# 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%3A99-4391 ## Record - **Collection:** Federal Register - **Document type:** Notice - **Published:** February 24, 1999 - **Citation:** 64 FR 9183 ## 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 January 30, 1999, through February 11, 1999. The last biweekly notice was published on February 10, 1999. Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently. 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 March 26, 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. [[Page 9184]] 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 document room for the particular facility involved. Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina Date of amendment request: January 28, 1999. Description of amendment request: The H. B. Robinson, Unit No. 2, Technical Specifications (TSs) are proposed to be changed to replace and add analytical methodologies used to determine acceptable core designs and provide inputs to methodologies that develop the core operating limits in the Core Operating Limits Report. 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. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed changes in a methodology have been previously generically reviewed and approved for use by the NRC for determining core neutronics design and gadolinimum oxide thermal conductivity. Analyzed events are assumed to be initiated by the failure of plant structures, systems, or components. The fuel design parameters developed in accordance with the new methodologies are bounded by the limitations in the NRC acceptance in its safety evaluations of the new methodologies. The topical reports associated with the new methodologies demonstrate that the integrity of the fuel will be maintained during normal operations and that design requirements preclude fuel rods containing gadolinium oxide from being limiting in accident and related safety analyses. The proposed change does not have a detrimental impact on the integrity of any plant structure, system, or component. The proposed change will not alter the operation of any plant equipment, or otherwise increase its failure probability. Therefore, the probability of occurrence for a previously analyzed accident is not significantly increased. The consequences of a previously analyzed accident are dependent on the initial conditions assumed for the analysis, the behavior of the fuel during the analyzed accident, the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. The proposed changes to methodology continues to meet applicable design and safety analyses acceptance criteria for neutronics design analysis and gadolinimum oxide thermal conductivity. The topical reports associated with the new methodologies demonstrate that the integrity of the fuel will be maintained as is assumed or is bounded initially in accident analyses and that design requirements preclude fuel rods containing gadolinimum oxide from being limiting in accident and related safety analyses. The proposed change does not affect the performance of any equipment used to mitigate the consequences of an analyzed accident. As a result, no analyses assumptions are violated and there are no adverse effects on the factors that contribute to offsite or onsite dose as the result of an accident. The proposed change does not affect setpoints that initiate protective or mitigative actions. The proposed change ensures that plant structures, systems, or components are maintained consistent with the safety analysis and licensing bases. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event. Therefore, the proposed change does not involve any increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve any physical alteration of plant systems, structures, or components. The proposed changes in methodology continue to meet applicable criteria for neutronics design analysis and assure that design requirements preclude fuel rods containing gadolinimum oxide from being limiting. The proposed change does not involve a physical alteration of the plant other than allowing for fuel design in accordance with NRC approved methodologies. No new or different equipment is being installed. No installed equipment is being operated in a different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions. As a result no new failure modes are being introduced. There are no changes in the methods governing normal plant operation, nor are the methods utilized to respond to plant transients altered. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does this change involve a significant reduction in a margin of safety? The margin of safety is established through the design of the plant structures, systems, and components, through the parameters within which the plant is operated, through the establishment of the setpoints for the actuation of equipment relied upon to respond to an event, and through margins contained within the safety analyses. The proposed change is to methodologies that continue to meet applicable criteria for neutronics design analysis and continues to assure that design requirements preclude fuel rods containing gadolinimum oxide from being limiting. The proposed change does not impact the condition or performance of structures, systems, setpoints, and components relied upon for accident mitigation. The proposed change does not significantly impact any safety analysis assumptions or results. Therefore, the proposed change does not result in 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. [[Page 9185]] Local Public Document Room location: Hartsville Memorial Library, 147 West College Avenue, Hartsville, South Carolina 29550. Attorney for licensee: William D. Johnson, Vice President and Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602. NRC Project Director: Cecil B. Thomas. Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois Date of amendment request: November 25, 1998. Description of amendment request: The proposed amendments would revise Improved Technical Specifications 3.8.4 and 3.8.9 to support on- line replacement of the Braidwood 125 Volt DC AT&T batteries with new Charter Systems Inc. batteries. 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? During the replacement of the existing batteries, a temporary battery bank will provide the same function as the AT&T batteries being removed. Even though this temporary battery will not be seismically mounted, due to its location in the Turbine Building, it is the safety related AT&T battery which was previously qualified and used to perform this function on Unit 1. While the temporary battery is being connected, the DC bus will be supplied by the existing crosstie with Unit 1. Similar crosstie conditions are allowed under the present Improved Technical Specifications. The DC system is normally supplied by the AC system through the ESF [Engineered Safety Feature] battery charger. The essential function of the DC system battery is to supply control power necessary to start and load the Diesel Generators. Once the Diesel Generators are on line, the DC system will be supplied via the battery charger. However, the ESF batteries have been sized for one hour to provide additional assurance that the critical DC loads are available in the event of a loss of a battery charger. During the 10 day Completion Time when the temporary battery and the ESF charger are supporting the bus, the ability of that DC Division to mitigate an event/accident is unchanged except for its ability to cope with a seismic event. However, the probability of a seismic event concurrent with the 10 day Completion Time is extremely small. During a seismic event, one DC division may be compromised, however, the unit has adequate DC power available in the form of the other division to mitigate all Design Basis accidents. This loss of one DC division is bounded by the loss of an entire AC division, a condition which the plant is currently evaluated to withstand. During the 8 hour Completion Time to connect and disconnect the temporary battery, there is no adverse impact on Unit 1. The compensatory measures to manually open the crosstie will ensure the Unit 1 DC battery can supply its required loads for the entire one hour duty cycle. The Unit 2 DC bus, which is crosstied, will be de- energized in the event of a Unit 2 accident based on the compensatory measures. This action would only be required if the associated Diesel Generator were to fail to re-energize its associated charger. This condition is consistent with the other crosstie scenarios currently permitted by the Technical Specifications. Thus, the 8 hour Completion Time is consistent with the two hour Completion Time with respect to the ability to safely shutdown the Unit. Only the duration of the Completion Time is different. Based on the above, the overall design, function, and operation of the DC system and equipment has not been significantly modified by these changes. The proposed changes do not affect any accident initiators or precursors and do not alter the design assumptions for the systems or components used to mitigate the consequences of an accident as analyzed in UFSAR Chapter 15. 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? During the replacement of the existing batteries, a temporary battery bank will provide the same function as the batteries being removed. Even though this temporary battery is not seismically mounted, it is the safety related AT&T battery which was previously qualified and utilized to perform this function on Unit 1. Because this temporary battery is identical to the battery that is currently installed, and will be connected and used in the same way, no new electrical or functional failure modes are created. The temporary battery will be located in the turbine building, which is non-seismic. The temporary battery will not be seismically mounted. Thus, a seismic failure of the batteries is possible. Since the temporary battery is located in the turbine building the potential for battery failure to initiate an accident is not present, and failure of the battery cannot create a different response from any previously postulated accident. Due to the location of the main generator in relationship to the temporary batteries, a turbine blade failure would not hit the battery unless it penetrated the turbine casing and ricocheted in the direction of the battery, which is an unlikely scenario due to the orientation of the temporary battery. Likewise, an unmitigated Outside Containment Steam Line Break of either unit would be interrupted by the successful closure of all MSIVs [Main Steam Isolation Valves] thereby leaving the battery and the DC bus intact and available. Also any affects of a postulated storm on the turbine building have been previously addressed and would not change as a result of the batteries being temporary located there. While the temporary battery is being connected, the DC bus will be supplied by the existing crosstie with Unit 1. To prevent any occurrence on Unit 2 from adversely affecting Unit 1, this crosstie will be manually disconnected based on specific criteria that may be indicative of a Unit 2 accident (specifically a Unit 2 LOOP). Once the crosstie is opened, the Unit 2 bus will be de-energized and the other Unit 2 division will be required to mitigate the accident. This loss of one DC division is bounded by the loss of one division (AC or DC), a condition which the plant is currently evaluated to withstand. The DC system and its equipment will continue to perform the same function and be operated in the same fashion. The proposed changes do not introduce any new accident initiators or precursors, or any new design assumptions for the systems or components used to mitigate the consequences of an accident. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated has not been created. Therefore, this proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does the change involve a significant reduction in a margin of safety? During the replacement of the existing batteries, a temporary safety related battery bank will perform the same function as the batteries being removed. Even though this temporary battery is not seismically mounted, it is the safety related battery which was previously qualified and used to perform this function on Unit 1 and is identical to the safety related battery that is currently installed. Therefore, it has the same capacity, margin and capability to fulfill the requirements of the Unit 2 DC bus as the existing qualified battery. The proposed replacement activity will not prevent the plant from responding to either a seismic event or design basis accident. In both cases, the design mitigation capability will be maintained. Due to the limited duration of the activity and the planned contingency actions, a significant reduction in the margin of safety will not result. While the temporary battery is being connected, the DC bus will be supplied by the existing crosstie with Unit 1. This condition is currently allowed for a limited time by the Improved Technical Specifications. The inherent design conservatism of the DC system and its equipment has not been altered. The DC system and its equipment will continue to be operated with the same degree of conservatism. Accordingly, there is no 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 [[Page 9186]] satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration. Local Public Document Room location: Wilmington Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481. 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 Project Director: Stuart A. Richards. Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois Date of amendment request: December 29, 1998. Description of amendment request: The proposed amendments would revise the Technical Specification Tables 3.3.1-1 and 3.3.2-1, to revise twelve Reactor Trip System and Engineered Safety Feature Actuation System Allowable Values. 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. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? These changes to the twelve AVs [Allowable Values] do not involve an increase in the probability of an accident previously evaluated. The AVs provide the basis for determining instrument channel operability and do not change the system function, or channel operation or calibration. Operation within the AV ensures the instrument channel's ability to provide the required reactor trip or engineered safety feature actuation signal during plant operation. In all cases, the proposed changes only make the twelve AVs more restrictive with respect to the current AVs, and do not effect the response characteristics of the instrumentation because actual trip setpoints are unchanged. There is no change being made to the approved design, nor is there any operational change being made which would increase the probability of occurrence of an accident previously evaluated. The RTS [Reactor Trip System] and ESFAS [Engineered Safety Feature Actuation System] systems which are actuated by the corresponding instrumentation setpoints will operate in the same manner as before and within their design limits. These changes to the twelve AVs do not involve an increase in the consequences of an accident previously evaluated. These changes have no effect on plant operation. There is no physical or operational change being made which would alter the sequence of events, plant response, or assumptions or conclusions of the affected analyses. The use of the AVs as a basis for determining instrument or channel operability does not change system operation or channel function. The proposed changes do not change the established trip setpoints for these functions. No design analyses have changed or will be affected. The twelve revised AVs are more restrictive than the current AVs and continue to ensure that the safety limits are not violated during anticipated transients, and that the consequences of design basis accidents remain acceptable. The change to the AVs does not degrade or prevent any actions from taking place in response to an accident. The use of NRC approved or endorsed methodology in developing the proposed AVs ensures that the present analytical limits for all accidents will be maintained. These proposed changes to the AVs for RTS and ESFAS instrumentation will continue to ensure that the associated RTS trip or ESFAS actuation signals will be generated when required within the bounds of the plant safety analyses. There is no change in the type or amount of any effluents released, and no change in either the onsite or offsite dose consequences as a result of this change. Therefore, based on this evaluation, this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? These proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes to the twelve AVs for RTS and ESFAS instrumentation will not affect the trip setpoints at which a reactor trip or engineered safety feature actuation is initiated. The trip setpoints contained in the Technical Requirements Manual are not being changed and will continue to be maintained. The only changes being made are to the AVs used as a basis for determining instrument channel operability. Because the trip setpoints are unchanged, RTS or ESFAS setpoint actuation is not affected by the revised AVs. An RTS trip or ESFAS actuation signal that may initiate between its trip setpoint and the associated AV is acceptable because an allowance has been made in the affected instrument uncertainty calculation to accommodate this deviation. It allows for potential drift while ensuring plant operation in a safe manner. Using this methodology provides plant operational flexibility and yet remains within the allowances accounted for in the various accident analyses. No new equipment is being installed, and no installed equipment is being operated in a new or different manner with these twelve AV changes. The revised AVs do not alter the intended design or operation of systems or instrument channels. As no physical plant equipment changes are being made, no new equipment failure modes are being introduced as a result of these proposed changes. There is no change in plant operation that affects previously evaluated failure modes and no change in plant response to a transient condition. These changes do not represent a new failure mode over what has been previously evaluated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the change involve a significant reduction in a margin of safety? There is no significant reduction in the margin of safety from these proposed changes. These proposed changes move twelve AVs closer to the trip setpoints compared to the existing AVs, which increases the margin of safety. An RTS trip or ESFAS actuation signal that may initiate between its trip setpoint and the associated AV is acceptable because an allowance has been made in the affected instrument uncertainty calculation to accommodate this deviation. The revised AVs have been calculated using NRC approved or endorsed methodology, which is consistent with existing safety analyses that define the margin of safety. Safety analyses assumptions and results are not affected. 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: For Byron, the Byron Public Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481. 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 Project Director: Stuart A. Richards. 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: January 21, 1999. Description of amendment request: This amendment request proposes to relocate Technical Specification (TS) Section 3/4.6.I to the Updated Final Safety Analysis Report (UFSAR) and plant procedures. TS Section 3/4.6.I contains reactor coolant chemistry limiting conditions for operation (LCO) and surveillance requirements (SR) for conductivity, chloride concentration and pH. [[Page 9187]] 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 simplify the TS, meet regulatory requirements for relocated TS's, and implement the recommendations of the Commission's Final Policy Statement on TS improvements. The Chemistry requirements will be relocated to the Updated Final Safety Analysis Report (UFSAR) and to applicable station procedures. Future changes to these requirements will be controlled by 10 CFR 50.59. The proposed changes are administrative in nature and do not involve any modification to any plant equipment or affect plant operation. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any previously evaluated accident. Consequently, this proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes are administrative in nature, do not involve any physical alterations to any plant equipment, and cause no change in the method by which any safety related system performs its function. Therefore, this proposed TS amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. Does the change involve a significant reduction in a margin of safety? The proposed amendment represents the relocation of current requirements which are based on generic guidance or previously approved provisions for other stations. The proposed changes are administrative in nature and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. The proposed changes have been evaluated and found to be acceptable for use at Quad Cities Nuclear Power Station. Since the proposed changes are administrative in nature, and are based on NRC accepted provisions which have been adopted at other nuclear facilities, and maintain the necessary levels of system reliability, 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: 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 Project Director: Stuart A. Richards. Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment request: January 28, 1999. Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) to correct Surveillance Requirement (SR) 3.7.13.4 and the associated Bases. This SR currently is incorrect and does not reflect the Fuel Handling Ventilation Exhaust System (FHVES) as designed. Specifically, the FHVES flow rate requirement has been inadvertently stated at half the design value (18,221 instead of 36,443 cfm [cubic feet per minute]). The proposed amendments would only revise the SR to the correct design value; no physical change to the FHVES design is involved. 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: First Standard Implementation of this amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. Approval of this amendment will have no effect on accident probabilities or consequences. The FHVES is not an accident initiating system; therefore, there will be no impact on any accident probabilities by the approval of this amendment. The design of the system is not being modified by this proposed amendment. The amendment merely aligns TS requirements with the existing design and function of the system. Therefore, there will be no impact on any accident consequences. Second Standard Implementation of this amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the plant which will introduce any new accident causal mechanisms. This amendment request does not impact any plant systems that are accident initiators; neither does it impact any accident mitigating systems. Third Standard Implementation of this amendment would not involve a significant reduction in a margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these fission product barriers will not be impacted by implementation of this proposed amendment. The FHVES is already capable of performing as designed. No safety margins will be impacted. Based upon the preceding analysis, Duke Energy has concluded that the proposed amendment does not involve a significant hazards consideration. The staff reviewed the licensee's analysis, and agrees 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: York County Library, 138 East Black Street, Rock Hill, South Carolina. Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina. NRC Project Director: Herbert N. Berkow. 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: December 16, 1998, supplemented January 25, 1999. Description of amendment request: The proposed amendments would completely replace the High Pressure Injection (HPI) section of the Improved Technical Specifications that were issued on December 16, 1998. The proposed changes would: (1) expand the applicability for the requirements regarding the third HPI pump, discharge crossover valves, and the HPI suction headers; (2) specify the HPI conditions and allowed times that require the discharge headers be cross-connected or separated; (3) incorporate limiting conditions for operation when specified equipment was inoperable during specified plant conditions; (4) specify changes in HPI system discharge path valve lineup when certain equipment is inoperable; (5) change the requirement to reduce reactor power when an HPI system is inoperable from 60 percent power to 75 percent power and specify the length of time operation may continue at this power level; (6) address the failure to cross-connect the HPI [[Page 9188]] discharge headers as an independent condition; (7) add a requirement to verify by administrative means that the Atmospheric Dump Valve flow path for each steam generator is operable every 12 hours under certain conditions; (8) add a requirement that the HPI pump and crossover valves be restored to operable status within 30 days; (9) delete the requirement to restore the capability to automatically actuate the HPI within 24 hours; (10) add a Required Action to reduce reactor power to less than or equal to 75 percent power within 3 hours in the event an HPI train cannot be actuated by automatic or manual means; (11) expand the Completion Time for restoring an inoperable HPI train to 72 hours; (12) require that Limiting Condition for Operation 3.0.3 be entered immediately if two HPI trains or two HPI (low pressure injection) -LPI flow paths are inoperable; (13) change the surveillance requirement to manually cycle open each LPI-HPI flow path discharge valve every 18 months to require that the HPI discharge crossover valves be cycled every 18 months; and (14) add or modify various administrative and Bases changes that support the proposed changes. The licensee supplied data resulting from risk-informed analyses that were performed in accordance with Regulatory Guides 1.174 and 1.177 to support the evaluation. 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: No. The proposed change do not involve a physical alteration of the plant. No new or different equipment is being installed, and no installed equipment is being operated in a new or different manner. No set points for parameters which initiate protective or mitigative action are being changed. The proposed changes do not have any impact upon the ability of the HPI [High Pressure Injection] System to add soluble poison to the Reactor Coolant System. The remaining potential impact is upon the ability to mitigate the consequences of a small break LOCA [Loss-of-Coolant Accident], which is addressed below. The small break LOCA is the limiting design basis accident with respect to HPI System operability requirements. The Technical Specification requirements for the HPI System are supported by a spectrum of small break LOCA analyses based on the approved Evaluation Model described in FTI [Framatome Technologies Incorporated] topical report BAW-10192PA. These small break LOCA analyses demonstrate that the acceptance criteria of 10 CFR 50.46 are satisfied. The requirements of LCO [Limiting Condition for Operation] 3.5.2 assure that flow can be provided via two HPI trains (i.e., one HPI train responds automatically upon an ESPS [Engineered Safeguards Protective System] signal, and the second HPI train is aligned within 10 minutes via operator actions in the Control Room) following a small break LOCA and a single active failure. The full power small break LOCA analyses supporting this proposed license amendment have been performed in accordance with the approved Evaluation Model described in FTI topical report BAW-10192P. If enhanced steam generator cooling is not credited in the accident analysis, two HPI trains are required to mitigate specific small break LOCAs with Thermal Power [less than or equal to] 75% RTP [Reactor Thermal Power]. However, if equipment not qualified as QA-1 (i.e., an ADV [Atmosphic Dump Valve] flow path for one steam generator) is credited for enhanced steam generator cooling, the safety analyses have determined that the capacity of one HPI train is sufficient to mitigate a small break LOCA on the discharge of the reactor coolant pumps if Thermal Power [less than or equal to] 75% RTP. An ADV flow path for each steam generator is credited as a compensatory measure in Actions B and C of LCO 3.5.2 to permit operation to continue with THERMAL POWER [less than or equal to] 75% RTP: a) for 30 days with an HPI pump of one or more HPI discharge crossover valve(s) inoperable; and b) for 72 hours with one HPI train inoperable. This provides additional defense-in-depth, because the ADV flow path for each steam generator is required to be operable while only one is needed to perform the function. Additionally, a risk-informed assessment (provided as Attachment 7 to Duke's license amendment request dated December 18, 1998) concluded that operating the plant in accordance with the Required Actions was acceptable. The proposed changes involve crediting an additional operator action (i.e., steaming that steam generator through an ADV flow path) that has not previously been reviewed and approved by the staff for licensing basis small break LOCA analyses. Additionally, while the EFW System has been credited in past SBLOCA [small break LOCA] analyses as described in responses to NUREG-0565, actions to raise steam generator levels to the loss of subcooled margin setpoint were only assumed in the smaller SBLOCAs. These operator actions have been included in the Emergency Operating Procedure (i.e., AP/1, 2, or 3/A/1800/001) for many years. The times for completing these operator actions (i.e., feeding a steam generator via EFW [Emergency Feedwater] and steaming that steam generator through an ADV flow path) are new to the small break LOCA analysis and the licensing basis, and are considered reasonable. Crediting the performance of these operator actions within the specified time frames in the SBLOCA analyses does not result in any substantive change to the operator's response to [an] SBLOCA. In summary, the technical analyses described in this license amendment justify the adequacy of this specification and assure that operability of the HPI System is maintained in a manner consistent with the requirements of the design basis accidents. Therefore, it is concluded that this amendment request will not significantly increase the probability or consequences of an accident previously evaluated. (2) Create the possibility of a new or different kind of accident from any kind of accident previously evaluated: No. The proposed changes do not involve a physical alteration of the plant. No new or different equipment is being installed, and no installed equipment is being operated in a new or different manner. No set points for parameters which initiate protective or mitigative action are being changed. As a result, no new failure modes are being introduced. The requirements of ITS [Improved Technical Specification] 3.5.2 continue to assure that operability of the HPI System is maintained in a manner consistent with the requirements of the design basis accidents. The requirements are supported by small break LOCA analyses which demonstrate that the acceptance criteria of 10 CFR 50.46 are satisfied. The proposed change involve crediting an additional operator action (i.e., steaming that steam generator through an ADV flow path) that has not previously been reviewed and approved by the staff for licensing basis small break LOCA analyses. Additionally, while the EFW System has been credited in past SBLOCA analyses as described in responses to NUREG-0565, actions to raise steam generator levels to the loss of subcooled margin setpoint were only assumed in the smaller SBLOCAs. These operator actions have been included in the Emergency Operating Procedure (i.e., AP/1, 2, or 3/ A/1800/001) for many years. The times for completing these operator actions (i.e., feeding a steam generator via EFW and steaming that steam generator through an ADV flow path) are new to the small break LOCA analysis and the licensing basis, and are considered reasonable. Crediting the performance of these operator actions within the specified time frames in the SBLOCA analyses does not result in any substantive change to the operator's response to [an] SBLOCA. Therefore, this proposed amendment will not create the possibility of any new or different kind of accident. (3) Involve a significant reduction in a margin of safety. No. The requirements of ITS 3.5.2 continue to assure that operability of the HPI System is maintained in a manner consistent with the requirements of the design basis accidents. The requirements are supported by small break LOCA analyses which demonstrate that the acceptance criteria of 10 CFR 50.46 are satisfied. These analyses were performed in accordance with the Evaluation Model described in FTI topical report BAW-10192P. Therefore, it is concluded that the proposed amendment request will not result in a significant decrease in the margin of safety. [[Page 9189]] 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: J. Michael McGarry, III, Winston and Strawn, 1200 17th Street, NW., Washington, DC. NRC Project Director: Herbert N. Berkow. Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania Date of amendment request: January 18, 1999 Description of amendment request: The proposed amendments would: (1) delete license condition 2.C.(3) from the Beaver Valley Power Station, Unit No. 1 (BVPS-1) operating license and delete some references to two-loop operation from BVPS-1 Technical Specifications (TSs); (2) revise BVPS-1 and Beaver Valley Power Station, Unit No. 2 (BVPS-2) TS 2.2.1, 3.3.2.1, associated tables 2.2-1 and 3.3.4, and associated bases, to use consistent format and wording between units; (3) revise BVPS-1 and BVPS-2 TS 2.2.1, 3.3.2.1, associated tables 2.2-1 and 3.3.4, and associated bases, to include revised nominal trip setpoints and allowable values which are more conservative than those currently listed; (4) delete or revise TS to reflect the current configuration of Unit 1 plant hardware; and (5) make miscellaneous editorial changes to BVPS-1 and BVPS-2 TS and associated Bases to define terms, revise formatting, modify titles, and add license numbers to pages. 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 [as modified by the NRC staff based upon information provided elsewhere in the licensee's submittal]. 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? This proposed amendment includes changes to nominal Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) trip setpoints and allowable values that have been determined with the use of an approved methodology. The new values ensure that all automatic protective actions will be initiated at or before the condition assumed in the safety analysis. This change, which includes modification of the requirements stated in Limiting Safety System Setting (LSSS) 2.2.1 and Limiting Condition for Operation (LCO) 3.3.2.1, will allow the nominal trip setpoints to be adjusted within the calibration tolerance band allowed by the setpoint methodology. There will be no adverse effect on the ability of the channels to perform their safety functions as assumed in the safety analyses. Since there will be no adverse effect on the trip setpoints or the instrumentation associated with the trip setpoints, there will be no significant increase in the probability of any accident previously evaluated. Other changes in trip system function, content and format are proposed based on the current configuration of the trip system hardware at Beaver Valley Power Station (BVPS) Unit No. 1. Similarly, since the ability of the instrumentation to perform its safety function is not adversely affected, there will be no significant increase in the consequences of any accident previously evaluated. Since the safety analysis is unaffected by this change there is no change in the consequences of any previously evaluated accident. The editorial changes do not affect plant safety. The administrative change, for BVPS Unit 1 only, pertaining to two loop operation and Reactor Coolant System isolation valve position, does not affect plant safety. The Technical Specification requirements in LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop operation and] ensure safe plant operation by properly controlling the operation and position of the reactor coolant loops and Reactor Coolant System isolation valves. [The administrative change to delete line item 7.d, pertaining to Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not affect plant safety because this function is not directly initiated by bus undervoltage. Rather, the automatic start of the motor-driven AFW pumps is accomplished by the combination of 1) Emergency Bus feed breaker opening 2) valid start signal from ESFAS, and 3) Emergency Diesel Generator (EDG) sequencer actuation. Requirements for these items are included in the ESFAS related TS, Table 3.3-3 and 3.3-4 items 7.a, 7.c, 7.e, and EDG related TS 4.8.1.1.2.b.3 (b). Therefore, since there is no change made to the plant hardware or its operation and requirements related to the AFW pump auto-start function are maintained elsewhere in the BVPS-1 TS, deleting line item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not change the probability or consequences of any accident previously evaluated.] Therefore, this change does not involve any significant increase in the probability of occurrence of any accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed amendment includes changes to the format and magnitudes of nominal trip setpoints and allowable values that preserve all safety analysis assumptions related to accident mitigation. The protection system will continue to initiate the protective actions as assumed in the safety analysis. The proposed changes to LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis. This proposed amendment does not involve additional hardware changes. Plant operation will not be changed. Other proposed changes are made so that the Technical Specifications more accurately reflect the plant-specific trip system hardware in BVPS Unit No. 1. Furthermore, the proposed changes do not alter the functioning of the RTS and ESFAS. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the change involve a significant reduction in a margin of safety? The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The proposed RTS and ESFAS trip setpoints are calculated with an approved methodology. The proposed changes to LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis. Therefore, the response of the RTS and ESFAS to accident transients reported in the Updated Final Safety Analysis Report is unaffected by this change. No additional hardware changes are involved. Therefore, accident analysis acceptance criteria are not affected. Other proposed changes are made so that the protection system Technical Specifications more accurately reflect the plant-specific trip system hardware in BVPS Unit No. 1. The editorial changes do not affect plant safety. The administrative change, for BVPS Unit 1 only, pertaining to two loop operation and Reactor Coolant System isolation valve position, does not affect plant safety. The Technical Specification requirements in LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop operation and] ensure safe plant operation by properly controlling the operation and position of the reactor coolant loops and Reactor Coolant System isolation valve. [The administrative change to delete line item 7.d, pertaining to Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not affect plant safety because this function is not directly initiated by bus undervoltage. Rather, the automatic start of the motor-driven AFW pumps is accomplished by the combination of (1) Emergency Bus feed breaker opening, (2) valid start signal from ESFAS, and (3) EDG sequencer actuation. Requirements for these items are included in the ESFAS related TS, Table 3.3-3 and 3.3-4 items 7.a, 7.c, 7.e, and EDG related TS 4.8.1.1.2.b.3 (b). Therefore, since there is no change made to the plant hardware or its operation and requirements related to the AFW pump auto-start function are maintained elsewhere in the BVPS-1 TS, [[Page 9190]] deleting line item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not involve a significant reduction in a margin of safety.] Therefore, operation of the facility in accordance with the proposed amendment 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: B.F. Jones Memorial Library, 663 Franklin Avenue, Aliquippa, PA 15001. Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037. NRC Project Director: S. Singh Bajwa. Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50- 458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana Date of amendment request: December 16, 1998. Description of amendment request: The licensee has proposed an amendment of Facility Operating License No. NPF-47, Appendix A-- Technical Specifications, Section 2.1.1.2, entitled ``Reactor Core [Safety Limits].'' The proposed amendment will change the two recirculation loop Minimum Critical Power Ratio (MCPR) limit from 1.13 to 1.12 and the single recirculation loop MCPR limit from 1.14 to 1.13. The revised limits are necessary to address the operation of Cycle 9 following the refueling outage which is scheduled to begin April 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: 1. The request does not involve a significant increase in the probability or consequences of an accident previously evaluated. The plant/cycle specific SLMCPRs have been calculated using methods identical to those used by General Electric (GE) to assess the SLMCPR for other Boiling Water Reactors (BWRs). Similar methods were used to determine the value of the SLMCPR for the previous cycle. These methods are within the existing design and licensing basis and cannot increase the probability or severity of an accident. The basis of the SLMCPR calculation is to ensure that greater than 99.9% of all fuel rods in the core avoid transition boiling and fuel damage in the event of the occurrence of Anticipated Operational Occurrences (AOO) or a postulated accident. The SLMCPR is used to establish the Operating Limit Minimum Critical Power Ratio (OLMCPR). Neither the SLMCPR nor the OLMCPR are initiators or affect initiators of an accident previously evaluated and therefore changes to the SLMCPR do not increase the probability of any accident previously evaluated. The proposed changes involve the use of an accepted methodology in calculating the SLMCPR and, since there is no change in the definition of the SLMCPR, these changes will not affect the consequences of any accident previously evaluated. In addition, the proposed changes do not involve any change in the way the plant is operated. Existing procedures will ensure that the SLMCPR is not violated. Therefore, these changes have no effect on the consequences of an accident. On these bases, there will be no increase in the probability or consequences of an accident previously analyzed as a result the proposed changes. 2. The request does not create the possibility of occurrence of a new or different kind of accident from any accident previously evaluated. The proposed changes consist of SLMCPR calculated from an accepted method of analysis that has been used by many BWRs. These changes do not involve any alteration of the plant and do not affect the plant operation. Neither the SLMCPR nor the OLMCPR can initiate an event, therefore a change to the SLMCPR does not create the possibility of occurrence of a new or different kind of accident from any accident previously evaluated. 3. The request does not involve a significant reduction in the margin of safety. The SLMCPR is a Technical Specification numerical value to ensure that 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated. The proposed SLMCPR change results from SLMCPR analysis using the accepted methods as identified in the Attachment. The margin of safety resides between the SLMCPR and the point at which fuel fails. Maintaining the MCPR above the proposed SLMCPR will maintain the margin of safety associated with GE's SLMCPR methodology. Existing plant procedures will continue to ensure that the SLMCPR is not violated. Therefore, this request 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: Government Documents Department, Louisiana State University, Baton Rouge, LA 70803. Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, N.W., Washington, D.C. 20005. NRC Project Director: John N. Hannon. Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: December 23, 1998. Description of amendment request: The proposed changes will modify the Limiting Condition for Operation for Technical Specifications 3.3.3.7.1 for the chlorine detection system at Waterford Steam Electric Station, Unit 3. A change in the alarm/trip setpoint from 3 parts per million (ppm) to 2 ppm is requested. Additionally, the proposed request corrects a typographical error in Table 3.3-4. 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: The chlorine detection system has no effect on the accidents analyzed in Chapter 15 of the Final Safety Analysis Report. Its only effect is on habitability of the control room, which will be enhanced by specifying a more conservative setpoint in the Technical Specifications (TS). Analysis using more conservative assumptions show that a setpoint of 2 parts per million (ppm) chlorine is acceptable. Correcting the typographical error on TS page 3/4 3-19 has no effect on the probability or consequences of an accident previously evaluated. 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: The proposed Technical Specification change in itself does not change the design or configuration of the plant. Using a more conservative setpoint performs the same function as the old setpoint, but it accomplishes this function with increased conservatism. Correcting the typographical error on TS page 3/4 3-19 will not create the possibility of a new or different type of accident from any accident previously evaluated. 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 [[Page 9191]] involve a significant reduction in a margin of safety? Response: The chlorine detection system has no effect on a margin of safety as defined by Section 2 of the Technical Specifications. Its only effect is on habitability of the control room, which will be enhanced by a more conservative setpoint provided by this change to the Technical Specifications. Correcting the typographical error on TS page 3/4 3-19 does not involve a significant reduction in a margin of safety. 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, LA 70122. Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 L Street N.W., Washington, D.C. 20005-3502. NRC Project Director: John N. Hannon. Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: January 25, 1999. Description of amendment request: The proposed change request will modify Technical Specification (TS) 3.5.1 to allow up to 72 hours to restore safety injection tank (SIT) to operable status if one SIT is inoperable due to boron concentration not within the limits or the inability to verify level and pressure. The proposed change would also allow up to 24 hours to restore SIT to operable status if one SIT is inoperable due to other reasons when Reactor Coolant System pressure is greater than or equal to 1750 psia. The ACTIONS for an inoperable SIT are being subdivided based on pressurizer pressure to be consistent with the current Waterford 3 requirements and applicability. Additionally, the Surveillance requirement to sample the SIT after a 1% volume increase is being changed to not be required if the source of the makeup is the refueling water storage pool. This amendment request is a collaborative effort of participating Combustion Engineering Owners Group members based on a review of plant operations, deterministic and design basis considerations, and plant risk, as well as previous generic studies and conclusions drawn by the NRC Staff and contained within NUREG-1366, ``Improvements to Technical Specifications Surveillance Requirements,'' and NUREG-1432, Revision 1, ``Standard Technical Specifications for Combustion Engineering (CE) Plants.'' TS Bases 3/4.5.1 will be revised to support above changes. 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: The Safety Injection Tanks (SITs) are passive components in the Emergency Core Cooling System. The SITs are not an accident initiator in any accident previously evaluated. Therefore, this change does not involve an increase in the probability of an accident previously evaluated. The SITs were designed to mitigate the consequences of Loss of Coolant Accidents (LOCA). These proposed changes do not affect any of the assumptions used in deterministic LOCA analyses. Hence the consequences of accidents previously evaluated do not change. In order to fully evaluate the affect of the SIT Allowed Outage Time (AOT) extension from 1 hour to 24 hours when one SIT is inoperable for reasons other than boron concentration or inability to measure level or pressure, probabilistic safety analysis (PSA) methods were utilized. The results of these analyses show no significant increase in the core damage frequency. As a result, there would be no significant increase in the consequences of an accident previously evaluated. These analyses are detailed in CE NPSD-994, Combustion Engineering Owners Group ``Joint Applications Report for Safety Injection Tank AOT/STI Extension.'' The proposed change to extend the AOT from 1 hour to 72 hours when unable to measure level or pressure is acceptable because SIT operability is not based on instrumentation availability. Therefore, this does not involve a significant increase in the consequences of an accident as evaluated and are endorsed by the Nuclear Regulatory Commission (NRC) in NUREG-1366, ``Improvements to Technical Specifications Surveillance Requirements.'' The inability to measure level or pressure is acceptable because the SIT instrumentation provides no safety actuation. The AOT extension from 1 hour to 72 hours, based upon boron concentration outside the prescribed limits does not involve a significant increase in the consequences of an accident as evaluated and approved by the NRC in NUREG-1432, ``Standard Technical Specifications for Combustion Engineering Plants.'' These changes are acceptable because the reduced concentration effects on core subcriticality during reflood are minor. The change in sampling requirements to not require sampling if the makeup source is of the same concentration limit as the SIT is acceptable as the concentration will remain within the TS limits. Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: The proposed change does not alter the design or configuration of the plant. It also does not alter the mitigation capabilities of any safety system or components. This change increases the AOTs for the condition of SIT inoperability. The boron concentration is maintained by make-up from a source of water with the required concentration of the SITs. 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: The proposed changes do not affect the limiting conditions for operation or their bases that are used in the deterministic analyses to establish the margin of safety. PSA and deterministic evaluations were used to evaluate these changes. The PSA evaluations demonstrated that the applicable changes are either risk neutral or risk beneficial. These evaluations are detailed in CE NPSD-994. The deterministic evaluations show that the SITs would be able to perform their safety function. These changes are consistent with NUREG-1366 and NUREG-1432. The margin of safety is not significantly affected by makeup from a source of the same concentration limit as the SIT or increase in the AOT for boron concentration of one SIT not within limits. 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, LA 70122. Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 L Street N.W., Washington, D.C. 20005-3502. NRC Project Director: John N. Hannon. [[Page 9192]] Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request: January 25, 1999. Description of amendment request: The proposed changes modify Technical Specifications Section 6.0 to remove certain administrative controls and instead rely on the change controls of 10 CFR 50.54(a)(3) and to add a requirement to Section 6.0 concerning the responsibilities of the General Manager Plant Operations. The requested changes are consistent with the Improved Standard Technical Specifications for Combustion Engineering plants, NUREG-1432. 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: The requested changes are purely administrative in nature. The proposed changes do not affect the operation of any structures, systems, or components or the assumptions of any accident analyses. The requested changes only affect Section 6.0 of the Waterford 3 Technical Specifications which describe the administrative controls to be implemented at the site. The requested changes either add an additional administrative requirement or remove quality assurance program details from the Technical Specifications. The details are being removed from the Technical Specifications and instead rely on the change controls of 10 CFR 50.54(a)(3). This submittal makes no changes to the regulatory controls governing changes. The requested changes are purely administrative in nature. Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: The proposed changes to the Technical Specification requirements are purely administrative in nature and do not involve a change in plant design or affect the configuration or operation of any structure, system, or component. 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: The proposed changes do not affect the operation of any structures, systems, or components or the assumptions of any accident analyses. The requested changes are purely administrative in nature. 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, LA 70122. Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 L Street N.W., Washington, D.C. 20005-3502. NRC Project Director: John N. Hannon. IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: January 22, 1999. Description of amendment request: The proposed amendment would revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) Section 4.3, ``Fuel Storage,'' by updating the criticality requirements (k-infinity and U-235 enrichment limits) for storage of fuel assemblies in the spent fuel racks. This change would allow for storage of nuclear fuel assemblies with new designs, including GE-12 with a 10X10 pin array. 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: After reviewing this proposed amendment, we have concluded: 1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. The probability of occurrence of the accident/abnormal conditions evaluated in UFSAR Section 9.1.2.3 is not significantly increased by this change because no modification in fuel handling equipment, fuel pool cooling equipment, fuel storage racks, or fuel handling practices is taking place. Only the k-infinity and enrichment limits for the stored fuel are being changed. The postulated accident/abnormal conditions evaluated in UFSAR Section 9.1.2.3 have been re-evaluated for the proposed changes in k-infinity and enrichment limits. The results demonstrate that the consequences are negligible. The analyses performed show that the requirement to maintain K-eff less than 0.95 (substantially subcritical) is satisfied for normal and postulated abnormal conditions using methods and assumptions that are consistent with the existing UFSAR. Seismic adequacy and structural integrity of the pool and racks are not affected by the introduction of GE-12 fuel. Local and bulk pool temperatures remain bounded by the current UFSAR analysis for fuel exposures with GE-12 fuel expected through two cycles of operation (i.e., through Cycle 18 operation). Based upon a scoping study comparing the hydraulic diameters of GE-10 and GE-12 fuel, large margins to pool boiling conditions at the final discharge exposures of GE-12 fuel will be maintained. Therefore, the consequences of the accident are not significantly increased by this change. 2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. No new types of accidents are being introduced because no modification in fuel handling equipment, fuel pool cooling equipment, fuel storage racks or fuel handling procedures is being made. The design basis function of the spent fuel racks is to maintain the fuel configuration substantially subcritical and within allowable temperatures under both normal and postulated abnormal conditions. This design basis function will be maintained with the proposed k-infinity and enrichment limits. 3. The proposed amendment will not involve a significant reduction in a margin of safety. The margin of safety is not significantly reduced. This margin is based on the requirement to limit the K-eff of fuel in the spent fuel racks to less than 0.95. The proposed changes in k-infinity and enrichment limits have been shown to meet this requirement, using methods and assumptions that are consistent with the existing UFSAR. Seismic adequacy and structural integrity of the pool and racks are not affected by the introduction of GE-12 fuel. Local and bulk pool temperatures remain bounded by the current UFSAR analysis for fuel exposures with GE-12 fuel expected through two cycles of operation (i.e., through Cycle 18 operation). Based upon a scoping study comparing the hydraulic diameters of GE-10 and GE-12 fuel, large margins to pool boiling conditions at the final discharge exposures of GE-12 fuel will be maintained. Based upon the above, we have determined that the proposed amendment will not involve a significant hazards consideration. 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: Cedar Rapids Public Library, 500 First Street, SE., Cedar Rapids, IA 52401. [[Page 9193]] Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869. NRC Project Director: Cynthia A. Carpenter. IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: October 15, 1998, as supplemented on December 21, 1998. Description of amendment request: The proposed amendment would revise the Duane Arnold Energy Center (DAEC) Technical Specifications (TS) by adding a new TS 3.7.9, ``Control Building/Standby Gas Treatment System (CB/SBGT) Instrument Air System.'' The proposed amendment would also revise (TS) 3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' Condition E, by adding a time limit for plant operation if a penetration flow path is isolated by a single purge valve with resilient seal. 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 will not involve a significant increase in the probability or consequences of an accident previously evaluated. The amendment is adding new requirements for the CB/SBGT Instrument Air System that are commensurate with the safety functions it supports and consistent with other support systems in the Technical Specifications. These requirements provide appropriate actions and time limits for plant operation with one or both CB/SBGT Instrument Air subsystems inoperable. The probability of an event while in this condition is low, and the consequences are bounded by the failure of the supported systems. The CB/SBGT Instrument Air System is not assumed to be an initiator of an analyzed event. The amendment is also adding a time limit for plant operation if a purge valve with resilient seal is used to satisfy TS 3.6.1.3 Required Action E.1 (isolate the affected penetration flow path). While primary containment integrity is provided by the purge valve, it is prudent to limit operation in this condition due to the potential for increased leakage from a single active failure. These additions will provide assurance that affected systems will be OPERABLE when required and as assumed in the design basis. This change will not physically alter the plant (no new or different type of equipment will be installed). This change will not alter the operation of process variables, structures, systems, or components as described in the safety analysis. This change will not alter assumptions relative to the mitigation of an accident or transient event. This change will not increase the probability of initiating, or the consequences of an analyzed event. (2) The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The amendment adds new requirements for the CB/SBGT Instrument Air System and adds a time limit for plant operation if a purge valve with resilient seal is used to satisfy TS 3.6.1.3 Required Action E.1. This change will not physically alter the plant (no new or different type of equipment will be installed). This change will not alter the operation of process variables, structures, systems, or components as described in the safety analysis. Thus, a new or different kind of accident will not be created. (3) The proposed amendment will not involve a significant reduction in a margin of safety. The amendment is adding new requirements for the CB/SBGT Instrument Air System to provide appropriate actions and time limits for plant operation with one or both CB/SBGT Instrument Air subsystems inoperable. The amendment is also adding a time limit for plant operation if a purge valve with resilient seal is used to satisfy TS 3.6.1.3 Required Action E.1 (isolate the affected penetration flow path). While primary containment integrity is provided by the purge valve, it is prudent to limit operation in this condition due to the potential for increased leakage from a single active failure in the remaining OPERABLE components. This change will not physically alter the plant (no new or different type of equipment will be installed). This change will not alter the operation of process variables, structures, systems, or components as described in the safety analysis. This change will not alter assumptions relative to the primary success path for mitigation of an accident or transient event. These additions will provide assurance that the accident mitigation functions will perform as assumed in the safety analysis. Thus, the margin of safety will not be reduced. 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: Cedar Rapids Public Library, 500 First Street, SE., Cedar Rapids, IA 52401. Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869. NRC Project Director: Cynthia A. Carpenter. Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date of amendment request: January 29, 1999. Description of amendment request: The amendment would revise the technical specifications (TS) to relocate three cycle-specific parameter limits; shutdown margin with Tcold>210 deg.F, moderator temperature coefficient, and minimum boric acid storage tank level versus concentration, to the Core Operating Limits Report (COLR). 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 safety analysis most impacted by a change to the negative Moderator Temperature Coefficient (MTC) limit is the Main Steam Line Break (MSLB) event. The Steam Line Break Cooldown curves for an MTC are calculated and then input to the cycle-specific MSLB analysis (if necessary) during the reload analysis process, using an NRC- approved methodology. The required/acceptable Shutdown Margin (SDM) is dependent upon the core loading pattern used (i.e., cycle- specific core physics parameters) and is largely dependent on the cycle-specific MTC and available scram worth. The SDM is determined based on the analysis of the Hot Zero Power (HZP) MSLB event in which the return-to-critical and return-to-power conditions are evaluated to provide acceptable results. With the ongoing changes in MTC as a result of core loadings for FCS and higher U-235 enrichments, the end-of-cycle MTC is becoming more negative than the present Technical Specifications limit. Since the MTC is fuel cycle specific and influences the required SDM, it is appropriate to move both of these values to the COLR, consistent with Generic Letter 88- 16. Note that no change to the SDM for Tcold 210 deg.F is being proposed. The cycle-specific reload analysis is performed for every operating cycle and the results, as incorporated into the COLR pursuant to the 10 CFR 50.59 process, are transmitted to the NRC. FCS will continue to provide COLR updates to the NRC. The relocation of the negative MTC and the ``BAST level versus BAST Concentration'' curves into the COLR, consistent with the NRC recommendations of Generic Letter 88-16, will not modify the methodology used in generating the limits, nor the manner in which they are implemented. These limits will continue to be determined by analyzing the same postulated events as previously analyzed. FCS will continue to operate within the limits specified in the COLR and will take the same corrective actions when or if these limits are exceeded as required by [[Page 9194]] current Technical Specifications. The potential increase of the absolute magnitude of the negative MTC with Shutdown Margin decrease is evaluated during the COLR reload analysis process in accordance with OPPD's NRC-approved topical report. Therefore, this proposed amendment is administrative in nature and has been concluded not to increase 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 changes to FCS Technical Specifications were the result of a recommendation from a Generic Letter. Future changes to the parameters being relocated to the COLR can only be performed with approved Reload Analyses. No new or different kind of accident is created by this administrative change because the actual operation of FCS remains unchanged. Therefore the possibility of an accident or malfunction of a different type than previously evaluated in the safety analysis report would not be created. 3. The proposed change does not involve a significant reduction in a margin of safety. As indicated above, the implementation of this proposed COLR change, consistent with the guidance of Generic Letter 88-16, makes use of the existing safety analysis methodologies and the resulting limits and setpoints for plant operation. Additionally, the safety analysis acceptance criteria for operation with this proposed amendment have not changed from the criteria used in the current reload analysis. Therefore, the margin of safety as defined in the bases of Technical Specifications is not reduced. 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: W. Dale Clark Library, 215 South 15th Street, Omaha, Nebraska 68102. Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L Street, N.W., Washington, DC 20005-3502. NRC Project Director: William H. Bateman. PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: January 12, 1999. Description of amendment request: The proposed change involves revising Technical Specification (TS) Section 3/4.4.2, ``Safety/Relief Valves,'' and TS Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2, to increase the allowable as-found main steam Safety Relief Valve (SRV) code safety function lift setpoint tolerance from plus or minus 1% to plus or minus 3%. This change will also require the as-left SRV code safety function lift setting to be set within plus or minus 1% of the specified nominal lift setpoint prior to reinstallation in the plant. In support of this proposed TS change, the required number of OPERABLE SRVs in Operational Conditions (OPCONs) 1, 2, and 3 will be changed from 11 to 12. The number of SRVs in each lift pressure grouping will remain the same. This proposed TS change does not alter the SRV nominal lift setpoints or the SRV lift setpoint test frequency currently specified by TS Section 3/4.4.2. The proposed change does not change the SRV testing commitment specified in LGS Updated Final Safety Analysis Report (UFSAR) Chapter 5.2.2.10, ``Inspection and Testing.'' 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 Technical Specifications (TS) changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed TS changes allow for an increase in the as-found main steam Safety Relief Valve (SRV) setpoint tolerance from plus or minus 1% to plus or minus 3%. The proposed changes also reduce the allowable number of SRVs to be out-of-service from three (3) to two (2). The proposed changes do not alter the SRV nominal lift setpoints or SRV lift setpoint test frequency. The actuation of an SRV is the precursor to the inadvertent opening of a SRV transient, as discussed in Updated Final Safety Analysis Report (UFSAR) Chapter 15.1.4. Increasing the allowable as-found SRV code safety function lift setpoint tolerance from plus or minus 1% to plus or minus 3% does have the potential for the minimum SRV simmer margin to be reduced from 113.3 psig to 89.9 psig. A reduction in simmer margin will not directly result in an increase of the probability on an inadvertent self actuation of an SRV. A reduction in simmer margin will reduce the seating force which may initiate leakage. However, this leakage is monitored and corrective actions can be implemented prior to progressing to the point of the potential of an inadvertent actuation. This reduction in SRV simmer margin has been evaluated by the SRV manufacturer and determined to be acceptable; therefore, the probability of an inadvertent SRV actuation remains unchanged. Actuation of an SRV is not a precursor for any other event evaluated in the Safety Analysis Report (SAR). The proposed TS changes have been evaluated on both a generic and plant specific basis. The NRC has approved the general approach of this change; however, implementation is contingent on several plant specific evaluations. The required plant specific analyses and evaluations included transient analysis of the anticipated operational transients (AOTs); analysis of the design basis overpressurization event; evaluation of the performance of high pressure systems, motor operated valves, and vessel instrumentation and associated piping; and evaluation of the containment response during Loss-of-Coolant Accident (LOCA) and hydrodynamic loads on the SRV discharge lines and containment. In addition to the plant specific analyses and evaluations required by the NRC, the following items were also considered: ECCS/LOCA [Emergency Core Cooling System] performance, SRV simmer margin, high pressure--low pressure interfaces, i.e., High Energy Line Break (HELB), Station Blackout (SBO), and Fire Safe Shutdown (FSSD), and the short term pressurization phase of an ATWS [anticipated transient without scram] event. These analyses and evaluations show that there is adequate margin to the design core thermal limits and reactor vessel pressure limits using the plus or minus 3% SRV code safety function lift setpoint tolerance and two (2) SRVs out-of-service. The analyses and evaluations also show that the operation of the high pressure injection systems will not be adversely affected, that SRV discharge piping stresses will not be exceeded, and that the containment response during a LOCA will be acceptable. Evaluations of the impact of the proposed change on the Equipment Important to Safety have been performed and no adverse conditions were identified. The reactor pressure vessel and attached systems and piping have been evaluated for the impact of this proposed TS change. A plant specific analysis has been performed which indicates that neither the American Society of Mechanical Engineers (ASME) Code upset limits or the TS Safety Limits for the reactor pressure vessel will be exceeded for the limiting event, i.e., Main Steam Isolation Valve (MSIV) closure with flux Scram. The reactor pressure vessel and attached piping design values will not be exceeded. The current high pressure--low pressure interface evaluation utilized nominal SRV setpoints, and therefore, is unaffected. Therefore, the probability of a malfunction of the reactor pressure vessel and attached systems and piping is not increased. The nuclear fuel has been evaluated for the impact of the proposed change. Plant specific analyses were performed which indicate that for all abnormal operational transients adequate margin to the limiting thermal limit parameter, i.e., Minimum Critical Power Ratio (MCPR), is maintained. Emergency Core Cooling System (ECCS)/LOCA performance is maintained adequate to meet the requirements of 10CFR50.46. Therefore, the probability of the malfunction of the nuclear fuel is not increased. The SRVs have been evaluated for the impact of the proposed TS changes. No physical changes to the SRVs will be made as a result of the proposed TS changes. Adequate simmer margin will be maintained with the increased tolerance to ensure that an inadvertent lifting of a SRV does not occur. [[Page 9195]] The increase in SRV discharge flow and reactor vessel pressure due to the potential for higher SRV lift setpoints are bounded by the SRV steam flows and reactor vessel pressure currently used in the evaluation of SRV discharge piping, quencher, quencher support, and hydrodynamic loads on the suppression pool and submerged structures; therefore, the probability of a malfunction of a SRV or associated components and structures is not increased. The Containment response during a LOCA has been evaluated for the impact of the proposed change. The major factor in the Containment response to a LOCA is the rate of reactor vessel water inventory loss. The rate of reactor vessel water inventory loss is mainly dependent on reactor decay heat which is not affected by the proposed change. Therefore, the probability of the malfunction of the Containment is not increased. The High Pressure Coolant Injection (HPCI) system has been evaluated for the impact of the proposed TS changes. The analysis determined that the HPCI system would not be capable of developing its design flowrate of 5600 gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint +3% tolerance) unless the HPCI turbine/ pump maximum rated speed was increased. However, increasing the HPCI turbine/pump maximum rated speed is prevented due to HPCI pump discharge piping overpressurization concerns. Further analysis has shown that the HPCI system is capable of meeting its required ECCS function design flowrate, and its required non-ECCS flowrate, without any change to the current system operating parameters. Therefore, the probability of a malfunction of the HPCI System is not increased. The Reactor Core Isolation Cooling (RCIC) system has been evaluated for the impact of the proposed change. The analysis determined that in order for the RCIC system to be capable of injecting its design flowrate of 600 gpm at a reactor pressure of 1205 psig (lowest SRV setpoint of 1170 psig +3% tolerance) the maximum rated speed of the RCIC turbine/pump is required to be increased from 4575 rpm to 4625 rpm. This increase in the RCIC turbine/pump maximum rated speed will reduce the margin to the overspeed trip from 123% to 122.1%. This reduction in the margin to the overspeed trip is acceptable due to the implementation of plant Modification P00210, ``RCIC System Startup Transient Improvement,'' which reduced the amount of turbine/pump speed overshoot during system startup. The RCIC overspeed trip setpoint will not be changed; therefore, a failure of the RCIC turbine/pump (missile hazard or system overpressurization) due to overspeed is not increased. All other RCIC System components will continue to operate within the currently specified design and operating limits. Therefore, the probability of a malfunction of the RCIC System is not increased. The Standby Liquid Control (SLC) system has been evaluated for the impact of the proposed change. The SLC system capability of shutting down the reactor during a postulated event in which all or some of the control rods cannot be inserted or during a postulated Anticipated Transient Without Scram (ATWS) event is not impacted by this proposed change. Therefore, the probability of a malfunction of the SLCS is not increased. The Control Rod Drive (CRD) system has been evaluated for the impact of the proposed change. The CRD system capability of controlling reactor power during normal plant operation and rapidly inserting control rod blades (Scram) during abnormal plant conditions is not impacted by the proposed change. Therefore, the probability of a malfunction of the CRD system is not increased. The Reactor Vessel Instrumentation System has been evaluated for the impact of the proposed change. The Reactor Vessel Instrumentation System will continue to be operated within the current design pressure/temperature requirements; therefore, the probability of a malfunction of the Reactor Vessel Instrumentation System is not increased. The LGS, Units 1 and 2, Generic Letter 89-10 Motor-Operated Valve (MOV) Program has been evaluated for the proposed change. The LGS MOV Program currently uses SRV nominal setpoints for differential pressure determinations for valves in which reactor pressure at the SRV setpoint is limiting. Use of nominal SRV setpoints is consistent with current industry practice. Therefore, the probability of a malfunction of a MOV is not increased. Reducing the number of SRVs allowed to be out-of-service does not make the consequences of a malfunction of a SRV more severe, since the number of SRVs required to maintain the reactor vessel within ASME Code and TS Safety Limits will be maintained OPERABLE. The proposed change does not result in any changes to the interactions of any system, structure, or component. All systems, structures, and components will continue to function as designed. Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated. 2. The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed TS changes allow for an increase in the as-found SRV setpoint tolerance from plus or minus 1% to plus or minus 3%. The proposed TS changes also reduce the allowable number of SRVs to be out-of-service from three (3) to two (2). Generic and plant specific analyses and evaluations indicate that the plant response to any previously evaluated event will remain unchanged. All plant systems, structures, and components will continue to be capable of performing their required safety function as required by event analysis guidance. The proposed TS changes do not alter the SRV nominal lift setpoints or SRV lift setpoint test frequency. The operation and response of the affected Equipment Important to Safety is unchanged. All systems, structures, and components will continue to be operated within acceptable operating and/or design parameters. No system, structure, or component will be subjected to a condition that has not been evaluated and determined to be acceptable using the guidance required for specific event analysis. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. The proposed TS changes do not involve a significant reduction in a margin of safety. The proposed TS changes allow for an increase in the as-found SRV setpoint tolerance from plus or minus 1% to plus or minus 3%. The proposed TS changes also reduce the allowable number of SRVs to be out-of-service from three (3) to two (2). The proposed TS changes do not alter the SRV nominal lift setpoints or SRV lift setpoint test frequency. The operation and response of the affected Equipment Important to Safety is unchanged. All systems, structures, and components will continue to be operated within acceptable operating and/or design parameters. While the calculated peak reactor vessel pressure for the ASME overpressure event and the ATWS Pressure Regulator Failure-Open (PREGO) event are higher than those calculated without the increase in setpoint tolerance, both are still within the respective licensing acceptance limits associated with these events. These licensing acceptance limits have been determined by the NRC to provide a sufficient margin of safety. The increase in the RCIC system turbine/pump maximum rated speed is within the capability of the system design. The reduction in the margin to the overspeed trip is not a reduction in the margin of safety, since the operation of the RCIC System has demonstrated minimal speed overshoot on system initiation due to the installation of plant Modification P00210, ``RCIC System Startup Transient Improvement.'' The inability of the HPCI system to be capable of injecting 5600 gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint of 1170 psig +3% tolerance) is not a reduction in the margin of safety, since analysis for events that would result in high reactor vessel pressure indicate that the HPCI System is capable of providing adequate coolant injection. The increase in SRV steam flow and reactor vessel pressure does not reduce the margin of safety associated with the SRVs and associated components and structures since the increased SRV steam flow rate and reactor vessel pressure are bounded by the current design analysis. The margin of safety for fuel thermal limits and 10CFR50.46 limits is unaffected by the proposed change. The margin of safety for the Containment is unaffected by the proposed change. The capability of the SLC system to perform its safety function during all required events, using the required guidance for event analysis, is maintained. Therefore, the proposed changes do not reduce the margin of safety provided by the SLC system. Therefore, these proposed TS changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this [[Page 9196]] 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: Pottstown Public Library, 500 High Street, Pottstown, PA 19464. Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, PA 19101. NRC Project Director: William M. Dean. PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: January 25, 1999. Description of amendment request: The proposed Technical Specification (TS) Change Request revises the TS Surveillance Requirement frequencies for Sections 4.8.1.1.2.e.1, 4.8.1.1.2.e.8.a, and 4.8.1.1.2.e.8.b for the Emergency Diesel Generator maintenance inspection outages, the 24-hour endurance run, and for the hot restart test from 18 to 24 months. 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 Technical Specifications (TS) changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The maintenance inspection interval change and the corresponding interval change for the associated 24 hour endurance test and hot restart test which are normally performed in conjunction with the diesel preventive maintenance overhaul inspections, as well as the programmatic improvements addressed here do not involve physical changes that would affect the ability of the EDGs [emergency diesel generators] to perform their safety function. The Emergency Diesel Generator System is not an accident initiator. The Surveillance Testing requirements of Technical Specification Section 3/4.8 will continue to verify the operability and reliability of the Emergency Diesel Generator system. The proposed changes do not affect the ability of the EDGs to mitigate the consequences of an accident, including the Loss of Coolant Accident (LOCA) coupled with Loss Of Offsite Power accident analyses as presented in Chapter 15 of the LGS [Limerick Generating Station] UFSAR [Updated Final Safety Analysis Report]. EDG unavailability due mostly to outage inspections is more than 2 times higher than EDG unplanned unavailability. An extension of the outage inspection frequency to 24 months will result in increased EDG availability to mitigate the consequences of a potential accident. When this program is taken in its entirety the extended maintenance intervals coupled with the defined enhancements is judged to result in an overall increase in EDG availability and reliability. Therefore, the probability or consequences of an accident previously evaluated is not increased. 2. The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The Emergency Diesel Generator system is not an accident initiator. The operation and design of the onsite emergency power system (including the EDGs) is not being changed; only the overhaul inspection interval coupled with the program improvements and the corresponding interval change for the associated 24 hour endurance test and hot restart test, (which are normally performed in conjunction with the diesel preventive maintenance overhaul inspections), are changed. The EDG system meets the single failure criteria at the EDG unit level, i.e., the SAR [safety analysis report] states that with one EDG failed or out-of-service, the standby AC system is capable of furnishing sufficient power for the minimum Class 1E load demand, assuming a limiting design basis accident has occurred. The proposed changes involve a routine preventive maintenance and inspection time interval change along with the corresponding surveillance test interval changes, and also include programmatic improvements to reduce the likelihood of a failure of an individual EDG unit; the proposed changes do not involve any physical design or operational changes that could create a malfunction extending beyond an individual EDG nor do they increase the potential for a common-mode EDG failure. Therefore, it is not possible to create a new or different type of accident through implementation of these changes. 3. The proposed TS changes do not involve a significant reduction in a margin of safety. The changes to bring the frequencies of the EDG overhaul, the 24 hour endurance test and the associated hot restart test into alignment with the current 2 year operating cycle, and the detailed programmatic changes to achieve conformance with the FMOG [Fairbanks Morse Owners Group] recommended maintenance program, will increase the reliability and availability of the EDG system. This will enhance the margin of safety as the amount of time the EDGs are out- of-service will decrease and the system will be single-failure proof for more clock hours when the nuclear reactor(s) are operating. The changes discussed here do not result in operation of the emergency diesel generator system nor any other plant system in a manner beyond their original design basis, and thus does not reduce any explicit or implicit Technical Specification 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: Pottstown Public Library, 500 High Street, Pottstown, PA 19464. Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, PA 19101. NRC Project Director: William M. Dean. Portland General Electric Company, et al., Docket No. 50-344, Trojan Nuclear Plant, Columbia County, Oregon Date of application for amendment: February 12, 1997. Brief description of amendment: The proposed amendment would delete a portion of the Trojan site from the 10 CFR 50 license when that portion of the site, designated for use as an independently licensed spent fuel storage installation (ISFSI), receives a part 72 license. 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. The NRC staff has reviewed the licensees' analysis against the standards of 10 CFR 50.92(c). The licensee's analysis is summarized below: The proposed changes would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change is administrative in nature and has no impact on the probability or consequences of accidents previously evaluated. The physical structures, systems, and components of the Trojan Nuclear Plant and the operating procedures for their use are unaffected by this proposed change. The proposed action would eliminate the ISFSI area from the Part 50 license when the Part 72 license is issued. The 10 CFR 72 licensing controls for the area will assure an adequate level of safety for the area during normal operation of the ISFSI and during abnormal events or accidents. Therefore the proposed Part 50 amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed action would eliminate the ISFSI area from the Part 50 license when the Part 72 license is issued. The proposed change is administrative in [[Page 9197]] nature and has no impact on plant systems, structures, or components or on any procedures for operating the plant equipment. The ISFSI will be separately licensed under Part 72 and physically separated from the Part 50 licensed structures and equipment. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated. The proposed changes do not involve reduction in the margin of safety. The Trojan Permanently Defueled Technical Specifications (PDTS) contain four limiting conditions of operation that address: 1) Spent Fuel Water Level, 2) Spent Fuel Pool Boron Concentration, 3) Spent Fuel Pool Temperature, and 4) Spent Fuel Pool load restrictions. These PDTS will remain in effect as long as spent fuel is stored in the Spent Fuel Pool, which is in accordance with their applicability statements. The ISFSI area is physically separated from the Spent Fuel Pool area and the Fuel Building and will have no effect on spent fuel water level, spent fuel pool boron concentration, spent fuel pool temperature, or loads over the Spent Fuel Pool. The proposed change is administrative and does not affect plant equipment, operating parameters, or procedures. Based on the above, the proposed change will not reduce the margin of safety. Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library, Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, Portland, Oregon 97207. Attorney for licensees: Leonard A. Girard, Esq., Portland General Electric Company, 121 S. W. Salmon Street, Portland, Oregon 97204. NRR Project Director: Seymour H. Weiss. Portland General Electric Company, et al., Docket No. 50-344, Trojan Nuclear Plant, Columbia County, Oregon Date of application for amendment: January 7, 1999. Brief description of amendment: The proposed amendment would allow loading and handling of spent fuel transfer and storage casks in the Trojan Fuel Building. 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. The NRC staff has reviewed the licensees' analysis against the standards of 10 CFR 50.92(c). The licensee's analysis is summarized below: The proposed changes would not involve a significant increase in the probability or consequences of an accident previously evaluated. With the permanent cessation of operations, the number of potential accidents was reduced to those types of accidents associated with the storage of irradiated fuel and radioactive waste storage and handling. Additional events were postulated for decommissioning activities due to the difference in the types of activities that were to be performed. The postulated accidents in the Defueled Safety Analysis Report (DSAR) are generally classified as: (1) radioactive release from a subsystem or component, (2) fuel handling accident and, (3) loss of spent fuel decay heat removal capability. The postulated events described in the Decommissioning Plan are grouped as: (1) decontamination, dismantlement, and materials handling events, (2) loss of support systems (offsite power, cooling water, and compressed air), (3) fire and explosions, and (4) external events (earthquake, external flooding, tornadoes, extreme winds, volcanoes, lightning, toxic chemical release). These types of accidents are discussed below. Radioactive release from a subsystem or component involves failure of a radioactive waste gas decay tank (WGDT) or failure of a chemical and volume control system holdup tank (HUT). For a failure of a WGDT, the radioactive contents are assumed to be principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes and trace quantities of halogens. For the failure of a HUT, the assumptions were full power operations with 1-percent failed fuel, 40 weeks elapsed since power operation, and 60,000 gallons of 120 deg. F liquid released over a 2-hour period. However, the WGDT's and HUT's are no longer active and have been emptied. Therefore, cask loading and transfer activities cannot increase the probability of occurrence of a failure or the consequence of a failure of the WGDT's or HUT's. The fuel handling accident involves a stuck or dropped fuel assembly that results in damage of the cladding of the fuel rods in one assembly and the release of gaseous fission products. Spent fuel handling and loading will involve moving the spent fuel assemblies one by one, from the Spent Fuel Pool to the baskets which will be located in the Cask Loading Pit. The fuel handling equipment will be the same as had been previously analyzed with the exception of special tools which will be used to manipulate failed fuel. These special tools will be similar in size and weight to the existing tools used for underwater manipulation and therefore will not present a new hazard. In addition, the same administrative controls and physical limitations imposed [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%3A99-4391. Public record. Not legal advice.