Source: https://www.federalregister.gov/documents/2001/09/05/01-22137/biweekly-notice-applications-and-amendments-to-facility-operating-licenses-involving-no-significant
Timestamp: 2018-10-19 13:02:47
Document Index: 583911258

Matched Legal Cases: ['art 2', '§\u200950', 'art] 20', 'art] 20', 'art] 20', 'art] 20', 'art 50']

A Notice by the Nuclear Regulatory Commission on 09/05/2001
As of the date of issuance and shall be implemented 30 days from the date of issuance.
66 FR 46473
46473-46485 (13 pages)
01-22137
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North Carolina
Consolidated Edison Company of New York, Inc., Docket No. 50-003, Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York
Exelon Generation Company, LLC, Docket Nos. 50-237, Dresden Nuclear Power Station, Unit 2, Grundy County, Illinois
PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric Station, Unit 1, Luzerne County, Pennsylvania
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna Nuclear Power Plant, Wayne County, New York
Notice of Issuance of Amendments to Facility Operating Licensess
https://www.federalregister.gov/d/01-22137 https://www.federalregister.gov/d/01-22137
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section Start Printed Page 46474189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or proposed to be issued from August 13, 2001 through August 24, 2001. The last biweekly notice was published on August 22, 2001 (66 FR 44161).
By October 5, 2001, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). 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.
If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.Start Printed Page 46475
Date of amendments request: July 26, 2001.
Description of amendments request: The proposed amendment modifies Administrative Controls Technical Specifications (TSs) 5.5.14.b and 5.5.14.b.2 such that they are consistent with Title 10 of the Code of Federal Regulations (10 CFR), § 50.59.
The proposed change replaces the word “involve” with “require” and deletes reference to the term “unreviewed safety question” consistent with 10 CFR 50.59. Deletion of the term “unreviewed safety question” was approved by the Nuclear Regulatory Commission with the revision to 10 CFR 50.59. Consequently, the probability of an accident previously evaluated is not significantly increased. Changes to the TS Bases are still subject to 10 CFR 50.59. As a result, the consequences of any accident previously evaluated are not significantly affected.
The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing plant operation. These changes are considered administrative changes and do not modify, add, delete, or relocate any technical requirements in the TS.
The proposed changes will not reduce the margin of safety because they have no effect on any safety analyses assumptions. Changes to the TS Bases are still subject to 10 CFR 50.59, including prior Nuclear Regulatory Commission approval if the criteria in 10 CFR 50.59(c)(2) are met. The proposed changes to TS 5.5.14 are considered administrative in nature based on the revision to 10 CFR 50.59.
Therefore, this proposed modification does not significantly reduce the margin of safety.
NRC Section Chief: Peter Tam (Acting).
Date of amendments request: July 27, 2001.
Description of amendments request: The proposed amendment modifies the conditions and required actions for the control room emergency ventilation system (CREVS) of Technical Specification (TS) 3.7.8 for Calvert Cliffs Nuclear Power Plant, Units Nos. 1 & 2. Note 2 is being added to TS 3.7.8 to specify CREVS train operability requirements during the movement of irradiated fuel assemblies. Associated Limiting Conditions for Operation (LCO) Action Statements F and G are also being modified to be consistent with the addition of Note 2.
The proposed amendment also modifies the conditions and required actions for the control room emergency temperature system (CRETS) of TS 3.7.9. The existing note in TS 3.7.9 is being modified to specify CRETS train operability requirements during the movement of irradiated fuel assemblies. LCO Action Statements C and D are also being modified to be consistent with the addition of Note 2.
The proposed changes will modify the conditions and required actions for the Control Room Emergency Ventilation System (CREVS) and the Control Room Emergency Temperature System (CRETS) to reflect the licensing basis for movement of irradiated fuel assemblies. The CREVS and CRETS mitigate the consequences of an accident and do not initiate an accident. The CREVS provides protection to the control room operators in the event of a radioactive release. The CRETS provides protection to the Control Room by maintaining the temperature below the required limit. Therefore, changing the Conditions, Required Actions, and Completion Times for the CREVS and CRETS does not increase the probability of an accident.
As described in the Updated Final Safety Analysis Report (UFSAR), the CREVS and CRETS mitigate the consequences of six accidents. All but the fuel handling accident are postulated to occur during Modes 1, 2, 3, or 4. The fuel handling accident is only Start Printed Page 46476postulated to occur during the movement of irradiated fuel assemblies. The changes proposed would only alter the response to the loss of one CREVS or CRETS train during the movement of irradiated fuel assemblies. Since a single failure is not required to be postulated during the response to a fuel handling accident, having one CREVS or CRETS train out-of-service during fuel movement would not result in a change to the ability of the CREVS or CRETS to mitigate the consequences of a design basis fuel handling accident. The loss of one CREVS or CRETS train during Modes 1, 2, 3, or 4 is covered by other Conditions, and those Conditions have not been changed by this request. Therefore, the ability of the CREVS or CRETS to respond to any design basis accident would not be diminished by this proposed change.
Therefore, the proposed Technical Specification changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes do not involve a change in the operation of the plant and no new accident initiation mechanism is created by the proposed changes. The operations of the CREVS or CRETS are not altered by the proposed changes. The proposed changes do not change the licensing basis requirements for the CREVS or CRETS response to the accidents described in the UFSAR. No plant changes will be made as a result of this request. No conditions have been created by this request that might result in a new accident that has not been previously analyzed. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Would not involve a significant reduction in the margin of safety.
The margin of safety created by the response of the CREVS or CRETS to various accidents has not been reduced by the proposed changes in the Conditions, Required Actions, or Completion Times. These changes merely clarify the Technical Specification so that the licensing basis is more accurately reflected. The fuel handling accident does not assume a single failure occurs during the plant response to the event; therefore, the loss of a single CREVS or CRETS train does not place the plant outside of the licensing basis. This would be reflected in the proposed changes. The changes do not alter the operation or response requirements of the CREVS or CRETS. The CREVS and CRETS will continue to respond to accidents as designed. Operators will continue to be protected as described in the UFSAR. Therefore, the margin of safety is not significantly reduced by these proposed changes.
Description of amendments request: The proposed amendment will add additional references to Technical Specification (TS) 5.6.5.b for the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 & 2. The references will allow the use of ZIRLOTM clad fuel rods in the Calvert Cliffs' reactor cores.
The proposed change allows the use of methods required for the implementation of ZIRLOTM clad fuel rods in Calvert Cliffs Unit Nos. 1 and 2 and the use of current versions of the ECCS [emergency core cooling system] performance evaluation models for large and small break LOCAs [loss-of-coolant accidents]. The use of updated analysis methodologies will not increase the probability of an accident because the plant systems will not be operated outside of design limits, no different equipment will be operated, and system interfaces will not change.
With ZIRLOTM material introduced in the reactor, cores will exist in which ZIRLOTM and Zircaloy-4 clad fuel rods are co-resident. Fuel rods clad with each material will be evaluated based on the approved topical report.
The use of the three additional methodologies will not increase the consequences of an accident because Limiting Conditions for Operation (LCOs) will continue to restrict operation to within the regions that provide acceptable results, and Reactor Protective System (RPS) trip setpoints will restrict plant transients so that the consequences of accidents will be acceptable. Also, the consequences of the accidents will be calculated using NRC accepted methodologies.
The cores that will exist with ZIRLOTM and/or Zircaloy-4 clad fuel in the reactor will not increase the consequences of an accident. Operation within the LCOs and RPS setpoints will continue to restrict plant transients so that the consequences of accidents will be acceptable.
The proposed change does not add any new equipment, modify any interfaces with any existing equipment, alter the equipment's function or change the method of operating the equipment. The proposed change does not alter plant conditions in a manner that could affect other plant components. The proposed change does not cause any existing equipment to become an accident initiator. The ZIRLOTM clad fuel rod design does not introduce features that could initiate an accident. Therefore, the proposed change does not create the possibility of a new or different [kind] of accident from any accident previously evaluated.
Safety Limits ensure that Specified Acceptable Fuel Design Limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences. All fuel limits and design criteria shall be met based on the approved methodologies defined in the topical reports. The RPS in combination with the LCOs, will continue to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure and thermal power level that would result in a violation of the Safety Limits. Therefore, the proposed changes will not involve a significant reduction in the margin of safety.
The safety analyses determine the LCO settings and RPS setpoints that establish the initial conditions and trip setpoints, which ensure that the Design Basis Events (Postulated Accidents and Anticipated Operational Occurrences) analyzed in the Updated Final Safety Analysis Report produce acceptable results. Also all fuel limits and design criteria shall be satisfied. The Design Basis Events that are impacted by the implementation of ZIRLOTM cladding will be analyzed using the NRC accepted methodology described in CENPD-404-P.
The change in the fuel rod cladding material and the use of the current ECCS performance evaluation models will not involve a reduction in the margin of safety because acceptable results for the impacted Design Basis Events will be maintained.
Therefore, the margin of safety is not significantly reduced by this proposed change.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.Start Printed Page 46477
NRC Section Chief: Peter Tam, Acting.
Date of amendments request: August 1, 2001.
Description of amendments request: The proposed amendments would revise the Technical Specifications (TS) to support a full-scope application of an Alternative Source Term (AST). The AST analyses were performed following the guidance in Regulatory Guide 1.183, “Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors,” dated July 2000, and Standard Review Plan Section 15.0.1, “Radiological Consequences Analyses Using Alternative Source Terms.” Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
The BSEP systems affected by implementation of the Alternative Source Term analyses and the relaxations associated with TSTF [Technical Specification Task Force]-51, Revision 2, are not initiators of any design basis accidents. Therefore, because design bases accident initiators are not being altered by adoption of the Alternative Source Term analyses and the relaxations associated with TSTF-51, Revision 2, the probability of an accident previously evaluated is not affected. The Alternative Source Term does not affect the design or normal operation of the facility. Rather, once the occurrence of the accident has been postulated, the Alternative Source Term is an input used to evaluate the consequences of an accident. Implementation of the Alternative Source Term has been evaluated for the limiting design basis accidents at BSEP, and it has been demonstrated that the dose consequences of those limiting design bases accidents are within the regulatory guidance provided by the NRC in Regulatory Guide 1.183 and Standard Review Plan Section 15.0.1. For a fuel handling accident, the AST analyses demonstrate acceptable doses, within regulatory limits, without credit for secondary containment or automatic isolation of the Control Room. As such, the consequences of an accident previously evaluated are not affected.
Based on the above, the proposed license amendments do not involve an increase in the probability or consequences of an accident previously evaluated.
The BSEP systems affected by implementing the Alternative Source Term changes and the changes associated with TSTF-51, Revision 2, do not alter any design bases accident initiators or create new types of accident precursors. In addition, these changes do not affect the design function or mode of operation of systems, structures, or components in the facility such that new equipment failure modes are created. Therefore, the proposed license amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The changes proposed are associated with the implementation of a new licensing basis for BSEP. Approval of the change from the original source term, developed in accordance with TID-14844, to a new Alternative Source Term, as described in NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants, Final Report,” dated February 1, 1995, is being requested. The results of the accident analyses, revised in support of the proposed license amendments, are subject to revised acceptance criteria. These analyses have been performed using conservative methodologies, as specified in Regulatory Guide 1.183. Safety margins have been evaluated and margin has been retained to ensure that the analyses adequately bound the postulated limiting event scenarios. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67, “Alternative source term,” and Regulatory Guide 1.183.
The proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the Control Room and Emergency Operations Facility/Technical Support Center, are within corresponding regulatory limits. Specifically, the margin of safety for these accidents is considered to be that provided by meeting the applicable regulatory limits, which for three of five event scenarios (i.e., the control rod drop accident, the fuel handling accident, and one of the two limits for a main steam line break accident), is conservatively set below the 10 CFR 50.67 limit. With respect to the Control Room personnel doses, the margin of safety is the difference between the 10 CFR 50.67 limits and the regulatory limit defined by 10 CFR 50, Appendix A, General Design Criterion 19.
Since the proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the Control Room are within corresponding regulatory limits, the proposed license amendments do not involve a significant reduction in a margin of safety.
Description of amendment request: The proposed amendment would revise Technical Specification (TS) 6.12, “High Radiation Area,” to delete the administrative requirements for the control of access to high radiation areas. The control of access to these areas is assured by plant radiation protection programs that comply with 10 CFR 20.1601 requirements by using the alternate method in Regulatory Guide 8.38, “Control of Access to High and Very High Radiation Areas of Nuclear Power Plants,” June 1993.
The proposed TS change is administrative in nature. It involves deleting specific requirements for complying with a subparagraph of 10 CFR [part] 20 for the purpose of controlling access to high radiation areas. Accident evaluations do not consider the effects of methods of controlling access to high radiation areas. The 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 TS change is administrative in nature. It involves deleting specific Start Printed Page 46478requirements for complying with a subparagraph of 10 CFR [part] 20. However, effective compliance with 10 CFR [part] 20 is mandated by [* * *] another IP1 [Indian Point Nuclear Generating Station, Unit 1] TS provision. The effectiveness of Con Edison [Consolidated Edison Company of New York, Inc.] compliance with 10 CFR [part] 20 is not adversely affected by this change. In addition, this change does not affect any design function for or the operation of any plant structure, system, or component.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that 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.
Date of amendment request: August 14, 2001.
Description of amendment request: The proposed amendments would revise Technical Specification Surveillance Requirement 3.3.5.2 by changing the Engineered Safeguards Protective System Analog Instrument channel functional test frequency from 31 days to 92 days.
Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the determination that this amendment request involves a No Significant Hazards Consideration by applying the standards established by the NRC regulations in 10 CFR 50.92. This ensures that operation of the facility in accordance with the proposed amendment would not:
No. This is a proposed change to the Technical Specification (TS) 3.3.5 Engineered Safeguards Protective System (ESPS) Analog Instrumentation, Surveillance Requirement (SR) 3.3.5.2 for the channel functional test. The proposed change to TS 3.3.5 ESPS Analog Instrumentation, SR 3.3.5.2 will extend the current 31 day surveillance frequency to a 92 day surveillance frequency. The proposed change does not alter the method of operating or configuration for any Structure, System, or Component.
No. The ESPS Analog Instrumentation provides the necessary actuation of the Engineered Safety Features based on the Reactor Coolant and/or Reactor Building pressure. The proposed revision to the frequency for SR 3.3.5.2 will not alter the actuation of the Engineered Safety Features. The channel functional testing of the ESPS Analog Instrumentation will continue to be performed in an acceptable timeframe following the implementation of the proposed change.
No. The proposed revision to the frequency for SR 3.3.5.2 will not impact the operation of the ESPS Analog Instrumentation. In addition, the channel functional testing of the ESPS Analog Instrumentation will continue to be performed in an acceptable timeframe following the implementation of the proposed change.
Date of amendment request: August 23, 2001.
Description of amendment request: The proposed amendment would revise the technical specifications to eliminate the requirement to move control element assembly #43 for the remainder of Cycle 15.
One function of the CEAs [control element assemblies] is to provide a means of rapid negative reactivity addition into the core. This occurs upon receipt of a signal from the Reactor Protection System. This function will continue to be accomplished with the approval of the proposed change. Typically, once per 92 days each CEA is moved at least five inches to prove operability. Operability of a CEA requires the CEA be trippable and free from mechanical binding, i.e., moveable. CEA #43 is operable. However, due to abnormal coil voltage on two of the five coils that move CEA #43, if CEA #43 were moved to perform the SR [surveillance requirement], it is possible that a drop rod incident could occur. The misoperation of a CEA, which includes a drop rod incident, is an abnormal occurrence and has been evaluated as part of the ANO-2 [Arkansas Nuclear One, Unit 2] accident analysis. The proposed change would eliminate the requirement to move CEA #43 every 92 days and therefore eliminate the potential of CEA misoperation, associated down power, and challenge to the plant.
If a reactor trip signal were generated, CEA #43 has been demonstrated to be operable and will drop into the core along with the remaining CEAs to ensure reactor shutdown. No modifications are proposed to the Reactor Protection System or associated Control Element Drive Mechanism Control System logic. The accident mitigation features of the plant are not affected by the proposed amendment.
CEA #43 is operable, both moveable and trippable. The proposed change will not introduce any new design changes or systems. If a reactor trip were generated, CEA #43 will drop into the core along with the remaining CEAs to ensure reactor shutdown. The proposed change does not establish a potential for a new accident precursor.
CEA #43 will continue to have the same capability to mitigate an accident as it had prior to approval of the proposed TS [technical specification] change. CEA #43 is moveable and trippable.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Section Chief: Robert A. Gramm.Start Printed Page 46479
Date of amendment request: July 18, 2001.
Description of amendment request: The submittal requests a change to Technical Specifications (TS) Definitions 1.12 and 1.25, the effect of which will be to allow either an allocated or a measured response time to be utilized for the sensors in the Reactor Protective System and Engineered Safety Features Actuation System instrument loops.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendment to Technical Specifications (TS) Definitions 1.12 and 1.25 allows substitution of an allocated sensor response time in lieu of measuring sensor response time. Response time testing is not an initiator of any accident previously evaluated. Further, overall system response time will continue to meet Technical Specification requirements. The allocated sensor response times allowed in lieu of measurement have been determined to adequately represent the response time of the components such that the safety systems utilizing those components will continue to perform their accident mitigation function as assumed in the safety analysis.
Therefore, this change does not involve a significant increase in the probability of consequences of any accident previously evaluated.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment to TS Definitions 1.12 and 1.25 allows the substitution of an allocated sensor response time in lieu of sensor response time testing for selected components. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The use of allocated response times in lieu of measured response times result in no physical change to the plant.
The proposed amendment does not involve a significant reduction in a margin of safety. The proposed amendment to TS 1.1, Definitions, allows the substitution of an allocated sensor response time in lieu of measured sensor response time for certain pressure sensors. The allocated pressure sensor response times allowed in lieu of measurement have been determined to adequately represent the response time of the components such that the safety systems utilizing those components will continue to perform their accident mitigation function as assumed in the safety analysis.
Attorney for licensee: N. S. Reynolds, Esquire, Winston and Strawn 1400 L Street NW., Washington, DC 20005-3502.
Date of amendment request: June 6, 2001.
Description of amendment request: The proposed amendment would revise the values of the Safety Limit for the Minimum Critical Power Ratio in Technical Specification 2.1.1.
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 Dresden Nuclear Power Station (DNPS), Unit 2, Cycle 18 such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.
Changing the SLMCPR does not increase the probability of an evaluated accident. The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operations. Therefore, no individual precursors of an accident are affected.
The proposed change revises the SLMCPR to protect the fuel during normal operation as well as during any transients or anticipated operational occurrences. 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 percent of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences) is met. Since the operability of plant systems designed to mitigate any consequences of accidents has not changed, the consequences of an accident previously evaluated are not expected to increase.
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 DNPS, Unit 2, Cycle 18.
The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0.1 percent of the rods are expected to be in boiling transition if the [minimum critical power ratio] 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 percent of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences) are met.
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 Start Printed Page 46480satisfied. Therefore, the NRC staff proposes to determine that the requested amendment involve no significant hazards consideration.
Description of amendment request: The proposed amendment would change the Technical Specification (TS) 5.5.7.a, b, and c, to update the Ventilation Filter Testing Program at Cooper Nuclear Station.
The District has evaluated each of the proposed TS changes in accordance with the criteria set forth in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.
The determination that the proposed changes do not involve a significant hazards consideration is based on an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the evaluation are presented below.
The application of the 1989 version of ASME N510 will not change any of the surveillance requirements for operability of the SGT or the CREF. The changes with respect to RG 1.52 are editorial in nature and will not result in any changes in surveillance requirements. Since SGT and CREF are ESF systems and not accident initiators the probability of an accident evaluated in the Updated Safety Analysis Report will not be increased. As such, the probability of occurrence for a previously analyzed accident is not significantly increased.
The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. This change does not affect the performance of any credited equipment. These details of testing are not analysis assumptions. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.
The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters within which the plant is operated. There are no setpoints, at which protective or mitigative actions are initiated, affected by this change. This change will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. The change does not result in alteration of the procedures which ensure the plant remains within analyzed limits, and no change is being made to the procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced. The change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. Sufficient equipment remains available to actuate upon demand for the purpose of mitigating an analyzed event. The proposed change, which replaces references to ASME N510-1980 with references to ASME N510-1989, is acceptable because the tests continue to require appropriated confirmation of the assumed function of the systems (and thereby assure continued operability), and more accurately presents acceptable testing conditions. The changes with respect to RG 1.52 are editorial in nature and do not change existing surveillances. There is no detrimental impact on any equipment design parameter, and the plant will still be required to operate within prescribed limits. Therefore, the change does not involve a significant reduction in the margin of safety.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power District, Post Office Box 499, Columbus, NE 68602-0499
Date of amendment request: May 31, 2001.
Description of amendment request: The proposed amendment would revise Technical Specification 2.1.1.2 to reflect the Unit 1 Cycle 13 ( U1C13) minimum critical power ratio (MCPR) safety limits.
No. The proposed change to the MCPR Safety Limit does not directly or indirectly affect any plant system, equipment, component, or change the way in which the plant is operated. Thus, this proposed amendment does not involve a significant increase in the probability of occurrence of an accident previously evaluated.
Prior to the startup of U1C13, licensing analyses are performed (using NRC approved methodology referenced in Technical Specification Section 5.6.5.b) to determine changes in the critical power ratio as a result of anticipated operational occurrences. These results are added to the MCPR Safety Limit values proposed herein to generate the MCPR operating limits in the U1C13 COLR [Core Operating Limits Report]. These limits could be different from those specified for the U1C12 COLR. The COLR operating limits thus assure that the MCPR Safety Limit will not be exceeded during normal operation or anticipated operational occurrences.
Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or different kind of accident from any previously analyzed?
No. The change to the MCPR Safety Limits and the U1C13 core loading which it supports does not directly or indirectly affect any plant system, equipment, or component (other than the core itself) and therefore does not affect the failure modes of any of these. Thus, the proposed changes do not create the possibility of a previously unevaluated operator error or a new single failure.
Therefore, this proposed amendment does not involve a possibility of a new or different kind of accident from any previously analyzed.
No. Since the proposed changes do not affect any plant system, equipment, or component, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. The proposed MCPR Safety Limits do not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections, because the MCPR Safety Limits calculated for U1C13 preserve the required margin of safety.Start Printed Page 46481
Therefore these changes do not involve a significant reduction in [a] margin of safety.
Date of amendment request: June 1, 2001, as supplemented June 13, 2001.
Description of amendment request: The proposed amendment would revise Technical Specification (TS) Section 3.7.1, Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat Sink (UHS), to address previously unidentified single failure vulnerabilities when one or more RHRSW subsystems are inoperable.
Implementation of the subject changes reduces the probability of occurrence and the probability of adverse consequences of accidents previously evaluated. Inclusion of the large array valves and the bypass valves to the Technical Specifications (TS) recognizes their importance to safe shutdown. The administrative controls that TS's invoke increases the probability that potential inoperability of these valves is controlled and managed in a manner commensurate with their risk significance.
Reducing the completion time for RHRSW subsystem inoperable conditions recognizes their importance to safe shutdown commensurate with their risk significance.
These changes do not affect the design or operation of the affected components/systems and serves to increase the level of administrative control for the UHS and RHRSW system that will help to ensure the ability to achieve safe shutdown.
The subject changes apply Technical Specification administrative controls to the UHS bypass and large array valves and shortens the completion times applicable to RHRSW inoperable conditions. The design and operation of the affected components and systems is not affected.
Application of these administrative controls does not involve a possibility of a new or different kind of accident from any previously analyzed.
Implementation of the subject changes increases the margin of safety since these changes add Technical Specification controls to components not currently addressed in the Technical Specifications and reduces the completion times for subsystems currently addressed in the Technical Specifications. These changes better account for the affected components/systems impact on safe shutdown.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, Inc., 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.
Date of amendment request: May 3, 2001.
Description of amendment request: The licensee proposed to amend Ginna Station Improved Technical Specifications (ITS) to reflect the design changes to the actuation circuitry associated with the Control Room Emergency Air Treatment System (CREATS). The proposed design changes consist of replacing the current diverse radiation monitors with two Gieger-Mueller (GM) tubes powered from two separate safety-related power supplies which are configured into two redundant actuation logic trains, including manual initiation. The design changes is intended to increase system reliability by providing redundancy and reducing spurious actuations. As a result of the proposed design changes, the licensee requested that the following changes be made to the Limiting Condition for Operation (LCO) 3.3.6 for the CREATS Actuation Instrumentation:
a. Add a new Condition to require immediately placing the CREATS in the emergency mode of operation upon the loss of two instrument channels/trains.
b. Add a new surveillance requirement involving a CHANNEL CHECK of the Control Room Radiation Intake Monitors.
c. Revise Table 3.3.6-1 to increase the number of trains of Manual Initiation, and Automatic Actuation Logic and Actuation Relays, from one train to two trains.
d. Extend the Completion Time of the Required Action for a loss of one channel/train from 1 hour to 7 days as the result of installing redundant channels/trains.
e. Revise Table 3.3.6-1 to remove reference to the Iodine, Noble Gas, and Particulate Control Room Radiation Intake Monitors. These monitors will be replaced by the two new GM tubes.
f. Revise Table 3.3.6-1 to replace the column heading “Trip Setpoint” with “Allowable Value.”
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff reviewed the licensee's analysis against the three standards of 10 CFR 50.92(c). The NRC staff's analysis is presented below:
The first standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed ITS changes listed above will not increase the probability of an accident previously evaluated because the CREATS actuation system is not an accident initiator as this system performs only mitigative functions. In particular, the system is designed to provide a protective environment from which the operators can control the plant following an uncontrolled release of radioactivity during a design-basis accident. The proposed design changes (increase system redundancy and reliability) and the ITS changes associated with LCO 3.3.6 (i.e., action statements for loss of instrument channels/trains, channel check requirements, etc.,) will only ensure that the CREATS will continue to perform its safety functions and that the consequences of an accident previously evaluated will not increase.
The second standard requires that operation of the unit in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed ITS changes listed above will not create the possibility of a new or different kind of Start Printed Page 46482accident from any accident previously evaluated because the CREATS actuation system is not an accident initiator as this system performs only mitigative functions. The proposed change creates no new functional interactions with existing plant equipment nor does it introduce any new failure mode or mechanisms which could lead to reactor core damage or fission product release.
The third standard requires that operation of the unit in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. The proposed ITS changes will not adversely affect the performance of the CREATS, nor will they affect the ability of the system to perform their intended functions. The reason being that the proposed amendment does not involve any new acceptance criteria, analytical limits, or evaluation models which could affect operator dose limits.
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 L Street, NW., Washington, DC 20005
NRC Section Chief: P. Tam, Acting.
Date of amendment request: April 25, 2001 as supplemented by letter dated July 31, 2001.
Brief description of amendments: The proposed license amendments would change the Technical Specifications (TS) to allow a one-time only change to TS 3.8.1, “AC Sources—Operating,” Action A.3, by extending the required Completion Time for restoration of an inoperable offsite circuit from 72 hours to 21 days.
The proposed one time Technical Specification Completion Time (CT) extension does not significantly increase the probability of occurrence of a previously evaluated accident because the startup transformer [ST] XST2 is not an initiator of previously evaluated accidents involving a loss of offsite power. The proposed changes to the Technical Specification CT do not affect any of the assumptions used in the deterministic or the Probabilistic Safety Assessment (PSA) analysis relative to loss of offsite power initiating event frequency.
The proposed one time Technical Specification CT extension will continue to provide assurance that the sources of power to 6.9 kV AC [kilovolt alternating current] buses perform their function when called upon. Extending the Technical Specification CT to 21 days does not affect the design of XST2, the operational characteristics of XST2, the interfaces between XST2 and other plant systems, the function, or the reliability of XST2. Thus, 6.9 kV AC components will be capable of performing either accident mitigation function and there is no impact to the radiological consequences of any accident analysis.
To fully evaluate the effect of the proposed change, PSA methods and deterministic analysis were utilized. The results of this analysis show no significant increase in the Core Damage Frequency.
The Maintenance Rule (a)(4) risk management program assesses risk based on plant status. It requires the consideration of other measures to mitigate consequences of an accident occurring while a ST is unavailable.
The proposed changes do not alter the operation of any plant equipment assumed to function in response to an analyzed event or otherwise increase its failure probability. Therefore, these changes do not involve a significant increase in the probability or consequences of any accident previously evaluated.
These proposed changes do not change the design, configuration, or method of operation of the plant. The proposed activity involves a change to the allowed plant mode for the performance of preventive maintenance that will ensure the inherent reliability of the XST2 Startup Transformer is maintained. No physical or operational change to the ST or supporting systems are made by this activity. Since the proposed change does not involve a change to the plant design or operation, no new system interactions are created by this change. The proposed Technical Specification change does not produce any parameters or conditions that could contribute to the initiation of accidents different from those already evaluated in the Final Safety Analysis Report.
The proposed change only addresses the time allowed to restore the operability of XST2. Thus the proposed Technical Specification change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change does not affect the Limiting Conditions for Operation or their Bases that are used in the deterministic analysis to establish any margin of safety. PSA evaluations were used to evaluate the proposed change, and these evaluations determined that the net changes are either risk neutral or risk beneficial. The proposed activity involves a one time change to Allowed Outage Times.
The proposed change does not involve a change to the plant design or operation and thus does not affect the design of the ST, the operation characteristics of the ST, the interfaces between the ST and other plant systems, or the function or reliability of the ST. Because ST performance and reliability will continue to be ensured by the proposed one time Technical Specification change, the proposed changes do not result in a reduction in the margin of safety.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036. NRC Section Chief: Robert A. Gramm.
Date of amendment request: August 7, 2001 (ET 01-0021).
Description of amendment request: The amendment would add the following to the Wolf Creek Generating Station (WCGS) Technical Specifications (TSs): (1) The phrase, “or if open, capable of being closed” to Limiting Condition for Operation (LCO) 3.9.4 for the equipment hatch, during core alterations or movement of irradiated fuel assemblies inside containment, and (2) the requirement to verify the capability to install the equipment hatch in a new Surveillance Requirement (SR) 3.9.4.2. Nothing is proposed to be deleted from the TSs. Existing SR 3.9.4.2 would be renumbered SR 3.9.4.3, but would not otherwise be changed. Item (1) will allow the equipment hatch to be open during the conditions stated above.
1. The proposed change does not involve a significant increase in the probability or Start Printed Page 46483consequences of an accident previously evaluated.
The proposed changes will allow the equipment hatch to be open during CORE ALTERATIONS and movement of irradiated fuel assemblies inside containment. The status of the equipment hatch during refueling operations has no affect on the probability of the occurrence of any accident previously evaluated. The proposed revision does not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. Since the consequences of a fuel handling accident inside containment with an open equipment hatch are bounded by the current analysis described in the USAR [Updated Safety Analysis Report for WCGS] and the probability of an accident is not affected by the status of the equipment hatch, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes do not create any new failure modes for any system or component, nor do they adversely affect plant operation. No new equipment will be added and no new limiting single failures will be created. The plant will continue to be operated within the envelope of the existing safety analysis.
Therefore, the proposed changes do not create a new or different kind of accident from any accident previously evaluated.
The previously determined radiological dose consequences for a fuel handling accident inside containment with the air lock doors open remain bounding for the proposed changes. These previously determined dose consequences were determined to be well within the limits of 10 CFR 100 and they meet the acceptance criteria of SRP [Standard Review Plan, NUREG-0800] section 15.7.4 and GDC [General Design Criteria of Appendix A to 10 CFR Part 50] 19.
Date of amendment request: January 24, 2001, as supplemented by letters dated July 18 and August 3, 2001.
Brief description of amendment: The request consists of a change to Technical Specification 3.6.1.3, “Primary Containment Isolation Valves (PCIVs),” to permit the operation of the Inclined Fuel Transfer System (IFTS) bottom valve after removal of the IFTS primary containment isolation blind flange while the containment is required to be operable.
Date of issuance: August 16, 2001.
Amendment No.: 117.
Date of initial notice in Federal Register: March 21, 2001 (66 FR 15921). The supplemental letters dated July 18 and August 3, 2001, provided additional information that did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (the Commission) staff's proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 16, 2001.
Date of amendment request: January 27, 2000, as supplemented by letters dated March 1, June 12, and July 26, 2001.
Brief description of amendments: The amendments allow the qualified condensate storage tank to be used for both units and defines new minimum volume requirements for the tank depending on whether Arkansas Nuclear One, Unit 1, Arkansas Nuclear One, Unit 2, or both units are aligned to the tank. The total volume requirements, the allowable alternative alignment for ANO-2, and other aspects of the Technical Specifications (TSs) are unaffected by the change.
Amendment Nos.: 214, 232.
Facility Operating License Nos. DPR-51 and NPF-6: Amendments revised the TSs.
Date of initial notice in Federal Register: May 30, 2001 (66 FR 29352).
The supplemental letters dated June 12 and July 26, 2001, provided additional information and revised TSs that did not expand the scope of the Start Printed Page 46484application or change the initial proposed no significant hazards consideration determination which addressed the original application and supplement dated March 1, 2001.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 16, 2001.
Date of amendment request: August 10, 2001.
Brief description of amendment: The proposed amendment will provide an alternative method for complying with the Limiting Conditions for Operation (LCO) requirements of Technical Specification 3.3.4.1, “End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation,” and require that an additional REQUIRED ACTION be added to CONDITION B as REQUIRED ACTION B.2.
Date of issuance: August 10, 2001.
Effective date: August 10, 2001.
Amendment No.: 148.
Facility Operating License No. NPF-29: Amendment revises the TS.
The Commission's related evaluation of the amendment, finding of emergency circumstances, and final determination of no significant hazards consideration, are contained in a Safety Evaluation dated August 10, 2001.
Date of application for amendments: September 1, 2000.
Brief description of amendments: Add a Technical Specification (TS) section regarding mechanical vacuum pump trip instrumentation.
Amendment Nos.: 186 and 181.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 21, 2001.
Date of application for amendments: April 16, 2001.
Brief description of amendments: The amendments change the reference in Technical Specification 5.5.6, “Inservice Inspection Program for Post Tensioning Tendons,” from Regulatory Guide 1.35, “Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments,” Revision 3, 1989, to a reference to Subsection IWL, “Requirements of Class CC Concrete Components of Light-Water Cooled Power Plants,” of Section XI, “Inservice Inspection,” of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, and delete the applicability of Surveillance Requirement (SR) 3.0.2 to TS Section 5.5.6. SR 3.0.2 allows the surveillance to be performed within 1.25 times the interval specified in the surveillance's frequency.
Amendment Nos.: 148 and 134.
Date of initial notice in Federal Register: June 12, 2001 (66 FR 31707).
Date of application for amendment: June 13, 2001.
Brief description of amendment: The amendment revises TS 5.3 to permit lead-test-assemblies to be used, regardless of clad material, as long as the Nuclear Regulatory Commission has generically approved the fuel assembly design for use in pressurized water reactors.
Date of issuance: August 13, 2001.
Date of initial notice in Federal Register: July 11, 2001 (66 FR 36342).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 13, 2001.
Date of amendment request: December 20, 2000, as supplemented by letters dated February 1 and 28, and June 12, 2001.
Brief description of amendments: The amendments revised the technical specification (TS) requirements and authorized revision of the Technical Requirements Manual provisions applicable when actions direct suspension of operations involving positive reactivity changes. The proposed changes remove the requirement not to make positive reactivity changes during certain plant conditions, and limit the reactivity changes that are allowed to those that will continue to assure appropriate reactivity limits are met. Related changes to the Bases were also made. In addition, an administrative TS change was made to remove a footnote regarding an alternate onsite emergency power source, which is no longer applicable.
Effective date: August 13, 2001.
Amendment Nos.: Unit 1-128; Unit 2-117.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Technical Specifications and authorized revision of the Technical Requirements Manual.
Date of initial notice in Federal Register: February 7, 2001 (66 FR 9387).
The February 1 and 28, and June 12, 2001, supplemental letters provided clarifying information that was within the scope of the original Federal Register notice and did not change the staff's initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 13, 2001.
Date of amendment request: December 20, 2000.Start Printed Page 46485
Brief description of amendments: The amendments delete License Condition 2.G, “Reporting to the Commission,” and Technical Specification 6.6.1.a, “Reportable Event Action.”
Effective date: The amendments are effective as of the date of their issuance.
Amendment Nos.: Unit 1—129; Unit 2—118.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments revised the Facility Operating Licenses and the Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR 31715).
Date of amendment request: February 28, 2001.
Brief description of amendments: The amendments revised the Technical Specifications (TS) to eliminate periodic response time testing requirements on selected sensors and selected protection channels, and modified TS Section 1.0 Definitions for “ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME” and “REACTOR TRIP SYSTEM (RTS) RESPONSE TIME” to provide for verification of response time for selected components. The associated Bases were also revised.
Date of issuance: August 21, 2001.
Effective date: The amendments are effective as of the date of their issuance and shall be implemented within 30 days from the date of issuance.
Amendment Nos.: Unit 1—130; Unit 2—119.
Date of initial notice in Federal Register: June 12, 2001 (66 FR 31716).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 21, 2001.
Dated at Rockville, Maryland, this 28th of August 2001.
[FR Doc. 01-22137 Filed 9-4-01; 8:45 am]