# Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/fr%3AX95-10927 ## Record - **Collection:** Federal Register - **Document type:** Notice - **Published:** September 27, 1995 - **Citation:** 60 FR 49929 ## Text NUCLEAR REGULATORY COMMISSION Biweekly Notice Applications and Amendments to Facility Operating LicensesInvolving 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 August 30, 1995, through September 15, 1995. The last biweekly notice was published on Wednesday, September 13, 1995 (60 FR 47613). 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 [[Page 49930]] 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 Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 October 27, 1995, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC and at the local public document room for the particular facility involved. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) the nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above. Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross- examine witnesses. If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment. A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Services Branch, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, by the above date. Where petitions are filed during the last 10 days of the notice period, it is requested that the petitioner promptly so inform the Commission by a toll-free telephone call to Western Union at 1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator should be given Datagram Identification Number N1023 and the following message addressed to (Project Director): petitioner's name and telephone number, date petition was mailed, plant name, and publication date and page number of this Federal Register notice. A copy of the petition [[Page 49931]] should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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, et al., Docket No. 50-324, Brunswick Steam Electric Plant, Unit 2, Brunswick County, North Carolina Date of amendment request: August 4, 1995 Description of amendment request: The proposed amendment will allow the loading and use of GE13 fuel assemblies in the Brunswick Steam Electric Plant (BSEP), Unit 2, during Cycle 12 operation. The use of GE13 fuel assemblies requires that the safety limit value for minimum critical power ratio be revised. This safety limit is established to maintain fuel cladding integrity. Use of GE13 fuel also requires an increase in the concentration of sodium pentaborate solution required by the Technical Specifications (TS) for the standby liquid control system. This change provides the additional shutdown reactivity necessary to permit use of this fuel. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated. Proposed Change 1: The proposed amendment will allow the loading and use of GE13 fuel assemblies in the Brunswick Unit 2 reactor core. The use of GE13 fuel assemblies requires that the safety limit minimum critical power ratio value also be revised. The safety limit minimum critical power ratio is established to maintain fuel cladding integrity. The GE13 fuel assembly design has been analyzed using methods that have been previously approved by the Nuclear Regulatory Commission and documented in General Electric Nuclear Energy's reload licensing methodology Topical Report (NEDE-24011-P-A-10, ``General Electric Standard Application for Reactor Fuel (GESTAR II)'' dated February 1991). The proposed revision of the safety limit minimum critical power ratio does not alter any plant safety-related equipment, safety function, or plant operations that could change the probability of an accident. The change does not affect the design, materials, or construction standards applicable to the fuel bundles in a manner that could change the probability of an accident. A methodology that has been previously reviewed and accepted by the Nuclear Regulatory Commission was used to derive the both existing and updated safety limit minimum critical power ratio value. The same methodology criteria have been applied to derive the existing safety limit minimum critical power ratio of 1.07 as that used to derive the updated safety limit minimum critical power ratio value of 1.09. The updated safety limit minimum critical power ratio assures that fuel cladding protection equivalent to that provided with the existing safety limit minimum critical power ratio value is maintained. This ensures that the consequences of previously evaluated accidents are not significantly increased. Proposed Change 2: The standby liquid control system provides a means of reactivity control that is independent of the normal reactivity control system. The standby liquid control system must be capable of assuring that the reactor core can be placed in a subcritical condition at any time during reactor core life. Technical Specification Figure 3.1.5- 1 specifies the acceptable range of concentrations and volumes for sodium pentaborate solution used as a neutron absorber (i.e., for reactivity control). The portion of the sodium pentaborate concentration range shown in Technical Specification Figure 3.1.5-1 applicable to the lower range of tank volumes is being revised to increase the required concentration of sodium pentaborate solution. This change is needed to account for the additional shutdown reactivity needed based on the planned use of GE13 fuel assemblies as reload fuel for the Unit 2 reactor core. Since the standby liquid control system is independent from the normal means of controlling reactor core reactivity and not used to control core reactivity during normal plant operations, the proposed revision to the sodium pentaborate concentration curve for the standby liquid control system does not alter any plant safety-related equipment, safety function, or plant operations that could change the probability of an accident. The current volume-concentration range of sodium pentaborate used in the standby liquid control system will achieve a sufficient concentration of boron in the reactor vessel to ensure reactor shutdown. Based on the increased reactivity of the new GE13 reload fuel assemblies, the required sodium pentaborate volume- concentration range is being revised to ensure sufficient neutron absorbing solution is available to achieve reactor shutdown; therefore, the consequences of an accident previously evaluated are not significantly increased. 2. The proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. Proposed Change 1: The GE13 fuel assembly has been designed and complies with the acceptance criteria contained in General Electric Nuclear Energy's standard application for reactor fuel (GESTAR-II), which provides the latest acceptance criteria for new General Electric fuel designs. The GE13 fuel assembly complies with GESTAR-II acceptance criteria that have been previously reviewed and accepted by the Nuclear Regulatory Commission. The similarity of the GE13 fuel design to the previously accepted GE11 fuel design, in conjunction with the increased critical power capability of the GE13 fuel design, ensure that no new mode or condition of plant operation is being authorized by the loading and use of the GE13 fuel type. The proposed revision of the safety limit minimum critical power ratio from 1.07 to 1.09 does not modify any plant controls or equipment that will change the plant's responses to any accident or transient as given in any current analysis. Therefore, the proposed change to allow the loading and use of the GE13 fuel type and the revision of the safety limit minimum critical power ratio value from 1.07 to 1.09 will not create the possibility for a new or different kind of accident from any accident previously evaluated. Proposed Change 2: As discussed above, the standby liquid control system provides a means of reactivity control that is independent of the normal reactivity control system and is capable of assuring that the reactor core can be placed in a subcritical condition at any time during reactor core life. The proposed revision to the sodium pentaborate concentration range does not modify the standby liquid control system or its controls, does not modify other plant systems and equipment, and does not permit a new or different mode of plant operation. As such, the proposed revision to the minimum pentaborate concentration value does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed license amendment does not involve a significant reduction in a margin of safety. Proposed Change 1: As previously discussed, the GE13 fuel assembly design has been analyzed using methods that have been previously approved by the Nuclear Regulatory Commission and documented in General Electric Nuclear Energy's reload licensing methodology Topical Report (NEDE- 24011-P-A-10, ``General Electric Standard Application for Reactor Fuel (GESTAR II)'' dated February 1991). The safety limit minimum critical power ratio value is selected to maintain the fuel cladding integrity safety limit (i.e., that 99.9 percent of all fuel rods in the core be expected to avoid boiling transition). [[Page 49932]] Appropriate operating limit minimum critical power ratio values are established, based on the safety limit minimum critical power ratio value, to ensure that the fuel cladding fuel integrity safety limit is maintained. The operating limit minimum critical power ratio values are incorporated in the Core Operating limits Report as required by Technical Specification 6.9.3.1. The new GE13 safety limit minimum critical power ratio value of 1.09 is based on the same fuel cladding integrity safety limit criteria at that for the GE11 safety limit minimum critical power ratio value of 1.07 (i.e., that 99.9 percent of all fuel rods in the core be expected to avoid boiling transition); therefore, the proposed change does not result in a significant reduction in the margin of safety. Proposed Change 2: As previously stated, the purpose of the standby liquid control is to inject a neutron absorbing solution into the reactor in the event that a sufficient number of control rods cannot be manually inserted to maintain subcriticality. Sufficient solution is to be injected such that the reactor will be brought from maximum rated power conditions to subcritical over the entire reactor temperature range from maximum operating to cold shutdown conditions. General Electric reactor fuel methodology establishes a fuel type dependent standby liquid control system shutdown margin to account for calculational uncertainties. General Electric calculations show that an in-vessel concentration of 660 ppm will provide an estimated standby liquid control system minimum shutdown margin of 4.1% delta k. To achieve an in-vessel concentration of 660 ppm, the acceptable range of standby liquid control system tank concentrations is being revised for the lower range of tank volumes. Thus, proposed revision of the standby liquid control system sodium pentaborate volume- concentration range ensures that there will not be a significant reduction in the amount of available shutdown margin and, therefore, not a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: University of North Carolina at Wilmington, William Madison Randall Library, 601 S. College Road, Wilmington, North Carolina 28403-3297. Attorney for licensee: R. E. Jones, General Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh, North Carolina 27602 NRC Project Director: David B. Matthews Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment request: April 10, 1995 Description of amendment request: The requested amendment would modify Technical Specification 4.6.4.3 to allow a reduction in the number of hydrogen mitigation system igniters that must be maintained Operable. This would allow removal of the hydrogen igniters in the incore instrument tunnel. 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: Criterion 1 The requested amendments will not involve a significant increase in the probability or consequences of an accident previously evaluated. No impact upon accident probabilities will be created, since the EHM System is not an accident initiating system. In addition, it has been demonstrated that based on the results of computer analysis, and the review of results of an external study performed for a similar type containment, that hydrogen concentrations in the cavity during degraded core accidents will remain within acceptable limits. No impact on the plant response to any accident will be created (either design basis or beyond-design basis). Criterion 2 The requested amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated. As stated previously, the EHM System is not an accident initiating system. No new accident causal mechanisms will be created as a result of deleting the affected igniters. Plant operation will not be affected by the proposed amendments and no new failure modes will be created. Criterion 3 The requested amendments will not involve a significant reduction in a margin of safety. No adverse impact upon any plant safety margins will be created. As shown previously, applicable computer analysis has successfully demonstrated that the affected igniters could be removed with no adverse consequences. No fission product barriers are being degraded. No change to the manner in which the units are operated is being made. Based upon the preceding analyses, Duke Power Company concludes that the requested amendments do 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: York County Library, 138 East Black Street, Rock Hill, South Carolina 29730 Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 South Church Street, Charlotte, North Carolina 28242 NRC Project Director: Herbert N. Berkow Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment request: September 1, 1995 Description of amendment request: Generic Letter 88-16 provided guidance on removing cycle-specific parameters which are calculated using NRC approved methodologies from Technical Specifications (TS). The parameters are replaced in TS with a reference to a named report which contains the parameters, and a requirement that the parameters remain within the limits specified in the report. The proposed changes incorporate NRC approved methodologies, approved revisions to previously approved methodologies, or republished versions of previously approved methodologies into Section 6.9 of the Catawba TS. For Catawba, the limits to which these methodologies are applied are explicitly listed in the TS. Since the proposed changes only incorporate NRC approved methodologies into the TS the licensee proposed that the changes are administrative in nature and can be assumed to have no impact, or potential impact, on the health and safety of the public. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: The proposed changes will not create a significant hazards consideration, as defined by 10 CRF 50.92, because: 1) The proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative in nature, and do not affect any system, procedure, or manipulation of any equipment which could affect the probability or consequences of any accident. 2) The proposed changes will not create the possibility of any new or different kind of accident from any accident previously evaluated. The proposed changes are administrative in nature, and cannot introduce any new failure mode or transient which could create any accident. [[Page 49933]] 3) The proposed changes will not involve a significant reduction in a margin of safety. The proposed changes are administrative in nature, and will not affect any operating parameters or limits which could result in a 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: York County Library, 138 East Black Street, Rock Hill, South Carolina 29730 Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 South Church Street, Charlotte, North Carolina 28242 NRC Project Director: Herbert N. Berkow Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina Date of amendment request: September 13, 1995 Description of amendment request: The proposed amendments modify the notation for the overpower delta-temperature (OPDT) reactor trip heatup setpoint penalty coefficient to be consistent with NUREG-0452, Revision 4, ``Standard Technical Specifications for Westinghouse Pressurized Water Reactors'' (STS). This change is necessary in order to allow implementation of the modification to reduce the reactor coolant system hot leg temperature as planned during the Unit 2 end-of- cycle 7 refueling outage. 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 required by 10CFR50.91, this analysis is provided concerning whether the requested amendments involve significant hazards considerations, as defined by 10CFR50.92. Standards for determination that an amendment request involves no significant hazards considerations are if operation of the facility in accordance with the requested 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. Criterion 1 The proposed amendments will not involve a significant increase in the probability or consequences of an accident previously evaluated. The amendments will have no impact whatsoever upon the probability of any accident being initiated, since the reactor trip system is an accident mitigating system. The amendments will have no adverse impact upon any accident consequences or upon the function of the OPDT setpoint. The reactor trip heatup setpoint penalty will continue to be applied anytime T-avg is greater than T [double prime] and will not be applied when T-avg is less than or equal to T [double prime]. This is consistent with the intent of this function. Criterion 2 The proposed amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated. The function of the OPDT setpoint will not be altered by the proposed changes. As stated previously, the reactor trip system is an accident mitigating system, so no new failure modes can be created. No change to any aspect of plant operation will result from NRC approval of the proposed amendments. Criterion 3 The proposed amendments will not involve a significant reduction in a margin of safety. The changes are necessary to allow full implementation of the T-hot reduction modification on Catawba Unit 2. The proposed changes are consistent with the terminology of both NUREG-0452, Revision 4 and NUREG-1431, Revision 1. OPDT setpoint behavior will not be adversely impacted by the proposed changes; therefore, no impact upon any plant safety margins will result. Based upon the preceding analyses, Duke Power Company concludes that the requested amendments do 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: York County Library, 138 East Black Street, Rock Hill, South Carolina 29730 Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 South Church Street, Charlotte, North Carolina 28242 NRC Project Director: Herbert N. Berkow Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Date of amendment request: March 29, 1995 Description of amendment request: The amendments would revise the Technical Specification 3.4.9.3 requirements for the Low Temperature Overpressure Protection (LTOP) system and update the heatup and cooldown curves. The intent of the proposed amendments is to enhance overpressure protection during low temperature operations. These enhancements can be fully implemented, improving startup and shutdown operation of McGuire Units 1 and 2. Specifically, these changes are categorized into five groups identified as follows: 1) Revisions to the LCO requirements, the Action Statements and the SR for the Reactor Coolant System Overpressure Protection System during low temperature conditions, 2) A reduction in the Reactor Coolant System (RCS) vent requirement from 4.5 square inches to 2.75 square inches, 3) The use of the Residual Heat Removal suction relief valve (1ND3 and 2ND3) for overpressure protection under restricted conditions. (RCS greater than 107 deg.F and cooldown rate less than 20 deg.F/hr; or RCS greater than 167 deg.F), 4) Revisions of the Pressure/Temperature curves to 16 EFPY, including the incorporation of the latest radiation surveillance capsule results and removal of instrumentation margins from the Technical Specification figures, and 5) Changes to format and consistency. 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 for each of the five groups listed above. FIRST STANDARD (Amendment would not) involve a significant increase in the probablility or consequences of an accident previously evaluated. 1) Revised LCO [limiting conditions for operation] and SR [surveillance requirements] for LTOP: The reduced maximum setpoint will prevent the violation of the 10 CFR 50 Appendix G pressure/temperature curves (as modified by the provisions of ASME Code Case N-514) during overpressure transients at low temperatures. Since the maximum setpoint is reduced, the peak pressure for LTOP [low-temperature overpressure protection] events will be reduced as well. Accordingly, the consequences of an LTOP event would not change as result of the proposed changes. The analysis performed to determine the setpoint is, in accordance with the methods used in previous evaluations, found acceptable by the NRC. The three possible transients evaluated are; 1) a mass input from an operable safety injection pump; 2) a mass input from an operable centrifugal charging pump; and 3) a heat input from a 50 deg.F temperature difference between the steam generators and the NC system. The LTOP setpoint of the PORV [power- operated relief valve] proposed by this technical specification change is not considered to be an initiator of any of these three transients. As such, the probability of an accident [[Page 49934]] previously evaluated would not be increased as a result of the proposed changes. Two additional conditions for operability of the LTOP system are defined (accumulator isolation and only one NV or NI pump operable) and new surveillance requirements are specified as well. They provide additional limitations, requirements and restrictions that currently do not exist within the technical specifications for McGuire. The incorporation of these proposed changes are consistent with what is specified within NUREG-1341. Therefore, these changes do not increase the probability of consequences of an accident previously evaluated. 2) Reduction in NC vent opening: The bases for the size of the vent to be established per the technical specifications is to ensure that the 10 CFR 50, Appendix G pressure/temperature limits are not exceeded during an LTOP event. The determination of the size of the opening continues to preserve the above design basis. The evaluation performed demonstrated that a 2.75 square inch opening would provide adequate overpressure protection for the combined capacity of a centrifugal charging pump and a safety injection pump. The only time that the vent path is to be established is when the PORVs may not be available. Defining the size of the vent is not considered to be an initiator of any LTOP events that have been previously evaluated. As such, this change in the size of the vent opening does not increase the probability of an overpressure event during low temperature conditions. The analysis performed verifies that the size opening specified is sufficient to mitigate the consequences of an LTOP event. Accordingly, the change in the size of the opening for the vent will not impact the consequences of LTOP events. 3) Use of RHR [residual heat removal] suction relief valves: By letter dated September 11, 1990, the NRC authorized the deletion of the RHR autoclosure interlock circuitry. A modification which removed the RHR system suction isolation valve autoclosure interlocks has been completed. As such, the RHR suction relief valve can be exposed to NC system pressure and would be available to mitigate LTOP events. The proposed amendments specify the necessary requirements and controls to ensure proper ND system alignments and conditions will exist to protect the pressure/temperature limits. This added relieving capacity will enhance the current LTOP system at McGuire in mitigating overpressure events at low temperatures. As such, the mitigation of previously evaluated LTOP events would be improved by the proposed technical specification changes. Further, the proposed changes would not esult in the initiation of an LTOP event or cause an overpressure transient. Accordingly, the proposed amendment would not involve an increase in the consequences or the probability of an accident previously evaluated. 4) Revised pressure/temperature curves to 16 EFPY [effective full-power year]: The proposed pressure/temperature curves, provided by this amendment request, satisfy all regulatory required material embrittlement considerations including: ASME Section XI Appendix G, 10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. In addition, the margins for instrument error have been removed from the curves. Instrument error will be administratively handled by incorporating them into the LTOP system setpoint selection calculations and into appropriate controlling procedures for unit operations. The proposed changes to the pressure/temperature curves are not considered to be an initiator of LTOP events. The changes to the curves proposed by this amendment request will not cause an LTOP event. The curves define the new limits that have been defined in accordance with regulatory requirements by which both units are to be operated within. Accordingly, the proposed amendment will not increase the probability or the consequences of previously evaluated accidents. 5) Format and consistency: The changes associated within this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, the proposed amendment request does not increase the probability or consequences of any previously evaluated accident. SECOND STANDARD (Amendment would not) create the possibility of a new or different kind of accident from any kind of accident previously evaluated. 1) Revised LCO and SR for LTOP: The only potential impact to plant systems, structures and components, as a result of the proposed changes associated with this group, would be the setting of the PORV low pressure setpoint. No other changes to plant systems, structures or components would occur. The proposed amendments, also, would not impact the plant operation. Although the value for the PORV pressure setting specified within the technical specification would be reduced per the proposed amendment, the actual settings of the PORV are now currently set low enough to comply with the proposed lower setpoint value. As such, the proposed lower setpoint would not require any changes to the plant nor how the plant is operated. The additional requirements for LTOP operability will not require any modifications to the plant nor how the plant is operated. Currently, when entering LTOP conditions, the accumulators are isolated and only one NV or NI pump is capable of injecting into the reactor vessel. these actions are currently controlled and are specified within the operating procedures for heatup and cooldown of the respective units. The proposed changes will now specify these current operating requirements within the technical specifications as well. Accordingly, the proposed revisions will not create a new or different kind of accident than what has already been previously evaluated. 2) Reduction in NC vent opening: The proposed changes to the technical specifications associated with this group involves the size of the vent opening. The proposed amendment reduces the size of the vent opening from 4.5 square inches to 2.75 square inches. The analysis that was performed has determined that the proposed size for the vent opening is adequate for overpressure events. Therefore, this proposed revision to the technical specifications will not result in a new or different kind of accident from any kind of accident previously evaluated. 3) Use of RHR suction relief valves; The proposed amendment associated with this group will specify the necessary requirements and controls to ensure the appropriate use of the RHR suction relief valve for overpressure protection. This added relieving capacity will enhance the current LTOP system in mitigating overpressure events during low temperature conditions. The analysis that has been performed demonstrates the adequacy of the RHR suction relief valve, in conjunction with a PORV, in mitigating overpressure events at low temperatures, assuming a worst case single failure as well. As such, the use of the RHR suction relief valve in the manner prescribed by the proposed technical specification amendment will not create a new or different kind of accident from those accidents that have been previously evaluated. 4) Revised pressure/temperature curves to 16 EFPY: The changes associated with this group, provide new heatup and cooldown curves for both Units 1 and 2, which will extend the service period from 10 EFPY to 16 EFPY and will remove the instrument error as well. The proposed [heatup] and cooldown curves were developed in accordance with all regulatory required material embrittlement criteria. Thus, operation of the units in accordance with the proposed new pressure/temperature curves will not create the possibility of a new or different kind of accident from those accident[s] that have been previously evaluated. 5) Format and consistency: The changes associated within this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, the proposed amendment will create the possibility of a new or different kind of accident from that previously evaluated. THIRD STANDARD (Amendment would not) involve a significant reduction in a margin of safety. 1) Revised LCO and SR for LTOP: This proposed change will reduce the maximum PORV setpoint such that, for LTOP events, the maximum pressure in the vessel would not exceed 110% of the pressure/temperature limits that have been established in accordance with ASME Appendix G. This is congruous with the provisions of ASME Code Case N-514. Currently, the maximum PORV setpoint for LTOP events ensure that the maximum pressure would not exceed 100% of the pressure/temperature curves. As such, the proposed change appears to involve a slight reduction in a margin of safety. Although the proposed change may involve a slight reduction in a margin of safety, the proposed change will provide an [[Page 49935]] equivalent margins of safety to the reactor vessel during LTOP transients and will satisfy the underlying purpose of 10 CFR 50.60 for fracture toughness requirements. By letter dated June 28, 1994, an exemption request and authorization to use ASME Code Case N-514 at McGuire was submitted to the NRC for review and approval. Approval for the use of the code case was granted on September 30, 1994. The proposed change to reduce the maximum PORV setpoint, coupled with the September 30, 1994 NRC approval for the use of Code Case N-514 satisfies current regulatory acceptance criteria. Therefore, the proposed change would not involve a significant reduction in a margin of safety. This change group, also, defines two additional conditions for the operability of the LTOP system (accumulator isolation and only one NV or NI pump operable) and proposes new surveillance requirements and restrictions that currently do not exist within the technical specifications for McGuire. The incorporation of these proposed changes are consistent with what is specified within NUREG- 1341. Therefore, these changes do not involve a significant reduction in a margin of safety. 2) Reduction in NC vent opening: The proposed changes to the technical specifications associated with this group involves the size of the vent opening. The proposed amendment reduces the size of the vent opening from 4.5 square inches to 2.75 square inches. The basis for the size of the vent to be established per the technical specifications is to ensure that the 10 CFR 50, Appendix G pressure/temperature limits are not exceeded during an LTOP event. The determination of the size of the opening continues to preserve the above design basis. The evaluation performed demonstrated that a 2.75 square inch opening would provide adequate overpressure protection for the combined capacity of a centrifugal charging pump and a safety injection pump. Accordingly, the proposed changes would not involve a significant reduction in a margin of safety. 3) Use of RHR suction relief valves: The proposed amendment associated with this group will specify the necessary requirements and controls to ensure the appropriate use of the RHR suction relief valves for overpressure protection. This added relieving capacity will enhance the current LTOP system in mitigating overpressure events during low temperature conditions. The analysis that has been performed demonstrates the adequacy of the RHR suction relief valve, in conjunction with a PORV, in mitigating overpressure events at low temperatures. Further, by letter dated September 11, 1990, the NRC approved amendments to delete a portion of the surveillance requirements regarding periodic verification that the RHR suction isolation valves automatically close on a RCS [reactor coolant system] signal less than or equal to 560 psig. This action, in effect, authorizes the removal of the RHR autoclosure interlock circuitry. As discussed within the NRC Safety evaluation for the amendment, the Commission and industry have recognized the safety benefits of removing the ACI [automatic closure and interlock] circuitry from the RHR system to minimize, and thus reduce the risk associated with loss of decay heat removal events. Therefore, the proposed amendments associated with this change group will not involve a significant reduction in a margin of safety. 4) Revised pressure/temperature curves to 16 EFPY: The changes associated with this group provide new heatup and cooldown curves for both Units 1 and 2, which will extend the service period from 10 EFPY to 16 EFPY and will relocate the instrument error as well. The proposed pressure/temperature curves provided by this amendment request satisfy all regulatory required material embrittlement considerations including; ASME Section XI Appendix G, 10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. The instrument error will be administratively handled by incorporating them into the LTOP system setpoint selection calculations and into the controlling procedures for unit operations. The relocation of the instrument error to licensee controlled documents is consistent with the NRC actions proposed within NUREG- 1431, new standard technical specifications for Westinghouse plants. As prescribed within NUREG-1431, the pressure/temperature limit curves are to be relocated to a licensee controlled document entitled ``Pressure Temperature Limit Report (PTLR)''. Changes to the heatup and cooldown curves would then be performed in accordance with 10 CFR 50.59 criteria. For the situation proposed by this amendment, updates and revisions of the instrument error associated with the pressure/temperature limit curves will be processed in a similar fashion. Thus, the proposed change to relocate the instrument error to licensee controlled documents is analogous with NRC acceptable practices. Accordingly, the proposed changes will not reduce a margin of safety. 5) Format and consistency: The changes associated within this group are considered to be administrative in nature. They do not affect station operability or require any modifications to the facility. Accordingly, there is no reduction in the margin of safety of the LTOP system due to the incorporation of these editorial/administrative changes. 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: Atkins Library, University of North Carolina, Charlotte (UNCC Station), North Carolina 28223 Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 South Church Street, Charlotte, North Carolina 28242 NRC Project Director: Herbert N. Berkow Florida Power and Light Company, et al., Docket Nos. 50-335 and 50- 389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: June 21, 1995 Description of amendment request: The proposed amendments will revise the action statements for a single inoperable Emergency Diesel Generator (EDG), TS 3.8.1.1.b, to extend the allowed outage time (AOT) from 72 hours to 7 days, and permit a 10 day AOT to be used once per refueling cycle. This proposal is a result of a cooperative study by participating Combustion Engineering Owners Group members which concluded that the proposed AOT extension improves plant operational flexibility while adequately controlling overall plant risk. 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) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendments for St. Lucie Unit 1 and Unit 2 will extend the action completion/allowed outage time (AOT) for a single inoperable Emergency Diesel Generator (EDG) from 72 hours to 7 days, with provisions for a 10 day AOT once per refueling cycle. The EDGs are designed as backup AC power sources for essential safety systems in the event of a loss of offsite power. As such, the EDGs are not accident initiators, and an extended AOT to restore operability of an inoperable diesel generator would not increase the probability of occurrence of accidents previously analyzed. The proposed technical specification revisions involve the AOT for a single inoperable EDG, and do not change the conditions, operating configuration, or minimum amount of operating equipment assumed in the plant safety analyses for accident mitigation. In addition, a Probability Safety Assessment (PSA) was performed to quantitatively assess the risk impact of the proposed amendment. The impact on the early radiological release probability for design basis events was also evaluated. It was concluded that the risk contribution from this proposed AOT is very small, and that the impact will be negligible. Therefore, operation of either facility in accordance with its proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. (2) Operation of the facility in accordance with the proposed amendment would not [[Page 49936]] create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendments will not change the physical plant or the modes of plant operation defined in either Facility License. The changes do not involve the addition or modification of equipment, nor do they alter the design of plant systems. Therefore, operation of either facility in accordance with its proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. (3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. The proposed amendments are designed to improve EDG reliability by providing flexibility in the scheduling and performance of preventive and corrective maintenance activities. The surveillance intervals or the operability requirements are not changed by the proposal; only the AOT for a single inoperable EDG will be extended. The proposed changes do not alter the basis for any technical specification that is related to the establishment of, or the maintenance of, a nuclear safety margin. Moreover, an integrated assessment of the risk impact of extending the AOT for a single inoperable EDG has determined that the risk contribution is very small and can be offset by improvements in EDG reliability. Therefore, operation of either facility in accordance with its proposed amendment would 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 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: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003 Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036 NRC Project Director: David B. Matthews Florida Power and Light Company, et al., Docket Nos. 50-335 and 50- 389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: June 21, 1995 Description of amendment request: The proposed amendments will revise TS 3.5.2 to allow up to 7 days to restore an inoperable Low Pressure Safety Injection train to operable status. This proposal is a result of a cooperative study by participating Combustion Engineering Owners Group members which concluded that an extension of the allowed outage time (AOT) from 72 hours to 7 days can improve plant operational flexibility and is risk beneficial. 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) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed amendments for St. Lucie Unit 1 and Unit 2 will extend the action completion/allowed outage time (AOT) for a single inoperable Low Pressure Safety Injection (LPSI) train from 72 hours to 7 days. A LPSI train is designed as a part of each Emergency Core Cooling System (ECCS) subsystem to supplement Safety Injection Tank (SIT) inventory during the early stages of mitigating a Design Basis Accident. As such, components of the LPSI system are not accident initiators, and an extended AOT to restore operability of an inoperable LPSI train would not increase the probability of occurrence of accidents previously analyzed. The safety analyses for both St. Lucie Units demonstrate that ECCS performance acceptance criteria are satisfied with only one of the two redundant ECCS subsystems operating during the postulated Design Basis Accident. The proposed technical specification revisions involve the AOT for a single inoperable LPSI train, and do not change the conditions assumed for the minimum amount of operating equipment needed for accident mitigation. Therefore, the consequences of an accident previously evaluated will not be significantly increased. In addition to the preceding evaluation, a Probabilistic Safety Analysis (PSA) was performed to quantitatively assess the risk impact of the proposed amendments. It was concluded from the results of that assessment that the risk contribution of the AOT extension is very small, and that the net impact of the proposed amendment can be risk beneficial. Therefore, operation of either facility in accordance with its proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. (2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendments will not change the physical plant or the modes of plant operation defined in either Facility License. The changes do not involve the addition or modification of equipment nor do they alter the design of plant systems. Therefore, operation of either facility in accordance with its proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. (3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. The margin of safety associated with the ECCS system is established by acceptance criteria for system performance defined in 10 CFR 50.46. The proposed amendments will not change this acceptance criteria nor the operability requirements for equipment that is used to achieve such performance as demonstrated in the plant safety analyses. Moreover, an integrated assessment of the risk impact of extending the AOT for a single inoperable LPSI train has concluded that the risk contribution is very small, LPSI system reliability can potentially be improved, and the net impact of the proposed change can be risk beneficial. Therefore, operation of either facility in accordance with its proposed amendment would 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 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: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003 Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036 NRC Project Director: David B. Matthews Florida Power and Light Company, et al., Docket Nos. 50-335 and 50- 389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: June 21, 1995 Description of amendment request: The proposed amendments will revise the action statements and certain surveillances of TS 3/4.5.1, Safety Injection Tanks (SIT). This proposal is based on the results of a cooperative study performed by participating Combustion Engineering Owners Group members which investigated the impact of a risk-based allowed outage time (AOT) extension, and also included recommendations for line-item TS improvements from NUREG-1366 and Generic Letter 93-05. 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) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The license amendments proposed for St. Lucie Units 1 and 2 incorporate certain line- [[Page 49937]] item Technical Specifications (TS) improvements for the Safety Injection Tanks (SIT), and include an extension of the required action completion/allowed outage time (AOT) from one hour to 72 hours to restore an inoperable SIT (that is still able to perform its safety function) to operable status. In addition, an AOT of 24 hours, based on risk assessment techniques, is proposed for an SIT that may be unable to perform its design function. The SITs are passive components of the Emergency Core Cooling System (ECCS). As such, they are not accident initiators for any transient evaluated in the plant safety analyses, and an extension of the AOTs for restoring an inoperable SIT to operable status would not increase the probability of occurrence of accidents previously analyzed. The SITs, in combination with other ECCS components, are used to mitigate the consequences of a loss of coolant accident. The TS revisions will provide a longer AOT for a single inoperable SIT, but do not involve a change to the ECCS configuration or method of operation. The proposed amendments will not change the conditions assumed for the minimum amount of operating equipment needed for accident mitigation. Therefore, the consequences of an accident previously evaluated will not be significantly increased. In addition to the preceding evaluation, a Probability Safety Assessment (PSA) was performed to quantitatively assess the risk impact of the 24 hour AOT proposal. The impact on the early radiological release probability for design basis events was also evaluated. It was concluded that the risk contribution from this AOT is very small, and that the impact is negligible. Therefore, operation of either facility in accordance with its proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. (2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendments will not change the physical plant or the modes of plant operation defined in either Facility License. The changes do not involve the addition or modification of equipment, nor do they alter the design of plant systems. Therefore, operation of either facility in accordance with its proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. (3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. The margin of safety associated with the ECCS system is established by acceptance criteria for system performance defined in 10 CFR 50.46. The proposed amendments will not change this criteria nor the operability requirements for equipment that is used to achieve such performance as demonstrated by the plant safety analyses. Moreover, an integrated assessment of the risk impact of allowing 24 hours to restore an inoperable SIT to operable status has concluded that this impact is very small, and can be offset by averting an unnecessary transition to the shutdown modes. Therefore, operation of either facility in accordance with its proposed amendment would 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 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: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003 Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036 NRC Project Director: David B. Matthews Florida Power and Light Company, et al., Docket Nos. 50-335 and 50- 389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: August 16, 1995 Description of amendment request: The revisions will modify Technical Specification 3.6.6.1, Shield Building Ventilation System (SBVS), to more effectively address the design functions performed by the SBVS for both the Shield Building (secondary containment) and the Fuel Handling Building. 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) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed license amendment for St. Lucie Unit 2 will clarify the Applicability and the Actions required by Technical Specification (TS) 3.6.6.1, and explicitly account for the dual purpose of the Shield Building Ventilation System (SBVS) to perform design functions for both the Shield Building (secondary containment) and the Fuel Handling Building. The proposed amendment is administrative in nature. The SBVS only operates when actuated by automatic control signals generated by systems detecting postulated accident conditions. The SBVS is not an accident initiator, the proposed TS changes do not involve any assumptions relative to accident initiators used in the plant safety analyses, and the amendment, therefore, will not impact the probability of occurrence for accidents previously analyzed. Relative to accident consequences, at least one train of the SBVS must operate to fulfill the design function of evacuating filtered air from the Shield Building during the postulated Loss of Coolant Accident; and likewise assumed in the analysis for the Fuel Handling Building during a fuel handling accident. The proposed changes simply remove elements of ambiguity from TS 3.6.6.1; do not reduce the existing operability requirements for the system; and provide further assurance that proper compensatory measures will be taken in the event one or both SBVS trains become inoperable. Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. (2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendment is administrative in nature and will not change the physical plant or the modes of plant operation defined in the facility license. The changes do not involve the addition or modification of equipment, nor do they alter the design or methods of operation of plant systems. Plant configurations that are prohibited by TS will not be created by this amendment. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. (3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. The proposed amendment will not change the SBVS operability requirements nor otherwise alter the basis for any technical specification that is related to the establishment of, or the maintenance of, a nuclear safety margin. The proposed changes are administrative in nature, and are designed to provide assurance that the SBVS capability to perform design functions assumed available in the safety analyses will remain available during the various plant operating modes. Therefore, operation of the facility in accordance with the proposed amendment would 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 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: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003 Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036 NRC Project Director: David B. Matthews [[Page 49938]] Florida Power and Light Company, et al., Docket Nos. 50-335 and 50- 389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida Date of amendment request: August 16, 1995 Description of amendment request: The proposed amendments revise St. Lucie Units 1 and 2 Technical Specifications to relocate selected Technical Specification Monitoring Instrumentation utilizing the Final Policy Statement on Technical Specification Improvement for Nuclear Power Reactors, 58 FR 39132, July 22, 1993. The proposed amendments also include relocation of Technical Specifications related to the Emergency and Security Plan review process utilizing the guidance contained in NRC Generic Letter 93-07, ``Modification of the Technical Specification Administrative Requirements for Emergency and Security Plans.'' 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) Operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes to the Selected Technical Specification Requirements Related to Instrumentation are administrative in nature in that the specifications for operation and surveillance of the selected Technical Specification instrumentation will be relocated from Appendix A of the facility operating license to the Updated Final Safety Analysis Report (UFSAR) for each unit. Once relocated, future changes will be controlled by 10 CFR 50.59 and the UFSARs updated pursuant to 10 CFR 50.71(e). Relocation of these requirements to the UFSAR is consistent with the NRC ``Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors'' published in the Federal Register (58 FR 39132) dated July 22, 1993. The selected Technical Specification instruments are not accident initiators nor a part of the success path(s) which function to mitigate accidents evaluated in the plant safety analyses. The proposed Technical Specification change does not involve any change to the configuration or method of operation of any plant equipment that is used to mitigate the consequences of an accident, nor do the changes alter any assumptions or conditions in any of the plant accident analyses. Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated. The Technical Specifications changes associated with Emergency Plan and Security Plan requirements are proposed in accordance with Generic Letter 93 07. The changes being proposed are administrative in nature and do not affect assumptions contained in plant safety analyses, the physical design and/or operation of the plant, nor do they affect Technical Specifications that preserve safety analysis assumptions. Therefore, operation of the facility in accordance with the proposed amendments would not affect the probability or consequences of an accident previously analyzed. (2) Use of the modified specification would not create the possibility of a new or different kind of accident from any previously evaluated. The proposed amendment to relocate the existing Technical Specification requirements for selected Technical Specification instrumentation to the UFSAR will not change the physical plant or the modes of plant operation defined in the Facility License. The change does not involve the addition or modification of equipment nor does it alter the design or operation of plant systems. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed amendments, in accordance with Generic Letter 93- 07, change the Technical Specifications to remove the audit of the emergency and security plans and implementing procedures from the list of responsibilities of the Facility Review Group. The changes being proposed are administrative in nature and will not change the physical plant or the modes of operation defined in the Facility License. The change does not involve the addition or modification of equipment nor does it alter the design or operation of plant systems. Therefore, operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated. (3) Use of the modified specification would not involve a significant reduction in a margin of safety. The proposed changes are administrative in nature in that operating and surveillance requirements for the selected Technical Specification instrumentation will be relocated from Appendix A of the facility license to the appropriate Updated Final Safety Analysis Report for each unit. These selected instruments are not used to actuate safety-related equipment, provide interlocks, or otherwise perform plant control functions. Conditions evaluated in plant accident and transient analyses do not involve these selected instruments. The proposed changes do not alter the basis for any technical specification that is related to the establishment of, or the maintenance of, a nuclear safety margin. Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. The proposed amendments, in accordance with Generic Letter 93- 07, change the Technical Specifications to remove the audit of the emergency and security plans and implementing procedures from the list of responsibilities of the Facility Review Group. The changes being proposed are administrative in nature and do not alter the bases for assurance that safety-related activities are performed correctly or the basis for any Technical Specification that is related to the establishment of or maintenance of a safety margin. Therefore, operation of the facility in accordance with the proposed amendments would 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 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: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003 Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800 M Street, N.W., Washington, DC 20036 NRC Project Director: David B. Matthews IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn County, Iowa Date of amendment request: July 21, 1995 Description of amendment request: The proposed amendment would make administrative changes to various sections of the Duane Arnold Energy Center (DAEC) Technical Specifications (TS). These changes replace a conditional surveillance if one emergency service water (ESW) pump or loop is determined to be inoperable (TS 4.8.E.2); credit successful emergency diesel generator (EDG) tests performed in the previous 24 hours (TS 4.8.E.2); clarify the requirements governing spent and new fuel storage in Section 5.5 of the DAEC TS; and eliminate the Operations Committee reviews of procedures in support of the DAEC Emergency Plan and Security Plan, as specified in Sections 6.5 and 6.8 of the TS. DAEC TS Section 4.8.E.2 states the surveillance requirement applicable when one ESW pump or loop is determined to be inoperable. This amendment request deletes the surveillance requirement to physically test the opposite train's EDG and replaces it with a requirement to verify OPERABILITY of the opposite train low pressure core and containment cooling systems and EDG. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the [[Page 49939]] licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. The proposed revision does not involve a significant increase in the probability or consequences of an accident previously evaluated. The changes are administrative in nature and are consistent with previously-published NRC guidance. The proposed revision does not change any accident analysis, plant safety analysis or calculations; degrade existing plant programs; or modify any functions of safety related systems or accident mitigation functions for which the DAEC has previously been credited. The proposed revision to the Surveillance Requirements will continue to assure OPERABILITY as required, but eliminate unnecessary operation of an EDG. 2. The proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed revision does not alter any plant parameters, revise any safety limit setpoint, or provide any new release pathways. In addition, the proposed revision does not modify the operation or function of any safety-related equipment, nor introduce any new modes of operation, failure modes, or physical changes to the plant. 3. The proposed revision does not involve a significant reduction in a margin of safety. The proposed revision does not alter any plant parameters, revise any safety limit setpoint, or provide any new release pathways. In addition, the proposed revision does not modify the operation or function of any safety-related equipment, nor introduce any new modes of operation, failure modes, or physical changes to the plant. The NRC staff has reviewed the licensee's analysis and, based on thisreview, 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, S.E., Cedar Rapids, Iowa 52401 Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis & Bockius, 1800 M Street, NW., Washington, DC 20036-5869 NRC Project Director: Gail H. Marcus Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska Date of amendment request: May 5, 1995, as revised by letter dated July 14, 1995 Description of amendment request: The proposed changes would amend the Cooper Nuclear Station (CNS) Technical Specifications (TS) sections 3/4.5.F.1, 3.5.F.2, 3.9.B.1, 3.9.B.2, 4.9.A.2, and the associated bases. These changes would revise the TS to: 1) verify that the redundant diesel generator is operable upon the loss of one diesel generator, and implement provisions to verify that the operable diesel generator does not have a common cause failure; 2) incorporate provisions to allow a modified start for the diesel generators; and 3) remove the requirement that the reactor power level be reduced to 25% of rated power upon loss of both diesel generator units or both incoming power sources (start-up and emergency transformers). In addition, the period of time allowed for continued reactor operation with both diesels inoperable would be reduced from 24 to two hours. 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: 10 CFR 50.91(a)(1), requires that licensee requests for operating license amendments be accompanied by an evaluation of significant hazards posed by the issuance of the amendment. NPPD has reviewed the proposed changes in accordance with 10CFR50.92 and concludes that the changes do not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a SHC because the changes would not: 1. Involve a significant increase in the probability or consequences of an accident previously analyzed. Proposed Revision 1: This proposed revision serves to ensure that an emergency diesel generator is always available to perform on demand and that lowering the number of demands to demonstrate operability reduces the probability of equipment failure. The required action no longer requires the redundant emergency diesel generator to be demonstrated operable immediately. Therefore, this requirement has been deleted from TS 4.5.F.1. The proposed change includes provisions to determine if the redundant diesel generator has been made inoperable by a common cause failure or perform a demonstration test. The redundant emergency diesel generator will remain in service during the entire period of inoperability of the out of service emergency diesel generator. If a common cause failure cannot be ruled out, the redundant diesel generator will be tested in accordance with the surveillance requirements of TS section 4.9.A.2.a.1 to assure operability. Since this proposed revision does not affect the design or negatively affect the performance of the diesel generators, the change will not result in an increase in the consequences or probability of an accident previously analyzed. This proposed revision will increase diesel generator reliability and availability, thereby increasing overall plant safety. Proposed Revision 2: This proposed revision only affects emergency diesel generator periodic testing. The diesel generators are not accident initiators and the method of testing the diesel generators cannot initiate an accident and therefore will not increase the probability of an accident. This change to the diesel generator testing method does not impact any Updated Safety Analysis Report (USAR) safety analysis. The proposed surveillances will still provide assurance that the diesel generators are available to mitigate the consequences of accidents previously evaluated. Thus the consequences of an accident previously evaluated are not increased. The revised periodic testing will still demonstrate that the emergency diesel generators are ready to perform their safety function. An overall improvement in diesel engine reliability and availability can be gained by performing diesel generator starts for surveillance testing using engine prelubes, warmups and other manufacturer recommended practices to reduce engine stress and wear. Since this proposed revision does not affect the design or negatively affect the performance of the diesel generators, the change will not result in an increase in the consequences or probability of an accident previously analyzed. This proposed revision will increase diesel generator reliability, thereby increasing overall plant safety. Proposed Revision 3: This proposed revision does not affect the operation of the emergency diesel generators or the incoming power sources (start-up and emergency transformers). Both the diesel generators and the incoming power sources function to mitigate the consequences of postulated accidents. As such, removing the requirement to reduce power level upon the loss of both redundant components in either of these systems does not create an increase in the probability of an accident. By eliminating this requirement, the potential for plant transients during power reduction to 25% are also eliminated. Eliminating this requirement will not increase the consequences of a postulated accident because the redundant components will remain available. Additionally, the loss of both offsite power sources condition becomes more restrictive by requiring a plant shutdown instead of notification within 24 hours. The proposed changes do not alter the conditions or assumptions in any of the Updated Safety Analysis Report (USAR) accident analyses. Since the USAR accident analyses remains bounding, the radiological consequences previously evaluated are not adversely affected by the proposed changes. Therefore, no significant increase in the probability or consequences of an accident previously analyzed would occur. The proposed rearrangement of information, and rewording of some the TS requirements are included to enhance usability and alleviate any possible confusion. These changes are strictly editorial have no impact, and do not alter technical content or meaning of the specifications. These editorial changes do not involve a significant increase in the probability or consequences of an accident previously analyzed. [[Page 49940]] 2. Create the possibility of a new or different kind of accident from any previously analyzed. Proposed Revision 1: Accidents involving loss of off-site power and single failure have been previously evaluated, and this proposed change does not impact any of those assumptions. This proposed revision does not introduce any new mode of plant operation or new accident precursors, involve any physical alterations to plant configurations, or make changes to system setpoints which could initiate a new or different kind of accident. Operation of the facility in accordance with the proposed revised changes does not create the possibility of a new or different kind of accident from any previously evaluated. Proposed Revision 2: This proposed revision only affects emergency diesel generator periodic testing. The diesel generators are not accident initiators and the method of testing the diesel generators cannot initiate an accident. This revision does not relieve the operation of the diesel generator from existing requirements and the diesel generators remain bounded by the assumptions in the USAR accident analysis. The method of testing provides assurance that the diesel generators are available when needed. The proposed revision does not involve any changes in setpoints, plant equipment, plant operation, protective functions, or the design basis of the plant. Therefore, a change in the method of starting the diesel generators during periodic testing would not create a different kind of accident than previously evaluated. Proposed Revision 3: This proposed revision does not add or change any equipment or logic, nor do the changes associated with this revision alter any system operability requirements. The proposed changes for this revision do not introduce any new failure modes for any plant system or component important to safety nor has any new limiting failure been identified as a result of the proposed revision. Since there are no changes to the function, or operation of any system, equipment, or component, the possibility of a new or different kind of accident is not created. The proposed rearrangement of information, and rewording of some [of] the TS requirements are included to enhance usability and alleviate any possible confusion. These changes are strictly editorial have no impact, and do not alter technical content or meaning of the specifications. These editorial changes do not create the possibility of a new or different kind of accident from any previously analyzed. 3. Involve a significant reduction in the margin of safety. Proposed Revision 1: This proposed revision does not result in an overall reduction in the margin of safety. The reduction in margin going from ``immediately'' testing an operable diesel generator to 24 hours to determine no common cause, is offset by the increase in margin resulting from increased diesel generator reliability and availability associated with implementing the vendor recommendations for testing and not exposing the diesel generator to potential grid disturbances when a diesel generator is found to be inoperable. No physical modification to the plant or change in the procedurally prescribed operator actions result from the proposed changes associated with this revision. Operation of the facility in accordance with the proposed revision does not involve a significant reduction in a margin of safety. Proposed Revision 2: This proposed revision is made to increase the reliability and availability of the emergency diesel generators thus enhancing the safety of the plant. Changing the way periodic testing of the diesel generators is conducted does not involve a reduction in safety. The test still demonstrates the ability of the diesel generator to start within the time required, and reach rated voltage and frequency as required in the accident analysis. The test also demonstrates the ability of the diesel generator to start reliably, carry the required load, and ensures the capabilities of the cooling system and other support systems are operable. Therefore, assurance that the diesel generators operate within the limits determined to be acceptable continues to be provided. Implementing manufacturer's recommendations to minimize stress and wear of the diesel engine does not involve a significant reduction in the margin of safety, but rather enhances safety. Proposed Revision 3: This proposed revision deletes the requirement to reduce reactor power level to 25% of rated power upon the loss of either both diesel generators or both incoming power sources. The elimination of this requirement will allow the plant to maintain the existing power level rather than subject the plant to an unnecessary transient. Maintaining the plant at the existing power level provides a more stable operating environment. The equipment and components of the diesel generators or the incoming power sources are not impacted in any way as a result of the proposed revisions. The margin of safety for the diesel generators and the incoming power sources are not significantly reduced since these systems are not altered in any way, and will continue to be surveillance tested as required. Assurance of operability is provided by the normal, scheduled surveillances which have been established at a sufficient interval to provide reasonable assurance of operability. Therefore, the proposed changes do not involve a significant reduction in the margin of safety. The proposed rearrangement of information, and rewording of some [of] the TS requirements are included to enhance usability and alleviate any possible confusion. These changes are strictly editorial have no impact, and do not alter technical content or meaning of the specifications. These editorial 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. The licensee's July 14, 1995, letter revised the proposed changes in their letter of May 5, 1995, to further limit the period of time that continued reactor operation would be allowed with both emergency diesel generators inoperable from 24 to two hours. This revision to the proposed changes is more restrictive and does not impact the licensee's analysis of the criteria of 10 CFR 50.92(c). Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Auburn Public Library, 118 15th Street, Auburn, Nebraska 68305 Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power District, Post Office Box 499, Columbus, Nebraska 68602-0499 NRC Project Director: William D. Beckner Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, Millstone Nuclear Power Station, Unit 1, New London County, Connecticut Date of amendment request: August 31, 1995 Description of amendment request: The proposed amendment modifies the definition of HOT SHUTDOWN and COLD SHUTDOWN to specify that the definitions are not applicable during the performance of an inservice hydrostatic and leak test (IHLT). Technical Specification Section 3.6.B and 4.6.B would be modified by adding Section 3.6.B.1.b and 4.6.B.1.b to identify the requirements that must be satisfied to consider the reactor in COLD SHUTDOWN during the performance of an IHLT. In addition, the proposed amendment will change temperature specific requirements on several pages to mode or condition specific requirements; make several editorial changes; and change the associated Bases. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has rovided its analysis of the issue of no significant hazards consideration, which is presented below: NNECO has reviewed the proposed changes in accordance with 10CFR50.92 and concluded that the changes do not involve a significant hazards consideration (SHC). The bases for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve an SHC because the changes would not: [[Page 49941]] 1. Involve a significant increase in the probability or consequences of an accident previously analyzed. The proposed changes will allow the reactor to be considered in COLD SHUTDOWN during an IHLT with the average reactor coolant temperature greater than 212 deg.F but less than 280 deg.F. The change to allow the reactor to be in COLD SHUTDOWN during the performance of IHLT will not increase the probability or consequences of an accident. The probability of a leak in the reactor pressure boundary during this testing is not increased by considering the reactor to be in COLD SHUTDOWN. The IHLT is performed near water solid, all control rods inserted, and with an appropriate availability of engineering safety features. The stored energy in the reactor core will be very low and the potential for failed fuel and a subsequent increase in coolant activity are minimal. In addition, secondary containment will be operable and capable of handling airborne radioactivity from leaks that could occur during the performance of an IHLT. Requiring secondary containment to be operable will further ensure that potential airborne radiation from leaks will be filtered by one or both trains of SBGT [standby gas treatment], thereby limiting releases to the environment. Therefore, the changes will not significantly increase the consequences of an accident. In the unlikely event of a large pressure boundary leak, the reactor vessel would rapidly depressurize, allowing one or both of the operable core spray systems to operate. Small system leaks would be detected by leakage inspections before significant inventory loss occurred, since leakage inspections are an integral part of the IHLT program. Based upon the above, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Create the possibility of a new or different kind of accident from any previously analyzed. The IHLT conditions remain unchanged. The potential for a system leak remains unchanged since the reactor coolant system is designed for temperatures exceeding 500 deg.F with similar pressures. The change in operable engineered safety features available to mitigate a postulated accident does not reduce the ability to safely mitigate a postulated accident. Adequate ECCS [emergency core cooling system] equipment will be available to mitigate a LOCA [loss of coolant accident] with an assumed single failure. Therefore, this will not create the possibility of a new or different kind of accident from any previously evaluated. 3. Will not involve a significant reduction in a margin of safety. The proposed changes will not have any significant impact on any design basis accident or safety limit. The various engineered safety features which are required by the proposed change will ensure appropriate mitigation of postulated events. Since the test is performed at a near water solid condition and at low decay heat values, no fuel damage is expected in case of an accident such as a LOCA. Nevertheless, secondary containment and the SBGT system will be maintained operable to process air-borne radioactivity from a steam leak that could occur during the performance of the IHLT. Therefore, the proposed change does not constitute a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Learning Resources Center, Three Rivers Community-Technical College, 574 New London Turnpike, Norwich, CT 06360. Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270. NRC Project Director: Phillip F. McKee Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear Power Station, Unit No. 2, New London, Connecticut Date of amendment request: August 31, 1995 Description of amendment request: The proposed change to the Millstone 2 Technical Specifications would remove the phrase ``other than Millstone Unit No. 2'' from Section 6.3.1 on page 6-2. 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: NNECO has reviewed the proposed change in accordance with 10CFR50.92 and concluded that the change does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve an SHC because the change would not: 1. Involve a significant increase in the probability or consequences of an accident previously analyzed. The proposed change does not affect any system or equipment of Millstone Unit No. 2. The proposed change does not affect the qualification of any of the licensed individuals involved in the day-to-day operation of Millstone Unit No. 2. The proposed change corrects a statement which could be interpreted such that an individual who once held a Millstone Unit No. 2 SRO [Senior Reactor Operator] license would not be eligible to be Operations Manager. Since this change does not affect any equipment or operating procedures, does not affect the level of expertise and training required for on-shift personnel, and does not reduce the level of expertise required of operations management, this change does not involve a significant increase in the probability or consequences of an accident previously analyzed. 2. Create the possibility of a new or different kind of accident from any previously analyzed. This change does not affect any equipment or operating procedures, does not affect the level of expertise and training required for on-shift personnel, and does not reduce the level of expertise required of operations management. Therefore, this change does not create the possibility of a new or different kind of accident. 3. Involve a significant reduction in the margin of safety. This change eliminates a phrase which could be interpreted to prevent an individual who had possessed a Millstone Unit No. 2 SRO license from becoming the Operations Manager. The training and experience necessary to possess a Millstone Unit No. 2 SRO license is equivalent to that of other PWRs. Therefore, this proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Local Public Document Room location: Learning Resources Center, Three Rivers Community-Technical College, 574 New London Turnpike, Norwich, CT 06360. Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, CT 06141-0270. NRC Project Director: Phillip F. McKee Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: July 28, 1995 Description of amendment request: The proposed amendment would eliminate the Technical Specifications requirements to perform 10 CFR 50, Appendix J, Type C hydrostatic testing on certain valves that are within closed systems and are assured a water seal following a Design Basis Accident. 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 TS changes do not involve a significant increase in the probability or [[Page 49942]] consequences of an accident previously evaluated. The primary containment (drywell and suppression pool) and the affected closed systems are accident mitigators not accident initiators. The proposed change to the scope of Appendix J, Type C testing does not affect the probability of the DBA [Design Basis Accident]. The valves will continue to be maintained in an operable state, and in their current design configuration. There is no correlation between the scope of Appendix J, Type C testing and accident probability. There are no physical or operational changes to the containment structure, system or components being made as a result of the proposed changes. Therefore, the consequences of a malfunction of equipment important to safety is not increased from those previously evaluated. The consequences of loss-of-coolant accidents (LOCAs) under the proposed change were considered where a single active failure of a containment isolation valve (CIV) or a passive failure of the closed system were reviewed, within the limits of the existing licensing basis. Under the existing licensing basis, a pipe rupture of the seismically qualified ECCS piping does not have to be assumed concurrent with the LOCA, except if it is a consequence of the LOCA. Consideration of consequential failures can be eliminated, since a LOCA inside containment is separated from the affected piping by the containment structure. Consideration of consequential failures of the ECCS piping from LOCAs outside containment are outside the Appendix J design considerations. A single active failure of the CIV, under the LOCA condition, can be accommodated since the closed and water sealed system piping remains as the leakage barrier. The ECCS passive failure criterion does require consideration of system leaks, but not pipe breaks, beyond the initiating LOCA. The capability to make-up water inventory to the suppression pool is adequate to ensure that postulated seat leakage and pipe leakage does not result in a condition that jeopardizes pool level. Make-up capability exists for the suppression pool via the Condensate Storage Tank and Ultimate Heat Sink Spray Pond. Operator actions to make-up the suppression pool are delineated in existing Operating Procedures. The subject valves are single isolation valves associated with lines that penetrate the primary containment, but are not connected directly to the primary containment atmosphere or the reactor coolant pressure boundary. This configuration is described in the LGS UFSAR, Section 6.2.4.3.1.3.1, which states ``the systems which the lines from the suppression pool connect to outside containment are closed systems meeting the appropriate requirements of closed systems.'' The integrity of these closed systems are also monitored and controlled in accordance with TS Section 6.8.4.a. Any leakage that may escape the confines of the closed system will be contained within the Reactor Building, treated by standby gas and radwaste systems, and, therefore, are within the existing LGS licensing bases. Finally, the affected penetrations will continue to be subjected to the periodic 10 CFR 50, Appendix J, Type A test (Integrated Containment Leakage Rate Test). The suppression pool level is designed and operated so that water level is maintained in accordance with current TS, and the associated bases. The supply of water in the suppression pool is assured for 30 days during all DBA, post-accident modes of operation. The lowest water level which the suppression pool will reach was analyzed, and it was determined that the affected lines will remain below this minimum level, thereby assuring a water seal. The valves will continue to be tested and maintained to ensure their operability, and the closed systems' integrity will continue to be monitored and controlled in accordance with TS 6.8.4.a and the performance of the periodic 10 CFR 50, Appendix J, Type A test. Therefore, the proposed changes will not 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 changes do not change the plant response to accident scenarios, and do not introduce new or different scenarios. The primary containment (drywell and suppression pool) and the affected closed systems are accident mitigators not accident initiators. The proposed change to the scope of Appendix J, Type C hydrostatic testing maintains the existing barriers to primary containment bypass leakage by the assurance that a water seal will be maintained for 30 days during all DBA, post-accident modes of operation. The valves will continue to be tested and maintained to ensure their operability, and the closed systems' integrity will continue to be monitored and controlled in accordance with TS 6.8.4.a. Therefore, the proposed changes cannot cause an accident, and the plant response to the design basis events is unchanged, whereby the change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed TS changes do not involve a significant reduction in a margin of safety. The water seal provided by the assurance of a minimum suppression pool level will prevent post-accident containment bypass leakage. Appendix J does not require air leak testing of the valves since the 30 day post-accident supply of water is maintained. In addition, the closed systems' integrity is monitored and controlled in accordance with TS 6.8.4.a. Any leakage that may escape the confines of the closed system will be contained within the Reactor Building, and is within the existing LGS licensing bases. Therefore, the 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 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, Pennsylvania 19464. Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and General Counsel, Philadelphia Electric Company, 2301 Market Street, Philadelphia, Pennsylvania 19101 NRC Project Director: John F. Stolz Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: July 28, 1995 Description of amendment request: The proposed amendments, which are consistent with the Improved Standard Technical Specifications (NUREG-1433), delete the operability and surveillance requirements involving secondary containment differential pressure instrumentation. 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. Deleting the operability and surveillance requirements for the secondary containment differential pressure instrumentation does not involve any changes to the design, function, or operation of any plant components or safety-related systems. There are no changes to the separation, redundancy, qualification, quality assurance or fire protection requirements for the associated components and systems, nor are there any new failure modes created. This activity only removes operability and surveillance requirements from the Technical Specifications for selected plant components associated with the secondary containment differential pressure trip functions. No credit for operation of these trip functions is taken in any design basis accidents valuated in the SAR [Safety Analysis Report]. Thesecomponents will be maintained in accordance with the plant preventive maintenance program. The failure of any of these components does not result in the occurrence of an accident. Consequently, there is no increase in the probability of occurrence of an accident previously evaluated in the SAR. The Outside Atmosphere to Reactor Enclosure Delta Pressure-Low and Outside Atmosphere To Refueling Area Delta Pressure-Low trip functions are not symptomatic of a design basis accident. No credit for operation of the trip functions is taken in any design basis accidents evaluated in the SAR. Neither failure of the differential pressure components nor failure to generate the associated trip functions affects the consequences of an accident previously evaluated in the SAR. The appropriate [[Page 49943]] accident prevention and mitigation actions are generated from other plant parameters symptomatic of an accident. Sufficient plant parameters symptomatic of a design basis accident are monitored to initiate the appropriate actions as evaluated in the SAR. Furthermore, all safety-related systems will still be able to perform all of their design basis safety-related functions. Consequently, there is no increase in the consequences of an accident previously evaluated in the SAR. Therefore, the proposed TS changes do not involve an increase in 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 failure of the differential pressure automatic isolation instrumentation components does not result in the occurrence of an accident. The failure to generate the associated trip functions does not result in the occurrence of an accident. This activity does not involve any changes to the design, function, or operation of any plant components or safety-related systems. There are no changes to the separation, redundancy, qualification, quality assurance or fire protection requirements for the associated components and systems. These components will be maintained in accordance with the plant preventative maintenance program. Consequently, there is no possibility of an accident of a different type than previously evaluated in the SAR. Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. The proposed TS changes do not involve a significant reduction in a margin of safety. The ability of secondary containment to minimize any ground level release of radioactive material which may result from any accident is not affected. Surveillance and operability requirements for secondary containment SGTS [Standby Gas Treatment System] and RERS [Reactor Enclosure Recirculation System] are not changed by this activity. Draw down time, leakage factors, secondary containment system ratings, and secondary containment system response to a LOCA [Loss of Coolant Accident] or refueling accident are not affected by this activity. SGTS and RERS initiation will continue to occur when plant parameters symptomatic of a LOCA or refueling accident exceed predetermined values. There are no changes to the inputs for the post-LOCA offsite dose analysis. Therefore, the 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 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, Pennsylvania 19464. Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and General Counsel, Philadelphia Electric Company, 2301 Market Street, Philadelphia, Pennsylvania 19101 NRC Project Director: John F. Stolz Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: July 28, 1995 Description of amendment request: The proposed amendment would modify Technical Specifications (TS) Surveillance Requirements 4.9.1.1, 4.9.1.2, 4.9.3, 4.9.5, and 4.9.8 to delete specific requirements to perform surveillances just prior to beginning or resuming core alterations or control rod withdrawal associated with refueling activities. This proposed TS change would delete the phrase ``incore instrumentation'' from the footnote in TS Section 3/4.9.5, ``Communications.'' 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 do not involve any physical changes to plant systems or equipment. The proposed TS changes only delete those Surveillance Requirements (SRs) pertaining to the performance of tests just prior to beginning or resuming core alterations or control rod withdrawal, and revises a footnote description to be consistent with the current TS definition of ``Core Alteration.'' The proposed TS changes do not revise any of the other applicable periodic SRs, or modify any procedural controls currently in place governing fuel handling operations. The periodic surveillance test frequencies provide adequate assurance that the equipment will remain in an operable condition. The normal periodic surveillance intervals bound those surveillance intervals for the tests that are being altered by this proposed TS change. In the event that one of the periodic surveillances has not been performed within the specified time interval, entry into the specified condition (i.e., performance of core alterations, control rod withdrawal, or handling of fuel or control rods) is not permitted as required by TS 4.0.4 until the surveillance has been satisfactorily completed. The consequences of an accident are not increased by the proposed TS changes, since the changes only involve revising the frequency of conducting surveillance tests. The method of operation or performance of plant structures, systems, or plant components are not affected by the proposed TS changes. The proposed TS changes will not impact the operation of any fuel handling equipment, and therefore, the potential for a Fuel Handling Accident as described in Section 15.7.4 of the LGS [Limerick Generating Station] Updated Final Safety Analysis Report (UFSAR) is not increased. In addition, any unexpected reduction of water level in the reactor cavity or fuel pool at the start of fuel handling or control rod handling will be immediately apparent to operators by direct observation. Plant procedures utilized by the refueling personnel require the suspension of core component transfers in the event of loss of water inventory. Therefore, the proposed TS change does not involve an increase in 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 only involve changes to the frequency in which the specified surveillance tests are performed. The proposed TS changes do not revise any of the other applicable periodic SRs, or modify any procedural controls currently in place governing fuel handling operations. The periodic surveillance test frequencies provide adequate assurance that the equipment will remain in operable condition. The periodic surveillance intervals bound those surveillance intervals for the tests that are being altered by this proposed TS change. The refueling interlock system combined with strict procedural controls provide multiple barriers to preclude an inadvertent criticality. The proposed TS changes do not involve any physical changes to plant systems or equipment. The proposed TS changes do not alter the configuration of the plant or the way that the plant is operated. The associated plant equipment will continue to function as designed. This equipment is not designed to perform any other function than it is presently capable of, and therefore, will not affect the operation of any other plant equipment. Therefore, the proposed TS change does 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 do not involve any physical changes to plant systems or equipment. The reactor will continue to be maintained subcritical during refueling operations and reactor water level will be maintained at the required level (i.e., above the vessel flange). The proposed TS changes do not affect the operation of other plant systems and equipment essential in maintaining reactor water temperature during refueling operations, or the capability in responding to a postulated Fuel Handling Accident. The proposed changes do not adversely affect reliability of the refueling interlocks or refuel platform communications equipment. [[Page 49944]] Since the proposed changes only impact the frequency in which certain surveillance tests are performed, and do not change the plant configuration or setpoints, there is substantial assurance that the reactor will be maintained subcritical during refueling. Therefore, the proposed TS changes do not involve a 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: Pottstown Public Library, 500 High Street, Pottstown, Pennsylvania 19464. Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and General Counsel, Philadelphia Electric Company, 2301 Market Street, Philadelphia, Pennsylvania 19101 NRC Project Director: John F. Stolz Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania Date of amendment request: July 28, 1995 Description of amendment request: The proposed amendment would revise Technical Specifications Table 4.3.1.1-1, ``Reactor Protection System Instrumentation Surveillance Requirements'', to reflect changes to the surveillance test frequency requirements for various Reactor Protection System [RPS] instrumentation. 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. In all of the applicable SAR [Safety Analysis Report] evaluated events, the IRM [Intermediate Range Monitor] and APRM [Average Range Power Monitor] instrumentation is credited for performing a mitigating function (i.e., initiating a scram), to terminate the transient prior to a safety limit being exceeded. The proposed TS changes do not alter the RPS configuration, or RPS instrumentation setpoints, nor do they change the manner in which the IRM and APRM instrumentation carry out the scram functions. Therefore the consequences of any potential malfunction of equipment important to safety will remain unchanged. In each case where a startup surveillance test requirement is proposed to be deleted, (i.e., IRM and APRM), the normal surveillance test frequency specified for the required Operational Condition remains unchanged (except for the APRM Upscale Setdown functional test). The startup surveillance requirement is conservati [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%3AX95-10927. Public record. Not legal advice.