Agency information collection activities: Operating licenses, amendments; no significant hazards considerations; biweekly notices,
[Federal Register: April 19, 2000 (Volume 65, Number 76)]
[Notices]
[Page 21034-21045]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr19ap00-129]
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
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Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any 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 March 25, 2000, through April 7, 2000. The last biweekly notice was published on April 5, 2000 (65 FR 17908).
Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.
The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.
Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.
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Written comments may be submitted by mail to the Chief, Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The filing of requests for a hearing and petitions for leave to intervene is discussed below.
By May 19, 2000, 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 filedin 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, the Gelman Building, 2120 L Street, NW., Washington, DC, 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 filedby the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has fileda petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross- examine witnesses.
If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.
If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must be filedwith the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Docketing and Services Branch, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: January 20, 2000.
Description of amendment request: The proposed amendment would revise the technical specifications (TSs) to extend the allowable completion times for the required actions associated with restoration of an inoperable emergency diesel generator (EDG), and permit the performance of the 24-hour EDG endurance run during Modes 1 and 2 (i.e., ``Power Operation'' or ``Startup''). A new requirement is proposed which
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will require verification of the opposite unit's EDGs when the affected EDG is inoperable.
Basis for proposed no significant hazards consideration determination. As required by 10 CFR 50.92(c), the staff's analysis of the issue of no significant hazards consideration is presented below:
Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes include the extension of the completion time for the EDGs from 72 hours to 14 days. In conjunction with the proposed change, a new required action is proposed to be incorporated into the TSs that will require verification of the operability of the opposite unit's EDGs while the affected EDG is inoperable. The proposed changes do not significantly increase the probability of occurrence of a previously evaluated accident because the EDGs are not initiators of accidents. Extending the completion times of the EDGs would not have any impact on the frequency of any accident previously evaluated and, therefore, the probability of a previously analyzed accident is unchanged. The proposed change to the completion time for EDGs will not result in any changes to the plant activities associated with EDG maintenance. The EDGs mitigate the consequences of previously evaluated accidents involving a loss of normal power, the safety-related buses and as such, the operability or availability of the EDGs could affect accident consequences. A configuration risk management program (CRMP) was developed and will be used to ensure that the risk impact of equipment out of service is appropriately evaluated prior to performing any maintenance activity. Increases in risk posed by potential combinations of equipment out of service during the EDG extended completion time will be managed under the CRMP. In addition, compensatory actions have been identified to mitigate an increase in risk. Procedures have been developed to implement the compensatory actions.
The proposed changes also include a change to the TS surveillance requirement related to the conduct of the 24-hour EDG endurance run. Specifically, the change would permit the endurance run to be performed during Modes 1 and 2. The test configuration to be used is consistent with the configuration currently used during the one-hour monthly EDG tests currently conducted.
The probability of an accident is not increased by performing the 24-hour endurance run in Modes 1 and 2 since the EDGs are used to support mitigation of the consequences of an accident. The failure of an EDG while testing is not an assumed initiator of a previously analyzed accident. The EDGs were designed to be tested by running in parallel with offsite power and design features such as protective devices were included. The proposed change does not affect parallel testing design features, the consequences of postulated failures during parallel testing, and postulated interactions with offsite power during parallel testing. If problems are encountered during testing, the EDG connection to the bus will be interrupted, allowing the offsite circuits to continue to supply the bus. Testing of the EDG does not affect the remainder of the safety-related equipment analyzed to mitigate the consequences of an accident. The control logic prevents potential damage of the emergency core cooling System (ECCS) equipment powered by the EDG to ensure that the ECCS equipment is available in the event of an actual safety injection with or without a Loss of Offsite Power (LOOP). Only one EDG per unit will be tested in parallel with the offsite sources at a time in order to prevent any grid disturbance from potentially affecting more than one EDG. Thus, during the test, the remaining EDG, which is capable of supplying power to mitigate all design basis accidents, will be available to respond normally to a start signal.
To fully evaluate the effect of the proposed EDG TS changes, probabilistic risk assessment (PRA) methods and deterministic analyses were utilized. The results of the risk analysis show no significant increase in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF).
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously analyzed.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes do not involve a physical change to the plant. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner except for the following. The electrical lineup for performing the 24-hour run will be the same as the lineup for performance of the one-hour run, which is routinely performed at least once per month for each EDG. The difference between these two surveillances is in their duration. There is no change being made to the parameters within which the plant is operated. There are no setpoints affected by this proposed change at which protective or mitigative actions are initiated. This proposed changes will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No alteration in the procedures, which ensure that the plant remains within analyzed limits, is being proposed, 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. Other than the changes in duration of EDG unavailability from 72 hours to 14 days and on-line testing from 60 minutes to 24 hours, the change does not alter assumptions made in the safety analysis and licensing basis.
Therefore, these proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Does the change involve a significant reduction in a margin of safety?
The proposed changes will extend the allowable Completion Times for the Required Actions associated with restoration of an inoperable EDG and allow the performance of the 24-hour endurance run at power. In conjunction with the proposed changes, a new required action is proposed to be incorporated into the TSs. The new action will require verification of the operability of the opposite unit's EDGs while the affected EDG is inoperable. These actions will be taken to ensure the availability of the remaining alternating current power sources to the affected engineered safety feature bus.
The CRMP will be used to ensure that the risk impact of equipment out of service is appropriately evaluated prior to performing any maintenance activity. Increase in risk posed by potential combinations of equipment out of service during the EDG extended completion time will be managed under the CRMP. In addition, compensatory actions have been identified to mitigate increase in risk. Procedures have been developed to implement the compensatory actions.
With regard to the proposed change for the 24-hour endurance run, the EDGs were designed to be tested by running in parallel with offsite power and, design features such as protective devices were included. The proposed change does not affect parallel testing design features, the consequences of postulated failures during parallel testing, and postulated interactions with offsite power during parallel testing. If problems are encountered during testing, the EDG connection to the bus will be interrupted allowing the offsite circuits to continue to supply the bus. Further, the EDG system design includes emergency override of the test mode for both accident conditions (safety injection) and loss of offsite power to permit a response to actual emergency signals and return control of the EDG to the automatic control system.
Therefore, implementation of the proposed changes will not involve a significant reduction in the margin of safety.
Based on the staff's analysis, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice President and General Counsel, Commonwealth Edison Company, P.O. Box 767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: March 6, 2000.
Description of amendment request: The proposed amendment would revise the Improved Technical Specification (ITS) Action Condition and Surveillance Requirement (SR) for the safety-related diesel-driven emergency feedwater pump (EFP-3). The ITS required
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inventory volume for lube oil would be revised to agree with the actual test values and are included in the ITS Action Condition, SR and Bases.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below.
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Does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The revised lube oil requirements are being made to ensure EFP-3 is capable of seven days of continuous operation. The proposed amendment provides the same functional requirement as previously approved. The EFW system is used for accident mitigation and is not an initiator of design basis accidents. Therefore, the probability of previously analyzed events is not affected by this change. No capability or design functions of EFP-3 or the emergency feedwater (EFW) system will change. The initial conditions for accidents that require EFW and accident mitigation capability of the EFW system will remain unchanged. Therefore, the proposed amendment will not increase the consequences of evaluated accidents.
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Does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The revised ITS Condition will ensure equipment is restored to operable status in accordance with previously approved timeframes and functional levels. The revised SR will assure the same functional requirement as the previously approved SR. Lube oil will be stored on-site and the lube oil inventory in the sump ensures adequate time to transfer the stored inventory into the engine. No new plant configurations or conditions are created by these revised ITS Conditions or SR. Therefore, the proposed amendment cannot create the possibility of an accident of a different type than previously evaluated in the Safety Analysis Report.
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Does not involve a significant reduction in the margin of safety.
The proposed ITS Condition and SR changes ensure adequate lube oil inventory is available to operate EFP-3 for seven days. The proposed changes replace the calculated lube oil inventory values with a more conservative value derived from actual test data for EFP-3. The revised SR ensures the same functional requirement for a seven-day supply of lube oil for EFP-3 as was previously approved. Similarly, the revised ITS Condition ensures the same functional level as currently approved by requiring that a reduced lube oil inventory of less than seven days but more than six days is restored to the seven-day level within 48 hours. The revised SR allows the lube oil inventory to be stored off engine. The inventory in the EFP-3 sump and auxiliaries provides sufficient time to permit the transfer of stored inventory into the engine. EFP-3 is designed to allow monitoring of lube oil level and addition of lube oil while the engine is operating. Based on the above, the revised ITS meets the same intent as the currently approved specifications. Therefore, there is no reduction in the margin of safety associated with the proposed ITS amendment.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida 33733-4042.
NRC Section Chief: Richard P. Correia.
Pacific Gas and Electric Company, Docket Nos. 50-275, Diablo Canyon Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California
Date of amendment requests: December 31, 1999, as supplemented by letter dated January 18, 2000.
Description of amendment requests: The amendment would revise Section 2.C.(1) of Facility Operating License No. DPR-80 for the Diablo Canyon Power Plant, Unit No. 1 to authorize operation at reactor core power levels not to exceed 3411 megawatts thermal (100 percent rated power). This amendment would also (1) revise the definition in Section 1.1 of the technical specifications (TS) of rated thermal power to reflect Unit 1 operation at the uprated reactor core power level, (2) change the reactor core safety limits in TS Figure 2.1.1-1 to reflect the current fuel type and provide additional margin for OTT and OPT setpoint calculations, and change the nominal full power Tavgin the OTT and OPT function in Notes 1 and 2 to TS Table 3.3.3-1.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
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The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
All previously evaluated accidents have been reviewed for the proposed increase in Unit 1 power rating, and these reviews are summarized in WCAP-14819, ``Pacific Gas and Electric Company Diablo Canyon Power Plant, Unit 1 3425 MWt [megawatt thermal] Uprating Program Licensing Report.'' The majority of the Diablo Canyon Power Plant (DCPP) accident analyses already bound the higher power rating of Unit 2 combined with the lower reactor coolant system (RCS) flow rate of Unit 1. Hence, the uprate has no impact on these previously evaluated accidents. This is also true of dose assessment, which remains based on the original 3568 MWt core source terms and is not impacted by the uprate.
The previously evaluated accidents that are impacted by the uprate are large break loss-of-coolant accident (LOCA), small break LOCA, the OTT/ OPT setpoint calculations, and accidental depressurization of the RCS. The large break LOCA was reanalyzed for uprated conditions using best estimate methodology. The reanalysis demonstrated no increase in consequence and was approved by the NRC in License Amendments 121 (Unit 1) and 119 (Unit 2). The small break LOCA was also reanalyzed, and continues to demonstrate a large margin to peak clad temperature limits. The current OTT/OPT setpoints are bounding for the Unit 1 uprated power conditions based on revising the reactor core safety limits in TS Figure 2.1.1-1 to credit the exclusive use of Vantage 5 fuel. The accidental RCS depressurization reanalysis shows that the departure from nucleate boiling ratio remains above the applicable limit value. In summary, no design or analysis acceptance criteria will be exceeded, the functional integrity of all plant systems are unaffected, and there is no impact on the integrity of the fission product barriers or assumed dose source terms. Therefore, the consequences of all previous evaluated accidents are not substantially increased.
It was determined that there would be no impact on any component reliabilities assumed in the PRA model, and therefore no impact on the resulting core damage frequency. The PRA model envelopes both units, based on using the originally higher rated Unit 2 power level.
The operation impacts of the proposed power increase were reviewed against the unit design capability, and it was determined that no system, structure, or component would exceed design conditions or loads. While the low pressure turbines see a small (less than 1.5 deg.F ) increase in temperature, the effect on missile generation probability is not significant. There is no significant increase in the probability of component failure, offsite power loss, or any other accident initiator. Therefore, the probability of all previously evaluated accidents is not substantially increased.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Normal operation will not be substantially impacted by increasing the Unit 1 licensed power rating to match Unit 2. Procedures will be essentially unchanged, or where changes are required, they will be made to more closely resemble those in effect at Unit 2. Training will communicate all impacts to personnel and the plant simulator will be updated to match the power level of both
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Units 1 and 2. There is, therefore, no possibility of a new or different kind of accident related to human performance.
Plant systems, structures, and components have been evaluated for the proposed uprate. Most have identical counterparts in operation at Unit 2 at this higher power level. A few are slightly different, such as the generator cooling system, and for these the design margins have been reviewed and found to be acceptable. Therefore, there is no possibility of a new or different kind of accident related to system, structure, or component performance.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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The proposed change does not involve a significant reduction in a margin of safety.
The proposed changes do not involve a significant reduction in a margin of safety because the margin of safety associated with plant parameters as verified by the results of the accident analyses are within acceptable limits. As mentioned, most analyses demonstrating adequate margins of safety already assume the higher thermal power rating of Unit 2 and bound Unit 1 at the uprated thermal power conditions. The few transients that are reanalyzed meet the applicable acceptance criteria.
The reactor core safety limits specified in TS Figure 2.1.1-1 envelope operation with both 17x17 standard and 17x17 Vantage 5 fuel. The proposed change revises the reactor core safety limits in Figure 2.1.1-1 to credit the exclusive use of Vantage 5 fuel. These revised safety limits will continue to satisfy fuel design criteria. The current OTT and OPT setpoints provide adequate margin to the revised reactor core safety limits at the uprated Unit 1 conditions, which include a slightly higher nominal full power Tavgin Notes 1 and 2 to ITS Table 3.3.3-1.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva Power and Light Company, and Atlantic City Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: August 11, 1999.
Description of amendment request: The proposed amendment clarifies the use of containment overpressure for ensuring adequate net positive suction head (NPSH) for the emergency core cooling system (ECCS) pumps.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
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The proposed License Amendment Request does not involve a significant increase in the probability or consequences of an accident previously evaluated.
This proposed license amendment request does not involve any physical changes to plant Structures, Systems, or Components (SSC), or how the SSC are operated, maintained, and tested. The proposed changes involve the acceptability of taking credit for a specific amount of containment overpressure following the initiation of an event. This credit involves the mitigation of an event, and not prevention or identification of an event. Credit for containment overpressure is not considered a precursor to any event.
Crediting containment overpressure does not turn an Anticipated Operational Occurrence (AOO) into an Abnormal Operational Transient (AOT) or a Design Basis Accident (DBA).
Calculations performed in support of the license amendment request provide a conservative estimate of the Minimum Containment Pressure Available (MCPA) following all design and licensing basis events for which some amount of containment overpressure is required. The NPSH calculations for the Residual Heat Removal (RHR) and Core Spray (CS) pumps include conservative assumptions and input values ensure that, barring beyond-design-basis loss of containment integrity, adequate NPSH is provided to the RHR and CS pumps for the entire duration of any of these events.
The proposed license amendment request makes a change to the PBAPS licensing basis to clearly define amount of containment overpressure allowed. This value is designated as the Containment Overpressure License (COPL). Conservative analyses have assured that the MCPA is always greater than this COPL for design basis events. Therefore, adequate NPSH is provided to the RHR and Core Spray pumps for all design and licensing basis events.
The evaluation for MCPA and NPSH includes the consideration for any one single active failure. The worst-case single active failure is the failure of one electrical division. There is no credible single active failure that can compromise the containment integrity. The evaluation for MCPA and NPSH does not place any restrictions on system operation following a design or licensing basis event. The analysis concludes that adequate NPSH will be available, even assuming the worst single active failure.
Therefore, the proposed license amendment request does not significantly increase the probability or consequences of an accident previously evaluated.
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The proposed License Amendment Request does not create the possibility of a new or different kind of accident from any accident previously evaluated.
This proposed license amendment request does not involve any physical changes to plant SSC, or how the SSC are operated, maintained, and tested. This proposed license amendment request involves the acceptability of taking credit for some amount of containment overpressure following the initiation of an event. This credit involves the mitigation of an event, and not prevention or identification of an event. Credit for containment overpressure is not considered a precursor to any event. Worst-case single active failure (i.e., loss of one electrical division) was considered in the assessment of MCPA and COPR [containment overpressure required]. The supporting calculations indicate that adequate NPSH is provided to the RHR and CS pumps for all design and licensing basis events, even with the worst single active failure.
Therefore, the proposed license amendment request does not create the possibility of a new or different kind of accident from any previously evaluated.
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The proposed License Amendment Request does not involve a significant reduction in a margin of safety.
The MCPA and NPSH analyses supporting this license amendment request include conservative assumptions and use conservative input values that are consistent with or bound the analytical limits of the PBAPS Technical Specifications. These analyses indicate that adequate NPSH margin is available for operation of the RHR and CS systems to meet their safety functions following any design or licensing basis event. This includes operation of RHR in Suppression Pool Cooling, Wetwell Spray, Drywell Spray, and Low Pressure Coolant Injection modes, and CS in Short Term and Long Term Spray Cooling. Therefore, the proposed license amendment request does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: James W. Clifford.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50- 362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego County, California
Date of amendment requests: January 19, 2000 (PCN-512).
Description of amendment requests: The amendment application proposes to revise the San Onofre Nuclear
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Generating Station, Units 2 and 3, technical specifications (TSs) Surveillance Requirement (SR) 3.0.3.
SR 3.0.3 allows compliance with the requirement to declare a limiting condition for operation not met to be delayed whenever it is discovered that a surveillance was not performed within its specified frequency (a missed surveillance). Presently, SR 3.0.3 allows a delay ``up to 24 hours or up to the limit of the specified Frequency, whichever is less.'' The licensee proposes to revise the allowable delay ``up to 24 hours or up to the limit of the specified Frequency, whichever is greater.''
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
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Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response: No.
The proposed change would extend the maximum allowable time for completing a Surveillance not performed within its specified Frequency (a missed Surveillance) without declaring the affected Limiting Condition For Operation (LCO) not met. The presently allowed time is up to 24 hours from the time of discovery or up to the limit of the specified Frequency, whichever is less. The proposed allowed time is up to 24 hours from the time of discovery or up to the limit of the specified Frequency, whichever is greater.
Surveillances are rarely missed. This is demonstrated by a limited review of Licensee Event Reports (LERs), which found very few occurrences of missed Surveillances, given the number of LERs submitted and the large number of Surveillances performed. Moreover, Surveillances, whether performed inside or outside the required Frequency, nearly always verify conformance with Technical Specification requirements. This is demonstrated by a survey of selected licensees regarding entries into Surveillance Requirement (SR) 3.0.3. As stated in Generic Letter 87-09, ``* * * the vast majority of surveillances do in fact demonstrate that systems or components are operable.'' As stated in the SR 3.0.3 Bases, ``* * * the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.''
Therefore, it is unlikely that plant equipment would be inoperable during the time period of up to 24 hours or up to the limit of the specified Frequency, whichever is greater, that would be allowed under the proposed change for the completion of a missed Surveillance.
If, upon discovery of a missed Surveillance, it is known that the Surveillance would fail, SR 3.0.1 would require that the affected LCO be declared not met and the appropriate Condition(s) entered.
Performance of some Surveillances carries with it a slight risk, either from making some plant equipment temporarily inoperable or from performing plant manipulations, or both. The increase in plant risk from performing such Surveillances, combined with the confidence that a Surveillance test will be satisfactory when performed, together provide justification for extending the current allowable time to up to 24 hours or up to the specified Frequency, whichever is greater.
The foregoing discussion demonstrates that the probability or consequences of any accident previously evaluated will not be significantly increased by the proposed change.
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Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
This amendment request is administrative in nature and does not involve any change to plant equipment. Therefore, it will not create the possibility of a new or different kind of accident from any previously evaluated.
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Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response: No.
This amendment request does not change the manner in which safety limits or limiting safety settings are determined.
As discussed above, Surveillances are rarely missed, and, when performed, Surveillances nearly always verify conformance with Technical Specification requirements, making it unlikely that plant equipment would be inoperable during the time period of up to 24 hours or up to the limit of the specified Frequency, whichever is greater, that would be allowed under the proposed change for the completion of a missed Surveillance.
Therefore, the proposed change will not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern California Edison Company, 2244 Walnut Grove Avenue, Rosemead, California 91770.
NRC Section Chief: Stephen Dembek.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 17, 2000.
Description of amendment request: Revise Technical Specification 3/ 4.7.4 to revise the surveillance requirements (SRs) 4.7.4.b.1 and 4.7.4.b.2 to incorporate the wording from the Westinghouse Standard Improved Technical Specifications (NUREG-1431) and to delete SR 4.7.4.b.3. SR 4.7.4.b.3 requires verifying at least once per 18 months that each screen wash booster pump and the traveling screen start automatically on a safety injection test signal. The licensee also proposed changes to the Technical Specifications Bases associated with the Technical Specification changes and administrative changes to the Bases Index.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
-
Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
NUREG 1431 related changes:
Incorporating the NUREG 1431 [Westinghouse Standard Improved Technical Specifications] wording for SR 4.7.4.b.1 and SR 4.7.4.b.2 does not significantly increase the probability of an accident because the surveillance testing of the Essential Cooling Water system has no effect on accident initiation probability. This change does not significantly increase the consequences of an accident because the surveillance requirements still provide adequate assurance that the Essential Cooling Water system can provide its design function.
Screen wash system changes:
Eliminating the requirement for the Essential Cooling Water traveling screens and screen wash booster pumps to start on a safety injection signal does not increase the probability of any accident previously evaluated. The traveling screens and the screen wash booster pumps have no potential for initiating an accident. Eliminating the requirement for the traveling screens and the screen wash booster pumps to start on a safety injection signal does not increase the consequences of any accident previously evaluated. A control system is provided to automatically start and stop the traveling screens during normal operation. A high differential water level sensed across any traveling screen alarms in the control room and automatically starts the screen wash booster pump and, after reaching adequate screen wash pressure, starts the traveling screen. A safety injection signal is not needed for this function. In addition, there are no circumstances associated with any event requiring a safety injection signal that would cause a high differential water level across the traveling screen.
The changes to the Bases Index are administrative and have no relevance to accident probability or consequences.
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Based on the above, STPNOC [STP Nuclear Operating Company] concludes that the proposed change does not increase the probability or consequences of an accident previously evaluated.
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Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
NUREG 1431 related changes:
Incorporation of the NUREG 1431 wording into the surveillance requirements does not create the possibility of a new or different kind of accident because the surveillance requirements are not substantially changed and do not involve any different operational configurations for the station.
Screen wash system changes:
Elimination of the requirement to start the traveling screen and screen wash booster pump on a safety injection signal will not create the possibility of a new or different kind of accident from any accident previously evaluated. As discussed above, the traveling screens and screen wash booster pump have no potential to initiate an accident. In addition, STPNOC is not proposing any different operational configurations for the station.
The changes to the Bases Index are administrative and have no relevance to accidents.
Based on the above, STPNOC concludes that the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Does the proposed amendment involve a significant reduction in a margin of safety?
NUREG 1431 related changes:
Incorporation of the NUREG 1431 wording for SR 4.7.4.b.1 and SR 4.7.4.b.2 does not significantly change the way the surveillance requirements will be performed. The Surveillance Requirements still provide adequate assurance that the Essential Cooling Water will perform its function. There is no change in the operational configuration of the plant. Consequently, the changes to these surveillance requirements do not significantly affect the margin of safety.
Screen wash system changes:
Elimination of the requirement for the traveling screen and screen wash booster pump to start on a safety injection signal will not prevent the traveling screen and screen wash booster pump to start when required. The systems will start automatically without the need for a safety injection signal. In addition, there is no design basis or mechanistic reason to postulate the need to automatically start the traveling screens or screen wash booster pump on a safety injection signal.
The changes to the Bases Index are administrative and have no relevance to the safety margin.
Based on the above, STPNOC concludes that the proposed change does not involve a significant decrease in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the request for amendments involves no significant hazards consideration.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey County, Kansas
Date of amendment request: March 31, 2000 (ET 00-0018).
Description of amendment request: The proposed amendment would modify the actions for Limiting Condition for Operation (LCO) 3.7.9, ``Ultimate Heat Sink (UHS),'' of the technical specifications (TSs). The proposed new Action A would allow the plant to operate with the plant inlet water temperature of the UHS above 90 deg.F, if the licensee verified the required cooling capacity within 4 hours and once per 12 hours thereafter, but that the plant would be shut down if the water temperature exceeded 94 deg.F. This would change the current requirement to shut down the plant if the inlet water temperature of the UHS exceeded the 90 deg.F. The time to shut down the plant is not being changed in the amendment request.
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 proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve any physical alteration of plant systems, structures or components. The proposed change provides an allowance for the plant to continue operation with [the] plant inlet water temperature [of the UHS] in excess of the current Technical Specification limit of 90 deg.F with the verification that required cooling capacity being maintained and [the plant inlet water] temperature 94 deg.F. The 94 deg.F limit is less than the design limit of 95 deg.F for associated plant components. The plant inlet water temperature is not assumed to be an initiating condition of any accident analysis evaluated in the Updated Safety Analysis Report (USAR). Therefore, the allowance for the [plant inlet] water temperature to be in excess of the current limit does not involve an increase in the probability of an accident previously evaluated in the USAR. The UHS supports OPERABILITY of safety related systems used to mitigate the consequences of an accident. Plant operation for brief periods with [the] plant inlet water temperature greater than 90 deg.F up to 94 deg.F will not adversely affect the OPERABILITY of these safety related systems and will not adversely impact the ability of these systems to perform their safety related functions. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated in the USAR.
-
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change does not involve any physical alteration of plant systems, structures or components. The temperature of the plant inlet water being greater than 90 deg.F but less than or equal to 94 deg.F (with the main cooling lake dam intact) does not introduce new failure mechanisms for systems, structures or components not already considered in the USAR. The 94 deg.F limit is less than the design limit of 95 deg.F for associated plant components. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
-
The proposed change does not involve a significant reduction in a margin of safety.
The proposed change will allow an increase in [the] plant inlet water temperature above the current Technical Specification limit of 90 deg.F for the UHS, provided [the] UHS temperature is maintained below 95 deg.F and that the required cooling capacity is verified maintained within 4 hours and once per 12 hours thereafter. Additionally, the plant inlet water temperature will be verified to be 94 deg.F once per 12 hours. The proposed change does not alter any safety limits, limiting safety system settings, or limiting conditions for operation, and the proposed changes provide continued assurance that with a plant inlet water temperature > 90 deg.F, the design temperature of safety related equipment are maintained within acceptable limits such that a safe shutdown of the plant can be performed. In addition, avoiding a plant transient during environmental conditions that could challenge the stability of the Electrical Power System offsets any perceptible reduction in the margin of safety as a result of the proposed change. Thus, the proposed change does not involve a significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Section Chief: Stephen Dembek.
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Previously Published Notices of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing
The following notice was previously published as a separate individual notice. The notice content was the same as above. It was published as an individual notice either because time did not allow the Commission to wait for this biweekly notice or because the action involved exigent circumstances. It is repeated here because the biweekly notice lists all amendments issued or proposed to be issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on the day and page cited. This notice does not extend the notice period of the original notice.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of application for amendments: March 29, 2000 (TS-402).
Brief description of amendments: Changes Technical Specification 3/ 4.6.4.1 ``Secondary Containment'' to permit maintenance on a secondary containment access door when one or more units are operating and the other door is closed.
Date of publication of individual notice in the Federal Register: April 6, 2000 (65 FR 18141)
Expiration date of individual notice: April 20, 2000.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of application for amendment: August 26, 1999, as supplemented on February 24, and March 14, 2000.
Brief description of amendment: This amendment revises Technical Specification 3/4.9.4, and its associated bases, to allow the personnel airlock and certain other containment penetrations to remain open during refueling operations provided specific administrative controls are met. This amendment is approved for use during refueling outage 9 and operating cycle 10.
Date of issuance: March 27, 2000.
Effective date: March 27, 2000.
Amendment No.: 97.
Facility Operating License No. NPF-63. Amendment revises the Technical Specifications.
Date of initial notice in Federal Register: October 6, 1999 (64 FR 54374).
The February 24, and March 14, 2000, submittals contained clarifying information only, and did not change the initial no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 27, 2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: May 20, 1999.
Brief description of amendments: The amendments changed the Technical Specification (TS) value for the minimum suppression chamber water level to a more conservative value.
Date of issuance: March 30, 2000.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 176 & 172.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR 46426).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 30, 2000.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of application for amendment: December 17, 1999, as supplemented March 8, 2000.
Brief description of amendment: The amendment revises Technical Specification (TS) 2.1, ``Safety Limits (SLs),'' changing the safety limit minimum critical power ratio limits in TS 2.1.1.2 to reflect the results of cycle-specific calculations performed for Fermi 2 operating Cycle 8.
Date of issuance: March 30, 2000.
Effective date: As of the date of issuance and shall be implemented prior to the startup from the seventh refueling outage.
Amendment No.: 138.
Facility Operating License No. NPF-43: Amendment revises the Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR 4269).
The March 8, 2000, letter provided clarifying information that was within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 30, 2000.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of application for amendment: July 30, 1999, as supplemented December 17, 1999, and March 1, 2000.
[[Page 21042]]
Brief description of amendment: The amendment revises Technical Specification (TS) 3.3.1.1, ``Reactor Protection System (RPS) Instrumentation,'' to reflect the activation of the automatic trip associated with the oscillation power range monitor (OPRM). The amendment also revises TS 3.4.1, ``Recirculation Loops Operating,'' to remove requirements related to the manual detection and suppression of core thermal-hydraulic instabilities because these actions are no longer necessary after the OPRM upscale function is activated.
Date of issuance: March 31, 2000.
Effective date: As of the date of issuance and shall be implemented prior to the startup from the seventh refueling outage.
Amendment No.: 139.
Facility Operating License No. NPF-43: Amendment revises the Technical Specifications.
Date of initial notice in Federal Register: November 3, 1999 (64 FR 59800).
The December 17, 1999, and March 1, 2000, letters provided clarifying information that was within the scope of the original Federal Register notice and did not change the staff's initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 31, 2000.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas
Date of amendment request: August 6, 1998, as supplemented by letter dated February 16, 2000.
Brief description of amendment: The amendment revises the minimum and the maximum concentration limits for the sodium hydroxide tank. The amendment also deletes the maximum specified tank volume and revises the minimum specified tank volume to refer to the parameter used in the safety analysis with no allowance for instrument uncertainty.
Date of issuance: March 28, 2000.
Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
Amendment No.: 206.
Facility Operating License No. DPR-51: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64 FR 6695).
The February 16, 2000, letter provided clarifying information that did not change the scope of the August 6, 1998, application and the initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 28, 2000.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas
Date of application for amendment: January 27, 2000.
Brief description of amendment: The amendment deleted the current requirements of Technical Specification (TS) 4.7.9.1.2.d, ``Source installed in the Boronometer,'' associated with the installed boronometer sealed source. The source was recently removed and stored, and the requirements of TS 4.7.9.1.2.d are no longer applicable.
Date of issuance: March 24, 2000.
Effective date: As of the date of issuance to be implemented within 30 days from the date of issuance.
Amendment No.: 212.
Facility Operating License No. NPF-6: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65 FR 9007).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 24, 2000.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas
Date of application for amendment: January 27, 2000.
Brief description of amendment: The amendment relocated the schedule for the withdrawal of reactor vessel material surveillance specimens, from the Technical Specifications to the Safety Analysis Report, pursuant to the guidance provided in Generic Letter 91-01, ``Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens From Technical Specifications.'' Changes to the related Bases were also made. In addition, the proposed change to the surveillance specimen removal schedule was approved.
Date of issuance: April 4, 2000.
Effective date: As of the date of issuance to be implemented within 30 days from the date of issuance.
Amendment No.: 213.
Facility Operating License No. NPF-6: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65 FR 9007).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 4, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: July 26, 1999, as supplemented by submittal dated December 7, 1999.
Brief description of amendment: This amendment permits implementation of 10 CFR Part 50, Appendix J, Option B, and reference Regulatory Guide 1.163, ``Performance-Based Containment Leak Test Program,'' dated September 1995, which specifies a method acceptable to the NRC for complying with Option B. These changes relate only to Type B and C (local) leakage rate testing. The use of Option B for Type A (integrated) leakage rate testing was approved on February 22, 1996, by License Amendment No. 205.
Date of issuance: March 28, 2000.
Effective date: Immediately as of its date of issuance and shall be implemented within 120 days.
Amendment No.: 240.
Facility Operating License No. NPF-3: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR 46437).
The letter of December 7, 1999, 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 March 28, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: September 9, 1999, as supplemented by submittal dated February 28, 2000.
Brief description of amendment: This amendment includes nine minor, unrelated revisions to the technical specifications (TSs). These revisions, which are minor in both content and safety significance, include
[[Page 21043]]
clarifications and editorial changes to the TSs.
Date of issuance: March 30, 2000.
Effective date: Immediately as of the date of issuance and shall be implemented within 90 days.
Amendment No.: 111.
Facility Operating License No. NPF-58: This amendment revised the Technical Specifications.
Date of initial notice in Federal Register: November 3, 1999 (64 FR 59803).
The 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 March 30, 2000.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: December 3, 1998.
Brief description of amendments: The amendments made administrative changes to several Technical Specifications to remove obsolete information, provide consistency between Unit 1 and Unit 2, provide consistency with the Standard Technical Specifications, provide clarification, and correct typographical errors.
Date of issuance: March 31, 2000.
Effective date: March 31, 2000, with full implementation within 30 days.
Amendment Nos.: 243 and 224.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 31, 1999 (64 FR 47535).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 31, 2000.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska
Date of amendment request: December 6, 1999, as supplemented March 17, 2000.
Brief description of amendment: Amendment to technical specifications changes the safety limit minimum critical power ratio (SLMCPR) from 1.06 to 1.08 for two recirculation loop operation and from 1.07 to 1.09 for single recirculation loop operation.
Date of issuance: March 31, 2000.
Effective date: March 31, 2000, to be implemented within 30 days.
Amendment No.: 182.
Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64 FR 73093).
The March 17, 2000, letter provided additional clarifying information that was within the scope of the original application and Federal Register notice and did not change the staff's initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 31, 2000.
No significant hazards consideration comments received: No.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: December 6, 1999, as supplemented by letters dated February 22 and March 14, 2000.
Brief description of amendment: The amendment modifies the Technical Specification (TS) surveillance requirements associated with ensuring a limited number of charging and high pressure safety injection pumps are incapable of injecting into the Reactor Coolant System when the plant is shutdown. In addition, the TS Bases are modified to address these changes.
Date of issuance: March 30, 2000.
Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
Amendment No.: 243.
Facility Operating License No. DPR-65: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR 4285).
The February 22 and March 14, 2000, supplemental letters provided clarifying information that did not change the staff's original no significant hazards consideration determination or expand the scope of the application as published.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 30, 2000.
No significant hazards consideration comments received: No.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: September 7, 1999.
Brief description of amendment: The amendment removes the current special exception which precludes applying the 18-month functional testing surveillance to the Steam Generator Hydraulic Snubbers for Technical Specification 3/4.7.8, ``Plant Systems, Snubbers.''
Date of issuance: March 31, 2000.
Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.
Amendment No.: 244.
Facility Operating License No. DPR-65: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR 4283).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 31, 2000.
No significant hazards consideration comments received: No.
Portland General Electric Company, et al., Docket No. 50-344, Trojan Nuclear Plant, Columbia County, Oregon
Date of application for amendment: January 29, 1998.
Brief description of amendment: The amendment deletes paragraph 2.D of Facility Operating License No. NPF-1 and revises the Permanently Defueled Technical Specifications (PDTS) by deleting PDTS 5.7.1.1(b). These changes remove the requirements for a security program at the 10 CFR part 50 licensed site once the spent nuclear fuel has been relocated to the 10 CFR part 72 licensed Independent Spent Fuel Storage Installation.
Date of issuance: April 6, 2000.
Effective date: April 6, 2000, to be implemented within 30 days after the transfer of the last cask of spent nuclear fuel from the spent fuel pool to the independent spent fuel storage installation is complete.
Amendment No.: 203.
Facility Operating License No. NPF-1: The amendment changes the Operating License and the Permanently Defueled Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64 FR 48865).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated April 6, 2000.
No significant hazards consideration comments received: No.
[[Page 21044]]
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: July 23, 1999, as supplemented September 13, 1999, and January 31, 2000.
Brief description of amendments: The amendments revised the Technical Specifications to remove the restriction on performing the 24-hour endurance run test of emergency diesel generators (EDGs) every 18 months only during shutdown. Additionally, for Salem Unit 1 only, a note associated with a one-time extension of a surveillance requirement was deleted.
Date of issuance: March 30, 2000.
Effective date: As of its date of issuance and shall be implemented within 60 days.
Amendment Nos.: 229 and 210.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 6, 1999 (64 FR 54380).
The January 31, 2000, supplement provided clarifying information that did not change the initial proposed no significant hazards consideration determination and did not expand the scope of the original application as published.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 30, 2000.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia
Date of application for amendments: October 4, 1999.
Brief description of amendments: The amendments revised the Technical Specifications 5.5.6, ``Prestressed Concrete Containment Tendon Surveillance Program,'' to incorporate three exceptions to Regulatory Guide (RG) 1.35, Revision 2, 1976. The exceptions concern the number of tendons detensioned, inspection of concrete adjacent to vertical tendons, and the time during which areas adjacent to tendons are inspected.
Date of issuance: March 27, 2000.
Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
Amendment Nos.: Unit 1-112; Unit 2-90.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 9, 2000 (65 FR 6411).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 27, 2000.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 29, 1999 (TS 99-04).
Brief description of amendments: The amendments delete Sequoyah Nuclear Plant Technical Specification (TS) monthly surveillance test on the auxiliary feedwater suction pressure switches.
Date of issuance: March 29, 2000.
Effective date: As of the date of issuance, to be implemented no later than 45 days after issuance.
Amendment Nos.: 253 and 244.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments revise the TS.
Date of initial notice in Federal Register: May 19, 1999 (64 FR 27325).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 29, 2000.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 30, 1999.
Brief description of amendments: Requirements related to containment isolation valves that were located in two different sections of the technical specifications were consolidated into one section. Also, conditions relating to or usage of a check valve as an isolation device was clarified.
Date of issuance: March 29, 2000.
Effective date: March 29, 2000.
Amendment Nos.: 254 and 245.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments revise the technical specifications.
Date of initial notice in Federal Register: October 6, 1999 (64 FR 54382).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 29, 2000.
No significant hazards consideration comments received: No.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: January 13, 2000.
Brief description of amendments: The amendments: (1) Revise Technical Specification 3.8.3 (Condition B and Surveillance Requirement (SR) 3.8.3.2) to increase the required emergency diesel generator (EDG) lube oil inventory values; (2) Revise SR 3.8.3.2, for EDG lube oil inventory, to add a note stating that the surveillance is not required to be performed until the diesel has been in shutdown greater than 10 hours; and (3) Delete the footnote associated with SR 3.8.4.7 which provided a ``one time only'' alternative to battery testing requirements.
Date of issuance: March 24, 2000.
Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
Amendment Nos.: 75 and 75.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65 FR 9012). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 24, 2000.
No significant hazards consideration comments received: No
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: January 13, 2000.
Brief description of amendments: The amendments add ``NOTE 3'' to Surveillance Requirement 3.3.1.10 to allow entry into MODES 2 or 1 without the performance of N-16 detector plateau verification until 72 hours after achieving equilibrium conditions at greater than or equal to 90 percent rated thermal power.
Date of issuance: March 24, 2000.
Effective date:E As of the date of issuance and shall be implemented within 30 days from the date of issuance.
Amendment Nos.: 76 and 76.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65 FR 9013)
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 24, 2000.
No significant hazards consideration comments received: No.
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TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: January 13, 2000.
Brief description of amendments: The amendments add ``NOTE 3'' to Surveillance Requirement 3.3.1.10 to allow entry into MODES 2 or 1 without the performance of N-16 detector plateau verification until 72 hours after achieving equilibrium conditions at greater than or equal to 90 percent rated thermal power.
Date of issuance: March 24, 2000.
Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.
Amendments Nos.: 76 and 76.
Facility Operating License Nos. NPF-84 and NPF-89: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65 FR 9013).
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated March 24, 2000.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri.
Date of application for amendment: December 3, 1999 (ULNRC-04158).
Brief description of amendment: The amendment changed Section 5.6.6, ``Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),'' of the improved Technical Specifications (ITS) that were issued on May 28, 1999, in Amendment No. 133. The current Technical Specifications (CTS) remain in effect until the ITS are implemented on or before April 30, 2000. The changes to the ITS approve the use of the PTLR by the licensee to make changes to the plant pressure temperature limits and low temperature over pressure protection limits without prior NRC staff approval, in accordance with Generic Letter 96-03, ``Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits,'' dated January 31, 1996. The changes (1) add the word criticality to ITS subsection 5.6.6.a as one of the reactor conditions for which RCS pressure and temperature limits will be determined, (2) add the phrase ``and COMS PORV,'' where COMS PORV stands for cold overpressure mitigation system power operated relief valve, to the introductory paragraph of ITS subsection 5.6.6.b to show that the analytical methods listed in the subsection are also the COMS PORV, and (3) replace the two documents listed in ITS subsection 5.6.6.b by the reference to the NRC letter that approves use of the PTLR and the Westinghouse topical report, WCAP-14040-NP-A, Revision 2, ``Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,'' dated January 1996, that provides the methodology that will be used by licensee in using the PTLR report. The current plant pressure temperature limits and low temperature overpressure protection limits are in the CTS and were approved in Amendment No. 124, which was issued April 2, 1998.
Date of issuance: March 24, 2000.
Effective date: March 24, 2000, to be implemented no later than April 30, 2000.
Amendment No.: 134.
Facility Operating License No. NPF-30: The amendment revised the Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64 FR 73101).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 24, 2000.
No significant hazards consideration comments received: No
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, Callaway County, Missouri
Date of application for amendment: January 14, 2000, as supplemented by letter dated February 17, 2000 (ULNRC-04172 and - 04187).
Brief description of amendment: The amendment revised several sections of the improved Technical Specification (ITSs) to correct 14 editorial errors made in either (1) the application dated May 15, 1997, (and supplementary letters) for the ITSs, or (2) the certified copy of the ITSs that was submitted in the licensee's letters of May 27 and 28, 1999. The ITSs were issued as Amendment No. 133 by the staff in its letter of May 28, 1999, and will be implemented by the licensee to replace the current TSs by April 30, 2000.
Date of issuance: March 27, 2000.
Effective date: March 27, 2000, to be implemented by April 30, 2000.
Amendment No.: 135.
Facility Operating License No. NPF-30: The amendment revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65 FR 9013) and February 25, 2000 (65 FR 10118).
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated March 27, 2000.
No significant hazards consideration comments received: No.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont.
Date of application for amendment: June 15, 1999, as supplemented on January 14, 2000.
Brief description of amendment: The amendment revises Technical Specifications (TSs) Sections 3.1/4.1 Reactor Protection System and 3.2/4.2 Protective Instrument Systems instrumentation, tables, and the associated Bases to increase the surveillance test intervals (STIs), add allowable out-of-service times (AOTs), replace generic emergency core cooling system actions for inoperable instrument channels with function-specific actions, and relocate selected trip functions from the TSs to a Vermont Yankee controlled document. In addition, revision to TS Section 3.1/4.1 Reactor Protection System and the associated Bases is proposed to remove the RUN Mode APRM Downscale/IRM High Flux/ Inoperative Scram Trip Function (APRM Downscale RUN Mode SCRAM). The submittal also proposes to implement editorial corrections and administrative changes that do not alter the meaning or intent of the requirements.
Date of Issuance: April 3, 2000.
Effective date: As of the date of issuance, and shall be implemented within 90 days.
Amendment No.: 186.
Facility Operating License No. DPR-28: Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: October 20, 1999 (64 FR 56535).
The Commission's related evaluation of this amendment is contained in a Safety Evaluation dated April 3, 2000.
No significant hazards consideration comments received: No
Dated at Rockville, Maryland, this 12th day of April 2000.
For the Nuclear Regulatory Commission. John A. Zwolinski, Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 00-9680Filed4-18-00; 8:45 am]
BILLING CODE 7590-01-P
