# Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/fr%3A99-21914 ## Record - **Collection:** Federal Register - **Document type:** Notice - **Published:** August 25, 1999 - **Citation:** 64 FR 46424 ## Text NUCLEAR REGULATORY COMMISSION Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations I. 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 July 31, 1999, through August 13, 1999. The last biweekly notice was published on August 11, 1999 (64 FR 43764). 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 [[Page 46425]] 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. Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administration 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 September 24, 1999, 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, the Gelman Building, 2120 L Street, NW., Washington, DC and at the local public document room for the particular facility involved. 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. 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 filed a 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 filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, or may be delivered to the Commission's 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 at the local public [[Page 46426]] document room for the particular facility involved. 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 amendment request: August 2, 1999. Description of amendment request: The proposed amendment would revise Technical Specification 6.2.2.e to require either the Operations Manager or an off-shift Operations superintendent to hold a senior reactor operator (SRO) license. This revision would delete the option which allows the Manager-Operations to have at one time held a Senior Reactor Operator License for a similar unit and replaces it with the requirement for an off-shift Operations superintendent who holds an SRO license to supervise shift work and licensed activities. 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: 1. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The change to Technical Specification 6.2.2.e to require the Manager-Operations or an off-shift Operations superintendent to hold an SRO license is administrative in nature and does not directly affect plant operations. The change does not physically alter the facility in any manner and, as such, does not affect the means in which any safety-related system performs its intended safety function. Therefore, there would be no increase in the probability or consequences of an accident previously evaluated. 2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. As stated above, the proposed change is administrative in nature. There is no physical alteration to any plant system, nor is there a change in the method in which any safety related system performs its function. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed amendment does not involve a significant reduction in the margin of safety. The proposed amendment does not reduce the margin of safety as defined in the Safety Analysis Report or the bases contained in the Technical Specifications. The requirement to have a licensed SRO management position responsible for plant operations is maintained within the proposed amendment. The proposed amendment is consistent with (1) 10 CFR 50.54(l), which requires individuals responsible for directing the licensed activities of licensed operators to hold an SRO license, (2) Revision 1 of NUREG-1431, ``Standard Technical Specifications Westinghouse Plants,'' and Technical Specification Traveler Form (TSTF) 65, Revision 1, and (3) the intent of ANSI/ANS- 3.1, ``Standard for Selection and Training of Personnel for Nuclear Power Plants,'' (September 1979 Draft). Therefore, the proposed 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. Local Public Document Room location: Cameron Village Regional Library, 1930 Clark Avenue, Raleigh, North Carolina 27605 Attorney for licensee: William D. Johnson, Vice President and Corporate Secretary, Carolina Power & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602 NRC Section Chief: Sheri R. Peterson. 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 amendment request: August 4, 1999. Description of amendment request: The proposed amendment would revise Technical Specification 6.9.1.6.2 to incorporate analytical methodology references which are used to determine core operating limits. The analytical methodologies to be referenced are documented in topical reports which have been accepted by the Nuclear Regulatory Commission for referencing in licensing applications. 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: 1. The proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes incorporate additional references to methodologies used to evaluate core operating limits. These methodologies have been approved for use by the NRC. Plant structures, systems, and components will not be operated in a different manner as a result of these proposed changes and no physical modifications to equipment are involved. Adding these references to the Core Operating Limits Report section of Technical Specifications does not increase the probability or consequences of an accident previously evaluated. 2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes incorporate additional references to methodologies used to evaluate core operating limits. These methodologies have been approved for use by the NRC. Plant structures, systems, and components will not be operated in a different manner as a result of these proposed changes and no physical modifications to equipment are involved. Adding these references to the Core Operating Limits Report section of Technical Specifications does not create the possibility of a new or different type of accident from any previously evaluated. 3. The proposed amendment does not involve a significant reduction in the margin of safety. The proposed changes incorporate additional references to methodologies used to evaluate core operating limits. These methodologies have been approved for use by the NRC. Plant structures, systems, and components will not be operated in a different manner as a result of these proposed changes and no physical modifications to equipment are involved. Adding these references to the Core Operating Limits Report section of Technical Specifications does not involve a 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. Local Public Document Room location: Cameron Village Regional Library, 1930 Clark Avenue, Raleigh, North Carolina 27605. Attorney for licensee: William D. Johnson, Vice President and Corporate Secretary, Carolina Power & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602. NRC Section Chief: Sheri R. Peterson. Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois Date of amendment request: May 20, 1999. Description of amendment request: The proposed amendments would revise Technical Specification (TS) 3.8.A to identify the specific Containment Cooling Service Water (CCSW) equipment required to support operation of the Control Room Emergency Ventilation System (CREVS). The proposed amendment would also [[Page 46427]] revise TS 3/4.5.C.2 to ensure that the suppression pool water level is adequate to prevent vortexing in the Low Pressure Coolant Injection and Core Spray pump suctions. 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: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: The proposed changes to the technical specifications provide clarity in the support system relationship and requirements for the CCSW system support of the CREVS operation. [Neither] [t]he CCSW system nor the CREVS system are assumed to be accident precursors for previously evaluated accident[s]. Therefore, the proposed changes have no effect on the probability or consequences of accidents previously evaluated. The proposed change to the allowable suppression chamber level does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change revises a Technical Specification acceptance value to [a] more conservative value and serves to ensure operability of equipment important to safety. By ensuring equipment availability, the probability or consequences of an accident previously evaluated are not increased. In addition, the proposed changes have no impact on any initial condition assumptions for accident scenarios. Onsite or offsite dose consequences resulting from an event previously evaluated are not affected by this proposed amendment request. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated because: The proposed changes do not create the possibility of a new or different kind of accident from that previously evaluated. The changes to the CCSW specifications more appropriate[ly] reflect the design requirements and clarify the support role of the CCSW system as it relates the CREVS. Neither the CCSW system nor the CREVS will be operated differently with the proposed change. Therefore new or different failure modes will not be created. Therefore, the possibility of new and different accidents has not been created with the proposed change. The proposed change to the suppression pool allowable level restores margin to the Technical Specifications and ensures equipment operability. The proposed change is conservative with respect to current requirements. The proposed amendment does not involve any plant physical changes that would create the possibility of a new or different kind of accident from any accident previously evaluated. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Involve a significant reduction in the margin of safety because: The proposed change to the CCSW technical specification will not result in a significant reduction in the margin of safety. The proposed change has greater consistency with the current design requirements for CSSW support of CREVS operation. Therefore, the margin of safety has been not been altered. [Therefore, the margin of safety has not been altered. SIC] The proposed changes for suppression pool level does not involve a significant reduction in a margin of safety. In fact, the proposed changes restore margin and ensure equipment operability. Since the changes maintain the necessary level of system reliability, they do not involve a significant reduction in the margin of safety. The proposed amendment for Dresden will not reduce the availability of systems required to mitigate accident conditions; therefore, the proposed 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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration. Local Public Document Room location: Morris Area Public Library District, 604 Liberty Street, Morris, Illinois 60450. 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. Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Date of amendment request: July 14, 1999. Description of amendment request: The proposed amendments would allow the units to operate at an uprated power level of 3489 MWt, an increase of 5 percent rated core thermal power. 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 change involve a significant increase in the probability or consequences of an accident previously evaluated? A. Evaluation of the Probability of Previously Evaluated Accidents The proposed power uprate imposes only minor increases in plant operating conditions. No change is made to the reactor operating pressure. Operation at uprated conditions will result in moderate flow increases in those systems associated with the turbine cycle in that steam flow increases by approximately six (6)% and feed flow increases by approximately six (6)%. The increase in flow in the carbon steel piping systems was evaluated for the effect on flow induced erosion and corrosion rates and it was confirmed that power uprate has no significant effect on flow induced erosion or corrosion. The affected systems are currently monitored by the Flow Accelerated Corrosion (FAC) program that addresses erosion and corrosion concerns. Continued monitoring of the systems provides a high level of confidence in the integrity of potentially susceptible high energy piping systems. Plant systems and components have been verified to be capable of performing their intended design functions at uprated power conditions. Where necessary, some components will be modified prior to implementation of uprated power conditions to accommodate the revised operating conditions. The review has concluded that operation at power uprate conditions will not affect the reliability of plant equipment, and that current Technical Specifications (TS) surveillance requirements ensure adequate monitoring of system operability. Systems continue to be operated in accordance with current design requirements under uprated conditions, therefore no new components or system interactions were identified that could lead to an increase in accident probability. Changes to reactor scram setpoints are such that no significant increase in scram frequency due to operation at uprated conditions will occur. B. Evaluation of the Consequences of Previously Evaluated Accidents The radiological consequences due to the Loss of Coolant Accident (LOCA) were calculated and are found to be below the applicable regulatory limits. The results are presented in Table 9-3 of Attachment E [of the July 14, 1999 submittal]. The LOCA radiological consequences have not significantly increased due to power uprate, and radiological consequences continue to meet established regulatory limits. The radiological evaluations for other non-LOCA Design Basis Accidents (DBAs) were also performed and the dose consequences for these events did not significantly increase. These changes are outlined in Section 9.2 of Attachment E and they demonstrate that LaSalle County Station (LCS), Units 1 and 2 still meets the applicable regulatory limits. Non-DBA Radiological Doses All of the other radiological releases discussed in Updated Final Safety Analysis Report (UFSAR) are either unchanged because they are not power-dependent, or increase approximately in linear proportion to the amount of the uprate. The dose consequences for all of the non-LOCA radiological release accident events did not significantly increase, and are bounded by the ``LOCA Radiological Consequences'' [[Page 46428]] events discussed above and were shown to meet the current dose acceptance limits. These events are discussed in Section 9.2 of Attachment E. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The configuration, operation and event response of the LCS, Units 1 and 2 systems, structures or components are [unchanged] by operation at uprated power conditions. Analysis of transient events has confirmed that the same transients remain limiting and that no transient event results in a new sequence of events that could lead to a new accident scenario. An increase in power level will not create a new fission product release path, or result in a new fission product barrier failure mode. The current fission product barriers consisting of the reactor fuel rod cladding, the reactor coolant pressure boundary, and the containment structure remain in place. Fuel rod cladding integrity is ensured by operating within thermal, mechanical, and exposure design limits, and was confirmed for a representative core by performance of transient and accident analysis. Cycle specific analysis will continue to be performed for each fuel reload to demonstrate the compliance with the applicable transient analysis criteria and to establish the cycle specific Minimum Critical Power Ratio (MCPR) safety limit and fuel operating limits. The integrity of the reactor coolant pressure boundary was confirmed by evaluation of the bounding overpressurization event and ensuring that the corresponding pressure remained below the American Society of Mechanical Engineers (AMSE) Boiler and Pressure Vessel (B&PV) Code, Section III, ``Rules for Construction of Nuclear Power Plant Components,'' overpressure protection requirements. Similarly, analysis of the primary containment structure has demonstrated under worst case design basis accident conditions that the containment structure remains below the containment design pressure. The effect of operation at uprated conditions on plant equipment has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified as a result of operating at uprated conditions. In addition, operation at power uprated conditions does not create any new sequence of events or failure modes that lead to a new type of accident. Plant modifications required to support implementation of power uprated conditions will be made to existing systems rather than by adding new systems of a different design, which might introduce new failure modes or accident sequences. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. Does the change involve a significant reduction in a margin of safety? The power uprate analysis for LCS, Units 1 and 2 assures that the power dependent safety margin will be maintained by meeting the appropriate regulatory criteria as prescribed by the applicable regulations. Similarly, factors of safety specified by application of the regulatory required design rules have been maintained, as have other acceptance criteria used to judge the acceptability of current plant operation. No change is required in the basic duel deign to achieve the uprated power levels, or to maintain current operating and safety margins. No increase in the allowable peak bundle power is requested as a result of operation at uprated conditions. The abnormal transients have been evaluated for a representative core configuration and confirmed that operation at uprated conditions does not have an adverse effect on the operating limit MCPR. No change to the Safety Limit MCPR results, thus the margin of safety as assured by the safety limit MCPR is maintained. The fuel operating limits related to heat generation rate would still be met at uprated conditions. Cycle specific analysis will continue to be performed for each fuel reload to demonstrate the compliance with the applicable transient analysis criteria and to establish the cycle specific safety limit and fuel operating limits. The Emergency Core Cooling System (ECCS)-LOCA performance has been evaluated at power uprated conditions using methodologies that have been approved by the NRC for 10CFR50.46, ``Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors,'' analysis. The current ECCS performance requirements were used in the power uprate analysis. The ECCS-LOCA analysis was conducted at 102% of the proposed uprated thermal power in accordance with regulatory guidance. The necessary analysis for operation of General Electric (GE) fuel under uprated conditions and the determination that the peak cladding temperature (PCT) remains below the 10CFR50.46 limit of 2200 deg.F have been performed. However, LCS Unit 2 currently contains a mixed core of GE and Siemens Power Corporation (SPC) fuel. LCS obtained [a] TS amendment that allows operation with SPC fuel, and approved the use of the SPC analytical methodology. The ECCS-LOCA analysis performed to support use of the SPC fuel was conducted at a power level that bounds 102% of the proposed uprated power level and determined that the PCT, for SPC fuel, remains below the 10CFR50.46 limit of 2200 deg.F. The analysis for both GE and SPC fuel types demonstrate all 10CFR50.46 criteria are met. Therefore, there is no reduction in margin with respect to maintaining ECCS performance. The margin of safety of the reactor coolant pressure boundary is maintained under power uprated conditions. The design pressure of the RPV and reactor pressure coolant pressure boundary remains at 1250 psig. The ASME B&PV Code allowable peak pressure is 1375 psig (i.e., 110% of design value), which is the acceptance limit for pressurization events. The limiting pressurization event is a Main Steam Isolation Valve (MSIV) closure with a failure of valve position scram and this event results in a calculated peak RPV pressure of 1332 psig at the bottom of the RPV. The peak pressure remains below the 1375 psig ASME limit. Therefore, there is no decrease in margin of safety in the reactor coolant pressure boundary. The margin of safety of the containment structure is maintained under power uprated conditions. The analyses were conducted using a newer NRC-reviewed methodology. The pre-uprated cases were run using the new methodology and the re-baselined cases were compared to the uprated cases. The short-term containment peak pressure analysis re- baseline result was 39.3 psig compared to the original analysis of 39.6 psig. At uprated conditions the peak containment drywell pressure would be 39.9 psig, and is below the design value of 45 psig. The long-term containment suppression pool temperature analysis re-baseline result was 190 deg.F compared to the original analysis result of 200 deg.F. At uprated conditions the analysis concluded that in the event of a LOCA, the calculated peak bulk suppression pool temperature would be 193 deg.F. This is less than the design temperature of the suppression pool of 275 deg.F, and the criteria used to ensure adequate Net Positive Suction Head (NPSH) to the ECCS pumps which is 212 deg.F. Therefore, power uprate does not challenge the structural integrity of the containment structure and ECCS NPSH is assured. Therefore, operation at power uprated conditions 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 requested amendments involve no significant hazards consideration. Local Public Document Room location: Jacobs Memorial Library, 815 North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, Illinois 61348-9692 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. Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois Date of amendment request: August 6, 1999 Description of amendment request: The proposed amendments would revise Technical Specification 3/4.6.4, ``Vacuum Relief'' to remove specific operability requirements related to position indication for the suppression chamber-drywell vacuum breakers. The amendments also reformat the action statements for inoperable vacuum breakers, increase the surveillance [[Page 46429]] interval for verifying that the vacuum breakers are closed, and delete the requirement to verify that the manual isolation valves are closed for an inoperable and open vacuum breaker. 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: Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed changes do not change the hardware configuration of the suppression chamber-drywell vacuum breakers, and the vacuum breakers are not considered an initiator in any accident scenario. The removal of specific indication requirements and the extension of the surveillance interval does not impact the ability of the vacuum breakers to perform their safety function. The vacuum breakers continue to meet their intended design function. The proposed changes do not impact the assumed source term for any analyzed accident. Therefore, no increases in the probability of an accident or consequences will result due to this proposed change. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes do not involve any physical alterations to the suppression chamber-drywell vacuum breakers, or cause any changes in the method by which the vacuum breakers or the containment vacuum relief system performs their associated design basis functions. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. Do the proposed changes involve a significant reduction in a margin of safety? The proposed changes do not impact the design function assumed for the containment vacuum relief system. The proposed changes do not require the vacuum breakers to operate in a condition not previously assumed in the facility accident analysis. The containment vacuum relief system will continue to operate and provide the protection assumed in the accident analysis. In order to limit bypass, the vacuum breakers are in a normally closed position. These vacuum breakers cannot be permanently placed in the open position. The proposed decrease in the surveillance frequency verifying the closed vacuum breakers will not increase the risk of the vacuum breakers being in the open position, since they will only open in response to a pressure differential or manual cycling. Therefore, the assurance of the operability of the containment vacuum breakers would be the same as provided under current Technical Specifications. The containment response analysis is unchanged, in that the vacuum breakers protect the containment structure, the peak containment pressure remains as calculated, and the vacuum breakers continue to maintain bypass leakage rates as assumed. Therefore this proposed change does not cause a 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 requested amendments involve no significant hazards consideration. Local Public Document Room location: Jacobs Memorial Library, 815 North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, Illinois 61348-9692. 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. Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois Date of amendment request: July 16, 1999. Description of amendment request: The proposed change to Technical Specification Section 3/4.7.D is to eliminate the limit for any one main steam line isolation valve (MSIV) leakage of less than or equal to 11.5 standard cubic feet per hour (scfh), and to replace that with an aggregate value of less than or equal to 46 scfh for all four MSIVs. 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 change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed changes to the Technical Specifications, Appendix A, modifies the allowed leakage limit to an aggregate value with no change to the total allowed leakage rate. This change does not affect either the automatic or manual features that would close the MSIVs. There are no physical changes to the plant and plant operations remain unchanged. Therefore, this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The safety function of the MSIVs is to provide a timely steam line isolation to mitigate the release of radioactive steam and limit reactor inventory loss under certain accident and transient conditions. The MSIVs are designed to automatically close whenever plant conditions warrant main steam line isolation. Changing the leakage limits to include an aggregate value does not affect the isolation function. No new equipment will be installed or utilized, and no new operating conditions will be initiated as a result of this change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. Does the change involve a significant reduction in a margin of safety? The total allowed leakage rate for all MSIVs remains unchanged at 46 scfh. Therefore, there will be no change in the types or significant increase in the amounts of any effluents released offsite, and, thus, the radiological analyses remain unchanged and within the guidelines of 10 CFR 100 and General Design Criteria 19. Therefore, these 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 standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration. Local Public Document Room location: Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021. 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. Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina Date of amendment request: July 27, 1999. Description of amendment request: The proposed amendments would add a surveillance requirement to verify the Keowee out-of-tolerance logic trips and blocks closure of the appropriate overhead or underground power path breakers. This logic is being added as part of a modification to provide voltage and frequency protection for the Keowee Hydro Units to protect them from being exposed to out-of-tolerance voltage and frequency. 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: [[Page 46430]] 1. Involve a significant increase in the probability or consequences of an accident previously evaluated: This change does not create any conditions or events, which lead to accidents previously, evaluated in the SAR. The Keowee Hydro units are used for mitigation of loss of power scenarios. The proposed changes do not change the current function of the Keowee Hydro Units. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated. The Keowee Hydro units and their role in the Oconee emergency power system currently meet the design/licensing basis requirements for the system. There is no adverse affect on containment integrity and no new release paths are created. The proposed changes do not cause any adverse effects to the Keowee single failure design or adversely affect the Keowee start time of 23 seconds. Therefore, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated: The Keowee Hydro units are used for mitigation of loss of power scenarios. No accidents new or different than already evaluated in the SAR are postulated as a result of the proposed change. No setpoints for parameters, which initiate protective or mitigative action, are being changed. Therefore, this proposed amendment does not create the possibility of any new or different kind of accident. 3. Involve a significant reduction in a margin of safety: The proposed change does not adversely affect any plant safety limits, set points, or design parameters. The change also does not adversely affect the fuel, fuel cladding, Reactor Coolant System, or containment integrity. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Duke has concluded, based on the above, that there are no significant hazards considerations involved in this amendment request. 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. Local Public Document Room location: Oconee County Library, 501 West South Broad Street, Walhalla, South Carolina. Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 17th Street, NW., Washington, DC. NRC Section Chief: Richard L. Emch, Jr. Energy Northwest, (formerly known as the Washington Public Power Supply System), Docket No. 50-397, WNP-2, Benton County, Washington Date of amendment request: July 29, 1999. Description of amendment request: The proposed amendment would change the applicability of Section 3.4.9 of the Technical Specifications (TS) from ``Mode 3 with steam drum pressure less than the RHR [residual heat removal] cut in permissive'' to ``Mode 3 with steam drum pressure less than 48 psig.'' Notes associated with TS Surveillance Requirements 3.4.9.1 and 3.5.1.2 would be changed to reflect the proposed 48 psig limit. 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: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. This change involves further restrictions on the use of RHR in the shutdown cooling mode of operation during hot shutdown conditions. Chapter 15 of the FSAR [Final Safety Analysis Report] defines the start of hot shutdown as the point when generated power is below one percent rated power. During entry into hot shutdown conditions the RHR system will be aligned in the Low Pressure Coolant Injection (LPCI) mode of operation. Thus, it will be aligned to provide water to the Reactor Pressure Vessel in the event the high pressure systems (HPCS and RCIC) are not able to perform this function. The change being proposed here has no impact on loss of coolant accidents (LOCAs) requiring mitigation using RHR aligned in the LPCI mode of operation. During the high pressure portion of the hot shutdown condition, intersystem (LOCAs) are a concern. The purpose of the RHR SDC Isolation Reactor Pressure--High (cut-in permissive) at 135 psig is to prevent over-pressurization of portions of the RHR system. This protection is not being modified by this change. The instrumentation that provides this protection will continue to function as designed. This change only impacts the applicability of Technical Specification 3.4.9 and when RHR SDC is required to be operable. During hot shutdown the reactor is normally cooled down through use of the main steam system and the condenser. Other means of cooling are also available using the reactor water cleanup system or a combination of emergency core cooling system (ECCS) pumps and safety relief valves (SRVs). The RHR system aligned in the SDC mode is used at the end of this cooling process to reach cold shutdown conditions of less than or equal to 200 deg.F. The change being proposed results in the RHR SDC being manually initiated at a lower pressure and temperature. This change will have no significant impact on the capability to cool the reactor. FSAR Chapter 15, ``Accident Analysis,'' describes two events associated with the RHR system. FSAR section 15.1.6, ``Inadvertent Residual Heat Removal Shutdown Cooling Operation,'' describes the impact of system operation during startup or cool-down when the reactor is near critical. The proposed change involves the point at which RHR is started in the SDC mode with the reactor sub-critical with control rods inserted. Therefore, there will be no change in the probability or consequences of this accident. FSAR section 15.2.9, ``Failure of Residual Heat Removal Shutdown Cooling,'' describes the failure of the RHR system to function in SDC mode. This evaluation assumes a failure of the SDC mode of operation but does not disable the remaining modes of RHR operation. The alternate shutdown cooling paths involve the use of the SRVs [safety relief valves] to establish a cooling path to the containment suppression pool. This evaluated accident does not result in any fuel failure. The proposed change will not result in any fuel failures. The evaluated accident does result in normal coolant activity being released to the suppression pool through the safety relief valves. The proposed activity will not result in a significant change in the release of this coolant activity. The proposed change will not cause a significant increase in the probability of a loss of SDC accident. This change proposes a delay in the use of SDC because of temperature limitations. During this time other means of decay heat removal would be used. This will result in a decrease in use of RHR in SDC mode and a decrease in the probability of failure of the system by restricting operation to be within analyzed temperature limits. The proposed change will not involve a significant increase in the consequences of the loss of shutdown cooling accident. The accident evaluated in the FSAR assumes SDC does not operate at any time and alternate means of cooling are evaluated. Section 15.2.9.6 states there is no fuel failure and release is limited to normal primary coolant activity to the suppression pool. The proposed change results in a short delay in the use of SDC because of temperature limitations. The accident described in FSAR section 15.2.9 bounds this condition and, as a result, there will be no increase in accident consequences. With multiple means of reactor water makeup and heat removal available the restriction in the use of RHR caused by this change will not result in a significant increase in the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change will not cause any new inadvertent shutdown cooling startup, loss of water inventory or loss of cooling accidents. New or different inadvertent RHR SDC startup accidents are not possible because this change is only a further restriction on when the system is operated. The LOCA accidents during Mode 3 are [[Page 46431]] bounded by the LOCAs defined for Modes 1 and 2. No new primary sytem LOCAs can be initiated because of this change. The purpose of the RHR cut-in permissive at 135 psig is to prevent overpressurization of portions of the RHR system that could cause an intersystem LOCA. This change will not result in a new or different kind of intersystem LOCA because this is only a further restriction on RHR SDC operation. The use of RHR in the SDC mode is restricted to operation at a lower pressure and temperature but other systems are available to remove the decay heat. No new or different accidents are created because of this change. The FSAR section 15.2.9 accident, ``Failure of Resident Heat Removal Shutdown Cooling,'' is bounding for all other accidents which postulate failure of the capability to remove decay heat. No additional accidents resulting in the loss of decay heat removal capability will be caused by this change. Therefore, the operation of WNP-2 in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed change does not involve a significant reduction in a margin of safety. The proposed amendment will increase the reliability of the RHR system when operated in shutdown cooling mode by providing assurance that the temperature limits of the piping and pipe supports will not be exceeded. The ability to protect against an intersystem LOCA is unchanged. The ability to remove decay heat from the reactor is not changed by this modification as alternate means of heat removal are available. Therefore, operation of WNP-2 in accordance with the proposed amendment will not involve a 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Richland Public Library, 955 Northgate Street, Richland, Washington 99352. Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502. NRC Section Chief: Stephen Dembek. Energy Northwest (formerly known as the Washington Public Power Supply System), Docket No. 50-397, WNP-2, Benton County, Washington Date of amendment request: July 29, 1999. Description of amendment request: The proposed amendment would revise Technical Specification Table 3.3.5.1-1, ``Emergency Core Cooling System (ECCS) Instrumentation Items 1.a, 2.a, 4.a and 5.a,'' to change the Reactor Vessel Water Level--Low Low Low, Level 1 allowable value from the current value of -148 inches to a new value of -142.3 inches. 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: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. This change involves the measurement of water level in the Reactor Pressure Vessel (RPV) used to initiate the ECCS. The accident evaluated for this condition is the spectrum of loss of coolant accidents (LOCA) severe enough to decrease the RPV water inventory by a significant amount. The additional uncertainty introduced because of harsh environmental effects could not be accommodated between the existing Technical Specification allowable value and the analytical limit. This uncertainty results in a requirement that the ECCS be initiated at a slightly higher water level than previously calculated. Therefore, operation of WNP-2 in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change will not create a new or different kind of accident since it only makes a small change in the RPV water level at which the ECCS is initiated. This change is in the conservative direction requiring a greater volume of water in the RPV to accommodate the uncertainty associated with the harsh environment of the water level sensors. The level indicating switches are located on instrument racks in the Reactor Building. The harsh environment in this building would have no impact on the initial trip needed to initiate the ECCS on loss of RPV level since conditions in the Reactor Building would be benign at the initial stages of the accident. Only if the Level 1 trip was reset and initiated after a significant period of time would the harsh environmental conditions have an impact on the accuracy of the level indicating switches. However, increasing the water level at which the ECCS is initiated results in a more conservative value that adequately includes post-accident harsh environment uncertainties and ensures that the associated analytical limit is met. Therefore, the operation of WNP-2 in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed change does not involve a significant reduction in a margin of safety. The proposed amendment increases the allowable value for water level in the RPV. This small increase will result in an increase in the margin of safety. A review of the plant settings for the Level 1 trip indicated that previous settings were within the new allowable value. Therefore, operation of WNP-2 in accordance with the proposed amendment will 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Richland Public Library, 955 Northgate Street, Richland, Washington 99352. Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502. NRC Section Chief: Stephen Dembek. Energy Northwest (formerly known as the Washington Public Power Supply System), Docket No. 50-397, WNP-2, Benton County, Washington Date of amendment request: July 29, 1999. Description of amendment request: The proposed amendment would revise Technical Specification Surveillance Requirement (SR) 3.5.2.2. This requirement verifies the adequacy of the water supply in the condensate storage tanks (CSTs) which support operation of the high pressure core spray (HPCS) system during Modes 4 and 5. Current Technical Specification SR 3.5.2.2 requires that CST water level be maintained above 13.25 feet in a single tank or above 7.6 feet in each tank if the suppression pool level is below its minimum level. It is proposed that the CST water level be maintained above 14.8 feet in a single tank or above 9.1 feet in each tank. 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: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. During Modes 4 and 5 HPCS may be required to provide water to the reactor vessel if the water level decreases. The revised condensate storage tank allowable levels increase the operating margins by providing an increased water inventory. The previously evaluated accident involving the loss of decay heat cooling inventory will not have an increase in probability because the inventory of water will be increased with the change being proposed. [[Page 46432]] The consequences of any accident involving the loss of decay heat cooling inventory will not change as the consequences are unaffected by the increased water inventory. Therefore, operation of WNP-2 in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change will not create a new or different kind of accident since it only increases the amount of water held in reserve to support reactor vessel inventory loss. The proposed change does not introduce any credible mechanisms for unacceptable radiation release nor does it require physical modification to the plant. The inventory of water in the CSTs will increase to support any loss of water inventory in the reactor vessel during shutdown. The proposed change modifies the monitored values for CST level. The plant has operated well within the existing allowable values. The increased margin provided by the increased level will assure no new or different kinds of accidents result from the proposed change. Therefore, the operation of WNP-2 in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed change does not involve a significant reduction in a margin of safety. The proposed amendment increases the allowable value for water level in the CSTs. This results in an increase in the inventory of water available for cooling and inventory control during reactor shutdown. This will result in an increase in the margin of safety. Therefore, operation of WNP-2 in accordance with the proposed amendment will 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Richland Public Library, 955 Northgate Street, Richland, Washington 99352. Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502. NRC Section Chief: Stephen Dembek. Energy Northwest (formerly known as the Washington Public Power Supply System), Docket No. 50-397, WNP-2, Benton County, Washington Date of amendment request: July 29, 1999. Description of amendment request: The proposed amendment request would revise Technical Specification Surveillance Requirement (SR) SR 3.8.4.6 of Technical Specification 3.8.4, ``DC Sources--Operating,'' and SR 3.8.5.1 of Technical Specification 3.8.5, ``DC Sources-- Shutdown.'' The proposed change to SR 3.8.4.6 would prohibit surveillance testing of Division 1, 2, and 3 125 and 250 volt DC, battery charger capacity during Modes 1, 2, and 3. However, credit could be taken for unplanned events that satisfied the surveillance requirement. The proposed change to SR 3.8.5.1 would include SR 3.8.4.6 as one of the surveillance tests that are not required to be performed. 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: 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change has no impact on previously analyzed accidents or transients, and has no effect on operation, capacity or surveillance test details of the DC system battery chargers. The change only imposes a mode restriction on performance of specified surveillance testing and allows taking credit for unplanned events that satisfy the surveillance. Therefore, operation of WNP-2 in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change has no effect on operation, capacity, or surveillance test details of the DC system battery chargers. The change only prohibits performing specified battery charger capacity surveillance testing from being implemented during Mode 1, 2, or 3 and allows taking credit for unplanned events that satisfy the surveillance. The proposed change to SR 3.8.4.6 of Technical Specification 3.8.4 and SR 3.8.5.1 of Technical Specification 3.8.5 are consistent with the wording previously evaluated and approved by the NRC in NUREG-1434 Rev. 1. Therefore, operation of WNP-2 in accordance with the proposed amendment will not create the possibility of a new or different kind of accident previously evaluated. 3. The proposed change does not involve a significant reduction in a margin of safety. The proposed change only imposes a mode restriction, prohibiting battery charger capacity surveillance testing from being performed during Modes 1, 2, and 3, allowing credit to be taken for unplanned events that satisfy the surveillance, and allowing such testing to be omitted under certain conditions during Modes 4 and 5 and during movement of irradiated fuel in secondary containment. Performance of this testing would remove a DC electrical power subsystem from service and could present a safety risk were an event to occur if the testing was performed in Modes 1, 2, and 3, or while DC service is required in other operating conditions. Therefore, operation of WNP-2 in accordance with the proposed amendment will 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 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Richland Public Library, 955 Northgate Street, Richland, Washington 99352. Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502. NRC Section Chief: Stephen Dembek. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of amendment request: July 20, 1998, as supplemented June 29, 1999. Description of amendment request: The amendment would incorporate the Technical Specification changes necessary for implementation of the Boiling Water Reactor Owners' Group Reactor Stability Long-Term Solution, Enhanced Option 1-A (E1A). E1A consists of modifications to the plant operating procedures and associated plant components that provide a means for reliably detecting and avoiding reactor instabilities. By letter dated February 25, 1998, the Nuclear Regulatory Commission (NRC) staff recognized E1A as a technically acceptable implementation of a long-term stability solution satisfying the requirements of NRC IE Bulletin 88-07, Supplement 1, and Generic Letter 94-02, ``Long Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors.'' 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: [[Page 46433]] 1. This request does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendments allow the implementation of the Enhanced Option I-A (E1A) long term solution to the neutronic/thermal- hydraulic instability issue. Current Technical Specification (TS) restrictions on power and flow conditions, number of operating recirculation loops and operator actions implemented to reduce the probability of neutronic/thermal-hydraulic instability are eliminated and new stability requirements consistent with NEDO- 32339-A, Supplement 4, Revision 1, are imposed. These requirements include restrictions on power and flow conditions and actions associated with the modified Average Power Range Monitor (APRM) flow biased scram and control rod block functions. Required actions include adherence to the boiling boundary limit stability control prior to entry and during operation in the region of the power and flow operating domain which is potentially susceptible to neutronic/ thermal-hydraulic instability in the absence of the stability control. In addition, the proposed amendments require operator actions based upon control room indications generated by a new Period Based Detection System (PBDS). The PBDS is designed to provide alarm indication that conditions consistent with a significant degradation in the stability performance of the reactor has occurred and the potential for imminent onset of neutronic/ thermal-hydraulic instability may exist. The PBDS also provides analog indication of the highest and second highest successive period confirmation count of all of the Local Power Range Monitors (LPRMs) monitored. This provides the plant operators with continuous indication of reactor stability operating conditions. The proposed amendments will permit operation in regions of the power and flow operating domain postulated to be susceptible to neutronic/thermal-hydraulic instability. Operation in these regions does not increase the probability of occurrence of initiators and precursors of previously analyzed accidents when neutronic/thermal- hydraulic instability is not possible. The proposed amendments permit the implementation of the features of the E1A solution which prevent neutronic/thermal-hydraulic instability including preemptive reactor scram upon entry into the regions of the power and flow operating domain most susceptible to neutronic/thermal-hydraulic instability. The E1A solution also requires implementation of stability control prior to entry into a region of the power and flow operating domain which is potentially susceptible, in the absence of stability control, to neutronic/thermal-hydraulic instability. The E1A solution prevents neutronic/thermal-hydraulic instability during operation in regions of the power and flow operating domain previously excluded from operation and therefore does not significantly increase the probability of a previously analyzed accident. Operation in the regions of the power and flow operating domain excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a result of anticipated operational occurrences. The severity of these transients may increase in the absence of operator actions due to the potential occurrence of neutronic/thermal-hydraulic instability as a result of operation in these regions. The proposed amendments will permit the implementation of the E1A long term solution to the stability issue. Required features of the E1A solution include adherence to a boiling boundary limit stability control prior to selection by the operator of APRM flow biased scram and control rod block function ``Setup'' setpoints which allow operation in a region of the power and flow operating domain potentially susceptible, in the absence of the stability control, to neutronic/thermal-hydraulic instability. Upon entry, as a result of an anticipated operational occurrence, into the region most susceptible to neutronic/thermal- hydraulic instability, the preemptive reactor scram prevents neutronic/thermal-hydraulic instability. Therefore, the consequences of an accident do not significantly increase while operating with the stability control met. After exiting the region requiring the stability control to be met, the setpoints can be manually reset to their normal values. Stability controls are required to be in place when setpoints are ``Setup''. As a backup E1A feature, the APRM flow biased setpoints automatically reset to their normal values above a pre-determined flow condition. This automatic reset to the more conservative setpoints ensures that the preemptive reactor scram will prevent operation as a result of an anticipated operational occurrence into the region most susceptible to neutronic/thermal-hydraulic instability should the operator not select the more conservative setpoints appropriate for operation following exit from the region requiring stability control. Other required E1A features, including the PBDS, control rod block alarms associated with entry into the region susceptible to neutronic/thermal-hydraulic instabilities in the absence of stability controls, and required operator actions, including manual reactor scram, help ensure prevention of neutronic/thermal-hydraulic instabilities. Therefore, the proposed amendments prevent the occurrence of neutronic/thermal-hydraulic instability as a consequence of an anticipated operational occurrence and do not significantly increase the consequences of any previously analyzed accident. 2. This request does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendments replace current restrictions on power and flow conditions with alternative restrictions which permit the implementation of the E1A long term stability solution. The current restrictions on the power and flow conditions and operating recirculation loops in the RUN mode do not automatically prevent the entry into regions of the power and flow operating domain most susceptible to neutronic/thermal-hydraulic instability and therefore the possibility of neutronic/thermal-hydraulic instability exists in the absence of operator action. The required features of the E1A solution implement a preemptive scram upon entry into the region most susceptible to neutronic/thermal-hydraulic instability, without operator action. The accessible operating domain allowed by the proposed amendments is a subset of the power and flow operating domain currently allowed. Current initiators and precursors of accidents and anticipated operational occurrences [cannot] occur with new or different initial conditions as a result of this change. Additionally, there are no new event initiators or precursors of accidents and anticipated operational occurrences created by this change. Therefore, the proposed amendments do not create the possibility of a new or different kind of accident from that previously evaluated. Concurrent with the implementation of the proposed amendments, a modified Flow Control Trip Reference (FCTR) card, the E1A FCTR card, and a new Period Based Detection System (PBDS) will be installed as required by the E1A solution. The function of the E1A FCTR card is to aid the operator in the identification of entry into regions of the power and flow operating domain potentially susceptible to neutronic/thermal-hydraulic instability in the absence of stability controls and to initiate a preemptive scram upon entry into the regions most susceptible to neutronic/thermal-hydraulic instability. This is accomplished by altering the existing values of setpoints of the APRM flow biased scram and the control rod block functions generated by the E1A FCTR card. The E1A FCTR card design includes components which may be susceptible to electromagnetic interference or other environmental effects. The plant specific environmental conditions (temperature, humidity, pressure, seismic, and electromagnetic compatibility) have been confirmed to be enveloped by the environmental qualification values for the E1A FCTR cards. Therefore, the potential for spurious scrams or common mode failures induced by environmental effects (e.g., electromagnetic interference) is considered negligible. The installation of the E1A FCTR card will therefore not create the possibility of a new or different kind of accident from any accident previously evaluated. The function of the PBDS is to provide the operator with an indication that conditions consistent with a significant degradation in the stability performance of the reactor has occurred and the potential for imminent onset of neutronic/thermal-hydraulic instability may exist. This is accomplished by the installation of a new PBDS card in the Neutron Monitoring System. The PBDS card takes inputs from individual local power range monitors and provides analog indication of the highest and second highest successive period confirmation count, provides a High Decay Ratio (Hi DR) and High-High Decay Ratio (Hi-Hi DR) alarms, and INOP status indication to the operator in the control room. These displays [cannot] create the possibility of a new or different kind of accident from any accident previously evaluated. The PBDS card design includes components which may be susceptible to electromagnetic interference or other environmental effects. However, the plant specific environmental conditions (temperature, humidity, pressure, seismic, [[Page 46434]] and electromagnetic compatibility) have been confirmed to be enveloped by the PBDS environmental qualification values. Therefore, the installation of the PBDS card will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. This request does not involve a significant reduction in a margin to safety. The proposed amendments permit the implementation of the E1A long term solution to the stability issue. Under certain conditions, existing BWR [boiling water reactor] designs are susceptible to neutronic/thermal-hydraulic instability. General Design Criterion (GDC) 12 of 10 CFR 50, Appendix A, requires thermal-hydraulic instability to be prevented by design or be readily and reliably detected and suppressed. When the design of the reactor system does not prevent the occurrence of neutronic/thermal-hydraulic instability, instability is an anticipated operational occurrence. GDC 10 of 10 CFR 50, Appendix A, requires that specified acceptable fuel design limits not be exceeded during anticipated operational occurrences. Analyses performed by the BWROG [Boiling Water Reactor Owners' Group] indicate that neutronic/thermal-hydraulic instability induced power oscillations could result in conditions exceeding the Minimum Critical Power Ratio (MCPR) Safety Limit (SL) prior to detection and suppression by the current design of the Neutron Monitoring System and Reactor Protection System. To ensure compliance with GDC 12 the BWROG developed Interim Corrective Actions (ICAs) to enhance the capability of the operator to readily and reliably detect and suppress neutronic/thermal- hydraulic instability. The BWROG ICAs also provided additional guidance for monitoring local power range monitors beyond the requirements of current TS 3.4.1 to ensure adequate margin to the onset of neutronic/thermal-hydraulic instability. Reliance on operator actions to comply with GDC 12 was accepted on an interim basis by the NRC pending final implementation of a long term solution to the stability issue. Neutronic/thermal-hydraulic instability is prevented by implementation of the E1A solution through the modified design of the Reactor Protection System (APRM [average power range monitor] flow biased scram) and the stability control prior to entry into a region of the power and flow operating domain which is potentially susceptible, in the absence of stability control, to neutronic/thermal-hydraulic instability. In addition, significant backup protection features, including the PBDS, control rod block alarms associated with entry into the region susceptible to neutronic/thermal-hydraulic instabilities in the absence of stability controls, and specified operator actions, including manual reactor scram, are required to be implemented. As a result, the margin to the onset of neutronic/thermal-hydraulic instability provided by the existing TS requirements and BWROG ICAs recommendations is not significantly reduced by the implementation of the E1A solution. The E1A solution assures compliance with GDC 12 by the prevention of neutronic/thermal-hydraulic instability and therefore precludes neutronic/thermal-hydraulic instability from becoming a credible consequence of an anticipated operational occurrence. The consequences of anticipated operational occurrences will not increase and the margin to the MCPR SL will not decrease upon implementation of the E1A solution. Therefore, the proposed amendments do 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. Local Public Document Room Location: Judge George W. Armstrong Library, 220 S. Commerce Street, Natchez, Mississippi 39120 Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502. NRC Section Chief: Robert A. Gramm. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of amendment request: May 6, 1999. Description of amendment request: The proposed amendments would change those Technical Specifications (TS) required to support Grand Gulf Nuclear Station (GGNS), Cycle 11 operation. The changes would include a change to the minimum critical power ratio safety limit (SLMCPR) that would reflect a decrease of the two recirculation loop SLMCPR limit from 1.11 to 1.09, and the single recirculation loop SLMCPR limit from 1.12 to 1.10. These values were developed with General Electric's cycle-specific SLMCPR methodology in GESTAR-II Amendment 25, which was recently approved by the Nuclear Regulatory Commission in a Safety Evaluation Report dated March 11, 1999. 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: I. The proposed change does not significantly increase the probability or consequences of an accident previously evaluated. The Minimum Critical Power Ratio (MCPR) safety limit is defined in the Bases to Technical Specification 2.1.1 as that limit which ``ensures that during normal operation and during Anticipated Operational Occurrences (AOOs), at least 99.9% of the fuel rods in the core do not experience transition boiling.'' The MCPR safety limit is re-evaluated for each reload and, for GGNS Cycle 11, the analyses have concluded that a two-loop MCPR safety limit of 1.09, based on the application of GE's [General Electric's] NRC-approved cycle-specific MCPR safety limit methodology demonstrates that this acceptance criterion is satisfied. For single-loop operation, a MCPR safety limit of 1.10, based on GE's [NRC-approved cycle-specific MCPR safety limit methodology, also demonstrates that this acceptance criterion is satisfied. Core MCPR operating limits are developed to support the Technical Specification 3.2 requirements and ensure these safety limits are maintained in the event of the worst-case transient. Since the MCPR safety limit will be maintained at all times, operation under the proposed changes will ensure at least 99.9% of the fuel rods in the core do not experience transition boiling. Therefore, these changes to the Minimum Critical Power Ratio (MCPR) safety limit do not affect the probability or consequences of an accident. GE's NRC-approved GESTAR-II cycle-specific MCPR safety limit methodology has been applied and has no effect on the probability or consequences of any accidents previously evaluated. As previously licensed, one exception to GESTAR is that the mis-oriented and mis- located bundle events will continue to be analyzed as accidents subject to the acceptance criteria in the current licensing basis. The design of the GE11 fuel bundles is such that the bundles are not likely to be mis-oriented or mis-located and the normal administrative controls will be in effect for assuring proper orientation and location. Therefore, the probability of a fuel loading error is not increased. This analysis ensures that postulated dose releases will not exceed a small fraction (10 percent) of 10CFR100 limits. Therefore, the probability or consequences of accidents previously evaluated are unchanged. II. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The GE11 fuel to be used in Cycle 11 is of a design compatible with fuel present in the core and used in the previous cycle. Therefore, the GE11 fuel will not create the possibility of a new or different kind of accident. The proposed changes do not involve any new modes of operation, any changes to setpoints, or any plant modifications. The proposed revised MCPR safety limits have been shown to be acceptable for Cycle 11 operation. Compliance with the applicable criterion for incipient boiling transition continues to be ensured. The proposed MCPR safety limits do not result in the creation of any new precursors to an accident. Therefore, the proposed changes do not create the possibility of a new or different type of accident from any accident previously evaluated. III. The proposed change does not involve a significant reduction in a margin of safety. [[Page 46435]] The MCPR safety limits have been evaluated in accordance with GE's NRC-approved cycle-specific methodology to ensure that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core are not expected to experience transition boiling. One exception to GESTAR is that the mis-oriented and mis-located bundle events will continue to be analyzed as accidents subject to the acceptance criteria in the current licensing basis. This analysis ensures that postulated dose releases for the worst case mis- oriented and mis-located bundle will not exceed a small fraction (10 percent) of 10CFR100 limits. On this basis, the implementation of this GE methodology 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. Local Public Document Room Location: Judge George W. Armstrong Library, 220 S. Commerce Street, Natchez, Mississippi 39120. Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502. NRC Section Chief: Robert A. Gramm. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of amendment request: June 23, 1999. Description of amendment request: The requested Technical Specification changes would revise those specifications associated with various engineered safety feature systems, which need no longer be credited following a design-basis fuel handling accident. The proposed changes affect conditions where irradiated fuel is handled in the primary or secondary containment, and also affect certain specifications related to performing core alterations. These changes are based on the revised analysis of the design-basis fuel handling accident for the Grand Gulf Nuclear Station. This requested change is consistent with the changes approved for the Perry Nuclear Power Plant Operating License (Amendment 102), and the industry-proposed change to the Technical Specification NUREGs, TSTF-51. 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: 1. The proposed changes do not significantly increase the probability or consequences of an accident previously evaluated. A new term to describe irradiated fuel is used to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis. Because the equipment affected by the revised operational conditions is not considered an initiator to any previously analyzed accident, inoperability of the equipment cannot increase the probability of any previously evaluated accident. The proposed requirements bound the conditions of the current design basis fuel handling accident analysis which concludes that the radiological consequences are within the acceptance criteria of NUREG 0800, Section 15.7.4 and General Design Criteria 19. Therefore, the proposed changes do not significantly increase the probability or consequences of any previously evaluated accident. Removing a one time only allowance granted by Amendment 129 to the Operating License that is no longer in affect is an administrative change. Therefore, the proposed change does not significantly increase the probability or consequences of any previously evaluated accident. Based on the above, neither the proposed changes to the Technical Specifications nor that to the Operating License significantly increase the probability or consequences of any accident previously evaluated. 2. The proposed changes would not create the possibility of a new or different kind of accident from any previous analyzed. The new term to describe irradiated fuel is used to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis. The proposed changes do not introduce any new modes of plant operation and do not involve physical modifications to the plant. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previous analyzed. Removing a one time only allowance granted by Amendment 129 to the Operating License that is no longer in affect is an administrative change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previous analyzed. Based on the above, neither the proposed changes to the Technical Specifications nor that to the Operating License create the possibility of a new or different kind of accident from any accident previously analyzed. 3. The proposed changes do not involve a significant reduction in a margin of safety. The new term to describe irradiated fuel is used to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis and are established such that the radiological consequences are at or below the current GGNS [Grand Gulf Nuclear Station] licensing limit. Safety margins and analytical conservatisms have been evaluated and are well understood. Substantial margins are retained to ensure that the analysis adequately bounds all postulated event scenarios. The proposed change only eliminates the unnecessary margin from the analysis. The current margin of safety is retained. Specifically, the margin of safety for the fuel handling accident is the difference between the 10CFR100 limits and the licensing limit defined by NUREG 0800, Section 15.7.4. With respect to the control room personnel doses, the margin of safety is the difference between the 10CFR100 limits and the licensing limit defined by 10CFR50, Appendix A, Criterion 19 (GDC 19). The additional margin between the calculated doses for the postulated events and the corresponding licensing limit provides no useful purpose. The proposed applicability continues to ensure that the whole- body and thyroid doses at both the control room and the exclusion area and low population zone boundaries are at or below the corresponding licensing limit. The margin of safety is unchanged; therefore, the proposed changes do not involve a significant reduction in a margin of safety. Removing a one time only allowance granted by Amendment 129 to the Operating License that is no longer in affect is an administrative change. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, neither the proposed changes to the Technical Specifications nor that to the Operating License result in a significant reduction in a margin of safety. Based on the above evaluation, operation in accordance with the proposed amendment involves no significant hazards considerations. 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. Local Public Document Room Location: Judge George W. Armstrong Library, 220 S. Commerce Street, Natchez, Mississippi 39120. Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502. NRC Section Chief: Robert A. Gramm. Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: October 6, 1998. [[Page 46436]] Description of amendment request: The proposed change modifies the requirement to perform a Moderator Temperature Coefficient (MTC) test near the end of each cycle. This request constitutes a lead-plant submittal, submitted by Waterford 3 on behalf of the Combustion Engineering Owners Group (CEOG). CE NPSD-911, Amendment 1, ``Analysis of Moderator Temperature Coefficients in Support of a Change in the Technical Specifications End of Cycle Negative MTC Limit'' dated January, 1998 is provided as an Attachment to the application. Specifically, the proposed change modifies Technical Specification (TS) 4.1.1.3.2c by adding a provision that eliminates the need to determine the MTC upon reaching two-thirds of core burnup if the results of the MTC tests required in TS 4.1.1.3.2a and 4.1.1.3.2b are within a specified tolerance. In addition, some editorial changes are proposed and the Bases change is included to support the changes in the TS. 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: 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Under the proposed change, compliance with the TS Limiting Condition for Operation is achieved through a surveillance program consisting of beginning-of-cycle (BOC) measurements, plant parameter monitoring, and end-of-cycle (EOC) MTC predictions. This change eliminates the performance of the 2/3 Cycle MTC Surveillance when the BOC MTC Surveillances are within a required tolerance of the design value. The probability and consequences of an accident previously evaluated will not be increased because this change does not modify any assumptions used in the input to the safety analyses. The current safety calculations will remain valid because the allowed range of MTC values will not change. The Combustion Engineering analysis CE NPSD-911 and CE NPSD-911 Amendment 1, demonstrate that if the startup test program has established that the core is operating as intended, and if the isothermal temperature coefficients measured at zero power during the cycle startup program, and at power prior to 40 EFPD [Effective Full Power Days], fall within the design value of plus or minus 0.16 x 10-4 delta k/k/ deg.F, then the end-of-cycle best estimate prediction will also be within plus or minus 0.16 x 10-4 delta k/k/ deg.F of true MTC. Removing the footnote that was applicable during Cycle 7 and providing a plus/minus for SR 4.1.1.3.2c is purely an administrative change. Therefore, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated. 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different type of accident from any accident previously evaluated? Response: No. Plant operation and plant parameter TS limits will remain unchanged. There are no new changes in plant design nor are any new failure modes introduced. CE NPSD-911 analysis determined that if the MTC at the beginning-of-cycle is within plus or minus 0.16 x 10-4 delta k/k/ deg.F of the design value then the MTC at the end-of-cycle will also be within plus or minus 0.16 x 10-4 delta k/k/ deg.F of the design value. Removing the footnote that was applicable during Cycle 7 and providing a plus/minus for SR 4.1.1.3.2c is purely an administrative change. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety? Response: No. The margin of safety will not be reduced because the range of allowed temperature coefficients will not be changed. The surveillance program consisting of beginning-of-cycle measurements, plant parameter monitoring, and end-of-cycle MTC predictions will ensure that the MTC remains within the range of acceptable values. Removing the footnote that was applicable during Cycle 7 and providing a plus/minus for SR 4.1.1.3.2c is purely an administrative change. 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 request involves no significant hazards consideration. Local Public Document Room Location: University of New Orleans Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122. Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn 1400 L Street NW., Washington, DC 20005-3502. NRC Section Chief: Robert A. Gramm. FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio Date of amendment request: July 26, 1999. Description of amendment request: The proposed amendment would make the following line-item Technical Specification (TS) improvements: (1) Relocate TS Section 3/4.3.3.2, Instrumentation--Incore Detectors; TS 3/4.3.3.9, Instrumentation--Waste Gas System Oxygen Monitor; and TS 3/4.4.7, Reactor Coolant System `` Chemistry, to the Updated Safety Analysis Report (USAR) Technical Requirements Manual (TRM); (2) Change to TS 3/4.11.2, Radioactive Effluents--Explosive Gas Mixture, and TS Bases 3/4.11.2, Explosive Gas Mixture, to reflect the above proposed relocation of TS 3/4.3.3.9; (3) Revise the requirements of TS 3/4.4.6.1, Reactor Coolant System Leakage--Leakage Detection Systems, to require one monitor (gaseous or particulate) of the containment atmosphere radioactivity monitoring systems to be operable, rather than requiring both systems to be operable simultaneously; and (4) Revise the requirements of TS 3/4.3.3.1, Radiation Monitoring Instrumentation, to be consistent with the above proposed revision to TS 3/4.4.6.1. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensees have provided their analysis of the issue of no significant hazards consideration, which is presented below: The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit Number 1, in accordance with these changes would: 1a. Not involve a significant increase in the probability of an accident previously evaluated because no accident initiator, conditions or assumptions are affected by the proposed revisions to Technical Specification (TS) 3/4.3.3.1, Radiation Monitoring Instrumentation, TS 3/4.3.3.2, Instrumentation--Incore Detectors; TS 3/4.3.3.9, Instrumentation--Waste Gas System Oxygen Monitor; TS 3/ 4.4.7, Reactor Coolant System--Chemistry; TS 3/4.11.2, Radioactive Effluents--Explosive Gas Mixture; and TS 3/4.4.6.1, Reactor Coolant System Leakage--Leakage Detection Systems, and their associated TS Bases. The requirements of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7 are proposed to be relocated from the TS to the DBNPS Updated Safety Analysis Report (USAR) Technical [[Page 46437]] Requirements Manual (TRM). These requirements would be relocated generally intact to the TRM whereby future changes would be subject to the regulatory controls of 10 CFR 50.59. These relocations are consistent with the NRC guidance provided in Generic Letter (GL) 95- 10, ``Relocation of Selected Technical Specifications Requirements Related to Instrumentation,'' or NUREG-1430, Revision 1, ``Standard Technical Specifications--Babcock and Wilcox Plants,'' dated April 1995. The proposed revision to TS 3/4.11.2, Radioactive Effluents-- Explosive Gas Mixture, and its Bases is an administration change to a reference necessitated by the proposed relocation of TS 3/4.3.3.9 to the USAR TRM. The proposed revision to TS 3/4.3.3.1 and TS 3/4.4.6.1 regarding the number of Reactor Coolant System (RCS) leakage detection monitors required and their allowed outage times is based upon the NRC's guidance of NUREG-1430, Revision 1. This proposed revision affects the TS only and does not reduce the number, diversity, or sensitivity of Reactor Coolant System leakage detection systems inside the containment building or as committed to in the DBNPS USAR. 1b. Not involve a significant increase in the consequences of an accident previously evaluated because no accident condition or assumption is affected by the proposed revisions. As described above, the revisions are consistent with the guidance of NRC GL 95- 10 or NUREG-1430, Revision 1. The proposed revisions, as described above, do not alter the source term, containment isolation, or allowable releases. The proposed changes, therefore, will not increase the radiological consequences of a previously evaluated accident. 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because no new accident initiators or assumptions are introduced by the proposed TS revisions. No new accident scenarios, transient precursors, failure mechanisms, or limiting failures are introduced as a result of the proposed changes. 3. Not involve a significant reduction in a margin of safety because the proposed revisions do not reduce or adversely affect the capabilities of any plant structures, systems or components. The proposed relocation of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7 to the USAR TRM is essentially an administrative change to the location and process by which these requirements are controlled and revised. Future revisions to these requirements relocated to the USAR TRM will be subject to the regulatory controls of 10 CFR 50.59. Therefore, these revisions will not result in a significant reduction in a margin of safety. The proposed revision to TS 3/4.11.2 and its Bases is administrative and reflects the relocation of TS 3/4.3.3.9 to the USAR TRM. Therefore, this revision will not result in a significant reduction in a margin of safety. The proposed revisions to TS 3/4.3.3.1 and TS 3/4.4.6.1 affect the number of containment atmosphere radioactivity monitors required by TS to be operable simultaneously. However, redundancy and diversity requirements are maintained in the TS for detecting Reactor Coolant System leakage. Although TS-allowed outage times are proposed to be increased consistent with NUREG-1430, Revision 1 guidance, related compensatory action requirements are also being increased. Furthermore, the DBNPS commitments made for complying with Regulatory Guide 1.45, May, 1973, ``Reactor Coolant Pressure Boundary Leakage Detection Systems,'' are not changed by the proposed revisions. Along with the applicable revised TS requirements, 10 CFR 50, Appendix B, Criterion XVI will require prompt corrective action for inoperable leakage detection systems. Accordingly, these proposed revisions will not result in 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. Local Public Document Room location: University of Toledo, William Carlson Library, Government Documents Collection, 2801 West Bancroft Avenue, Toledo, OH 43606. Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Section Chief: Anthony J. Mendiola. FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio Date of amendment request: July 26, 1999. Description of amendment request: The proposed amendment would change the Technical Specifications to adopt the performance-based 10 CFR Part 50, Appendix J, Option B approach for Type B and C containment leakage rate testing, and to relocate certain details of the tests into a Containment Leakage Testing Program. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensees have provided their analysis of the issue of no significant hazards consideration, which is presented below: The Davis-Besse Nuclear Power Station has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with these changes would: 1a. Not involve a significant increase in the probability of an accident previously evaluated because accident initiators, conditions, or assumptions are not affected by the proposed changes. The proposed changes to the Technical Specifications and Bases implement 10 CFR [Part] 50 Appendix J Option B for Type B and C Local Leak Rate Testing, based on the guidance of Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test Program.'' Provided that components have performed satisfactorily on a historical basis, this guidance permits the use of extended testing frequencies. These proposed changes do not affect accident initiators, conditions, or assumptions. 1b. Not involve a significant increase in the consequences of an accident previously evaluated because the proposed changes do not change the source term or total allowable releases. With the exception of the proposed increase in the containment air lock leakage limits, the proposed changes do not affect the total allowable containment leakage rates presently specified in the Technical Specifications. Although the air lock leakage limits are proposed to be increased, the accident analyses are based on the current TS allowable maximum bypass leakage, which is not proposed to be changed. Therefore, increases in leakage limits for individual components, such as the air locks and their door seals, which are constituents of bypass leakage, will have no effect on the radiological consequences described in the accident analyses. The proposed TS changes relating to implementation of 10 CFR [Part] 50 Appendix J Option B may result in a small, but acceptable increase in post-accident containment leakage, due to the increased probability that due to generally increased intervals between tests, an unacceptable leakage rate could go undetected for a longer length of time. NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' September, 1995, which provided the technical basis for the 10 CFR [Part] 50 Appendix J Option B rulemaking, provides a detailed evaluation of the expected leakage and its consequences and concludes that increased test frequencies are workable without significant risk impacts. 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because no new accident initiators or assumptions are introduced by the proposed changes. The proposed changes do not affect the methodology used in conducting containment leak rate testing. The proposed changes do not involve a change to the plant design or operation and, therefore, will not introduce any new or different failure modes or initiators. 3. Not involve a significant reduction in a margin of safety. The proposed changes relating to implementation of 10 CFR [Part] 50, Appendix J, Option B do not significantly affect the allowable containment leakage rates presently specified in the Technical Specifications. The Technical Specifications, under the proposed changes, will continue to ensure containment reliability by periodic testing performed in full compliance with 10 CFR [Part] 50, Appendix J. [[Page 46438]] 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. Local Public Document Room location: University of Toledo, William Carlson Library, Government Documents Collection, 2801 West Bancroft Avenue, Toledo, OH 43606. Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Section Chief: Anthony J. Mendiola. FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio Date of amendment request: July 28, 1999. Description of amendment request: The proposed amendment would change Technical Specification (TS) Section 3/4.7.5.1, ``Ultimate Heat Sink,'' to allow operation on Modes 1 through 4 with an Ultimate Heat Sink water temperature of less than or equal to 90 deg.F, instead of the current limit of less than or equal to 85 deg.F. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensees have provided their analysis of the issue of no significant hazards consideration, which is presented below: The Davis-Besse Nuclear Power Station has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with these changes would: 1a. Not involve a significant increase in the probability of an accident previously evaluated because no accident initiators, conditions, or assumptions are significantly affected by the proposed change. The proposed change would increase the allowable Ultimate Heat Sink (UHS) water temperature, as specified in TS LCO 3.7.5.1.b, from less than or equal to 85 deg.F to less than or equal to 90 deg.F. This water is used by the Service Water System to provide cooling to equipment that is used to mitigate accidents such as a Large Break Loss of Coolant Accident. This increase in Service Water temperature has been evaluated and the proposed change does not result in the operation of equipment important to safety outside their acceptable operating ranges. 1b. Not involve a significant increase in the consequences of an accident previously evaluated because the proposed change does not change the source term, containment isolation, or allowable releases. The proposed increase in the Service Water System temperature has been evaluated with respect to the containment and equipment used to mitigate the consequences of accidents previously evaluated. These evaluations have determined that there are no significant increases in consequences. 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because no new accident initiators or assumptions are introduced by the proposed 5 deg.F increase in UHS temperature. The proposed change does not result in installed equipment being operated outside their design operating ranges. No new or different equipment failure modes or mechanisms are introduced by the proposed change. 3. Not involve a significant reduction in a margin of safety because the proposed 5 deg.F increase in UHS temperature does not result in significant changes to the initial conditions contributing to accident severity or consequences. 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. Local Public Document Room location: University of Toledo, William Carlson Library, Government Documents Collection, 2801 West Bancroft Avenue, Toledo, OH 43606. Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037 . NRC Section Chief: Anthony J. Mendiola. FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit 1, Lake County, Ohio Date of amendment request: June 17, 1999. Description of amendment request: The proposed amendment modifies multiple surveillance requirements to support implementation of a 24- month operating cycle. 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 amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. A. Frequency Extensions The proposed Technical Specification (TS) changes involve a change in the surveillance testing intervals to facilitate a change in the Perry Nuclear Power Plant (PNPP) operating cycle from 18 months to 24 months. The proposed TS changes do not physically impact the plant, nor do they impact any design or functional requirements of the associated systems. That is, the proposed TS changes do not degrade the performance of, or increase the challenges to, any safety systems assumed to function in the accident analysis. The proposed TS changes do not impact the TS surveillance requirements themselves, or the way in which the surveillances are performed. In addition, the proposed TS changes do not introduce any accident initiators, since no accidents previously evaluated have, as their initiators, anything related to the frequency of surveillance testing. Also, evaluation of the proposed TS changes demonstrated that the availability of equipment and systems required to prevent or mitigate the radiological consequences of an accident are not significantly affected because of other, more frequent testing that is performed, the availability of redundant systems and equipment, or the high reliability of the equipment. Since the impact on the systems is minimal, it is concluded that the overall impact on the plant accident analysis is negligible. Furthermore, a historical review of surveillance test results and associated maintenance records indicated that there was no evidence of any failures that would invalidate the above conclusions. Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated. B. Allowable Value Changes The proposed changes in Allowable Values for the instrumentation include in Table 3.3.8.1-1 Items d and e of the Technical Specifications are the result of application of the Perry Instrument Setpoint Methodology (ISM) using plant specific drift values. Application of this methodology results in Allowable Values which more accurately reflect total instrumentation loop accuracy as well as that of test equipment and calculated drift between surveillances. The proposed changes will not result in any hardware changes. The instrumentation is not assumed to be an initiator of any analyzed event. Existing operating margin between plant conditions and actual plant setpoints is not significantly reduced due to these changes. The role of the instrumentation is in mitigating and thereby limiting the consequences of accidents. The Allowable Values have been developed to ensure that the design and safety analysis limits will be satisfied. The methodology used for the development of the Allowable Values ensures the affected instrumentation remains capable of mitigating design basis events as described in the safety analyses and that the results and radiological consequences described in the safety analyses remain bounding. Additionally, the proposed change does not alter the plant's ability to detect and mitigate events. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. [[Page 46439]] C. Frequency Reductions to Semiannual The proposed Technical Specification (TS) changes involve a change in the surveillance testing intervals from 18 months to either 6 months or quarterly. The shorter frequencies are based on PNPP specific results of setpoint drift evaluations. The proposed more restrictive TS changes do not physically impact the plant, nor do they impact any design or functional requirements of the associated systems. That is, the proposed TS changes do not degrade the performance of, or increase the challenges to, any safety systems assumed to function in the accident analysis. The proposed TS changes do not impact the TS surveillance requirements themselves, or the way in which the surveillances are performed. In addition, the proposed TS changes do not introduce any accident initiators, since no accidents previously evaluated have, as their initiators, anything related to the frequency of surveillance testing. The proposed TS frequencies will demonstrate that the equipment and systems required to prevent or mitigate the radiological consequences of an accident are continuing to meet the assumptions of the setpoint evaluation, on a more frequent basis. Since the impact on the systems is minimal, and the assumptions of the safety analyses will be maintained, it is concluded that the overall impact on the plant accident analysis is negligible. Furthermore, a historical review of surveillance test results and associated maintenance records indicated that there was no evidence of any failures that would invalidate the proposed test frequencies. Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated. The proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. A. Frequency Extensions The proposed TS changes involve a change in the surveillance testing intervals to facilitate a change in the PNPP operating cycle length. The proposed TS changes do not introduce any failure mechanisms of a different type than those previously evaluated, since there are no physical changes being made to the facility. No new or different equipment is being installed. No installed equipment is being operated in a different manner. As a result, no new failure modes are being introduced. In addition, the surveillance test requirements themselves, and the way surveillance tests are performed, will remain unchanged. Furthermore, a historical review of surveillance test results and associated maintenance records indicated there was no evidence of any failures that would invalidate the above conclusions. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated. B. Allowable Value Changes The proposed changes are the result of application of the ISM using plant specific drift values and do not create the possibility of a new or different kind of accident from any accident previously evaluated. This is based on the fact that the method and manner of plant operation is unchanged. The use of the proposed Allowable Values does not impact safe operation of PNPP in that the safety analysis limits will be maintained. The propose Allowable Values involve no system additions or physical modifications to systems in the station. These Allowable Values were revised to ensure the affected instrumentation remains capable of mitigating accidents and transients. Plant equipment will not be operated in a manner different from previous operation, except that setpoints may be changed. Since operational methods remain unchanged and the operating parameters have been evalu [Text truncated at 120,000 characters. The full text is on the page linked above.] --- Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A99-21914. Public record. Not legal advice.