves, which are not being modified by the proposed change in pressurizer water level.
Date of amendment request: June 5, 2002.
Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to implement the alternate source term methodology for the fuel-handling accident analysis. Specifically, the proposed amendment would revise TS 3.9.3, “Containment Penetrations,” to: (1) Permit the equipment hatch opening and the personnel air lock doors to be capable of being closed during movement of irradiated fuel, (2) allow use of administrative controls for unisolating containment penetrations during movement of irradiated fuel, (3) delete the containment purge and containment pressure relief requirements and associated surveillances with the reactor subcritical for less than 550 hours, and (4) eliminate the TS applicability “during core alterations.” In this regard, the proposed amendment would adopt TS Task Force (TSTF) Standard TS Change Travelers TSTF-68, “Containment Personnel Airlock Doors Open During Fuel Movement,” TSTF-312, “Administratively Control Containment Penetrations,” and, in part, TSTF-51, “Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations.” The proposed amendment would also relocate the requirements in TS 3.7.13, “Fuel Storage Building Emergency Ventilation System,” and TS 3.3.8, “Fuel Storage Building Emergency Ventilation System Actuation Instrumentation,” to the licensee-controlled Technical Requirements Manual.
The proposed change involves the reanalysis of a fuel handing accident (FHA) in containment and in the fuel storage building. The new analysis, based on the Alternate Source Term (AST) in accordance with 10 CFR [Code of Federal Regulations] 50.67, will replace the existing analysis based on methodologies and acceptance criteria in place when Indian Point 3 was originally licensed. As a result of the new analysis, changes to the Technical Specifications are proposed which take credit for the new analysis results.
The proposed changes to the technical specifications modify requirements regarding containment closure during movement of irradiated fuel assemblies in containment and relocate requirements for the fuel storage building emergency ventilation system from the technical specifications to a licensee controlled document. The proposed changes do not involve physical modifications to plant equipment and do not change the operational methods or procedures used for moving irradiated fuel assemblies. As such, there are no accident initiators affected by the proposed amendment. The revised requirements apply only when the plant is in a refueling condition (Mode 6), and specifically only when irradiated fuel is being moved. Previously evaluated accidents with the plant in other conditions ranging from cold shutdown (Mode 5) through power operation (Mode 1) are not affected. The AST methodology is used to evaluate a[n] FHA that is postulated to occur during fuel movement activities in the containment building and the fuel storage building. The analysis follows the guidance of the NRC Regulatory Guide 1.183 and uses the acceptance criteria of the NRC Standard Review Plan (NUREG 0800) for offsite doses and General Design Criteria 19 for control room personnel. The analysis demonstrates that the dose consequences meet regulatory acceptance criteria. The accident analysis conservatively assumes that the containment building and the fuel storage building, including ventilation filtration systems for those building[s] does not diminish or delay the assumed fission product release. The analysis does take credit for, and technical specifications enforce, the presence of 23 feet of water over the irradiated fuel while fuel movement activities are being performed. The analysis also takes credit for, and the technical specification bases enforce a fuel decay time of at least 84 hours. In addition, administrative controls are put in place to provide for closure of containment openings in the event of a[n] FHA. Use of an alternate analysis method does not affect fuel parameters or the equipment used to handle the fuel. The proposed changes to the technical specifications reflect assumptions made in the analysis.
The proposed amendment involves the use of an alternate analysis methodology for the evaluation of the dose consequences from a[n] FHA that is postulated to occur in either the containment building or the fuel storage building (FSB). The analysis demonstrates that containment closure conditions and operation of the containment purge filtration system are not required to maintain dose consequence within regulatory limits following a postulated FHA in containment. Therefore the new analysis supports proposed changes to requirements for containment closure during movement of irradiated fuel assemblies in containment. The analysis results also demonstrate that operation of the fuel storage building emergency ventilation system is not required to maintain dose consequences within regulatory limits following a postulated FHA in the FSB. The containment closure components (e.g., equipment hatch, personnel airlock doors, and various containment penetrations) and filtration systems are not accident initiators. The proposed changes do not involve the addition of new systems or components nor do they involve the modification of existing plant systems. The proposed changes do not affect the way in which a[n] FHA is postulated to occur.
The existing dose analysis methodology and assumptions demonstrates that the dose consequences of a[n] FHA are within regulatory limits for whole body and thyroid doses as established in 10 CFR 100. The alternate dose analysis methodology and assumptions also demonstrates that the dose consequences of a[n] FHA are within regulatory limits. The limits applicable to the alternate analysis are established in 10 CFR 50.67 in conjunction with the TEDE (total effective dose equivalent) acceptance directed in Regulatory Guide 1.183. The acceptance criteria for both dose analysis methods have been developed for the Start Printed Page 45568purpose of evaluating design basis accidents to demonstrate adequate protection of public health and safety. An acceptable margin of safety is inherent in both types of acceptance criteria.
Description of amendment request: The proposed amendment would change the requirements associated with handling irradiated fuel and performing core alterations. Specifically, the changes would eliminate operability requirements for secondary containment when handling recently irradiated fuel and during core alterations. The amendment would also revise the requirements associated with equipment whose performance is not credited in the new calculations.
The proposed TS [Technical Specifications] changes do not modify the design or operation of equipment used to move spent fuel or to perform core alterations. Because the equipment affected by the change is not an initiator to any previously analyzed accident, the proposed change cannot increase the probability of any previously analyzed accident.
The conservative re-analysis of the fuel handling accident concludes that radiological consequences are within the acceptance criteria in Regulatory Guide 1.183 and 10 CFR 50.67. The results of the core alteration events, other than the fuel handling accident, remain unchanged from the original design-basis, which showed that these events do not result in fuel cladding damage or radioactive release. The radiological analysis uses the same FHA [fuel handling accident] source activity previously accepted in the design-basis FHA analysis. The same source activity is used with the guidance in the Regulatory Guide 1.183, Appendix B and the passive release/transport path, which does not take the dose mitigation credit of engineered safeguards including secondary containment and CREVAS [Control Room Emergency Ventilation] Systems.
The proposed post-FHA activity transport path is passive in nature and it does not take the credit of dose mitigation functions previously credited in the design-basis FHA analysis. The proposed changes do not introduce any new modes of plant operation and do not involve physical modifications to the plant.
The proposed changes revise the FitzPatrick TS to establish operational conditions where specific activities represent situations during which significant radioactive releases can be postulated. These new operational conditions are consistent with the proposed design-basis accident analysis and are established such that the radiological consequences are less than the regulatory allowable limits. Safety margins and analytical conservatisms are retained to ensure that the analysis adequately bounds all postulated event scenarios. The selected assumptions and release models provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensates for large uncertainties in facility parameters, accident progression, radioactive material transport and atmospheric dispersion. The proposed TS applicability statements continue to ensure that the TEDE [Total Effective Dose Equivalent] at the control room and the exclusion area and low population zone boundaries are below the corresponding regulatory allowable limits in 10 CFR 50.67(b)(2).
Date of application for amendments: May 24, 2002.
Description of amendment request: Exelon Generation Company, LLC, the licensee, is proposing changes to the Peach Bottom Atomic Power Station, Units 2 and 3 (PBAPS), Operating Licenses and Technical Specifications associated with an increase in the licensed power level. The changes involve a proposed 1.62 percent increase in the licensed reactor core thermal power level (an increase in reactor power level from 3,458 megawatts thermal to 3,514 megawatts thermal). These changes result from increased accuracy of the feedwater flow and temperature measurements to be achieved by utilizing high accuracy ultrasonic flow measurement instrumentation. This results in a more accurate determination of reactor core thermal power level. The basis for this change is consistent with the revision, issued in June 2000, to Appendix K to Part 50 of Title 10 of the Code of Federal Regulations, allowing operating reactor licensees to use an uncertainty factor of less than 2 percent of rated reactor thermal power in analyses of postulated design-basis loss-of-coolant accidents.
Response: No. The comprehensive analytical efforts performed to support the proposed uprate conditions included a review and evaluation of all components and systems that could be affected by this change. Evaluation of accident analyses confirmed the effects of the proposed uprate are bounded by the current dose analyses. All systems will function as designed, and all performance requirements for these systems have been evaluated and found acceptable.
The primary loop components (reactor vessel, reactor internals, control rod drive housings, piping and supports, recirculation pumps, etc.) continue to comply with their applicable structural limits and will continue to perform their intended design functions. Thus, there is no increase in the probability of a structural failure of these components.
All of the [Nuclear Steam Supply System] NSSS systems will still perform their intended design functions during normal and accident conditions. The balance of plant [(BOP)] systems and components continue to Start Printed Page 45569meet their applicable structural limits and will continue to perform their intended design functions. Thus, there is no increase in the probability of a structural failure of these components. All of the NSSS/BOP interface systems will continue to perform their intended design functions. The safety relief valves and containment isolation valves meet design sizing requirements at the uprated power level.
Because the integrity of the plant will not be affected by operation at the uprated condition, it is concluded that all structures, systems, and components required to mitigate a transient remain capable of fulfilling their intended functions. The reduced uncertainty in the flow input to the core thermal power uncertainty measurement allows most of the current safety analyses to be used, with small changes to the core operating limits, to support operation at a core power of 3514 megawatts thermal (MWt). Other analyses performed at a nominal power level have either been evaluated or re-performed for the 1.62% increased power level. The results demonstrate that the applicable analysis acceptance criteria continue to be met at the 1.62% uprate conditions. As such, all PBAPS Updated Final Safety Analysis Report (UFSAR) Chapter 14 accident analyses continue to demonstrate compliance with the relevant event acceptance criteria. Those analyses performed to assess the effects of mass and energy releases remain valid. The source terms used to assess radiological consequences have been reviewed and determined to bound operation at the 1.62% uprated condition.
Response: No. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. All systems, structures, and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed changes have no adverse effects on any safety-related system or component and do not challenge the performance or integrity of any safety related system.
Response: No. Operation at the uprated power condition does not involve a significant reduction in a margin of safety. Analyses of the primary fission product barriers have concluded that all relevant design criteria remain satisfied, both from the standpoint of the integrity of the primary fission product barrier and from the standpoint of compliance with the required acceptance criteria. As appropriate, all evaluations have been performed using methods that have either been reviewed and approved by the NRC, or that are in compliance with regulatory review guidance and standards.
Description of amendment request: The proposed change revises the safety limit minimum critical power ratio for Unit 1 Cycle 18 for two loop operation and single loop operation.
The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences. Limits have been established consistent with NRC approved methods to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change conservatively establishes the safety limit for the minimum critical power ratio (SLMCPR) for Quad Cities Nuclear Power Station (QCNPS), Unit 1, Cycle 18 such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation. The proposed change does not involve any modifications of the plant configuration or allowable modes of operation. The proposed change to the SLMCPR assures that safety criteria are maintained for QCNPS, Unit 1, Cycle 18.
The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1% of the rods are expected to be in boiling transition if the MCPR limit is not violated. The proposed change will ensure the appropriate level of fuel protection. Additionally, operational limits will be established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation.
This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences) are met.
NRC Section Chief: Anthony J. Mendiola. Start Printed Page 45570
Description of amendment request: The amendment would revise the Improved Technical Specifications (ITS) 3.3.8 and associated bases, “Emergency Diesel Generator (EDG) Loss of Power Start (LOPS),” by changing the completion time for required action D.2 from 12 to 36 hours. The amendment also corrects a typographical error in ITS 3.3.8 and clarifies the discussion in Bases Section B 3.3.8 for Actions D.1 and D.2 to recognize the applicability of ITS 3.3.8 in MODES 5 and 6.
The proposed license amendment revises the Required Time to place the plant in MODE 5 if an inoperable loss of voltage Function for the emergency diesel generator (EDG) loss of power start (LOPS) cannot be restored to OPERABLE status, corrects a typographical error in the Section Number of ITS 3.3.8, and clarifies the wording of ITS Bases Section B 3.3.8 for Action D.1 and D.2 regarding the applicability of the specification during MODES 5 and 6.
The EDG LOPS is intended to protect engineered safeguards equipment from damage due to sustained undervoltage conditions, and to ensure rapid restoration of power to the engineered safeguards electrical buses in the event of a loss of offsite power. The EDG LOPS is not an initiator of any design basis accident. The design functions of the EDG LOPS and the initial conditions for accidents that require an EDG LOPS will not be affected by the change. Therefore, the change will not increase the probability or consequences of an accident previously evaluated.
The proposed amendment involves no changes to the design functions or operation of the EDG LOPS. Editorial corrections, clarification of the wording in Bases Section B 3.3.8, or changing the Required Completion Time for placing the plant in MODE 5 when an inoperable loss of voltage function cannot be restored will not introduce any new failure mechanisms, malfunctions or accident initiators. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change corrects a typographical error, clarifies the wording of Bases Section B 3.3.8 for Actions D.1 and D.2, and revises the required Completion Time to place the plant in MODE 5. The revised Completion Time will allow the plant to be shutdown in an orderly fashion without challenging plant systems or plant cooldown limits. The proposed change does not change the design or operation of the EDG LOPS, and does not impact the ability of the EDG LOPS to perform its design functions. Thus, the proposed amendment will not result in a reduction in the margin of safety.
Description of amendment request: The proposed amendments would delete requirements from the Technical Specifications (TSs) (and, as applicable, other elements of the licensing bases) to maintain a Post Accident Sampling System (PASS). Licensees were generally required to implement PASS upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to PASS were imposed by Order for many facilities and were added to or included in the TSs for nuclear power reactors currently licensed to operate. However, lessons learned and improvements implemented over the last 20 years have shown that the information obtained from PASS can be readily obtained through other means, or is of little use in the assessment and mitigation of accident conditions.
The Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on December 27, 2001 (66 FR 66949) on possible amendments to eliminate PASS, 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 March 20, 2002 (67 FR 13027). The licensee affirmed the applicability of the NSHC determination in its application dated June 7, 2002. The NSHC determination is restated below.
The regulatory requirements for the PASS can be eliminated without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the Start Printed Page 45571consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. The elimination of the PASS will not prevent an accident management strategy that meets the initial intent of the post-TMI-2 accident guidance through the use of the SAMGs, the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).
Description of amendment request: The proposed amendment would revise TS 3.8.1, “AC Sources—Operating,” to allow portions of Surveillance Requirement (SR) 3.8.1.5 to be performed with the units in Mode 1, 2, 3 or 4. This proposed amendment is consistent with changes made to NUREG-1431, Standard Technical Specifications, Westinghouse Plants, by Technical Specification Task Force (TSTF) Traveler, TSTF-283, Revision 3.
The standby emergency power sources are primarily a support system for systems required to be operable for accident mitigation. SR 3.8.1.5 demonstrates the standby emergency power source operation, during a loss of offsite power actuation test signal in conjunction with an Engineering Safeguards Feature (ESF) actuation signal. The proposed amendment only changes the allowed operating Modes in which portions of this surveillance may be performed. Performing portions of the surveillance in Mode 1, 2, 3, or 4 will require an assessment to determine that plant safety is maintained or will be enhanced.
The possibility for a new or different type of accident from any accident previously evaluated is not created as a result of this amendment. These changes do not introduce any new or different normal operation or accident initiators. Performing the surveillance in Mode 1, 2, 3, or 4 will require an assessment to determine that plant safety is maintained or will be enhanced.
The standby emergency power sources are primarily a support system for systems required to be operable for accident mitigation. SR 3.8.1.5 demonstrates the standby emergency power source operation, during a loss of offsite power actuation test signal in conjunction with an ESF actuation signal. Performing the surveillance in Mode 1, 2, 3, or 4 will require an assessment to determine that plant safety is maintained or will be enhanced. There are no new or significant changes to the initial conditions contributing to accident severity or consequences. The proposed amendment will not otherwise affect the plant protective boundaries, will not cause a release of fission products to the public, nor will it degrade the performance of any other structures, systems or components (SSCs) important to safety. Therefore, allowing a portion of the surveillance to be performed in Mode 1, 2, 3, or 4, will not result in a significant reduction in the margin of safety.
Description of amendment request: The proposed amendment revises Technical Specifications (TSs) 3/4.3.5, allowing the automatic operation of the atmospheric steam relief valves during Mode 2 to maintain secondary side pressure at or below an indicated steam generator pressure of 1225 psig during startup and shutdown of the reactors.
The proposed change only provides another method of controlling the SG PORVs [steam generator power-operated relief valves] under specified operating conditions. The operating conditions in Specification 3/4.3.5 remain unchanged. No change is required to plant design since the proposed method of control is already part of the plant's configuration. The proposed method of control is the same method of control Start Printed Page 45572normally required by the specification in Modes 1 and 2. The proposed method of control will not impact the accident analysis assumptions or results. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed method of controlling the SG PORVs is the same method that these valves are controlled in Modes 1 and 2 by the specification under normal conditions. The proposed change will allow the setpoint of these valves to be adjusted to support startup and shutdown activities. The adjustment of the setpoint is restricted so that the accident analysis is not impacted. No change to the design of the valves or plant configuration is required to implement the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
The proposed change that will allow for an additional method of controlling the SG PORVs during startup and shutdown activities is consistent with the operating restrictions for the current method of valve control. The accident analysis assumptions and results will remain unaffected. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis, & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
Description of amendment request: The proposed amendment revises the near-end of life (EOL) Moderator Temperature Coefficient (MTC) Surveillance Requirements by placing a set of conditions on core operation.
The probability or consequences of accidents previously evaluated in the UFSAR [updated 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.
The margin of safety associated with the acceptance criteria of any accident is unchanged. The proposed change will have no affect on the availability, operability, or performance of the safety-related systems and components. A change to a surveillance requirement is proposed, but the limiting conditions for operation required by the Technical Specifications are not changed.
The Technical Specifications 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 the 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 [technical specification] is not reduced and the proposed change does not involve a significant reduction in a margin of safety.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, & Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
Date of amendment request: May 24, 2002.
Description of amendment request: The proposed amendments would allow Mode 2 (startup) operation with two, rather than three, intermediate range monitor channels per trip system.
The intermediate range monitors (IRMs) monitor neutron flux levels in the reactor core during startup. The IRM detectors are capable of generating a trip signal during a continuous rod withdrawal error in the startup range. However, the IRMs perform no function related to the probability of occurrence of a previously evaluated accident. Also, the IRM trip signal is not necessary to mitigate the limiting control rod withdrawal error. The limiting case assumes the trip signal is generated from the safety-related average power range monitor (APRM). Therefore, the consequences of this previously evaluated abnormal operating transient are not increased.
The proposed change reduces the number of required operable IRM channels per trip system from three to two. However, the manner in which the actuation logic functions and the systems respond are unaffected by the proposed change. Furthermore, the IRMs will continue to perform their design function of core monitoring during startup and mitigating nonlimiting transient events postulated to occur during startup. Therefore, the proposed change cannot create the possibility of a new or different kind of accident from any previously evaluated.
The Bases for Units 1 and 2 Technical Specification Table 3.3.1.1-1 state the “IRMs are capable of generating trip signals that can be used to prevent fuel damage resulting from abnormal operating transients in the intermediate power (startup) range.” The proposed change ensures the IRMs will still effectively mitigate these events. The most significant source of reactivity change is due to a control rod withdrawal error. With the proposed change, the IRMs will continue to Start Printed Page 45573provide protection against rod withdrawal errors, and peak fuel energy depositions will remain below the 170 cal/gm threshold criterion defined in the Technical Specifications Bases. Therefore, the proposed change does not reduce a margin of safety.
Date of amendment request: March 19, 2002, as supplemented on June 3, 2002.
Description of amendment request: The proposed Technical Specification changes involve the removal of the existing scram function and Group 1 isolation valve closure functions of the Main Steam Line Radiation Monitors (MSLRM). An explicit requirement for periodic functional test and calibration of the MSLRM is added to maintain operability of the mechanical vacuum pump (MVP) isolation function. This proposed no significant hazards consideration determination replaces in its entirety the notice published in the Federal Register on May 14, 2002 (67 FR 34495).
There is no accident analysis that relies on the high radiation scram of the reactor protection system and its removal has no impact on the consequences of accidents previously evaluated. The results of the control rod drop accident analysis remain within approved guidelines, thus any potential increase in consequences would not be considered significant.
The proposed changes to the plant involve limited changes to protective circuitry, but do not involve any plant hardware changes that could introduce any new failure modes. The changes will not affect non-MSLRM scram and isolation functions. In addition, the MSLRMs will remain active for other trip/isolation functions, and these monitors will still alarm in the control room to alert operators to off-normal conditions.
The proposed change involves the elimination of the scram and Group I isolation signal from the MSLRMs. Operation under the proposed change will not change any plant operation parameters, nor any protective system setpoints other than removal of these functions. The effects of the control rod drop accident without the MSLRM scram and isolation signal results in doses which remain well within 10 CFR Part 100, “Reactor Site Criteria,” limits.
Brief description of amendment request: The proposed amendment would revise Technical Specifications Section 4.13.A, “Inspection Requirements,” to allow the use of the optimum eddy current probe size when performing steam generator tube inspections. The proposed amendment would also correct several grammatical errors.
Date of publication of individual notice in Federal Register: June 25, 2002 (67 FR 42806).
Brief description of amendment: The amendment makes editorial and administrative corrections to Technical Specifications (TS) Section 3.3, Start Printed Page 45574“Instrumentation,” and eliminates minor discrepancies between TS Section 3.3 and other plant licensing basis documents.
Date of issuance: June 25, 2002.
Date of initial notice in Federal Register: December 26, 2001 (66 FR 66463). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 25, 2002.
Date of application for amendments: December 13, 2001.
Brief description of amendments: The amendments revise Item d of TS 5.5.11, “Ventilation Filter Testing Program (VFTP),” to lower the maximum allowable differential pressure across the engineered safety features ventilation systems units when tested at the specified system flow rates.
Date of issuance: June 18, 2002.
Effective date: June 18, 2002, and shall be implemented within 60 days of the date of issuance.
Amendment Nos.: Unit 1-142, Unit 2-142, Unit 3-142.
Date of initial notice in Federal Register: February 5, 2002 (67 FR 5325). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 18, 2002.
Date of application for amendment: November 19, 2001, as supplemented March 27, 2002
Brief description of amendment: The amendment revises Technical Specification 5.5.16 to eliminate the requirement to perform post-modification containment integrated leakage rate testing following replacement of the Unit 2 steam generators.
Date of issuance: June 27, 2002.
Effective date: As of the date of issuance to be implemented following the Unit 2 refueling and steam generator replacement outage in spring 2003.
Date of initial notice in Federal Register: March 19, 2002 (67 FR 12599).
The March 27, 2002, supplemental letter provided clarifying information that 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 June 27, 2002.
Date of application for amendment: February 21, 2002.
Brief description of amendment: The amendment authorizes changes to the Updated Final Safety Analysis Report (UFSAR) and the Technical Requirements Manual to eliminate the chlorine detection function from the control center heating, ventilation and air conditioning system. Changes to the UFSAR are subject to the requirements of 10 CFR 50.59; however, the changes were submitted to the Nuclear Regulatory Commission for review and approval since they involve the elimination of an automatic action.
Date of issuance: June 26, 2002.
Facility Operating License No. NPF-43: Amendment revises the UFSAR and TRM.
Date of initial notice in Federal Register: April 16, 2002 (67 FR 18643). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 26, 2002.
Date of application for amendment: May 24, 2001.
Brief description of amendment: The amendment deletes License Condition 2.C.(11), which required inspection of the low-pressure turbine discs during the second refueling outage and specified that the frequency of subsequent inspections should be in accordance with the turbine manufacturer's recommendations. License Condition 2.C.(11) is no longer applicable to Fermi 2.
Facility Operating License No. NPF-43: Amendment revises the License.
Date of initial notice in Federal Register: December 12, 2001 (66 FR 64288). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 26, 2002.
Date of application of amendments: June 21, 2000, as supplemented by letters dated April 30 and May 20, 2002.
Brief description of amendments: The amendments authorize changes to the Updated Final Safety Analysis Report Section 10.4.7, “Emergency Feedwater System.”
Date of Issuance: June 11, 2002.
Amendment Nos.: 325/325/326.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: Amendments authorized changes to the UFSAR.
Date of initial notice in Federal Register: July 26, 2000 (65 FR 46008). The supplement dated April 30 and May 20, 2002, provided clarifying information that did not change the scope of the June 21, 2000, application nor the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 11, 2002.
Date of application for amendment: April 16, 2001, as supplemented by letters dated November 8, 2001, and February 11, 2002.
Brief description of amendment: The amendment authorizes the licensee to modify the Final Safety Analysis Report (FSAR) to allow an unisolable drain line between the reactor core isolation cooling and the control rod drive/condensate pump rooms and identify the pump room doors and penetration seals that are not watertight. In addition, the change documents the minimum acceptable safe shutdown equipment.
Date of issuance: June 19, 2002.
Effective date: June 19, 2002, and shall be implemented in the next periodic update to the FSAR in accordance with 10 CFR 50.71(e).
Facility Operating License No. NPF-21: The amendment revises the FSAR.
Date of initial notice in Federal Register: May 16, 2001 (66 FR 27175). The November 8, 2001 and February 11, 2002, supplemental letters 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 amendment is contained in a Safety Evaluation dated June 19, 2002.
Date of application for amendment: April 11, 2002.
Brief description of amendment: The amendment revises Technical Specification Surveillance Requirement (SR) 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed Surveillance. The delay period is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours or up to the limit of the specified Start Printed Page 45575Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”
Effective date: June 27, 2002.
Date of initial notice in Federal Register: May 14, 2002 (67 FR 34485). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 27, 2002.
Brief description of amendment: The amendment revises Surveillance Requirement (SR) 3.0.3 to extend the delay period before entering a Limiting Condition for Operation, following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”
Date of issuance: June 10, 2002.
Effective date: As of the date of issuance and shall be implemented in conjunction with the implementation of Amendment No. 215.
Date of initial notice in Federal Register: April 30, 2002 (67 FR 21287). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 10, 2002.
Date of application for amendments: August 1, 2001.
Brief description of amendments: These amendments revise Limerick Generating Station's Units 1 and 2 Technical Specifications by deleting Section 6.4, “Training.”
Date of issuance: June 14, 2002.
Amendment Nos.: 160/122.
Date of initial notice in Federal Register: October 31, 2001 (66 FR 55018). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 14, 2002.
Date of application for amendment: April 18, 2002.
Brief description of amendment: The proposed amendment revises Surveillance Requirement (SR) 3.0.3 to extend the delay period, before entering a Limiting Condition for Operation, following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”
Date of initial notice in Federal Register: May 14, 2002 (67 FR 34487). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated June 26, 2002.
Brief description of amendment request: The amendment would replace referenced control requirements for access to high radiation areas with the actual requirements of 10 CFR Part 20, and would replace the existing Three Mile Island Nuclear Station, Unit 2, Technical Specifications (TS) Section 6.11 with the wording contained in Three Mile Island Nuclear Station, Unit 1, TS Section 6.12.
Date of initial notice in Federal Register: April 2, 2002 (67 FR 15623). The Commission's related evaluation of the amendment is contained in a safety evaluation dated June 27, 2002.
Date of application for amendment: June 18, 2001, as supplemented by letters dated January 30, and March 1, 2002.
Brief description of amendment: The amendment revises (1) the reference point for reactor vessel level instrumentation specifications to use instrument “zero” instead of “top of active fuel;” (2) simplifies the safety limits and limiting safety system settings to eliminate specifications that are unnecessary, outdated, or redundant to other Technical Specifications (TSs); (3) changes the reactor coolant system pressure safety limit from 1335 psig to 1332 psig to correct a minor calculation error; and (4) makes corresponding TS Bases changes.
Date of initial notice in Federal Register: July 25, 2001 (66 FR 38764). The supplements dated January 30 and March 1, 2002, 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 amendment is contained in a Safety Evaluation dated June 11, 2002.
Date of application for amendments: January 10, 2002.
Brief description of amendments: The amendments revise Surveillance Requirement (SR) 3.0.3 to extend the delay period before entering a Limiting Condition for Operation following a missed surveillance. The delay period is extended from the current limit of “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is less” to “* * * up to 24 hours or up to the limit of the specified Frequency, whichever is greater.” In addition, the following requirement is added to SR 3.0.3: “A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.”
Effective date: June 19, 2002, shall be implemented within 30 days from the date of issuance.
Amendment Nos.: Unit 1-153; Unit 2-153.
Facilit Operating License Nos. DPR-80 and DPR-82: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 5, 2002 (67 FR 10014). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 19, 2002.Start Printed Page 45576
Date of amendment request: May 22, 2002, as supplemented by letters dated June 10, and June 14, 2002.
Brief description of amendment: This amendment revises Technical Specification (TS) TS 5.5.2.11.f.1.h, “Steam Generator (SG) Tube Surveillance Program,” to more clearly delineate the scope of the SG tube inspection required in the tubesheet region. This TS change will apply only to Cycle 12 (Unit 2) and Cycle 11 (Unit 3) operations.
Date of issuance: June 17, 2002.
Effective date: June 17, 2002, to be implemented within 30 days from the date of issuance.
Amendment Nos.: Unit 2—189 ; Unit 3—180.
Public comments requested as to proposed no significant hazards consideration: Yes (67 FR 38150 dated May 31, 2002). The notice provided an opportunity to submit comments on the Commission's proposed no significant hazards consideration determination. No comments have been received. The notice also provided for an opportunity to request a hearing by July 1, 2002, but indicated that if the Commission makes a final no significant hazards consideration determination any such hearing would take place after issuance of the amendment. The Commission's related evaluation of the amendment, finding of exigent circumstances, consultation with the State of California and final determination of no significant hazards consideration are contained in a Safety Evaluation dated June 17, 2002. The June 10, and June 14, 2002, supplemental letters 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.
Brief description of amendments: The proposed change revised the Technical Specification (TS) 3.7.3, “Feedwater Isolation Valves (FIVs) and Associated Bypass Valves,” to adopt the NUREG-1431, “Standard Technical Specifications for Westinghouse Plants,” Revision 2 version of the specification. The requirements of revised TS 3.7.3 added, among other things, operability and suitable surveillance requirements for Feedwater Control Valves and Associated Bypass Valves and allowed for the extended out-of-service time for one or more FIVs. In addition, a footnote which allowed a one-time extension for Condition A Completion Time, has been deleted because it is no longer applicable.
Date of issuance: June 20, 2002.
Amendment Nos.: NPF-87, Amendment No. 97 and NPF-89, Amendment No. 97.
Date of initial notice in Federal Register: May 14, 2002 (67 FR 34492). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 20, 2002.
Dated at Rockville, Maryland, this 1st day of July 2002.
2. U.S. Nuclear Regulatory Commission (NRC) letter to Consolidated Edison, “Order to Authorize Decommissioning and Amendment No. 45 to License No. DPR-5 for Indian Point Unit 1 (TAC No. M59664),” dated January 31, 1996.
[FR Doc. 02-16956 Filed 7-8-02; 8:45 am]