Source: https://www.federalregister.gov/documents/2001/09/19/01-23209/biweekly-notice-applications-and-amendments-to-facility-operating-licenses-involving-no-significant
Timestamp: 2017-08-17 04:39:18
Document Index: 628041476

Matched Legal Cases: ['art 2', 'art 100', 'art 100', 'art 100', 'art 100', 'art 50', 'art 100', 'art 50', 'art 100', 'art 21', 'art 50', 'art 50', 'art 20']

48283-48295 (13 pages)
Tennessee Valley Authority (TVA), Docket Nos. 50-260 and 50-296, Browns Ferry Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Docket No. 72-8, Calvert Cliffs Independent Spent Fuel Storage Installation, Calvert County, Maryland
Exelon Generation Company, PSEG Nuclear LLC, and Atlantic City Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, York County, Pennsylvania
https://www.federalregister.gov/d/01-23209 https://www.federalregister.gov/d/01-23209
This biweekly notice includes all notices of amendments issued, or proposed to be issued from August 27, 2001 through September 7, 2001. The last biweekly notice was published on September 5, 2001 (66 FR 46473).
The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Start Printed Page 48284involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.
By October 19, 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.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be Start Printed Page 48285granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Description of amendment request: As a follow-up response to a commitment identified in the Nuclear Regulatory Commission (NRC) staff letter dated December 22, 2000, “Completion of Licensing Action for Generic Letter (GL) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions,” Entergy Operations Inc., (Entergy, the licensee) has proposed to revise their Waterford Steam Electric Station, Unit 3 (Waterford 3) Final Safety Analysis Report (FSAR) to resolve the ten containment penetrations susceptible to thermally induces overpressurization through an evaluation, detailed analysis, or installation of physical modifications prior to startup from the spring 2002 refueling outage. Entergy determined a change to Waterford 3's license basis, through procedural controls, risk analysis, and engineering analysis, for seven penetrations, as discussed in this license basis change request. Permanent resolution to the GL 96-06 issues for the remaining three penetrations could be satisfied through the installation of physical modifications.
The proposed FSAR change reflects the use of administrative procedural controls to ensure these seven containment penetrations (two 4-inch diameter Steam Generator Blowdown penetrations and five-1/2 inch diameter Process Sampling penetrations) contain fluid at temperatures representative of Reactor Coolant, and the very low probability for overpressurization failure of containment penetrations during Mode 4 plant operation as a permanent solution to the GL 96-06 issue. The engineering analysis determined these seven containment penetrations met the acceptance criteria for allowed stresses contained in ASME [American Society of Mechanical Engineers] Section III Code, [Boiler and Pressure Vessel Code] Appendix F 1995. The result of the risk analysis is such that the very small change in LERF (Large Early Release Frequency], on the order of 1×10−9 per reactor year, remained well below the 1×10−7 ΔLERF guideline for a small change given in Regulatory Guide 1.174. The negligible reduction in LERF that would be achieved by adding thermal relief valve overpressure protection is not risk significant and is too small to justify the addition of the relief valves.
With respect to the probability or the consequences of an accident previously evaluated in the FSAR, the proposed deviation to the existing ASME Section III Code, Class 2 design provisions and operating requirements for the seven containment penetrations would not significantly increase the probability of an accident since the administrative procedural controls are being provided to: (1) minimize penetration heat-up and over-pressurization during a small window of vulnerability, approximately 1% per year of Mode 4 plant operation; and (2) minimize process fluid cooldown during normal plant operation by closing the containment isolation valves for the five sample penetrations when process fluid samples are obtained and the laboratory sample valves downstream of the CIV [containment isolation valves] are closed or flow through the penetration is stopped. Also the results of engineering analyses showed that the containment penetrations may exceed ASME Section III, Subsection NC 3500 Code required yield stresses and experience plastic deformation, but would not catastrophically fail; therefore, the penetrations would retain their ability to perform their safety function and maintain containment integrity.
On this basis, the proposed changes are not considered to constitute a significant increase in the probability or consequences of an accident due to:
Administrative controls to minimize penetration heat-up and over-pressurization during the small window of vulnerability
The seven containment penetrations retaining their ability to perform their safety function and maintaining containment integrity in accordance with engineering analyses performed that met acceptance criteria for allowed stresses contained in ASME Section III Code, Appendix F 1995, and
The low risk significance of overpressurization failure of the seven containment penetrations during a DBA [Design Basis Accident] while the plant is in Mode 4.
The proposed changes will not significantly affect the results of any accident previously evaluated. The accident mitigation features of the plant are not significantly affected by these proposed changes. The proposed changes do not add or modify any existing equipment.
The change proposes a deviation to the existing ASME, Section III, Class 2 license basis requirements for portions of the Steam Generator Blowdown System, Primary Sampling System, and Secondary Sampling System that penetrate the containment as a permanent solution to the GL 96-06 issues. This change involves recognition of the acceptability of administrative procedural controls to minimize penetration heat-up and over-pressurization during the small window of vulnerability, approximately 1% per year for Mode 4 plant operation. Added assurance is provided through the engineering analysis performed on these penetrations that determined allowable stresses did not exceed the ASME Section III Code, Appendix F 1995 pipe stress values. Therefore, the change would not contribute to the possibility of, or be the initiator for any new or different kind of accident.
The proposed change does not alter the configuration of the plant. There has been no physical change to plant systems, structures, or components.
The proposed change does not involve a significant reduction in margin of safety. The existing licensing basis for Waterford 3, with respect to the ASME Section III, Subsection NC-3621.2 provisions for portions of the Steam Generator Blowdown System, Primary Sampling System, and Secondary Sampling System that penetrate the containment, is to ensure piping that has the potential to experience pressurization due to trapped fluid expansion shall be designed to withstand the increased pressure or have provisions for relieving the excess pressure piping. With the acceptance of this proposed deviation to the license basis, it will be recognized that the seven containment penetrations have administrative procedural controls to minimize penetration heat-up and over-pressurization during the small window of vulnerability, approximately 1% per year for Mode 4 plant operation. Added assurance is also provided through the engineering analysis performed on these penetrations that Start Printed Page 48286determined stresses did not exceed the ASME Section III Code, Appendix F 1995 pipe stress values and predicted the penetration piping would experience plastic deformation, but would not catastrophically fail. Therefore, the penetrations would retain their ability to perform their safety function and maintain containment integrity. This deviation to license basis requirements for these seven containment penetrations is not considered to constitute a significant decrease in the margin of safety.
Description of amendment request: The proposed amendments delete requirements from the Technical Specifications (TSs) 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. 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 NRC staff issued a notice of opportunity for comment in the Federal Register on August 11, 2000 (65 FR 49271) 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 October 31, 2000 (65 FR 65018). The licensee affirmed the applicability of the following NSHC determination in its application dated August 13, 2001.
The PASS was originally designed to perform many sampling and analysis functions. These functions were designed and intended to be used in post-accident situations and were put into place as a result of the TMI-2 accident. The specific intent of the PASS was to provide a system that has the capability to obtain and analyze samples of plant fluids containing potentially high levels of radioactivity, without exceeding plant personnel radiation exposure limits. Analytical results of these samples would be used largely for verification purposes in aiding the plant staff in assessing the extent of core damage and subsequent offsite radiological dose projections. The system was not intended to and does not serve a function for preventing accidents and its elimination would not affect the probability of accidents previously evaluated.
The elimination of PASS related requirements will not result in any failure mode not previously analyzed. The PASS was intended to allow for verification of the extent of reactor core damage and also to provide an input to offsite dose projection calculations. The PASS is not considered an accident precursor, nor does its existence or elimination have any adverse impact on the pre-accident state of the reactor core or post-accident confinement of radionuclides within the containment building.
NRC Section Chief: Timothy G. Colburn, Acting. Start Printed Page 48287
Description of amendment request: Florida Power and Light Company (FPL) requests to amend Facility Operating Licenses DPR-67 for St. Lucie Unit I and NPF-16 for St. Lucie Unit 2 by revising Technical Specifications (TS) relating to positive reactivity additions while in shutdown modes. The proposed changes clarify TS involving positive reactivity additions to the shutdown reactor, and would allow small, controlled, safe insertions of positive reactivity while in shutdown modes. The proposed changes conform closely to an NRC approved generic change for Standard Technical Specifications, known as TSTF-286 Rev. 2, which revises most actions requiring “Suspend operations involving positive reactivity additions” to allow minimum reactivity additions due to temperature fluctuations or operations, which are necessary to maintain fluid inventory within the required shutdown margin or refueling boron concentration, as applicable.
The proposed TS changes revise actions that either require suspension of operations involving positive reactivity additions or preclude reduction in boron concentration less than the reactor coolant system (RCS). Reactivity excursions are analyzed events. The proposed changes limit positive reactivity additions into the RCS such that the required shutdown margin (SDM) or refueling boron concentration continue to be met. Reactivity changes performed during shutdown modes are currently governed by strict administrative controls. Although the proposed changes will allow procedural flexibility with regards to RCS temperature and boron concentration, these operations will still be under administrative control. The changes proposed by these amendments are within the scope and assumptions of the existing analyses. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed TS revisions relate to positive reactivity additions while in shutdown modes of operation. Reactivity excursions are analyzed events. The operational flexibility allowed in these proposed license amendments will be performed under strict administrative controls in order to limit the potential for excess positive reactivity addition. Although the existing procedural controls will need modification, no new or different operational failure modes would be introduced by these changes.
Additionally, implementation of these proposed changes do not require any physical plant modifications, so no new or different hardware related failure modes are introduced. The changes proposed by these amendments are within the scope and assumptions of the existing analyses. Therefore, operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes conform closely to the industry and NRC approved TSTF-286, Rev. 2 and relate to small, controlled, safe insertions of positive reactivity additions while in shutdown modes. These changes revise actions that either require suspension of operations involving positive reactivity additions, or prohibit RCS boron concentration reduction. The proposed changes provide operational flexibility while controlling positive reactivity additions in order to preserve the required SDM or refueling boron concentration. The proposed changes to provide for continued safe reactor operations, while also limiting any potential for excess positive reactivity addition. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.
Date of amendment request: June 22, 2001, as supplemented August 24, 2001.
Description of amendment request: The proposed amendment would revise the St. Lucie Unit 2 Technical Specification (TS) 3.9.4, Containment Penetrations. TS 3.9.4.a. requires that the containment equipment door be closed during core alterations or movement of irradiated fuel within containment. TS 3.9.4.b. requires a minimum of one door in each airlock to be closed during core alterations or movement of irradiated fuel within containment. The proposed change to TS 3.9.4.a. would allow the containment equipment door to be open during core alterations and movement of irradiated fuel in containment provided: (a) The equipment door is capable of being closed with four bolts within 30 minutes, (b) the plant is in MODE 6 with at least 23 feet of water above the reactor pressure vessel flange, and (c) a designated crew is available at the equipment door to close the door. The capability to close the containment equipment door includes the requirements that the door is capable of being closed and that any cables or hoses across the equipment door have quick-disconnects to ensure the door is capable of being closed in a timely manner. The proposed change to TS 3.9.4.b would allow both doors of each containment airlock to be open during core alterations and movement of irradiated fuel in containment provided: (a) At least one door of each open containment airlock is capable of being closed, (b) the plant is in MODE 6 with at least 23 feet of water above the reactor pressure vessel flange, and (c) a designated individual is available outside each open containment airlock to close the door. The capability to close the containment airlock door includes the requirement that the door is capable of being closed and that any cables or hoses across the airlock door have quick-disconnects to ensure the door is capable of being closed in a timely manner.
The proposed change to TS 3.9.4 would allow the containment equipment door and both doors of each containment airlock to be open during fuel movement or core alterations. Currently, the equipment door is closed with four (4) bolts and a single door on each containment airlock is closed during fuel movement or core alterations to prevent the escape of radioactive material in the Start Printed Page 48288event of an in-containment fuel handling accident. Neither the containment equipment door nor either of the containment airlock doors is an initiator of an accident. Whether the containment equipment door or both doors of the containment air locks are open or closed during fuel movement and core alterations has no affect on the probability of any accident previously evaluated. Allowing the containment equipment door and the containment airlock doors to be open during fuel movement or core alterations does not significantly increase the consequences from a fuel handling accident. The calculated offsite doses are well within the limits of 10 CFR part 100. In addition, the calculated doses are larger than the expected doses because the calculation does not incorporate the closing of the containment equipment door or the containment airlock doors after the containment is evacuated, which would be much less than the two hours assumed in the analysis. The proposed change would significantly reduce the dose to workers in containment in the event of a fuel handling accident by reducing the time required to evacuate the containment. The changes being proposed do not affect assumptions contained in other plant safety analyses or the physical design of the plant, nor do they affect other Technical Specifications that preserve safety analysis assumptions. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously analyzed.
The proposed change to Technical Specification 3.9.4, “Containment Building Penetrations,” affects a previously evaluated fuel handling accident. The new Fuel Handling Accident Analysis assumes that all of the iodine and noble gases that become airborne escape and reach the exclusion boundary and low population zone with no credit taken for filtration, the containment building barrier or for decay or deposition. Since the proposed change does not involve the addition or modification of equipment nor does it alter the design of plant systems and the revised analysis is consistent with the Fuel Handling Accident Analysis, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The margin of safety as defined by 10 CFR part 100 has not been significantly reduced. The calculated dose is well within the limits given in 10 CFR part 100 or NUREG 0800. The proposed change does not alter the bases for assurance that safety-related activities are performed correctly or the basis for any Technical Specification that is related to the establishment of or maintenance of a safety margin. Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
Description of amendment request: The proposed amendment to the Cooper Nuclear Station (CNS) Operating License DPR-46 would revise the design basis accidents (DBA) radiological assessment methodology for offsite and control room radiological doses, and the associated supporting Technical Specifications (TS).
The proposed revisions to the CNS DBA radiological assessment methodology for offsite and control room doses, and the associated supporting TS changes, do not involve initiators or precursors of accidents previously evaluated. Furthermore, these changes do not affect the design, function, or modes of operation of systems, structures, or components within the facility. Therefore, the proposed radiological assessment calculational methodology revisions and TS changes do not involve a significant increase in the probability of an accident previously evaluated in the Updated Safety Analysis Report (USAR).
The proposed revisions to the CNS DBA radiological assessment methodology for offsite and control room doses, and the associated supporting TS changes, do not affect the design, function or modes of operation of systems, structures or components in the facility. The calculation revisions utilize conservatively lower accident mitigation system filter efficiency assumptions and incorporate plant specific accident mitigation system operating parameter and design assumptions. Due to the changes in the calculational methodology and assumptions, and an increase in the postulated accident source term, the calculated radiological dose consequences of each DBA have changed and in some cases increased. In each case, however, the calculated radiological dose consequences are within the exclusion area boundary (EAB) and low population zone (LPZ) radiological dose acceptance criteria specified in 10 CFR part 100 and the control room dose acceptance criteria discussed in General Design Criterion (GDC) 19 of 10 CFR part 50, Appendix A. Therefore, the proposed revisions to the radiological assessment methodology, and associated TS changes, do not involve a significant increase in the consequences of an accident previously evaluated in the USAR.
The proposed revisions to the CNS DBA radiological assessment methodology for offsite and control room doses, and the associated supporting TS 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. No new or different type of plant equipment is installed by the revised radiological assessment calculational methodology or changes to the TS. Neither the calculations nor the TS changes introduce changes to existing design parameters governing normal plant operation or new plant operating modes. No new types of accident initiators or precursors are created by the proposed revisions. Therefore, the proposed revisions to radiological assessment methodology and the proposed changes to the TS do not create the possibility of a new or different kind of accident previously evaluated in the USAR.
The proposed revisions to the CNS DBA radiological assessment methodology for offsite and control room doses, and the associated supporting TS changes, do not affect the design, function or mode of operation of systems, structures or components in the facility. These proposed TS changes are consistent with the criteria of 10 CFR 50.36(c)(2)(ii) for TS content.
The proposed revisions will not result in any challenges to plant equipment, fuel integrity, or the reactor coolant system pressure boundary. Due to the changes in the calculational methodology and assumptions, and an increase in the postulated accident source term, the calculated radiological dose consequences of each design basis accident have changed and in some cases increased. In each case, however, the calculated radiological dose consequences are within the EAB and LPZ radiological dose acceptance criteria specified in 10 CFR part 100 and the control room dose acceptance criteria discussed in GDC 19 of 10 CFR part 50, Appendix A. Therefore, the proposed revisions to the radiological assessment methodology, and associated TS changes, do not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three Start Printed Page 48289standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Description of amendment request: The proposed amendment would change the Cooper Nuclear Station (CNS) Technical Specification (TS) 5.5.10.b.2 to replace the phrase, “A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59” with the phrase “A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.”
The proposed change deletes the reference to unreviewed safety question as defined in 10 CFR 50.59. Deletion of the definition of unreviewed safety question was approved by the NRC with the revisions to 10 CFR 50.59. Consequently, the probability of an accident previously evaluated is not significantly increased. Changes to the TS Bases are still evaluated in accordance with 10 CFR 50.59. As a result, the consequences of any accident previously evaluated are not significantly affected. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change will not reduce the margin of safety because it has no direct effect on any safety analyses assumptions. Changes to the TS Bases that result in meeting the criteria in revised 10 CFR 50.59 (c)(2) will still require NRC approval pursuant to 10 CFR 50.59. This change is administrative in nature as discussed by the NRC in FR (Volume 64, Number 191, Pages 53582-53617) dated October 4, 1999, docketing the change to 10 CFR 50.59. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Date of amendment request: April, 12, 2001.
Description of amendment request: The amendment request would modify the Cooper Nuclear Station (CNS) Technical Specifications Surveillance Requirement (SR) 3.6.1.3.8 to relax the SR frequency by allowing a representative sample of Excess Flow Check Valves (EFCVs) to be tested every 18 months, such that each EFCV will be tested once every 10 years.
The current SR frequency requires each reactor instrumentation line EFCV to be tested every 18 months. The EFCVs at CNS are designed to close automatically in the event of a line break downstream of the valve. This proposed change allows a reduced number of EFCVs to be tested every 18 months. Industry operating experience, documented in BWR [Boiling Water Reactor] Owners' Group Topical Report NEDO-32977-A [“Excess Flow Check Valve Testing Relaxation,” dated June 2000], concludes that a change in surveillance test frequency has a minimal impact on the reliability for these valves. A failure of an EFCV to isolate cannot initiate previously evaluated accidents. Furthermore, neither the EFCV actuation test, nor the frequency of testing is considered an initiator of any analyzed event. Therefore, there is no increase in the probability of occurrence of an accident as a result of this proposed change.
The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and 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. The installed restricting orifice on each associated instrument line provides assurance that any instrument line break will limit offsite doses to substantially below 10 CFR part 100 values. Neither the EFCV actuation test, nor the frequency of testing is an analysis assumption. Therefore, there is no increase in the previously evaluated consequences of the rupture of an instrument line and there is no potential increase in the radiological consequences of an accident previously evaluated as a result of this change.
This proposed change allows a reduced number of EFCVs to be tested each operating cycle. No other changes in requirements are being proposed. Industry operating experience as documented in [BWR Owners' Group Topical Report NEDO-32977-A] provides supporting evidence that the reduced testing frequency will not affect the high reliability of these valves. The potential failure of an EFCV to isolate as a result of the proposed reduction in test frequency is bounded by the previous evaluation of an instrument line pipe break. This change will not physically alter the plant (no new or different type of equipment will be installed). This change will not alter the operation of process variables, structures, systems, or components as described in the safety analysis. Thus, a new or different kind of accident will not be created.
The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. EFCV design, operation, and flow actuation criteria remain unaffected by this change. Restricting orifices for each associated instrument line remains available to mitigate an instrument line break. The proposed change, which impacts the frequency of testing EFCVs is acceptable because the tests continue to require appropriate confirmation of the assumed function of the system (and thereby assure continued operability), and has been shown to reflect an acceptable frequency for detecting failures. There is no detrimental impact on any other 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.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Start Printed Page 48290
Description of amendment request: The amendment would change the Seabrook Station Technical Specifications (TSs) Index, TS 3/4.9.3 (“Decay Time”), TS 3/4.9.4 (“Containment Building Penetrations”), and TS 3/4.9.9 (“Containment Purge And Exhaust Isolation System”). The amendment would also change Bases 3/4.9.3, Bases 3/4.9.4, and Bases 3/4.9.9 for consistency with the proposed TS changes. These changes are consistent with the improved Standard Technical Specifications (STS) for Westinghouse plants.
The proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not adversely affect accident initiators or precursors nor do they adversely alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. In addition, the proposed changes do not adversely affect the manner in which the plant responds in normal operation, transient or accident conditions nor do they change any of the procedures related to operation of the plant. Though a portion of the proposed change to TS 3/4.9.4 appears to be a relaxation to the current licensing basis, North Atlantic has incorporated administrative conservatism into TS 3/4.9.4 to assure the proposed changes, in conjunction with other TS required surveillance testing, do not alter or prevent the ability of structures, systems and components (SSCs), in particular the Containment Purge and Exhaust Isolation System, to perform its intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR).
The proposed changes do not adversely affect the source term, containment isolation or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated in the Seabrook Station UFSAR. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.
Therefore, it is concluded that these proposed revisions to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not involve a significant increase in the probability or consequence of an accident previously evaluated.
This proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not adversely affect the operation nor do they change the design basis of any plant system or component during normal or accident conditions. The proposed changes do not include any physical changes to the plant. In addition, the proposed changes do not adversely affect the function or operation of plant equipment or introduce any new failure mechanisms such that the design basis is adversely affected. The current licensing basis allows penetration isolation by manual or automatic means. The plant equipment will continue to respond per the design and analyses and there will not be a malfunction of a new or different type introduced by the proposed changes that creates the possibility of a new or different kind of accident.
The proposed changes do not modify the facility nor do they adversely affect the plant's response to normal, transient or accident conditions. The changes do not introduce a new mode of plant operation. While these changes may afford North Atlantic operational flexibility, the changes are an enhancement and do not affect plant safety. The plant's design and design basis are not revised and the current safety analyses remains in effect.
Thus, these proposed revisions to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not adversely affect the safety margins established through Limiting Conditions for Operation, Limiting Safety System Settings and Safety Limits as specified in the Technical Specifications nor is the plant design revised by the proposed changes. The current licensing basis allows penetration isolation by manual or automatic means.
Though a portion of the proposed change to TS 3/4.9.4 appears to be a relaxation to the current licensing basis, North Atlantic has incorporated administrative conservatism into TS 3/4.9.4 to ensure the proposed changes, in conjunction with other TS required surveillance testing, offset any potential minimal reduction in the margin of safety. North Atlantic believes that the proposed change to TS 3/4.9.4 is more conservative than that currently allowed in the improved STS, NUREG-1431, Revision 2.
Thus, it is concluded that these proposed revisions to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not involve a significant reduction in a margin of safety.
Date of amendment request: August 15, 2001.
Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to (1) reflect the replacement of Monticello's licensed operator initial and requalification training programs with an accredited systems approach to training program and (2) relocate the existing TS requirements for procedures, records, and reviews to the operational quality assurance plan.
The proposed changes are administrative in nature and compliance with applicable regulatory requirements will continue to be maintained. The proposed changes do not involve any change to the configuration or alter existing system relationships. In addition, the proposed changes do not alter the conditions or assumptions in any of the previous accident analyses thus, the radiological consequences previously evaluated are not adversely affected by the proposed changes.
Therefore, the probability or consequences of an accident previously evaluated are not affected by the proposed amendment.
2. The proposed amendment will not create the possibility of a new or different kind of accident from any previously analyzed.
The proposed changes are administrative in nature and compliance with applicable regulatory requirements will continue to be maintained. The proposed changes do not involve any change to the configuration or method of operation of any plant equipment. Accordingly, no new failure modes have been introduced for any plant system or component important to safety nor has any new limiting single failure been identified as a result of the proposed changes. Also, there Start Printed Page 48291will be no changes in types or increases in the amounts of any effluents released offsite.
Therefore, the possibility of a new or different kind of accident from any accident previously evaluated will not be created.
The proposed changes are administrative in nature and do not involve any change in the methodology or method of operation of any plant equipment. The proposed changes do not involve any change to the configuration or alter existing system relationships. The appropriate controls to provide continued assurance of compliance to applicable regulatory requirements has been maintained.
Description of amendment request: The proposed amendment would amend the licenses to change the required implementation date for previously issued Amendment No. 184 to Facility Operating License NPF-14 and Amendment No. 158 to Facility Operating License NPF-22. The proposed amendment would not alter any of the requirements of the Susquehanna Steam Electric Station (SSES) Unit 1 and 2 Technical Specifications (TSs). The previously issued amendments incorporate long-term power stability solution instrumentation into the SSES Unit 1 and 2 TSs. When implemented, these amendments will incorporate into the TSs the licensee's final response to GL 94-02, “Long Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors.” Specifically, these amendments will, in part, add TS requirements related to the operating power range monitoring (OPRM) system. The licensee stated that recently identified deficiencies in the OPRM trip setpoint methodology, as documented in a General Electric 10 CFR part 21 report issued on June 29, 2001, have adversely affected its ability to implement the subject amendments. Therefore, the licensee requested that the required implementation date for Amendment No. 184 to License No. NPF-14 and Amendment No. 158 to License No. NPF-22 be revised to become effective no later than November 1, 2003.
The proposed amendment implementation date extension is administrative in nature and does not require any physical plant modifications, physically affect any plant systems or components, or entail changes in plant operation. The resulting consequences of transients and accidents will remain within the NRC approved criteria. Therefore, the proposed action does not involve an increase in the probability or consequences of an accident previously evaluated.
The proposed amendment implementation date extension is administrative in nature and does not require any physical plant modifications, physically affect any plant systems or components, or entail changes in plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Description of amendment request: The proposed amendments would revise the reactor vessel pressure-temperature (P-T) limits depicted in Technical Specification Figure 3.4.9-1 for each unit. In addition, pursuant to 10 CFR 50.12, TVA is requesting an exemption from the requirements of 10 CFR part 50, Appendix G, to allow the use of American Society of Mechanical Engineers (ASME) Code Case N-640 as a basis for these revised curves. Code Case N-640, “Alternative Requirement Fracture Toughness for Development of P-T Limit Curves for ASME Boiler and Pressure Vessel Code Section XI, Division 1,” permits the use of the plane strain fracture toughness (KIc) curve instead of the crack arrest fracture toughness (KIa) curve for reactor pressure vessel materials in determining the P-T limits. The exemption request is being reviewed separately.
The proposed Units 2 and 3 change deals exclusively with the reactor vessel pressure-temperature (P-T) curves which define the permissible regions for operation and testing. Failure of the reactor vessel is not considered as a design basis accident. Through the design conservatisms used to calculate the P-T curves, reactor vessel failure has a low probability of occurrence and is not considered in the safety analyses. The proposed changes adjust the reference temperature for the limiting material to account for irradiation effects and provide the same level of protection as previously evaluated and approved. The adjusted reference temperature calculations were performed using the guidance contained in Regulatory Guide 1.99, Revision 2, and ASME Section XI Code Case N-640 to reflect use of the operating limits to 19.5 Effective Full Power Years (EFPY). These changes do not alter or prevent the operation of equipment required to mitigate any accident analyzed in the BFN Final Safety Analysis Report. Therefore, this change does not increase the probability or consequences of any previously evaluated accident. Start Printed Page 48292
The proposed change to the Units 2 and 3 reactor vessel P-T curves does not involve a modification to plant equipment. No new failure modes are introduced. There is no effect on the function of any plant system, and no new system interactions are introduced by this change. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed curves conform to the guidance contained in Regulatory Guide 1.99, Revision 2, and maintain the safety margins specified in 10 CFR 50, Appendix G. Therefore, the proposed amendment does not involve a reduction in a margin of safety.
Date of amendment request: August 7, 2001 (TS-01-04).
Description of amendment request: The proposed amendment would add a new condition and associated actions to the Technical Specification Limiting Condition for Operation (LCO) 3.8.1, “AC Sources Operating,” to allow one Diesel Generator (DG) be out of service for 14 days.
The emergency DGs are designed as backup AC power sources in the event of loss of offsite power. The proposed AOT [allowed outage time] does not change the conditions, operating configurations, or minimum amount of operating equipment assumed in the safety analysis for accident mitigation. No changes are proposed in the manner in which the DGs provide plant protection or which create new modes of plant operation. In addition, a Probabilistic Safety Analysis (PSA) evaluation concluded that the risk contribution of the AOT extension is non-risk significant. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not introduce any new modes of plant operation or make physical changes to plant systems. Therefore, extension of the allowable AOT for DGs does not create the possibility of a new or different accident.
The DGs are designed as backup AC power sources in the event of loss of offsite power. The proposed AOT does not change the conditions, operating configurations, or minimum amount of operating equipment assumed in the safety analysis for accident mitigation. No changes are proposed in the manner in which the DGs provide plant protection or which create new modes of plant operation. In addition, a PSA evaluation concluded that the risk contribution of the AOT extension is non-risk significant. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
Date of amendment request: August 20, 2001.
Description of amendment request: The proposed change to the Technical Specifications (TSs) would revise certain requirements associated with demonstrating the operability of alternate trains when redundant equipment is made or found to be inoperable. The TSs revised include: 4.4.B, 4.5.A.2, 4.5.A.3, 4.5.A.4, 4.5.B.2, 4.5.C.2, 4.5.C.3, 4.5.D.2, 4.5.D.3, 4.5.E.2, 4.5.F.2, 4.5.H.1, 4.7.B.3.c, 4.10.B.1, and 4.10.B.3.b.2. Some format and typographical errors are also being corrected.
Because changing surveillance test requirements does not change the probability of accident precursors, this proposed change does not affect the probability of an accident previously evaluated. Since other periodic and post-maintenance surveillance requirements ensure that the operability of systems and components is maintained, there is no significant increase in the consequences of accidents previously evaluated.
Furthermore, the removal of the additional surveillance testing from the Technical Specifications would result in a decrease in the probability of equipment failure because the excessive testing causes unnecessary wear on the safety-related equipment and unnecessary challenges to safety systems. Reduced testing may also eliminate the potential for human error associated with system alignments and misdirection of attention from monitoring and directing plant operations.
Administrative changes to the Technical Specifications do not alter any technical requirements, and as such, do not increase the probability or consequences of accidents.
Therefore, the proposed change will not increase the probability or consequences of any accident previously evaluated.
Reduced surveillance testing does not create new or different kinds of accidents since modes of operation are unchanged and additional accident precursors are not introduced. System operability requirements and design bases remain the same, and reactor operations are unchanged. Since system and component testing only involves the assurance of operability, reduced testing does not introduce mechanisms that may contribute to the possibility of new or different kinds of accidents.
Administrative changes to the Technical Specifications do not alter any technical requirements, and as such, do not create the possibility of new or different kinds of accidents.
The proposed change will not decrease operability requirements, nor reduce the equipment required during various plant conditions. An acceptable level of testing exists in other Technical Specification requirements to demonstrate system and component operability. There are no changes to system or component operability requirements; therefore, systems and Start Printed Page 48293components will be available to provide existing margins of safety. The same systems and components with the same performance levels assumed in safety analyses will still be available to mitigate consequences of postulated accidents.
Administrative changes to the Technical Specifications do not alter any technical requirements, and as such, have no effect on margins of safety.
Date of application for amendment: March 29, 2001, as supplemented by letters dated June 27, 2001, and July 24, 2001.
Brief description of amendment: The amendment revised the reactor coolant system heatup, cooldown, and inservice leak hydrostatic test limitations for the reactor coolant system to a maximum of 29 effective full power years in accordance with Title 10 of the Code of Federal Regulations, Part 50, Appendix G. These pressure-temperature (P-T) limits are contained in TMI Unit 1 Technical Specification (TS) 3.1.2. In addition, the amendment revised the low-temperature overpressure protection (LTOP) requirements in TSs 3.1.12 and 4.5.2 to reflect the revised P-T limits. These changes will allow operation of two reactor coolant pumps in a single loop during LTOP conditions.
Date of issuance: September 6, 2001.
Date of initial notice in Federal Register: July 25, 2001 (66 FR 38758).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 6, 2001.
Date of application for amendments: November 22, 1999, as supplemented by letters dated October 4 and November 10, 2000, and May 18, 2001.
Brief description of amendments: The amendments authorize revisions to the Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis Report and Independent Spent Fuel Storage Installation Updated Safety Analysis Report to incorporate changes associated with the aircraft hazards analysis due to increased “random” military flights in the vicinity of these facilities. These changes constitute an unreviewed safety question as defined in 10 CFR 50.59 and 10 CFR 72.48.
Date of issuance: August 29, 2001.
Amendment Nos.: 246 and 221.
Renewed Facility Operating License Nos. DPR-53 and DPR-69 and Materials License No. SNM-2502: Amendments revised licenses.
Date of initial notice in Federal Register: December 29, 1999 (64 FR 73085).
The supplemental letters dated October 4 and November 10, 2000, and May 18, 2001, provided clarifying information that did not change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated August 29, 2001.
Date of application for amendment: December 11, 2000.
Brief description of amendment: The amendment revises the Technical Specifications (TSs) to incorporate editorial revisions, clarifications, and corrections. Specifically, the amendment: (1) Provides updated information and corrections to the TS cover page, table of contents, and list of figures, (2) revises TS 4.5.E, “Control Room Air Filtration System,” to remove an incorrect system test description and provide consistent test values for system flow rate and filter efficiency, (3) revises TS 6.2.1.a, “Facility Management and Start Printed Page 48294Technical Support,” to reference the Quality Assurance Program Description as the location of the documentation rather than the Updated Final Safety Analysis Report, (4) revises TS 6.9.1.7, “Monthly Operating Report,” to change the recipient of the Monthly Operating Report, and (5) corrects the periodicity of the Radioactive Effluent Release Report from semi-annual to annual in TS 6.15, “Offsite Dose Calculation Manual” and TS 6.16, “Major Changes to Radioactive Liquid, Gaseous and Solid Waste Systems.” In addition, the amendment revises TS Figure 5.1-1B concerning the indicated vent location associated with Indian Point Unit 3 (IP3). The labels for the IP3 plant vent and the machine shop were reversed.
Date of initial notice in Federal Register: February 21, 2001 (66 FR 11057).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 29, 2001.
Date of application for amendment: April 23, 2001, as supplemented June 25, June 29, and July 19, 2001.
Brief description of amendment: The amendment revises pressure-temperature limit curves and cold overpressure protection limits.
Date of issuance: August 27, 2001.
Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance. August 27, 2001.
Date of initial notice in Federal Register: July 11, 2001 (66 FR 36340).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated August 27, 2001.
Date of application for amendments: April 3, 2001.
Brief description of amendments: The amendments revised the PBAPS Units 2 and 3 Technical Specifications (TSs) to incorporate Technical Specification Task Force (TSTF) Item 258, Revision 4. TSTFs are changes to the improved standard TS that were initiated by the nuclear power industry and submitted to the NRC staff. TSTF-258, Revision 4, revises TS Section 5.0, Administrative Controls, to delete specific TS staffing requirements for licensed Reactor Operators (ROs) and Senior Reactor Operators (SROs), relocate the working hour limits to a plant procedure, clarify requirements for the Shift Technical Advisor position, add regulatory definitions for ROs and SROs, revise the Radioactive Effluent Controls Program to be consistent with the intent of 10 CFR Part 20, and revises radiological area control requirements for high radiation areas to be consistent with 10 CFR 20.1601(c).
Date of issuance: August 30, 2001.
Amendments Nos.: 240 and 243.
Date of initial notice in Federal Register: June 12, 2001 (66 FR 31708).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 30, 2001.
Date of application for amendment: March 7, 2001, as supplemented April 25, June 20, and July 16, 2001.
Brief description of amendment: The amendment revised the Improved Technical Specifications (ITS) 5.6.2.20, “Containment Leakage Rate Testing Program” to allow a one-time interval increase for the Type A Integrated Leakage Rate Test for no more than 5 years.
Date of initial notice in Federal Register: April 2, 2001 (66 FR 17967). The supplemental letters 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 August 30, 2001.
Brief description of amendments: The amendment approves changes to the Updated Final Safety Analysis Report (UFSAR) regarding the modeling of the pressurizer heater operation and spray effectiveness as they relate to certain transients that are analyzed for pressurizer overfill. Specifically, the amendment approves a change to the moderator temperature coefficient currently in the UFSAR assumed as an initial condition for the loss of all nonemergency alternating current power and loss of normal feedwater transients.
Date of issuance: August 23, 2001.
Facility Operating License No. DPR-74: Amendment revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: September 20, 2000 (65 FR 56953).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated August 23, 2001.
Date of application for amendment: January 18, 2001, as supplemented April 20, 2001.
Brief description of amendment: The amendment revises the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications (TSs) 3.10.m to increase the minimum reactor coolant flow from Start Printed Page 4829585,500 gallons per minute (gpm) flow per loop to 93,000 gpm flow per loop.
Date of issuance: September 5, 2001.
The April 20, 2001, supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated September 5, 2001.
Brief description of amendment: The amendment deleted items 3 and 4 from Section 5.15, “Post-Accident Radiological Sampling and Monitoring,” of the Fort Calhoun Station, Unit No. 1 Technical Specifications, and thereby eliminates the requirements to have and maintain the post-accident sampling system (PASS).
Effective date: August 29, 2001, and shall be implemented within 120 days from the date of issuance.
Facility Operating License No. DPR-40. The amendment revised the Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR 38765).
Date of application for amendment: April 11, 2001, as supplemented June 13, 2001.
Brief description of amendment: The amendment revises the Hope Creek Technical Specifications (TSs) to relax the frequency for testing of excess flow check valves (EFCVs). Specifically, TS surveillance requirement 4.6.3.4 has been changed to revise required testing of EFCVs from once per 18 months for all valves to a test of a representative sample each 18 months such that all valves are tested once in 10 years.
Date of issuance: August 28, 2001.
Date of initial notice in Federal Register: May 30, 2001 (66 FR 29361).
The June 13, 2001, 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 August 28, 2001.
Date of application for amendments: September 22, 2000.
Brief Description of amendments: These amendments revise the Facility Operating Licenses ( FOLs) and the Technical Specifications (TS) to remove obsolete license conditions, make editorial changes in the FOLs, and implement associated changes to the TS and Bases.
Effective date: August 30, 2001.
Amendment Nos.: 227 and 227.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments change the License and Technical Specifications.
Date of initial notice in Federal Register: November 1, 2000 (65 FR 65351).
Brief description of amendment: The amendment (1) decreases the allowable values for Function 8, pressurizer pressure-low and pressurizer pressure-high, in Table 3.3.1-1, “Reactor Trip System Instrumentation,” and (2) increases the allowable value for Function 1.d, pressurizer pressure-low for safety injection, in Table 3.3.2-1, “Engineered Safety Feature Actuation System Instrumentation.”
Effective date: August 30, 2001, and shall be implemented prior to entry into Mode 3 in the restart from refueling outage 12 scheduled for the Spring 2002.
Date of initial notice in Federal Register: May 2, 2001 (66 FR 22035).
Dated at Rockville, Maryland, this 10th day of September, 2001.
[FR Doc. 01-23209 Filed 9-18-01; 8:45 am]