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| | number = ML052780034 | | | number = ML052780034 |
| | issue date = 10/20/2014 | | | issue date = 10/20/2014 |
| | title = Current Facility Operating License NPF-39, Technical Specifications, Revised 08/15/2018 | | | title = Current Facility Operating License NPF-39, Technical Specifications, Revised 08/06/2024 |
| | author name = Dean W | | | author name = Dean W |
| | author affiliation = NRC/NRR/DORL | | | author affiliation = NRC/NRR/DORL |
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| | contact person = Boska J, NRR/DORL | | | contact person = Boska J, NRR/DORL |
| | document type = License-Operating (New/Renewal/Amendments) DKT 50, Technical Specifications | | | document type = License-Operating (New/Renewal/Amendments) DKT 50, Technical Specifications |
| | page count = 412 | | | page count = 383 |
| }} | | }} |
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| =Text= | | =Text= |
| {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO 50-352 LIMERICK GENERATING STATION, UNIT 1 RENEWED FACILITY OPERATING LICENSE Renewed License No. NPF-39 | | {{#Wiki_filter:}} |
| : 1. The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
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| A. The application for renewed license filed by Exelon Generation Company, LLC (Exelon Generation Company or the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. Construction of the Limerick Generating Station, Unit 1 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-106 and the application, as amended, the provisions of the Act and the regulations of the Commission; C. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission (except as exempted from compliance in Section 2.D. below);
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| D. There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D. below);
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| E. The licensee is technically qualified to engage in the activities authorized by this renewed license in accordance with the Commission's regulations set forth in 10 CFR Chapter I; F. The licensee has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; G. The issuance of this renewed license will not be inimical to the common defense and security or the health and safety of the public; Renewed License No. NPF-39
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| H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of this Renewed Facility Operating License No. NPF-39, subject to the conditions for protection of the environment set forth in the Environmental Protection Plan attached as Appendix B, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and
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| : 1. The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70.
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| J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21 (a)(1), and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21 (c),
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| such that there is reasonable assurance that the activities authorized by the renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations.
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| : 2. Based on the foregoing findings, the Partial Initial Decisions issued by the Atomic Safety and Licensing Board dated March 8, 1983, August 29, 1984, May 2, 1985 and July 22, 1985, and the Decision of the Appeal Board dated September 26, 1984, regarding this facility, and approval by the Nuclear Regulatory Commission in its Memorandum and Order dated August 8, 1985, the license for Fuel Loading and Low Power Testing, License No. NPF-27, issued on October 26, 1984, is superseded by Renewed Facility Operating License NPF-39 hereby issued to the Exelon Generation Company (the licensee), to read as follows:
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| A. This renewed license applies to the Limerick Generating Station, Unit 1, a boiling water nuclear reactor and associated equipment, owned by Exelon Generation Company. The facility is located on the licensee's site in Montgomery and Chester Counties, Pennsylvania on the banks of the Schuylkill River approximately 1.7 miles southeast of the city limits of Pottstown, Pennsylvania and 21 miles northwest of the city limits of Philadelphia, Pennsylvania, and is described in the licensee's Final Safety Analysis Report, as supplemented and amended, and in the licensee's Environmental Report-Operating License Stage, as supplemented and amended.
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| B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Exelon Generation Company:
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| (1) Pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use, and operate the facility at the designated location in Montgomery and Chester Counties, Pennsylvania, in accordance with the procedures and limitations set forth in this renewed license; Renewed License No. NPF-39
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| (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.
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| C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2. D. below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
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| ( 1) Maximum Power Level Exelon Generation Company is authorized to operate the facility at reactor core power levels not in excess of 3515 megawatts thermal (100% rated power) in accordance with the conditions specified herein and in Attachment 1 to this license. The items identified in Attachment 1 to this renewed license shall be completed as specified. Attachment 1 is hereby incorporated into this renewed license.
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| (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 231, are hereby incorporated into this renewed license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
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| Renewed License No. NPF-39 Amendment No. 231
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| (3) Fire Protection (Section 9.5, SSER-2, -4)*
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| Exelon Generation Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evaluation Report dated August 1983 through Supplement 9, dated August 1989, and Safety Evaluation dated November 20, 1995, subject to the following provision:
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| The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
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| The Information contained on FOL pages 5 and 6 were intentionally omitted.
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| *The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
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| Renewed License No. NPF-39
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| (16) Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 230, are hereby incorporated into this renewed license.
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| Exelon Generation Company shall operate the facility in accordance with the Additional Conditions.
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| (17) Exelon Generation Company shall provide to the Director of the Office of Nuclear Reactor Regulation a copy of any application, at the time it is filed, to transfer (excluding grants of security interests or liens) from Exelon Generation Company to its direct or indirect parent, or to any other affiliated company, facilities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company's consolidated net utility plant, as recorded on Exelon Generation Company's books of account.
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| (18) Exelon Generation Company shall have decommissioning trust funds for Limerick, Unit 1, in the following minimum amount, when Limerick, Unit 1, is transferred to Exelon Generation Company:
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| Limerick, Unit 1 $94,127,446 (19) The decommissioning trust agreement for Limerick, Unit 1, at the time the transfer of the unit to Exelon Generation Company is effected and thereafter, is subject to the following:
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| (a) The decommissioning trust agreement must be in a form acceptable to the NRC.
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| (b) With respect to the decommissioning trust fund, investments in the securities or other obligations of Exelon Corporation or affiliates thereof, or their successors or assigns are prohibited. Except for investments tied to market indexes or other non-nuclear sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.
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| (c) The decommissioning trust agreement for Limerick, Unit 1, must provide that no disbursements or payments from the trust shall be made by the trustee unless the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.
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| Renewed License No. NPF-39 Amendment No. 230
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| (d) The decommissioning trust agreement must provide that the agreement can not be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.
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| (e) The appropriate section of the decommissioning trust agreement shall state that the trustee, investment advisor, or anyone else directing the investments made in the trust shall adhere to a prudent investor standard, as specified in 18 CFR 35.32(a)(3) of the Federal Energy Regulatory Commissions regulations.
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| (20) Exelon Generation Company shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application for approval of the transfer of Limerick, Unit 1, renewed license and the requirements of the Order approving the transfer, and consistent with the safety evaluation supporting the Order.
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| (21) Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:
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| (a) Fire fighting response strategy with the following elements:
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| : 1. Pre-defined coordinated fire response strategy and guidance
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| : 2. Assessment of mutual aid fire fighting assets
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| : 3. Designated staging areas for equipment and materials
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| : 4. Command and control
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| : 5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:
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| : 1. Protection and use of personnel assets
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| : 2. Communications
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| : 3. Minimizing fire spread
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| : 4. Procedures for implementing integrated fire response strategy
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| : 5. Identification of readily-available pre-staged equipment
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| : 6. Training on integrated fire response strategy
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| : 7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:
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| : 1. Water spray scrubbing
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| : 2. Dose to onsite responders Renewed License No. NPF-39
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| (22) The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.
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| (23) Upon implementation of Amendment No. 188 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 4.7.2.2.a, in accordance with TS 6.16.c.(i), the assessment of CRE habitability as required by Specification 6.16.c.(ii), and the measurement of CRE pressure as required by Specification 6.16.d, shall be considered met. Following implementation:
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| (a) The first performance of SR 4.7.2.2.a, in accordance with Specification 6.16.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 4.0.2, as measured from September 16, 2004, the date of the most recent successful tracer gas test, as stated in the December 10, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.
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| (b) The first performance of the periodic assessment of CRE habitability, Specification 6.16.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 4.0.2, as measured from September 16, 2004, the date of the most recent successful tracer gas test, as stated in the December 10, 2004 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.
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| (c) The first performance of the periodic measurement of CRE pressure, Specification 6.16.d, shall be within 24 months, plus the 180 days allowed by SR 4.0.2, as measured from September 16, 2004, the date of the most recent successful pressure measurement test, or within 180 days if not performed previously.
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| (24) The information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d), is henceforth part of the UFSAR which will be updated in accordance with 10 CFR 50.71(e). As such, the licensee may make changes to the programs and activities described in the UFSAR supplement, without prior Commission approval, provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
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| Renewed License No. NPF-39
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| (25) The licensee's UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and as revised in accordance with license condition 2.C.(24),
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| describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).
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| (a) Exelon Generation Company shall implement those new programs and enhancements to existing programs no later than April 26, 2024.
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| (b) Exelon Generation Company shall complete those activities designated for completion prior to the PEO, as noted in Commitment Nos. 18, 19, 20, 22, 23, 24, 28, 29, 30, 38, 39, 40, 41, 42, 43, and 47, of Appendix A of NUREG-2171, Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2, no later than April 26, 2024, or the end of the last refueling outage prior to the period of extended operation, whichever occurs later.
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| (c) Exelon Generation Company shall notify the NRC in writing within 30 days after having accomplished item (a) above and include the status of those activities that have been or remain to be completed in item (b) above.
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| D. The facility requires exemptions from certain requirements of 10 CFR Part 50.
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| These include (a) exemption from the requirement of Appendix J, the testing of containment air locks at times when the containment integrity is not required (Section 6.2.6.1 of the SER and SSER-3), (b) exemption from the requirements of Appendix J, the leak rate testing of the Main Steam Isolation Valves (MSIVs) at the peak calculated containment pressure, Pa, and exemption from the requirements of Appendix J that the measured MSIV leak rates be included in the summation for the local leak rate test (Section 6.2.6 of SSER-3), (c) exemption from the requirement of Appendix J, the local leak rate testing of the Traversing Incore Probe Shear Valves (Section 6.2.6 of the SER and SSER-3).
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| These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.
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| Therefore these exemptions are hereby granted pursuant to 10 CFR 50.12 and 50.47(c). With the granting of these exemptions the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provision of the Act, and the rules and regulations of the Commission.
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| Renewed License No. NPF-39
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| E. Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
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| The combined set of plans 1 , submitted by letter dated May 17, 2006, is entitled:
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| "Limerick Generating Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2." The set contains Safeguards Information protected under 10 CFR 73.21.
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| Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 204 and modified by License Amendment No. 218.
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| F. Deleted G. The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
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| H. This renewed license is effective as of the date of issuance and shall expire at midnight on October 26, 2044.
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| FOR THE NUCLEAR REGULATORY COMMISSION IRA!
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| William M. Dean, Director Office of Nuclear Reactor Regulation Attachments/Appendices:
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| : 1. Attachments 1-2
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| : 2. Appendix A - Technical Specifications
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| : 3. Appendix B - Environmental Protection Plan
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| : 4. Appendix C - Additional Conditions Date of Issuance: October 20, 2014 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
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| Renewed License No. NPF-39 Amendment No. 218
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| ATTACHMENT 1 To-NPF 39 The attachment identifies items which must be completed to the satisfaction of the staff in accordance with the operational modes or conditions identified below.
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| : 1. OUTSTANDING ITEM TO BE ACCOMPLISHED PRIOR TO INITIALLY INERTING THE CONTAINMENT AS REQUIRED BY TECHNICAL SPECIFICATIONS 3/4.6.6.3 AND 3/4.10.5 Complete the modifications to the liquid nitrogen vaporization facility and the containment inerting system described in the licensee's letter dated September 26, 1984 or alternate modifications determined acceptable following an evaluation per 10 CFR 50.59 (IE Bulletin 84-01)
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| In the event alternate modifications are used, describe these modifications and their supporting bases in a report to NRC Region I within thirty days of their implementation.
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| : 2. OUTSTANDING ITEMS TO BE CORRECTED BY THE FIRST REFUELING OUTAGE
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| : a. Seal the conduits to instruments in the pipe tunnel. (Inspection Report 50-352/84-27, Item 04)
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| : b. Complete the actions for Construction Deficiency Report (84-00-10 "Water accumulation in diesel fuel oil tanks.")
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| Renewed License No. NPF-39
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| ATTACHMENT 2 To NPF-39 This attachment identifies the shift operating staff experience requirements.
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| At all times the plant is in an operating condition other than cold shutdown or refueling, the licensee shall have a licensed senior operator on each shift who has had at least six months of hot operating experience on a same type plant, including at least six weeks at power levels greater than 20% of full power, and who has had startup and shutdown experience. For those shifts where such an individual is not available on the plant staff, an advisor shall be provided who has had at least four years of power plant experience, including two years of nuclear plant experience, and who has had at least one year of experience on shift as a licensed senior operator at a similar type facility. Advisors, as a minimum, shall be trained on plant procedures, technical specifications and plant systems, and shall be examined on these topics at a level sufficient to assure familiarity with the plant. These advisors or suitably qualified replacements shall be retained until at least one of the senior operators on each shift has the required experience. The NRC shall be notified at least 30 days prior to the release of any special assigned advisors.
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| Renewed License No. NPF-39
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| MTWI'iJf AlITHORI- Y FILE COPY NUREG-1 149 DO NOT REMOVE Technical Specifications Limerick Generating Station, Unit No. 1 Docket No. 50-352 Appendix "A" to License No. NPF-39 Issued by the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1985
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| _s
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| DEFIN TION SECTION
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| : 1. 0 DEF IN IT IONS
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| : 1. 1 ACT I ON ....................................................... 1-1 1.2 AVERAGE PLANAR EXPOSURE ...................................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ................... 1-1 1.4 CHANNEL CALIBRATION .......................................... 1-1 1.5 CHANNEL CHECK ................................................ 1-1 1.6 CHANNEL FUNCTIONAL TEST ...................................... 1-1
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| : 1. 7 CORE ALTERATION .............................................. 1-2 1.7A CORE OPERATING LIMITS REPORT ................................. 1-2 1.8 CRITICAL POWER RATIO ......................................... 1-2 1.9 DOSE EQUIVALENT I-131. ....................................... 1-2 1.9a DOWNSCALE TRIP SETPOINT (DTSP) ............................... 1-2 1 . 9b DRAIN TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 2 1.10 (DELETED) .................................................... l-2a 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ........... l-2a 1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME .... 1-3 1.13 (DELETED) .................................................... 1-3 1.14 (DELETED) .................................................... 1-3 1.15 FREQUENCY NOTATION ........................................... 1-3 1.15a HIGH (POWER) TRIP SETPOINT (HTSP) ............................ 1-3 1.16 IDEN TI FI ED LEAKAGE ........................................... 1-3 1.16a INTERMEDIATE (POWER) TRIP SETPOINT (ITSP) .................... 1-3 1.17 ISOLATION SYSTEM RESPONSE TIME ............................... 1-3 1.18 LIMITING CONTROL ROD PATTERN ................................. 1-3 1.19 LINEAR HEAT GENERATION RATE .................................. 1-3 1.20 LOGIC SYSTEM FUNCTIONAL TEST ................................. 1-4 LIMERICK - UNIT 1 Amendment No. JJ, J-7., ~. +74, 227
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| Q"W}JD.."JQ~N~S==
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| SECTION DEFT NITIONS (Conti nued) PAGE 1.20a LOW (POWER) TRIP SETPOINT (LTSP) ............................. 1-4 1.2l (DELETED) .................................................... 1-4 1.22 r~EMBER(S) OF THE PUBLIC ...................*..............*... 1-4 1.22a MAPFAC(F) - (MAPLHGR FLOW FACTOR) ..........*......*.......... 1-4 1.22b MAPFAC(P) (POWER DEPENDENT MAPLHGR MULTIPLIER) *.**.....*..* 1-4 1.23 MINIMUM CRITICAL POWER RATIO (MCPR) .......................... 1-4 1.24 OFFSITE DOSE CALCULATION MANUAL ...................*.......... 1-4 1.25 OPERABLE - OPERABILITY ...................................**.. 1-4 1.26 OPERATIONAL CONDITION - CONDITION .......*.......*............ 1-5 1.27 PHYSICS TESTS .....*.....*.....................*..*****.....*. 1-5
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| : 1. 28 PRESSURE BOUNDARY LEAKAGE ....................*.*..*.....**.*. 1- 5 1.29 PRIMARY CONTAINMENT INTEGRITy ................................ 1-5 1.30 PROCESS CONTROL PROGRAM ..*..............*...............*.... 1-5 1.31 PURGE - PURGING *..**...*................*........***.*....... 1-6 1.32 RATED THERMAL POWER .......................................... 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITy ............ 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME ...................... 1-6 1.35 RECENTLY IRRADIATED FUEL ..................................... 1-6 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITy ..........**.. 1-6
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| : 1. 37 REPORTABLE EVENT ....*.............................*....**.... 1-7
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| : 1. 37a RESTRICTED AREA *...**..*...*............................**.** 1-7 1.38 (DELETED) *...***..*.*.....**............................*..*. 1-7 1.39 SHUTDOWN MARGIN .............................................. 1-7 1.40 SITE BOUNDARy .................................*.............. 1-7 1.41 SOURCE CHECK *..*....*.............................*..*....*.. 1-7 LIMERICK - UNIT 1 ii Amendment No. JJ,4S,~,ee,~,207
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| INDEX DEFINITIONS SECTION DEFINITIONS (Continued) -PAGE
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| ~~~.. ... ..... ... .. ...... .... .......
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| 1.42 - STAGGERED TEST BASIS . .... 1-8 1.43 THERMAL POWER .. 1-8 1.43A TURBINE BYPASS SYSTEM RESPONSE TIME .1-8 1.44 UNIDENTIFIED LEAKAGE .. 1-8 1.45 UNRESTRICTED AREA .. 1-8 1.46 VENTILATION EXHAUST TREATMENT SYTEM .1-8 1.47 VENTING ......... 1-8 Table 1.1, Surveillance Frequency Notation ......................... 9 Table 1.2, Operational Conditions;................................... 1-10 LIMERICK - UNIT : Amendment No. 66 FED .. 10 1-9
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| INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow . .2-1 THERMAL POWER, High Pressure and High Flow . .2-1 Reactor Coolant System Pressure .. 2-1 Reactor Vessel Water Level .2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints .2-3 Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints .2-4 BASES 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow .B 2-1 THERMAL POWER, High Pressure and High Flow .B 2-2 Left Intentionally Blank .B 2-3 Left Intentionally Blank .B 2-4 Reactor Coolant System Pressure .B 2-5 Reactor Vessel Water Level .B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints. B 2-6 LIMERICK - UNIT 1 iv Amendment No.',9 33 OCT 3 0 198q
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| INDEX K.) LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE!REQUIREMENTS SECTION PAGE
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| - 3/4.'0 APPLICABILITY............................. ;...
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| - 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS -'
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| 3/4.1.1 SHUTDOWN MARGIN. ............ . ..... 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES . ....... .3/4 1-2 3/4.1.3 CONTROL RODS - , -
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| Control Rod Operability ........................... , 3/4 1-3 Control Rod Maximum Scram InsertonTimes . . 3/4 1-6
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| . .. r ..
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| Control Rod Average Scram Insertion Times........:' 3/4 1-7 Four Control Rod Group Scram Insertion Times............ 3/4 1-8 Control Rod Scram Accumulators........................... 3/4 1-9
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| ' Control Rod Drive Coupling ....... 3/4 1-11 Control Rod PositionInaication ............... 3/4 1-13 Control Rod Drive Housing Support . .3/4 1-15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS
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| ---,RodWorth Minimizer.....................;................ 3/4 1-16 Rod Block Monitor ........... 3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM . . ... 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Solution Temperature/Concentration Requirements....... . ........ 3/4 1-21 Figure 3.1.5-2 Deleted (LEFT BLANK INTENTIONALLY).. 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ............... 3/4 2-1 Information on pages 3/4 2-2 thru 3/4 2-6c has been INTENTIONALLY OMITTED, refer to note on page 3/4 2-2 ... 3/4 2-2
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| - LIMERICK'- UNIT 1 v Amendment No.,2f, 37 MAY 15 1990
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| , eIra INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE POWER DISTRIBUTION LIMITS (Continued) 3/4.2.2 (DELETED)..3/4 2-7 I 3/4.2.3 MINIMUM CRITICAL POWER RATIO .3/4 2-8 Information on pages 3/4 2-10 thru 3/4 2-11 has been INTENTIONALLY OMITTED, refer to Note on page 3/4 2-10.... 3/4 ; 2-10 3/4.2.4 LINEAR HEAT GENERATION RATE .3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION .3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation .... 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times .3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements..3/4 3-7 I-LIMERICK UNIT - 1 vi Amendment No. 7, 10, fl, %7, 66 1.3 -, 15u
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION INSTRUMENTATION (Continued) 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION .......................... 3/4 3-9 Table 3.3.2-1 Isolation Actuation Instrumentation .......................... 3/4 3-11 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints ................ 3/4 3-18 Table 3.3.2-3 Isolation System Instrumen-tation Response Time ..................... 3/4 3-23 Table 4.3.2.1-1 Isolation Actuation Instrumen-tation Surveillance Requirements ........................... 3/4 3-27 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION .............................................. 3/4 3-32 Table 3.3.3-1 Emergency Core Cooling System Actuation Instrumentation ................ 3/4 3-33 Table 3.3.3-2 Emergency Core Cooling System Actuation Instrumentation Setpoints ................................ 3/4 3-37 Table 3.3.3-3 Emergency Core Cooling System Response Times ........................... 3/4 3-39 Table 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation Surveillance Requirements .............. 3/4 3-40 3/4.3.3.A REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC)
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| INSTRUMENTATION .............................................. 3/4 3-4la Table 3.3.3.A-l RPV Water Inventory Control (WIC)
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| Instrumentation ......................... 3/4 3-4lb Table 3.3.3.A-2 RPV Water Inventory Control (WIC)
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| Instrumentation Set points ............... 3/4 3-4ld Table 4.3.3.A-l RPV Water Inventory Control (WIC)
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| Instrumentation Surveillance Requirements3/4 3-4le 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation .......... 3/4 3-42 Table 3.3.4.1-1 ATWS Recirculation Pump Trip System Instrumentation ................. 3/4 3-43 Table 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation Setpoints .............................. 3/4 3-44 Table 4.3.4.1-1 (Deleted) .............................. 3/4 3-45 End-of-Cycle Recirculation Pump Trip System Instrumentation .............................................. 3/4 3-46 LIMERICK - UNIT 1 Vii Amendment No. JJ, ~ . 227
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)
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| Table 3.3.4.2-1 End-of-Cycle Recirculation Pump Trip System Instrumentation ............ 3/4 3-48 Table 3.3.4.2-2 End-of-Cycle Recirculation Pump Trip Setpoints ......................... 3/4 3-49 Table 3.3.4.2-3 End-of-Cycle Recirculation Pump Trip System Response Time .............. 3/4 3-50 Table 4.3.4.2.1-1 (Deleted) ............................ 3/4 3-51 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION ............................................... 3/4 3-52 Table 3.3.5-1 Reactor Core Isolation Cooling System Actuation Instrumenta-tion ..................................... 3/4 3-53 Table 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation Setpoints ................................ 3/4 3-55 Table 4.3.5.1-1 (Deleted) ............................... 3/4 3-56 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION ............................. 3/4 3-57 Table 3.3.6-1 Control Rod Block Instrumenta-tion ...................................... 3/4 3-58 Table 3.3.6-2 Control Rod Block Instrumenta-tion Setpoints ............................ 3/4 3-60 Figure 3.3.6-1 SRM Count Rate vs Signal-to-Noise Ratio .................................... 3/4 3-60b Table 4.3.6-1 Control Rod Block Instrumenta-tion Surveillance Requirements ............ 3/4 3-61 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation .......................... 3/4 3-63 Table 3.3.7.1-1 Radiation Monitoring Instrumentation ........................ 3/4 3-64
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| . LIMERICK - UNIT I viii Amendment No. 34,186
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)
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| Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance Requirements ......................... 3/4 3-66 The information from pages 3/4 3-68 through 3/4 3-72 has been intentionally omitted. Refer to note on page 3/4 3-68 ............... 3/4 3-68 The information from pages 3/4 3-73 through 3/4 3-75 has been intentionally omitted. Refer to note on page 3/4 3-73 .............. 3/4 3-73 Remote Shutdown System Instrumentation and Controls ......... 3/4 3-76 Table 3.3.7.4-1 Remote Shutdown System Instrumentation and Controls ......... 3/4 3-77 Table 4.3.7.4-1 (Deleted) ............................ 3/4 3-83 Accident Monitoring Instrumentation ......................... 3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-tation ............................... 3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instrumenta-tion Surveillance Requirements ....... 3/4 3-87 Source Range Monitors ....................................... 3/4 3-88 This information from page 3/4 3-89 has been intentionally omitted.
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| Refer to note on page ....................................... 3/4 3-89 Chlorine Detection System ................................... 3/4 3-90 Toxic Gas Detection System .................................. 3/4 3-91 DELETED; Refer to note on page .............................. 3/4 3-92 LIMERICK - UNIT 1 ix Amendment No. 48, -76, 4-04, -,186
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| l!iQ.Ll LIMITING .C.OlilllllilliS. £OR filEB.hlliltl AND Sl!fil/ULLANCI RE,QUlREMENJS.
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| .SECT ION JNSTRUMENTATION (Continued)
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| (Deleted) ...................................................... 3/4 3-97 The information from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted. Refer to note on page 3/4 3-98 ....................... 3/4 3-98 The information from pages 3/4 3-103 through 3/4 3-108 has been intentionally omitted. Refer to note on page 3/4 3-103 ...................... 3/4 3-103 3/4.3.8 (Deleted) The information on pages 3/4 3-110 and 3/4 3-111 has been intentionally omitted.
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| Refer to note on page 3/4 3-110 ............... 3/4 3-110 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION ........................................... 3/4 3-112 Table 3.3.9-1 Feedwater/Main Turbine Trip System Actuation Instrumentation .............. 3/4 3-113 Table 3.3.9-2 Feedwater/Main Turbine Trip System Actuation Instrumen-tation Setpoints .............................. 3/4 3-114 Table 4.3.9.1-1 (Deleted) ..................................... 3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops ............................................. 3/4 4-1 LIMERICK - UNIT l X Amendment No. 48, MO, M, lw, '* 228
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILlANCE REOUIREMENTS SECTION REACTOR COOLANT SYSTEM (Continued)
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| Figure 3.4.1.1-1 Deleted 3/4 4-3 Jet Pumps 3/4 4-4 Recirculation Pumps 3/4 4-5 Idle Recirculation Loop Startup 3/4 4-6 3/4.4.2 SAFETY/RELIEF VALVES 3/4 4-7 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems 3/4 4-8 Operational Leakage 3/4 4-9 Table 3.4.3.2-1 Deleted 3/4 4-11 3/4.4.4 (Deleted) The information from pages 3/4 4-12 through 3/4 4-14 has been intentionally omitted.
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| Refer to note on page 3/4 4-12 3/4 4-12 3/4.4.5 SPECIFIC ACTIVITy 3/4 4-15 Table 4.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program , 3/4 4-17 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System '" 3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs. Reactor Vessel Pressure 3/4 4-20 Table 4.4.6.1.3-1 Deleted 3/4 4-21 Reactor Steam Dome 3/4 4-22 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES 3/4 4-23 3/4.4.8 (DELETED) 3/4 4-24 LIMERICK - UNIT 1 xi Amendment No. +e+,++4,+++,~,199
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION REACTOR COOLANT SYSTEM (Continued) 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown ............................................... 3/4 4-25 Cold Shutdown .............................................. 3/4 4-26 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING ........................................... 3/4 5-1 3/4.5.2 REACTOR PRESSURE VESSEL (RPV)
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| WATER INVENTORY CONTROL (WIC) .............................. 3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER ........................................ 3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity .............................. 3/4 6-1 Primary Containment Leakage ................................ 3/4 6-2 Primary Containment Air Lock ............................... 3/4 6-5 MSIV Leakage Alternate Drain Pathway ....................... 3/4 6-7 Primary Containment Structural Integrity ................... 3/4 6-8 Drywel l and Suppression Chamber Internal Pressure .......... 3/4 6-9 Drywell Average Air Temperature ............................ 3/4 6-10 Drywell and Suppression Chamber Purge System ............... 3/4 6-11 3/4.6.2 DEPRESSURIZATION SYSTEMS Suppression Chamber ........................................ 3/4 6-12 Suppression Pool Spray ..................................... 3/4 6-15 Suppression Pool Cooling ................................... 3/4 6-16 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES ....................... 3/4 6-17 LIMERICK - UNIT 1 xii Amendment No. J.J, +G+, J..4e, 227
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| INDEX I
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| LIMITING COND-TIONS FOR OPERATION AND SURVEILLANCE REQUIRMENTS SECTION PAGE CONTAINMENT SYSTEMS (Continued) 3/4.6.4 VACUUM RELIEF Suppression Chamber - Drywell Vacuum Breakers ...... ............. 3/4 6-44 3/4.6.5 SECONDARY CONTAINMENT Reactor Enclosure Secondary Containment Integrity ..... .......... 3,4 6-46 Refueling Area Secondary Containment Integrity ...... ............. 3/4 6-47 Reactor Enclosure Secondary Containment Automatic Isolation Valves ................................................ 3/4 6-48 Refueling Area Secondary Containment Automatic Isolation Valves ................................................ 3/4 6-50 Standby Gas Treatment System - Common System ...... ............... 3/4 6-52 Reactor Enclosure Recirculation System ....... ................... 3/4 6-55 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Deleted ......................... : 3/4 6-57 Krywell Hydrogen Mixing System ................................... 3/4 6-58 Drywell and Suppression Chamber Oxygen Concentration ..... ....... 3/4 6-59 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS Residual Heat Removal Service Water System - Common System ...................................................... 3/4 7-1 Emergency Service Water System - Common System ..... ............. 3/4 7-3 Ultimate Heat Sink . .............................................. 3/4 7-5 LIMERICK - UNIT 1 X'lil Amendment No. 2X7, 44, 10-5, 173
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued) 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM ........................................................ 3/4 7-6 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM ......................... 3/4 7-9 3/4.7.4 SNUBBERS ...................................................... 3/4 7-11 Figure 4.7.4-1 Sample Plan 2) For Snubber Functional Test .......................... 3/4 7-16 3/4.7.5 SEALED SOURCE CONTAMINATION ................................... 3/4 7-17 3/4.7.6 DELETED; Refer to note on page ................................ 3/4 7-19 3/4.7.7 DELETED; Refer to note on page ................................ 3/4 7-19 3/4.7.8 MAIN TURBINE BYPASS SYSTEM .................................... 3/4 7-33 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. Sources - Operating ................................... 3/4 8-1 Table 4.8.1.1.2-1 DELETED .............................. 3/4 8-8 A.C. Sources - Shutdown .................................... 3/4 8-9 3/4.8.2 D.C. SOURCES D.C. Sources - Operating ................................... 3/4 8-10 LIMERICK - UNIT I xiv Amendment No. -34, 4-, 5, 4-g4,189
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| ELECTRICAL POWER SYSTEMS (Continued)
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| Table 4.8.2. Battery Surveillance Requirements ........................ 3/4 8-13 J.C. Sources Shutdown .................................... 3/4 8 14 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS Distribution Operating ................................... 3/4 8-15 Distribution Shutdown .................................... 3/4 8 18 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES (Deleted) .................................................. 3/4 8 21 (Deleted) .................................................. 3/4 8-27 Reactor Protecti on System El ectri c Power Moni tori ng ........ 3/4 8 28 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH ........................................ 3/4 9 3/4.9.2 INSTRUMENTATION ............................................ 3/4 9 3 3/4.9.3 CONTROL ROD POSITION ....................................... 3/4 9-5 3/4.9.4 DECAY TIME ................................................. 3/4 9-6 3/4.9.5 COMMUNICATIONS ............................................. 3/4 9 7 3/4.9.6 REFUELING PLATFORM ......................................... 3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL ..................... 3/4 9 10 3/4.9.8 WATER LEVEL REACTOR VESSEL ............................... 3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL ...................... 3/4 9 12 L:MERICK UNIT 1 xv Amendment No. YJ, 209
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
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| > SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal ............................ 3/4 9-13 Multiple Control Rod Removal. 3/4 9-15 3/4.9.11 - RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level . ............ ... .. ................. 3/4 9-17 Low Water Level .... .. .3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY .. 3/4 10-1 3/4.10.2 ROD WORTH MINIMIZER .. 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRANS TN..3/4 10-3 3/4.10.4 RECIRCULATION LOOPS. . . ..... 3/4 10-4 3/4.10.5 OXYGEN CONCENTRATION .. 3/4 10-5 3/4.10.6 TRAINING STARTUPS A R T U PS................
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| ....... 3/4 10-6 3/4.10.7 RESERVED- CURRENTLY NOT-USED.;..-3/4-10-7 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING . .3/4 10-8 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS The information from pages 3/4 11-1 through 3/4 11-6 has been intentionally omitted. Refer to note on page 3/4 11-1.............; 3/4 11-1
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| ----Liquid Holdup Tanks .. ;..-;- -- 3/411-7 .. . ..
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| 3/4.11.2 GASEOUS EFFLUENTS ..
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| The information from pages 3/4 11-8 through 3/4 11-14 has been intentionally omitted. Refer to note on page 3/4 11-8 .............. 3/4 11-8 LIMERICK - UNIT 1 xvi *Amendment No. r7, 33x ;g9 133 JAN 1 2 1999
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| INDEX LIMlll.liG .C.ONDIIIONS_f.OR.D£ERAIION 1ilill. 5UR1!£1.LLANC£ R£.QUIR.Ei1Etl1S.
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| SECTION RADIOACTIVE EFFLUENTS (Continued)
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| The information from page 3/4 11-15 has been intentionally omitted.
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| Refer to note page 3/4 11-15 ............................... 3/4 11-15 Main Condenser ............................................. 3/4 11-16 The information on page 3/4 11-17 has been intentionally omitted. Refer to note on this page .................................................. 3/4 11-17 3/4.11.3 (Deleted) The information on pages 3/4 11-18 through 3/4 11-20 has been intentionally omitted. Refer to note on page 3/4 11-18.
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| 3/4.11.4 (Deleted) .................................................. 3/4 11-18 3/4.12 (Deleted) The information on pages 3/4 12-1 through 3/4 12-14 has been intentionally omitted.
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| Refer to note on page 3/4 12-1 ................... 3/4 12-1 LIMERICK - UNIT 1 xvii Amendment No. ~. 228
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| BASES SECTION 3/4.0 APPLICABILITY .................................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN ............................................. B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES ........................................ B 3/4 1-1 3/4.1.3 CONTROL RODS ................................................ B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTROLS ................................ B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM ............................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 2-1 3/4.2.2 (DELETED) ................................................... B 3/4 2-2 LEFT INTENTIONALLY BLANK ............................................... B 3/4 2-3 3/4.2.3 MINIMUM CRITICAL POWER RATIO ................................ B 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE ................................. B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ................... B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION ......................... B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ............................................. B 3/4 3-2 3/4.3.3.A REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC) INSTRUMENTATION ............................... B 3/4 3-2a 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ........... B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION ............................................. B 3/4 3-4 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION ........................... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation ..................... B 3/4 3-5 LIMERICK - UNIT 1 xviii Amendment No.+, JJ, ~. ~. 227
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| BASES SECTION INSTRUMENTATION (Continued)
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| (Deleted) .................................................. B 3/4 3-5 (Deleted) .................................................. B 3/4 3-5 Remote Shutdown System Instrumentation and Controls ........ B 3/4 3-5 Accident Monitoring Instrumentation ........................ B 3/4 3-5 Source Range Monitors ...................................... B 3/4 3-5 (Deleted) ................*................................. 8 3/4 3-6 Chlorine and Toxic Gas Detection Systems ................... B 3/4 3-6 (Deleted) .................................................. B 3/4 3-6 (Deleted) .................................................. B 3/4 3-7 (Deleted) .................................................. B 3/4 3- 7 (Deleted) .................................................. B 3/4 3-7 3/4.3.8 (Deleted) .................................................. B 3/4 3-7 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION ............................................ B 3/4 3- 7 Bases Figure B 3/4.3-1 Reactor Vessel Water Level ....................... B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM ....................................... B 3/4 4-1 3/4.4.2 SAFETY/RELIEF VALVES ....................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems .................................. B 3/4 4-3 Operational Leakage ........................................ B 3/4 4-3 3/4.4.4 CHEMISTRY .................................................. B 3/4 4-3 LIMERICK - UNIT 1 xix Amendment No. ,4.g, w, W, A,
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| -l-04 , ++7- , 1-J..l. , -l-lM, 228
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| BAE SECTION REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY ............................................ B 3/4 4-4 3/4.4.6 PRESSURE/TEMPERATURE LIMITS .................................. B 3/4 4-4 Bases Table B 3/4.4.6-1 Reactor Vessel Toughness ..................... B 3/4 4-7 Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) At 1/4 T As A Function of Service Life .......................... B 3/4 4-8 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES ............................. B 3/4 4-6 3/4.4.8 (DELETED) .................................................... B 3/4 4-6 3/4.4.9 RESIDUAL HEAT REMOVAL ........................................ B 3/4 4-6 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/ 4 . 5 . 1 EC CS - 0PERA TI NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 5 -1 3/4 5.2 REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC) ........................................... B 3/4 5-3a 3/4.5.3 SUPPRESSION CHAMBER ..................................... B 3/4 5-4 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity ........................... B 3/4 6-1 Primary Containment Leakage ............................. B 3/4 6-1 Primary Containment Air Lock ............................ B 3/4 6-1 MSIV Leakage Control System ............................. B 3/4 6-1 Primary Containment Structural Integrity ................ B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure .............................................. B 3/4 6-2 Drywel l Average Air Temperature ......................... B 3/4 6-2 Drywel l and Suppression Chamber Purge System ............ B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS ................................ B 3/4 6-3 LIMERICK - UNIT 1 xx Amendment No. ;n, 9-9-,
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| Associated with Amendment 216 227
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| BASES SECT ION ................................................................. PAGE CONTAINMENT SYSTEMS (Continued) 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES ................ B 3/4 6-4 3/4.6.4 VACUUM RELIEF ........................................ B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT .............................. B 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL .............. B 3/4 6-6 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS - COMMON SYSTEMS .............. B 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -
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| COMMON SYSTEM ........................................ B 3/4 7-1 a 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM ................ B 3/4 7-1 cl 3/4.7.4 SNUBBERS ............................................. B 3/4 7-2 3/4.7.5 SEALED SOURCE CONTAMINATION ......................... B 3/4 7-3 3/4.7.6 (Deleted) .................................. ......... B 3/4 7-4 3/4.7.7 (Deleted) ........................................ B 3/4 7-4 3/4.7.8 MAIN TURBINE BYPASS SYSTEM ........................... B 3/4 7-5 3/4.8 ELECTRI CAL POWER SYSTEM 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION SYSTEMS ................................. B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES .............. B 3/4 8-3
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| .3/4.9 REFUELI NG OPERATIONS 3/4.9.1 REACTOR MODE SWITCH .................................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION .............. ........................ B 3/4 9-1 3/4.9.3 CONTROL ROD POSITION ................................. B 3/4 9-1 3/4.9.4 DECAY TIME ........................................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS ............ ....................... B 3/4 9-1
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| .LIMERICK - UNITI1 xxi Amendment No.' 2-, 4.9, 4, 4-94,188
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| INDEX
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| ,BASES U~SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING PLATFORM............................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL........... B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL........ B 3/4 9-2 3/4.9.10 CONTROL ROD REMOVAL............................... B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.... B 3/4 9-2 3/4.10 SPECIA 1 TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY............. B 3/410-1 3/4.10.2 ROD WORTH MINIMIZER.............................. B 3/410-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS................... B 3/410-1 3/4.10.4 RECIRCULATION LOOPS.............................. B 3/410-1 3/4.10.5 OXYGEN CONCENTRATION............................. B 3/4 10-1 3/4.10.6 TRAINING STARTUPS................................ B 3/410-1 3/4.10.7 RESERVED - CURRENTLY NOT USED..... ............ B 3/410-1 I
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| -3/4.10.8 --INSERVICE LEAK AND HYDROSTATIC TESTING.-.-......-.-.-. -B3/4 10-2 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS The information on page B 3/4 11-1 has been intentionally omitted. Refer to note on this page. B 3/4 11-1 (Deleted) ........................ B 3/4 11-2 Liquid Holdup Tanks............................... B 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS tDel eted)D--^--~@>-- . . .-.-. .. - .-
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| . . .'. ;. B 3/4 11-2 The information on page B 3/4 11-3 has been intentionally omitted. Refer to note on this page. B 3/4 11-3 (Deleted)........................................ B 3/4 11-4 LIMERICK - UNIT 1 xxii Amendment No. 4T1, 46, 133 JAN 1 2 1999
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| SECTION RADIOACTIVE EFFLUENTS (Continued)
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| (Deleted) The information from page B 3/4 11-4 has been intentionally omitted.
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| Refer to note on page B 3/4 11-4 ............. B 3/4 11-4 Main Condenser .......................................... B 3/4 11-5 (Deleted) ............................................... B 3/4 11-5 3/4.11.3 (Deleted) ............................................... B 3/4 11-5 3/4.11.4 (Deleted) ............................................... B 3/4 11-5 3/4.12 (Deleted) The information from pages B 3/4 12*1 through B 3/4 12-2 has been inten-tionally omitted. Refer to note on page B 3/4 12-1 .............................. B 3/4 12-1 LIMERICK* UNIT 1 xx iii Amendment No. 48, 22~
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| INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area .. 5-1 Figure 5.1.1-1 Exclusion Area .5-2 Low Population Zone .. 5-1 Figure 5.1.2-1 Low Population Zone .5-3 Maps Defining UNRESTRICTED AREAS and SITE BOUNDARY for Radioactive Gaseous and Liquid Effluents .5-1 Figure 5.1.3-la Map Defining UNRESTRICTED AREAS for Radioactive Gaseous and Liquid Effluents ................ 5-4 Figure 5.1.3-lb Map Defining UNRESTRICTED AREAS for Radioactive Gaseous and Liquid Effluents .5-5 (Deleted)..................;.................... 5-1 The figure on page 5-6 has been intentionally omitted. Refer to note on page 5-6 .5-6 5.2 CONTAINMENT Configuration................................................... 5-1 Design Temperature and Pressure .5-1 Secondary Containment .5-7 5.3 REACTOR CORE Fuel Assemblies .5-7 Control Rod Assemblies .5-7 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature .5-7 Volume......................................................... 5-8 5.5 FUEL STORAGE Criticality .5-8 LIMERICK - UNIT 1 xxiv Amendment No. 48
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| .aL uLt, .' 99
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| INDEX K-) DESIGN FEATURES SECTION PAGE FUEL STORAGE (Continued)
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| Drainage..................................................... 5-8 Capacity..................................................... 5-8 5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT.......................... 5-8 Table 5.6.1-1 Component Cyclic or Transient Limits..... 5-9 K>
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| LIMERICK - UNIT 1 xxv
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| INDEX ADMINISRBATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY ............................................. 6-1 6.12 ORGANIZATION ............................................. 6-1 6.2.1 OFFSITE AND ONSITE ORGANIZATION ........... ...................... 6-1 Figure 6.2.1-1 Deleted ........................... 6-3 6.2.2 UNIT STAFF ................... .................... 6-2 Figure 6.2.2-1 Deleted .6-4 Table 6.2.2-1 Minimum Shift Crew Composition .6-5 6.2.3 DELETED; Refer to note on page ................................ 6-6 6.2.4 SHIFT TECHNICAL ADVISOR ............. ................... 6-6 6.3 UNIT STAFF QUALIFICATIONS ............... ................. 6-6 6.4 DLETED .................................6-7 6.5 DELETED 6.5.1 Deleted Deleted .... 6-7 Deleted................................................... 6-7 Deleted................................................... 6-7 Deleted................................................... 6-7 Deleted .... 6-7 Deleted................................................... 6-8 Deleted................................................... 6-9 ALMRC NT1xv mnmn J.23 ,9,4~ 7 e.,
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| INDEX ADMINISTRATIIVE CONTROLS SECTION PAGE 6.5.2- Deleted Deleted ................ ................................... 6-9 Deleted ................................................... .. 6-9 Deleted ................................................ 6-10 Deleted .. ".............................................. 6-10 Deleted .................................................... 6-10 Deleted ................................................ 6-10 Deleted ................................................ 6-10 Deleted .................................................... 6-12 6 .5.3 Deleted .................. .................................. 6-12 6.6 REPORTABLE EVENT ACTION ............................................ 6-12a 6.7 SAFETY LIMIT VIOLATION ......................................... 6-12a 6.8 PROCEDURES AND PROGRAMS ..................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ............................................ 6-15 Startup Report .............................................. 6-15 Annual Reports .......................................... 6-15 Monthly Operating Reports .................................. 6-16 Annual Radiological Environmental Operating Report ......... 6-16 Annual Radioactive Effluent Release Report ................. 6-17 CORE OPERATING LIMITS REPORTS ............................ 6-18a 6.9.2 SPECIAL REPORTS ......................................... 6-18a 6 .10 DELETED ........................................................... 6-19 6.11 RADIATION PROTECTION PROGRAM .. ................................... 6-20 6.12 HIGH RADIATION AREA ................................................ 6-20 LIMERICK - UNIT 1 xxvi i Amendment No. -;, 4-;, 44, 4-7-, 192
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| ADMINISTRATIVE CONTROLS SECTION PAGE 6.13 PROCESS CONTROL PROGRAM (PCP) .................................... 6-21 6.14 OFFSITE DOSE CALCULATION MANUAL (QDCMl ........................... 6-22 6.15 (Deleted) ........................................................ 6-22 6.16 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM ....................... 6-22 6.17 SAFETY FUNCTION DETERMINATION PROGRAM (SFDP) ..................... 6-23 LIMERICK - UNIT 1 xxviii Amendment No . .4g, -+/-gg, 219
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| SECTION 1.0 DEFINITIONS
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| 1.0 DEFINITIONS
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| -> tThe following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.
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| ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.
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| AVERAGE PLANAR EXPOSURE 1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
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| AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
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| CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or
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| -r trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
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| CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possi~ble, comparison of the channel indication and/or status.with other indications and/or status derived from independent instrument channels measuring the same parameter.
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| CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
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| : a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.
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| : b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
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| The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.
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| LIMERICK - UNIT 1 1-1
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| DEFI IT CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:
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| a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell.
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| Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
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| CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.1.12. Plant operation within these limits is addressed in individual specifications.
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| CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
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| DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same inhalation committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The inhalation committed effective dose equivalent (CEDE) conversion factors used for this calculation shall be those listed in Table 2.1 of Federal Guidelines Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," ORNL, 1989, as described in Regulatory Guide 1.183. The factors in the column headed "effective" yield doses corresponding to the CEDE.
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| DOWNSCALE TRIP SETPOINT (DTSPl 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.
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| DRAIN TIME 1.9b The DRAIN TIME is the time it would take for the water inventory in and above the Reactor Pressure Vessel (RPV) to drain to the top of the active fuel (TAF) seated in the RPV assuming:
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| a) The water inventory above the TAF is divided by the limiting drain rate; bl The limiting drain rate is the larger of the drain rate through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths LIMERICK - UNIT 1 1-2 Amendment No. J+, -i-&, &-7-, +74, +&§., 227
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| DEFI IT DRAIN TIME (Continued) susceptible to a common mode failure (e.g., seismic event, loss of normal power, single human error), for all penetration flow paths below the TAF except:
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| : 1. Penetration flow paths connected to an intact closed system, or isolated by manual or automatic valves that are locked, sealed, or otherwise secured in the closed position, blank flanges, or other devices that prevent flow of reactor coolant through the penetration flow paths;
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| : 2. Penetration flow paths capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation; or
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| : 3. Penetration flow paths with isolation devices that can be closed prior to the RPV water level being equal to the TAF by a dedicated operator trained in the task, who is in continuous communication with the control room, is stationed at the controls, and is capable of closing the penetration flow path isolation device without offsite power.
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| c) The penetration flow paths required to be evaluated per paragraph b) are assumed to open instantaneously and are not subsequently isolated, and no water is assumed to be subsequently added to the RPV water inventory; d) No additional draining events occur; and e) Realistic cross-sectional areas and drain rates are used.
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| A bounding DRAIN TIME may be used in lieu of a calculated value.
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| 1.10 (Deleted)
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| EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
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| LIMERICK - UNIT 1 l-2a Amendment No. :J+, ~. &+, +/-+4, ~ . 227
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| DEFINITIONS END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.12 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:
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| : a. Turbine stop valves, and
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| : b. Turbine control valves.
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| This total system response time consists of two components, the instrumen-tation response time and the breaker arc suppression time. These times may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
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| 1.13 (Deleted) 1.14 (Deleted)
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| FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
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| HIGH (POWER) TRIP SETPOINT CHTSP) l.15a The high power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable above 85% reactor thermal power.
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| IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:
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| : a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or
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| : b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.
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| INSERVICE TESTING PROGRAM l.16a The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).
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| INTERMEDIATE (POWER) TRIP SETPOINT (ITSP) l.16b The intermediate power trip setpoint associated with the Rod Block Monitor (RBM)I rod block trip setting applicable between 65% and 85% reactor thermal power.
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| ISOLATION SYSTEM RESPONSE TIME 1.17 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.
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| Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
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| LIMITING CONTROL ROD PATTERN 1.18 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, OR MCPR.
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| LINEAR HEAT GENERATION RATE 1.19 LINEAR HEAT GENERATION RATE CLHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.
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| LIMERICK - UNIT 1 1-3 Amendment No. ee, 225
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| DEFINITIONS LOGIC SYSTEM FUNCTIONAL TEST 1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through.and including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.
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| LOW (POWER) TRIP SETPOINT (LTSP) 1.20a The low power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable between 30% and 65% reactor thermal power.
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| 1.21 (Deleted)
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| MEMBER(S) OF THE PUBLIC 1.22 MEMBER OF THE PUBLIC means any indiyidual except when that individual is receiving an occupational dose.
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| MAPFAC(F)-(MAPLHGR FLOW FACTOR) 1.22a A core flow dependent multiplication factor used to flow bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.
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| MAPFAC(P)-(POWER DEPENDENT MAPLHGR MULTIPLIER) 1.22b A'core power dependent multiplication factor used to power bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.
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| MINIMUM CRITICAL POWER RATIO (MCPR) 1.23 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core (for each class of fuel). Associated with the minimum critical power ratio is a core flow dependent (MCPR(F)) and'core power dependent (MCPR(P)) minimum critical power ratio.
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| OFFSITE DOSE CALCULATION MANUAL 1.24 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.
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| OPERABLE - OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s) and when all necessary attendant instrumentation, controls, 'electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).
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| LIMERICK - UNIT I 1-4 Amendment No. 48, 46, ;-9,187
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| :DEFINITIONS
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| ',OPERATIONAL CONDITION --CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant tempera-ture as specified in Table 1.2.
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| 'PHYSICS TESTS
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| '1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1)described in Chapter .14 of the FSAR, (2)authorized under the provisions of 10 CFR 50.59, or (3)otherwise approved by the Commission.
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| PRESSURE BOUNDARY LEAKAGE 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall.
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| PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:
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| : a. All primary containment penetrations required to be closed during accident conditions are either:
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| : 1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or
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| : 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, K> except for valves that are opened under administrative control as permitted by Specification 3.6.3.
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| : b. All primary containment equipment hatches are closed and sealed.
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| : c. The primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.
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| : d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.
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| : e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
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| : f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or O-rings, is OPERABLE.
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| PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the solidification or dewatering and packaging of radioactive wastes results in a waste package with properties that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for transportation to the disposal site and receipt at the disposal site.
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| With solidification or dewatering, the PCP shall identify the process parameters influencing solidification or dewatering, based on laboratory scale and full scale testing or experience.
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| Amendment No. 48, 66, 146 I TMFnTr Llr1Lnlr%
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| 11TT Uil I 1
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| I 1
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| *1-D' OCT 1 8 2000
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| PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
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| RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3515 MWt.
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| REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:
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| a All reactor enclosure secondary containment penetrations required tobe closed during accident conditions are either:
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| : 1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or
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| : 2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.
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| h All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.
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| : c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
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| : d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.
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| : e. At least one door in each access to the reactor enclosure secondary containment is closed, except when the access opening is being used for entry and exit.
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| £ The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
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| : g. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.la, except as indicated by the footnote for Specification 4.6.5.1.la.
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| REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
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| RECENTLY IRRADIATED FUEL 1.35 RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours.
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| REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:
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| : a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:
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| LIMERICK - UNIT 1 1-6 Amendment No. JJ.~.~.~.~.~.-2-2-0.
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| 229
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| REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)
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| : 1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or
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| : 2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.2.
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| All refueling floor secondary containment hatches and blowout panels are closed and sealed.
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| The standby gas treatment system is in compliance with the requirements of specification 3.6.5.3.
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| d At least one door in each access to the refueling floor secondary containment is closed, except when the access opening is being used for entry and exit.
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| The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
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| £ The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.l.2a, except as indicated by the footnote for Specification 4.6.5.l.2a.
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| REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
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| RESTRICTED AREA 1.37a RESTRICTED AREA means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. RESTRICTED AREA does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a RESTRICTED AREA.
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| 1.38 (Deleted)
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| SHUTDOWN MARGIN (SOM) 1.39 SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:
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| : a. The reactor is xenon free;
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| : b. The moderator temperature is~ 68°F, corresponding to the most reactive state; and
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| : c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.
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| SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-la.
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| SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
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| LIMERICK - UNIT 1 1-7 Amendment No. 48,66,105,185,187, 207,215,220,229
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| DEFINITIONS STAGGERED TEST BASIS 1.42 A STAGGERED TEST BASIS shall consist of:
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| : a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.
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| : b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
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| THERMAL POWER 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
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| TURBINE BYPASS SYSTEM RESPONSE TIME 1.43A The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required position.
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| The response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured.
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| UNIDENTIFIED LEAKAGE 1.44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
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| UNRESTRICTED AREA 1.45 UNRESTRICTED AREA means an area, access to which is neither limited nor controlled by the licensee.
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| VENTILATION EXHAUST TREATMENT SYSTEM 1.46 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
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| VENTING 1.47 VENTINGshall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or
| |
| ,other operating condition,-in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
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| LIMERICK - UNIT 1 1-8 Amendment No. -521,187
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| | |
| DEFINITIONS TABLE-1.1 SURVEILLANCE FREOVENCY NOTATION
| |
| .NOTATION FREQUENCY S At least once per 12 hours.
| |
| 0 At least once per 24 hours.
| |
| ; De At least once per 7 days.
| |
| M At leastonceper 31 days.
| |
| Q At least once per 92 days.
| |
| SA At least once per. 184 days.-
| |
| A At least once per 366 days.
| |
| E ' At least once per 18 months (550 days).
| |
| R (Refueling Interval) At least once per 24 months (731 days).
| |
| S/U ; Prior to each reactor startup.
| |
| P Prior to each radioactive release.
| |
| N.A. Not applicable.-
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| LIMERICK - UNIT 1 1-9 Amendment No. 71 i
| |
| w4-'),bLL 67,iq
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| | |
| DEFINITIONS TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE
| |
| : 1. POWER OPERATION Run Any temperature
| |
| : 2. STARTUP Startup/Hot Standby Any temperature
| |
| : 3. HOT SHUTDOWN Shutdown# *** > 2000 F
| |
| : 4. COLD SHUTDOWN Shutdown# ## ***
| |
| * 200 0 F ****
| |
| : 5. REFUELING* Shutdown or Refuel** # NA
| |
| #The reactor mode switch may be placed in the Run or Startup/Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.
| |
| ##The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.
| |
| *Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
| |
| **See Special Test Exceptions 3.10.1 and 3.10.3.
| |
| ***The reactor mode switch may be placed in the Refuel position while a single control rod is being moved provided that the one-rod-out interlock is I OPERABLE.
| |
| ****See Special Test Exception 3.10.8.
| |
| LIMERICK - UNIT 1 1-10 Amendment No. 88, -i-es, 149 APR o0 2001
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| | |
| SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
| |
| | |
| 2.0 SAFETY LIMITS AND l IMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 700 psia or core flow less than 10% of rated flow.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 700 psia or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1.
| |
| THERMAL POWER. High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.10 for two recirculation loop operation and shall not be less than 1.14 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 700 psia and core flow greater than 10% of rated flow.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| With MCPR less than 1.10 for two recirculation loop operation or less than 1.14 for single recirculation loop operation and the reactor vessel steam dome pressure greater than 700 psia and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1.
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| REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.
| |
| ACTION:
| |
| With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with the reactor coolant system pressure less than or equal to 1325 psig within 2 hours and comply with the requirements of Specification 6.7.1.
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| LIMERICK - UNIT 1 2-1 Amendment No. +,J.G,-+/--1-+/-,~.~.
| |
| ++-O,MJ,~.-2-cl.
| |
| 222
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| | |
| SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued)
| |
| REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.
| |
| APPLICABILITY: OPERATIONALCONDITIONS 3, 4, and 5 ACTION:
| |
| With the reactor vessel water level at or below the top of-the active irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required. Comply with the requirements of Specification 6.7.1.
| |
| LIMERICK - UN I2 2-2
| |
| | |
| SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.
| |
| APPLICABILITY: As shown in Table 3.3.1-1.
| |
| ACTION:
| |
| With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable* and apply the applicable ACTION statement requirement I of:Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.
| |
| *The APRM Simulated Thermal Power - Upscale Functional Unit need not be declared inoperable upon entering single reactor recirculation loop operation provided that the flow-biased setpoints are adjusted within 6 hours per Specification 3.4.1.1.
| |
| LIMERICK - UNIT 1 2-3 Amendment No. 141 APR 1 2 2000
| |
| | |
| TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES
| |
| : 1. Intermediate Range Monitor, Neutron Flux-High ~ 120/125 divisions ~ 122/125 divisions of full scale of full scal e
| |
| : 2. Average Power Range Monitor:
| |
| : a. Neutron Flux Upscale (Setdown) ~ 15.0% of RATED THERMAL POWER 5 20.0% of RATED THERMAL POWER
| |
| : b. Simulated Thermal Power - upscale:
| |
| Two Recirculation Loop Operation ~ 0.65 W+ 61.7% and ~ 0.65 W+ 62.2% and 5 116.6% of RATED 5 117.0% of RATED THERMAL POWER THERMAL POWER
| |
| - Sinqle Recirculation Loop Operation*** ~ 0.65 (W-7.6%) + 61.5% and 5 0.65 CW-7.6%) + 62.0% and 5 116.6% of RATED 5 117.0% of RATED THERMAL POWER THERMAL POWER
| |
| : c. Neutron Flux - Upscale 118.3% of RATED 118.7% of RATED THERMAL POWER THERMAL POWER
| |
| : d. In 0 pe rat i ve N.A. N.A.
| |
| : e. 2-0ut-Of-4 Voter N.A. N.A.
| |
| : f. OPRM Upscale **** N.A.
| |
| : 3. Reactor Vessel Steam Dome Pressure - Hi 5 1096 pSlg 5 1103 pSig
| |
| : 4. Reactor Vessel Water Level - Low, Level ~ 12.5 inches above instrument ~ 11.0 inches above zero* instrument zero
| |
| : 5. Main Steam Line Isolation Valve - Closure 5 8% closed 5 12% closed
| |
| : 6. DELETED DELETED DELETED
| |
| : 7. Drywell Pressure - High ~ 1. 68 psi g 5 1.88 psi g
| |
| : 8. Scram Di scharge Vol ume Water Level - Hi
| |
| : a. Level Transmitter 5 260' 9 5/8" elevation** ~ 261' 5 5/8" el evati on
| |
| : b. Float Switch ~ 260' 9 5/8" elevation** ~ 261' 5 5/8" elevation
| |
| : 9. Turbine Stop Valve - Closure ~ 5% closed ~7% closed
| |
| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low ~ 500 pSlg ~ 465 psig
| |
| : 11. Reactor Mode Switch Shutdown Position N.A. N.A.
| |
| : 12. Manual Scram N.A. N.A.
| |
| * See Bases Figure B 3/4.3-1.
| |
| ** Equivalent to 25.45 gallons/scram discharge volume.
| |
| *** The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W~ 7.6%. For flows W< 7.6%, the (W-7.6%) term is set equal to zero.
| |
| **** See COLR for OPRM od based detection algorithm trip setpoints. OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power ~ 29.5% and recirculation drive flow < 60%.
| |
| LIMERI - UN IT 1 2-4 Amendment No. +,JQ,~,~,~,+4+/-,~.
| |
| 201
| |
| | |
| SECTIONS 3.0 and 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
| |
| | |
| 3/4.0 APPLICABILITY LIMITINk.-CONDlIIQN___ EQR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met, except as provided in Specifications 3.0.S and 3.0.6.
| |
| 3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals, except as provided in Specifications 3.0.5 and 3.0.6. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
| |
| 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the Specification does not apply by placing it, as applicable, in:
| |
| : a. At least STARTUP within the next 6 hours,
| |
| : b. At least HOT SHUTDOWN within the following 6 hours, and
| |
| : c. At least COLD SHUTDOWN within the subsequent 24 hours.
| |
| Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Speci fi cations .
| |
| This Specification is not applicable in OPERATIONAL CONDITION 4 or 5.
| |
| 3.0.4 When a Limiting Condition for Operation is not met, entry into an OPERATIONAL CONDITION or other specified condition in the Applicability shall only be made:
| |
| : a. When the associated ACTION requirements to be entered permit continued operation in the OPERATIONAL CONDITION or other specified condition in the Applicability for an unlimited period of time; or
| |
| : b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the OPERATIONAL CONDITION or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or
| |
| : c. When an allowance is stated in the individual value, parameter, or other Specification.
| |
| This Specification shall not prevent changes in OPERATIONAL CONDITIONS or other specified conditions in the Applicability that are required to comply with ACTION requirements or that are part of a shutdown of the unit.
| |
| LIMERICK - UNIT 1 3/4 0-1 Amendment No. li, 169, :l.-9, 226
| |
| | |
| 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION (Continued) 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONs may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to the second premise of Specification 3.0.1 and is an exception to Specification 3.0.2 (i.e., to not comply with the applicable ACTION(s)) for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
| |
| 3.0.6 When a supported system Limiting Condition for Operation is not met solely due to a support system Limiting Condition for Operation not being met, the ACTIONS associated with this supported system are not required to be entered. Only the support system Limiting Condition for Operation ACTIONs are required to be entered. This is an exception to the second premise of Specification 3.0.l and is an exception to Specification 3.0.2 (i.e., to not comply with the applicable ACTION(s)) for the supported system. In this event, an evaluation shall be performed in accordance with Specification 6.17, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate ACTIONs of the Limiting Condition for Operation in which the loss of safety function exists are required to be entered.
| |
| When a support system's ACTION directs a supported system to be declared inoperable or directs entry into ACTIONs for a supported system, the applicable ACTIONS shall be entered in accordance with Specification 3.0.1.
| |
| LIMERICK - UNIT 1 314 0-la Amendment No. 219 I
| |
| | |
| APPLICABILITY SURYEILLANCE REQUIREMENTS_
| |
| 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other specified conditions in the Applicability for individual Limiting Conditions for Operation, unless otherwise stated in the Surveillance Requirement. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the Limiting Condition for Operation.
| |
| Failure to perform a Surveillance within the specified Surveillance time interval and allowed extension per Specification 4.0.2, shall be failure to meet the Limiting Condition for Operation except as provided in Specification 4.0.3.
| |
| Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
| |
| 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance time interval with a maximum allowable extension not to exceed 25%
| |
| of the surveillance interval.
| |
| 4.0.3 If it is discovered that a Surveillance was not performed within its specified Surveillance time interval and allowed extension per Specification 4.0.2, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Surveillance time interval, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.
| |
| If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION requirements must be entered.
| |
| When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION requirements must be entered.
| |
| 4.0.4 Entry into an OPERATIONAL CONDITION or other specified condition in the Applicability of a Limiting Condition for Operation shall only be made when the Limiting Condition for Operation's Surveillance Requirements have been met within their Surveillance time interval, except as provided in Specification 4.0.3. When a Limiting Condition for Operation is not met due to its Surveillance Requirements not having been met, entry into an OPERATIONAL CONDITION or other specified condition in the Applicability shall only be made in accordance with Specification 3.0.4.
| |
| This provision shall not prevent entry into OPERATIONAL CONDITIONS or other specified conditions in the Applicability that are required to comply with ACTION requirements or that are part of a shutdown of the unit.
| |
| 4.0.5 Inseryice Inspection and Inservice Testing Program The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.S5a(f). The Inservice Inspection Program is the licensee program that fulfills the requirements of 10 CFR 50.55a(g).
| |
| The provisions of SR 4.0.2 and SR 4.0.3 do not apply to the INSERVICE TESTING PROGRAM unless there is a specific SR referencing usage of the program.
| |
| LIMERICK - UNIT 1 3/4 0-2 Amendment No. :H::,3-8, 1:r5,16r,1:U9, 1:94, 2-5,226
| |
| | |
| APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)
| |
| LIMERICK - UNIT 1 3/4 0-3 Amendment No. l:l:,49, ~.3:79:, 194,225
| |
| | |
| 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION.
| |
| 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:
| |
| : a. 0.38% Ak/k with the highest worth rod analytically determined, or
| |
| : b. 0.28% Mk/k with the highest worth rod determined by test.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4,--and 5.
| |
| ACTION:
| |
| With the SHUTDOWN MARGIN less than specified:
| |
| : a. In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| b.- In OPERATIONAL CONDITION .3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that could reduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4, establish SECONDARY CONTAINMENT INTEGRITY within 8 hours.
| |
| : c. In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS-and other activities that could reduce the SHUTDOWN MARGIN and insert all insertable control rods within 1 hour. Establish SECONDARY CONTAIN-MENT-INTEGRITY within 8 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than specified at any time during the fuel cycle:
| |
| : a. By measurement, prior to or during the first startup after each refueling.
| |
| : b. By measurement, within 500 MWD/T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.
| |
| : c. Within 12 hours after detection of a withdrawn control rod that is immovable, as a result of excessive friction or mechanical inter-ference, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod.
| |
| _<JKJ LIMERICK - UNIT 1 3/4 1-1
| |
| | |
| REACTIVITY CONTROL SYSTEMS 3/4,1.2 REACTIVITY ANOMALIES 3.1.2 The reactivity difference between the actual core keN and the predicted core Keff shall not exceed 1% ;'.k/k.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1 and 2.
| |
| fleT ION:
| |
| With the reactivity difference exceeding 1% ~~/k:
| |
| : a. Within 12 hours perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.
| |
| : b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| 4.1.2 The reactivity difference between the actual core k~f and the predicted core k~f shall be verified to be less than or equal to 1% ~k/k:
| |
| : a. During the first startup following CORE ALTERATIONS, and
| |
| : b. At least once per 31 effective full power days during POWER OPERATION.
| |
| : c. The provisions of Specification 4.0.4 are not applicable.
| |
| LIMERICK - UNIT 1 3/4 1-2 Amendment No. B, '2fJ7
| |
| | |
| REACTIVITY CONTROL SYSTEMS 3/4.1.3 CONTROL RODS CONTROL ROD OPERABILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods and scram discharge volume vent and drain valves shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3***
| |
| ACTION:
| |
| : a. With one withdrawn control rod inoperable due to being immovable, as a result of excessive friction or mechanical interference, or known to be untrippable:
| |
| : 1. Within 1 hour:
| |
| a) Verify that the inoperable withdrawn control rod is separated from all other inoperable withdrawn control rods by at least two control cells in all directions.
| |
| b) Disarm the associated directional control valves** either:
| |
| : 1) Electrically, or
| |
| : 2) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : 2. Restore the inoperable withdrawn control rod to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. With one or more control rods trippable but inoperable for causes other than addressed in ACTION a, above:
| |
| : 1. If the inoperable control rod(s) is withdrawn, within 1 hour:
| |
| a) Verify that the inoperable withdrawn control rod(s) is separated from all other inoperable withdrawn control rods by at least two control cells in all directions, and b) Demonstrate the insertion capability of the inoperable with-drawn control rod(s) by inserting the control rod(s) at least one notch by drive water pressure within the normal operating range*.
| |
| * The inoperable control rod may then be withdrawn to a position no further withdrawn than its position when found to be inoperable.
| |
| ** May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
| |
| *** OPERATIONAL CONDITION 3 is only applicable to the scram discharge volume vent and drain valves.
| |
| LIMERICK - UNIT I 3/4 1-3 Amendment No. 1-6B, 178
| |
| | |
| REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| Otherwise, insert the inoperable withdrawn control rod(s) and disarm the associated directional control valves** either:
| |
| a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| : 2. If the inoperable control rod(s) is inserted, within 1 hour disarm the associated directional control valves** either:
| |
| a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : c. With more than 8 control rods inoperable, be in at least HOT SHUTDOWN within 12 hours.
| |
| : d. With one or. more scram discharge volume (SDV) vent or drain lines with one valve inoperable, restore the inoperable valve(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours*** and in COLD SHUTDOWN within the following 24 hours.
| |
| : e. With one or more SDV vent or drain lines with both valves inoperable, isolate the associated line within 8 hours **** or be in at least HOT SHUTDOWN within the next 12 hours*** and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by: II
| |
| : a. Verifying each valve to be open,* and I
| |
| : b. Cycling each valve through at least one complete cycle of full travel.
| |
| * These valves may be closed ihtermittently for testing under administrative controls.
| |
| ** May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
| |
| * Separate Action entry is allowed for each SDV vent and drain line.
| |
| ****An isolated line may be unisolated under administrative control to allow draining and venting of the SDV.
| |
| LIMERICK - UNIT I 3/4 1-4 Amendment No. --
| |
| 7, 449, 9,186
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 4.1.3.1.2 When above the preset power level of the RWM, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:
| |
| : a. In accordance with the Surveillance Frequency Control Program, and
| |
| : b. Within 24 hours from discovery that a control rod is immovable as a result of excessive friction or mechanical interference.
| |
| 4.1.3.1.3 All control rods shall be demonstrated OPERABLE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.6, and 4.1.3.7.
| |
| 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by demonstrating:
| |
| : a. The scram discharge volume drain and vent valves OPERABLE in accordance with the Surveillance Frequency Control Program, by verifying that the drain and vent valves;
| |
| : 1. Close within 30 seconds after receipt of a signal for control rods to scram, and
| |
| : 2. Open when the scram signal is reset.
| |
| : b. Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT I 3/4 1-5 Amendment No. 4-0, .74, P8, 4-68, 4-7-8,186
| |
| /
| |
| | |
| REACTIVITY CONTROL SYSTEMS \
| |
| CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| With the maximum scram insertion time of one or more control rods exceeding 7 seconds:
| |
| : a. Declare the control rod(s) with the slow insertion time inoperable, and
| |
| : b. Perform the Surveillance Requirements of Specification 4.1.3.2c. at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.
| |
| Otherwise, be in at least HOT SHUTDOWN within 12 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:
| |
| : a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER with reactor coolant pressure greater than or equal to 950 psig, following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days.
| |
| : b. For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "1" or "2" as follows:
| |
| l.a Specifically affected individual control rods shall be scram time tested at zero reactor coolant pressure and the scram insertion time from the fully withdrawn position to notch position 05 shall not exceed 2.0 seconds, and 1.b Specifically affected individual control rods shall be scram time tested at greater than or equal to 950 psig reactor coolant pressure prior to exceeding 40% of RATED THERMAL POWER.
| |
| : 2. Specifically affected individual control rods shall be scram time tested at greater than or equal to 950 psig reactor coolant pressure.
| |
| : c. For at least 10% of the control rods, with reactor coolant pressure greater than or equal to 950 psig, on a rotating basis, and in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 1-6 Amendment No. 94, 4-9186
| |
| | |
| REACTIVITY CONTROL SYSTEMS CONTROL ROD AVERAGE SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.3 The average scram insertion time of all OPERABLE control rods from the fully withdrawn position, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:
| |
| Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 0.43 39 0.86 25 1.93 05 3.49 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| With the average scram insertion time exceeding any of the above limits, be in at least HOT SHUTDOWN within 12 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.1.3.3 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.
| |
| LIMERICK - UNIT 1 3/4 1-7
| |
| | |
| REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve sole-noids as time zero, shall not exceed any of the following:
| |
| Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 0.45 39 0.92 25 2.05 5 3.70 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.-
| |
| ACTION:
| |
| With the average scram insertion times of control rods exceeding the above limits:
| |
| : a. Declare the control rods with the slower than average scram insertion times inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group, and
| |
| : b. Perform the Surveillance Requirements of Specification 4.1.3.2c. at least once per 60 days when operation is continued with an average scram insertion time(s) in excess of the average scram insertion time limit.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| SURPVEILLIANCE-REUITREMNTIS .
| |
| 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.
| |
| LIMERICK - UNIT 1 3/4 1-8 Amendment No. 169
| |
| | |
| REACTIVITY CONTROL SYSTEMS
| |
| * CONTROL ROD SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*.
| |
| ACTION
| |
| : a. In OPERATIONAL CONDITION 1 or 2:
| |
| : 1. With one control rod scram accumulator inoperable, within 8 hours:
| |
| a) Restore the inoperable accumulator to OPERABLE status, or b) Declare the control rod associated with the inoperable accumulator inoperable.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : 2. With more than one control rod scram accumulator inoperable, declare the associated control rods inoperable and:
| |
| a) If the control rod associated with any inoperable scram accumulator is withdrawn, immediately verify that at least one control rod drive pump is operating by verifying that control rod charging water header pressure is i1400 psig or by inserting at least one withdrawn control rod at least one notch. If no control rod drive pump is operating and:
| |
| : 1) If reactor pressure is k9OO psig, than restart at least one control drive pump within 20 minutes or place the reactor mode switch in the shutdown position, or
| |
| : 2) If reactor pressure is <900 psig, then place the reactor mode switch in the Shutdown position.
| |
| b) Insert the inoperable control rods and disarm the associated control valves either:
| |
| : 1) Electrically, or
| |
| : 2) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| Otherwise, be in at least HOT SHUTDOWN within 12 hours.
| |
| : b. In OPERATIONAL CONDITION 5*:
| |
| : 1. With one withdrawn control rod with its associated scram accumulator inoperable, insert the affected control rod and disarm the associated directional control valves within one hour, either:
| |
| a) Electrically, or b Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| *At least the accumulator associated with each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| LIMERICK - UNIT 1 3/4 1-9 Amendment No. 39,143 MAY 222000
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS
| |
| : 2. With more than one withdrawn control rod with the associated scram accumulator inoperable or no control rod drive pump oper-ating, immediately place the reactor mode switch in the Shutdown position.
| |
| 4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying that the indicated pressure is greater than or equal to 955 psig unless the control rod is inserted and disarmed or scrammed.
| |
| L'IMERICK - UNIT 1 3/4 1-10 Amendment No. 3-1, 3,9;-;4, 4-43, 4-69, 186
| |
| | |
| REACTIVITY CONTROL-SYSTEMS CONTROL ROD DRIVE COUPLING LIMITING CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITIONS 1 and 2 with one control rod not coupled to its associated drive mechanism, within 2 hours:
| |
| : 1. If permitted by the RWM, insert the control rod drive mechanism to accomplish recoupling and verify recoupling by withdrawing the control rod, and:
| |
| a) Observing any indicated response of the nuclear instrumenta-tion, and b) Demonstrating that the control rod will not go to the over-travel position.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : 2. If recoupling is not accomplished on the first attempt.or, if not permitted by the RWM, then until permitted by the RWM, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves** either:
| |
| a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours either:
| |
| : 1. Insert the control rod to accomplish recoupling and verify recoup-ling by withdrawing the control rod and demonstrating that the control rod will not go to the overtravel position, or
| |
| : 2. If recoupling is not accomplished, insert the control rod and disarm the associated directionalcontrol valves** either:
| |
| a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| * At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| ** May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
| |
| LIMER T CK - UNIT 1 3/4 1.-11 Amendment No. -, 4, 14-64, 192
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.6 Each affected control rod shall be demonstrated to be coupled to its drive mechanism by observing any indicated response of the nuclear instrumen-tation while withdrawing the control rod to the fully withdrawn position and then verifying that the control rod drive does not go to the overtravel position:
| |
| : a. Prior to reactor criticality, after completing CORE ALTERATIONS that could have affected the control.rod drive coupling integrity,
| |
| : b. Anytime the control rod is withdrawn to the "Full out" position in subsequent operation, and
| |
| : c. Following maintenance on or modification to the control rod or control rod drive system which could haveaffected the control rod drive coupling integrity.
| |
| LIMERICK - UNIT 1 3/4 1-12 Amendment No. 4-2-4, 192
| |
| | |
| REACTIVITY CONTROL SYSTEMS CONTROL ROD POSITION INDICATION LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod position indication system shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 1 or 2 with one or more control rod position indicators inoperable, within 1 hour:
| |
| : 1. Determine the position of the control rod by using an alternate method, or:
| |
| a) Moving the control rod, by single notch movement, to a position with an OPERABLE position indicator, b) Returning the control rod, by single notch movement, to its original position, and c) Verifying no control rod drift alarm at least once per 12 hours, or
| |
| : 2. Move the control rod to a position with an OPERABLE position indicator, or
| |
| : 3. When THERMAL POWER is:
| |
| a) Within the preset power level of the RWM, declare the control rod inoperable.
| |
| b) Greater than the preset power level of the RWM, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves** either:
| |
| : 1) Electrically, or
| |
| : 2) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. In OPERATIONAL CONDITION 5* with a withdrawn control rod position indicator inoperable, move the control rod to a position with an OPERABLE position indicator or insert the control rod.
| |
| *At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| **May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
| |
| LIMERICK - UNIT 1 3/4 1-13 Amendment No. 4-7, 169
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod position indication system shall be determined OPERABLE by verifying:
| |
| : a. In accordance with the Surveillance Frequency Control Program that the position of each control rod is indicated,
| |
| : b. That the indicated control rod position changes during the movement of the control rod drive when performing Surveillance Requirement 4.1.3.1.2, and
| |
| : c. That the control rod position indicator corresponds to the control rod position indicated by the "Full out" position indicator when performing Surveillance Requirement 4.1.3.6b.
| |
| LIMERICK - UNIT 1 3/4 1-14 Amendment No. 186
| |
| | |
| REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE HOUSING SUPPORT LIMITING CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be *inplace.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
| |
| ACTION:
| |
| With the control rod drive housing support not in place, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within thefollowing 24 hours.
| |
| SURVEILLANCE REQUIREMENTS I 4.1.3.8 The control rod drive housing support'shall be verified to be in place by a visual inspection prior to startup any time it has been disassembled or when maintenance has'been performed in the control rod.drive housing support area.
| |
| I . . . I I .I . I .
| |
| LIMERICK - UNIT 1 3/4 1-15
| |
| | |
| REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS I<
| |
| ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The rod worth minimizer (RWM) shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*, **, when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER. I ACTION:
| |
| : a. With the RWM inoperable after the first 12 control rods are fully withdrawn, operation may continue provided that control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or technically qualified member of the unit technical staff.
| |
| : b. With the RWM inoperable before the first 12 control rods are fully withdrawn, one startup per calendar year may be performed provided that control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or technically qualified member of the unit technical staff.
| |
| : c. Otherwise, with the RWM inoperable, control rod movement shall not be permitted except by full scram.***
| |
| *See Special Test Exception 3.10.2.
| |
| **Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
| |
| ***Control rods may be moved, under administrative control, to permit testing associated with demonstrating OPERABILITY of the RWM.
| |
| LIMERICK - UNIT 1 3/4 1-16 Amendment No. XX,17 I MAR 2 2 1989
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.4.1 The'RWM shall be demonstrated OPERABLE:
| |
| : a. In OPERATIONAL CONDITION-2 within 8 hours prior to withdrawal of control rods for the purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 1 hour after RWM automatic initia-,
| |
| tion when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.
| |
| : b. InOPERATIONAL CONDITION 2 within 8 hours prior to withdrawal of control rods for the purpose'of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequencecontrol rod. --
| |
| : c. In OPERATIONAL CONDITION 1 within 1 hour after RWM automatic initiation when reducing THERMAL POWER, by. verifying the rod block function by demonstrating inability to withdraw an'out-of-sequence control rod.. -
| |
| d.- By verifying that the' controll-rod patterns and sequence input to the RWM computer-are correctly' loaded following any loading of the program into the computer.
| |
| 3.1.4.2 DELETED - -
| |
| 4.-1.4.2 -DELETED - : --
| |
| . ; .. . .. I . I .
| |
| LIMERICK - UNIT 1 3/4 1-17 Amendment No. 17 I MAR 2 2 1989
| |
| | |
| REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION 3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER with MCPR less than 1.70, or THERMAL POWER greater than or equal to 90% of rated with MCPR less than 1.40.
| |
| ACTION:
| |
| : a. With one RBM channel inoperable:
| |
| : 1. Verify that the reactor is not operating on a LIMITING CONTROL ROD PATTERN, and
| |
| : 2. Restore the inoperable RBM channel to OPERABLE status within 24 hours.
| |
| Otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour.
| |
| : b. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within 1 hour.
| |
| SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a:
| |
| : a. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION for the OPERATIONAL CONDITIONS specified in Table 4.3.6-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.6-1.
| |
| : b. CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when the reactor is operating on a LIMITING CONTROL ROD PATTERN.
| |
| LIMERICK - UNIT 1 3/4 1-18 Amendment No. 4g&, 186
| |
| | |
| REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM 3.1.5 The standby liquid control system shall be OPERABLE and consist of the following:
| |
| : a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,
| |
| : b. In OPERATIONAL CONDITION 3, a minimum of one pump and corresponding flow path.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 ACTION:
| |
| : a. With only one pump and corresponding explosive valve OPERABLE, in OPERATIONAL CONDITION 1 or 2, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying that:
| |
| : 1. The temperature of the sodium penta borate solution is within the limits of Figure 3.1.5-1.
| |
| : 2. The available volume of sodium pentaborate solution is at least 3160 gallons.
| |
| : 3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.
| |
| LIMERICK - UNIT 1 3/4 1 19 Amendment No.
| |
| | |
| REACTIVITY CONTROL SYSTEMS
| |
| : b. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Verifying the continuity the explosive charge.
| |
| : 2. Determining by chemical analysis and calculation* that the available weight of Boron 10 is greater than or equal to 185 lbs; the concentration of sodium pentaborate in solution is less than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied:
| |
| x x 0 ~ 1 13% wt. 29 atom % 86 gpm where C Sodium penta borate solution (% by weight)
| |
| Q Two pump flowrate, as determined per surveillance requirement 4.1.5.c.
| |
| E Boron 10 enrichment (atom % Boron 10)
| |
| : 3. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
| |
| : 4. Verifying that no more than two pumps are aligned for automatic operation.
| |
| : c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1230 +/- 25 psig is met.
| |
| : d. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Initiating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of the batch success fully fired. All injection loops shall be tested in 3 operating cycles.
| |
| : 2. Verify all heat-treated piping between storage tank and pump suction is unblocked.**
| |
| : e. Prior to addition of Boron to storage tank verify sodium penta borate enrichment to be added is ~ 29 atom % Boron 10.
| |
| * This test shall also be performed anytime water or boron is added to the solu tion or when the solution temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours after water or boron addition or solution temperature is restored.
| |
| ** This test shall also be performed whenever suction piping temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours after solution temperature is restored.
| |
| LIMERICK - UNIT 1 3/4 1-20 Amendment No. ~,~,~,9+,~,~,
| |
| ~,201
| |
| | |
| 80 13.8 I
| |
| t 75 f ........
| |
| .i- '-_
| |
| I . . .
| |
| i ... , ... H 70 _ . . . 1_ OPERATING RANGE .......
| |
| I . . . . . . . . . . . . . ....
| |
| 7 , . ...
| |
| : 1. . . . .
| |
| 65 1 7
| |
| ac ____.
| |
| II
| |
| :E: I C- i 4 I,, ., . _.
| |
| If 60 .L OPERAT .IGiLIMIT -
| |
| II
| |
| .I..
| |
| II I, , , , , _
| |
| 55 10 11 12 13 14 l5 CONCENTRATION, *0BY WEIGHT SODIUM PENTABORATE SOLUTION TEMPERATURE/CONCENTRATION REQUIREMENTS FIGURE 3.1.5-1 LIMERICK - UNIT :-3/4 1-21 .Amendment No. 22 l.
| |
| JUN 8 1989
| |
| | |
| II THIS PAGE LEFT BLANK INTENTIONALLY I LIMERICK - UNIT 1 3/4 1-22 Amendment No. 22 I-,
| |
| JUN 8 1989 I
| |
| | |
| 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall be within limits based on applicable APLHGR limit values which have been determined by approved methodology for the respective fuel and lattice types.
| |
| When hand calculations are required, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) as shown in the CORE OPERATING LIMITS REPORT (COLR). During operation, the APLHGR for each fuel type shall not exceed the above values multiplied by the appropriate reduction factors for power and flow as defined in the COLR.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
| |
| ACTION:
| |
| With an APLHGR exceeding the limiting value, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limiting val ue
| |
| : a. In accordance with the Surveillance Frequency Control Program,
| |
| : b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
| |
| : c. Initially and in accordance with the Surveillance Frequency Control Program when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
| |
| : d. The provisions of Specification 4.0.4 are not applicable.
| |
| LIMERICK - UNIT 1 3/4 2-1 Amendment No. .7,30, 3-7, 66, 186
| |
| | |
| . 1'K Figures onipages -
| |
| 3/4 2-2 thru 3/4 2-6c have been removed from Technical Specifications, and relocated to the CORE OPERATING LIMITS REPORT.
| |
| Technical Specifications pages 3/4 2-3 thru 3/4 2-6c have been INTENTIONALLY OMITTED.
| |
| e LIMERICK - UNIT 1 3/4 2-2 Amendment No. 37 *: I MdA? 1 5 1990
| |
| | |
| POWER DISTRIBUTION LIMITS
| |
| ,; t , ,.
| |
| Section 3/4.2.2 (DELETED)
| |
| I . -
| |
| I.
| |
| INFORMATION CONTAINED ON THIS PAGE HAS BEEN
| |
| * I DELETED I ;.: ., -
| |
| , . . I I I .:i . ...-I I .; : - ( i,
| |
| , r, .
| |
| -'. I ,.
| |
| I . .. . I I 1
| |
| ,.,. -: ', ' '!,p
| |
| ...I LIMERICK - UNIT I 3/4 2-7 Amendment No. 7, 10, 66 FB 10 1994.
| |
| | |
| POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the rated MCPR limit adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT, provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2 and the main turbine bypass system is OPERABLE per Specification 3.7.8, with:
| |
| r= (iAve TB)
| |
| TA - TB where:
| |
| TA = 0;86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, rB = 0.672 + 1.65 IN1 I '/2(0.016),
| |
| In E Ni i-I n
| |
| rave NIT
| |
| - n I Ni i=1 n = number of surveillance tests performed to date in cycle, NP = number of active control rods measured in the jth surveillance test, r= average scram time to notch 39 of all rods measured in the sth surveillance test, and N1 - total number of active rods measured in Specification 4.1.3.2.a.
| |
| APPLICABILITY:
| |
| OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
| |
| Amendment No. 17, f7, %Z,66 LIMERICK - UNIT 1 3/4 2-8 F-3 I e 1994'
| |
| | |
| POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION
| |
| : a. With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue provided that, within 1 hour, MCPR is determined to be greater than or equal to the rated MCPR limit as a function of the average scram time (shown in the CORE OPERATING LIMITS REPORT) EOC-RPT inoperable curve, adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT.
| |
| : b. With MCPR less than the applicable MCPR limit adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours or reduce THERMAL POWER to less than 25%
| |
| of RATED THERMAL POWER within the next 4 hours.
| |
| : c. With the main turbine bypass system inoperable per Specification 3.7.8, operation may continue provided that, within 1 hour, MCPR is determined to be greater than or equal to the rated MCPR limit as a function of the average scram time (shown in the CORE OPERATING LIMITS REPORT) main turbine bypass valve inoperable curve, adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT.
| |
| SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:
| |
| : a. = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2a, and during reactor startups prior to control rod scram time tests in accordance with Specification 4.1.3.2.b.1.b, or
| |
| : b. T as defined in Specification 3.2.3 used to determine the limit within 72 hours of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit, including application of the MCPR(P) and MCPR(F) factors as determined from the CORE OPERATING LIMITS REPORT.
| |
| : a. In accordance with the Surveillance Frequency Control Program,
| |
| : b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
| |
| : c. Initially and in accordance with the Surveillance Frequency Control Program when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.
| |
| : d. The provisions of Specification 4.0.4 are not applicable.
| |
| LIMERICK - UNIT 1 3/4 2-9 Amendment No. 1--, 9, -, -&9, 66, -9, 186
| |
| | |
| Figures on pages 3/4 2-10 thru 3/4 2-11 have been removed from Technical Specifications, and relocated to the CORE OPERATING LIMITS REPORT.
| |
| Technical Specifications pages 3/4 2-10a thru 3/4 2-11 have been INTENTIONALLY OMITTED.
| |
| `LIMERICK
| |
| * UNIT 1 3/4 2-10 Amendment No. ;, ,9,;0, 37 MAY 1 5 MD I
| |
| | |
| POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) for each fuel type shall not exceed the value in the CORE OPERATING LIMITS REPORT.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
| |
| ACTION:
| |
| With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:
| |
| : a. In accordance with the Surveillance Frequency Control Program, I
| |
| : b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
| |
| : c. Initially and in accordance with the Surveillance Frequency Control Program when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.
| |
| : d. The provisions of Specification 4.0.4 are not applicable.
| |
| LIMERICK - UNIT 1 3/4 2-12 Amendment No. 7, 3;, 186
| |
| | |
| 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.
| |
| APPLICABILITY: As shown in Table 3.3.1-1.
| |
| ACTION:
| |
| Note: Separate condition entry is allowed for each channel.
| |
| : a. With the number of OPERABLE channels in either trip system for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, within one hour for each affected functional unit either verify that at least one* channel in each trip system is OPERABLE or tripped or that the trip system is tripped, or place either the affected trip system or at least one inoperable channel in the affected trip system in the tripped condition.
| |
| : b. With the number of OPERABLE channels in either trip system less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) or the affected trip system** in the tripped conditions within 12 hours.
| |
| : c. With the number of OPERABLE channels in both trip systems for one or more Functional Units less than the Minimum OPERABLE Channels per Trip System required by Table 3.3.1-1, place either the inoperable channel(s) in one trip system or one trip system in the tripped condition within 6 hours**.
| |
| : d. lf within the allowable time allocated by Actions a, b or c, it is not desired to place the inoperable channel or trip system in trip (e.g., full scram would occur), Then no later than expiration of that allowable time initiate the action identified in Table 3.3.1-1 for the applicable Functional Unit.
| |
| *For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, at least two channels shall be OPERABLE or tripped. For Functional Unit 5, both trip systems shall have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or tripped. For Function 9, at least three channels per trip system shall be OPERABLE or tripped.
| |
| **For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, inoperable channels shall be placed in the tripped condition to comply with Action b. Action c does not apply for these Functional Units.
| |
| LIMERICK - UNIT 1 314 3-1 Amendment No . .§.J,++/-,-+/-4-+/-,-+/--++,~. 219
| |
| | |
| 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown in Table 4.3.1.1-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.1.1-1.
| |
| 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program, except Table 4.3.1.1-1 Functions 2.a, 2.b, 2.c, 2.d, 2.e and 2.f. Functions 2.a, 2.b, 2.c, 2.d, and 2.f do not require separate LOGIC SYSTEM FUNCTIONAL TESTS. For Function 2.e, tests shall be performed in accordance with the Surveillance Frequency Control Program. LOGIC SYSTEM FUNCTIONAL TEST for Function 2.e includes simulating APRM and OPRM trip conditions at the APRM channel inputs to the voter channel to check all combinations of two tripped inputs to the 2-Out-Of-4 voter logic in the. voter channels.
| |
| 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit in accordance with the Surveillance Frequency Control Program. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times the frequency specified in the Surveillance Frequency Control Program where N is the total number of redundant channels in a specific reactor trip system.
| |
| LIMERICK - UNIT 1 3/4 3-1a Amendment No. 44-, --7;, 186
| |
| | |
| CF TA .11 REACTOR PROTECTION S STEM INSTRUMENTATION
| |
| (
| |
| APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION
| |
| : 1. Intermediate Range Monitorsb):
| |
| : a. Neutron Flux - High 2 3 1 3(i), 4(i) 3 2 5(i) 3(d) 3
| |
| : b. Inoperative 2 3 1 3(i), 4(i) 3 2 5(i) 3(d) 3
| |
| : 2. Average Power Range Monitor',):
| |
| : a. Neutron Flux - Upscale (Setdown) 2 3(m) 1
| |
| : b. Simulated Thermal Power - Upscale 1 3(m) 4
| |
| : c. Neutron Flux - Upscale 1 3(m) 4
| |
| : d. Inoperative 1, 2 3(m) 1
| |
| : e. 2-Out-Of-4 Voter 1, 2 2 1
| |
| : f. OPRM Upscale 1(o)(p) 3(m) 10 I
| |
| : 3. Reactor Vessel Steam Dome Pressure - High 1, 2(f) 2 1
| |
| : 4. Reactor Vessel Water Level - Low, Level 3 1, 2 2 1
| |
| : 5. Main Steam Line Isolation Valve-Closure 1(g) 1/valve 4 LIMERICK - UNIT 1 3/4 3-2 Amendment No. -8,44, 444, 4-49,177
| |
| | |
| C TABLE 3.3.1-1 (Continued)
| |
| REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION
| |
| : 6. DELETED DELETED DELETED DELETED I
| |
| : 7. Drywell Pressure - High 1, 2(h) 2
| |
| : 8. Scram Discharge Volume Water Level - High
| |
| : a. Level Transmitter 1, 2 2 5(i) 2
| |
| : b. Float Switch 1, 2 2 5(i) 2
| |
| : 9. Turbine Stop Valve - Closure 1(j) 4(k)
| |
| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 1(j) 2(k)
| |
| : 11. Reactor Mode Switch Shutdown Position 1, 2 2 3, 4 2 5 2
| |
| : 12. Manual Scram 1, 2 2 3, 4 2 5 2
| |
| | |
| TABLE 3.3.1-1 (Continued)
| |
| REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 Be in at least HOT SHUTDOWN within 12 hours.
| |
| ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position with; n 1 hour.
| |
| ACTION 3 Suspend all operations involving CORE ALTERATIONS a insert all insertable control rods within 1 hour.
| |
| ACTION 4 Be in at least STARTUP within 6 hours.
| |
| ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours or in at least HOT SHUTDOWN within 12 hours.
| |
| ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure until the function is automatically bypassed, within 2 hours.
| |
| ACTION 7 Verify all insertable control rods to be inserted within 1 hour.
| |
| ACTION 8 Lock the reactor mode switch in the Shutdown position within 1 hour.
| |
| ACTION 9 Suspend all operations involving CORE ALTERATIONS, and insert all insertable control rods and lock the reactor mode switch in the Shutdown position within 1 hour.
| |
| ACTION 10 a. the condition exists due to a common-mode OPRM deficiency*, then initiate alternate method to detect and suppress thermal-hydraulic instability oscillations within 12 hours AND restore required channels to OPERABLE status within 120 days,
| |
| : b. Reduce THERMAL POWER to < 25% RATED THERMAL POWER within 4 hours.
| |
| * Unanticipated characteristic of the instability detection algorithm or equipment that renders all OPRM channels inoperable at once.
| |
| LIMERICK - UNIT 1 3/4 3-4 Amendment No. +4l,+49,~,200
| |
| | |
| TABLE 3.3.1-1 (Continued)
| |
| REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours for required surveillance without plaCing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
| |
| (b) This function shall be automatically bypassed when the reactor mode switch is in the Run pOSition.
| |
| (c) DELETED (d) The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 6 IRMs.
| |
| (e) An APRM channel is inoperable if there are less than 3 LPRM inputs per level or less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputs to the APRM channel have been bypassed since the last APRM calibration (weekly gain calibration).
| |
| (f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
| |
| (g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
| |
| (h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.
| |
| (i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| (j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 29.5% of RATED THERMAL POWER.
| |
| (k) Also actuates the EOC RPT system.
| |
| (1) DELETED (m) Each APRM channel provides inputs to both trip systems.
| |
| (n) DELETED (0) With THERMAL POWER ~ 25% RATED THERMAL POWER. The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is ~ 29.5% and recirculation drive flow is < 60%. The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is
| |
| < 29.5% or recirculation drive flow is ~ 60%.
| |
| (p) A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must OPERABLE for an OPRM channel to be OPERABLE.
| |
| LIMERICK - UNIT 1 3/4 3 5 Amendment No. 4+/-,~,+4+ ,201
| |
| | |
| !.AB( 3.12(
| |
| REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT (Seconds)
| |
| : 1. Intermediate Range Monitors:
| |
| : a. Neutron Flux - High N.A.
| |
| : b. Inoperative N.A.
| |
| : 2. Average Power Range Monitor*:
| |
| : a. Neutron Flux - Upscale (Setdown) N.A.
| |
| : b. Simulated Thermal Power - Upscale N.A.
| |
| : c. Neutron Flux - Upscale N.A.
| |
| : d. Inoperative N.A.
| |
| : e. 2-Out-Of-4 Voter <0.05*
| |
| : f. OPRM Upscale N.A.
| |
| : 3. Reactor Vessel Steam Dome Pressure - High *0.55
| |
| : 4. Reactor Vessel Water Level - Low, Level 3 *1.05#
| |
| : 5. Main Steam Line Isolation Valve - Closure <0.06
| |
| : 6. DELETED DELETED
| |
| : 7. Drywell Pressure - High N.A.
| |
| : 8. Scram Discharge Volume Water Level - High
| |
| : a. Level Transmitter N.A.
| |
| : b. Float Switch N.A.
| |
| : 9. Turbine Stop Valve - Closure *0.06
| |
| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low *0.08**
| |
| : 11. Reactor Mode Switch Shutdown Position N.A.
| |
| : 12. Manual Scram N.A.
| |
| * Neutron detectors, APRM channel and 2-Out-Of-4 Voter channel digital electronics are exempt from response time testing. Response time shall be measured from activation of the 2-Out-Of-4 Voter output relay. For applications of Specification 4.3.1.3, the redundant outputs from each 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APR outputs are considered to be separate channels, so N = 8. Testing of OPRM and APRM outputs shall alternate.
| |
| ** Measured from start of turbine control valve fast closure.
| |
| # Sensor is eliminated from response time testing for the RPS circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.
| |
| LIMERICK - UNIT I 3/4 3-6 Amendment No. 89, 432, 444, 177
| |
| | |
| TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATI CHANNEL FUNCTIONAL EL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK(n) TESTCn) CALIBRATIONCa)(n) SURVEILLANCE REQUIRED
| |
| : 1. ate Range Monitors:
| |
| : a. Neutron Flux - Hi Cb) 2 (j) 3(i), 4(i), 5(i)
| |
| : b. Inoperative N.A. (j ) N.A. 2, 3(i), 4(i), SCi)
| |
| : 2. Average Power Range MonitorCf):
| |
| : a. Neutron Flux - Upscale (Setdown) (b) (1) 2
| |
| : b. Simulated Thermal Power - Upscale (e) (d), (g), 1 (0), (p)
| |
| : c. Neutron Flux - Upscale (d) 1
| |
| : d. tive N.A. N.A. 1, 2
| |
| : e. 2-0ut-Of-4 Voter N.A. 1, 2
| |
| : f. OPRM Upscale (e) (c)(g) 1(m)
| |
| : 3. Reactor Vessel Steam Dome Pressure High 1, 2(h)
| |
| : 4. Reactor Vessel Water Level Low, Level 3 1, 2
| |
| : 5. Main Steam Line Isolation Valve - Closure N.A. 1
| |
| : 6. DELETED
| |
| : 7. Pressure - High 1, 2
| |
| : 8. Scram Discharge Volume Water Level High
| |
| : a. Level Transmitter 1, 2, 5(i)
| |
| : b. Float Switch N.A. 1, 2, 5(i)
| |
| LIMERICK - UNIT 1 3/4 3-7 Amendment No. 4+/-,~,gg,+49 ,l&&,
| |
| 201
| |
| | |
| TABLE 4.3.1.1-1 (Continued)
| |
| REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNE FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECKCn) TEST( n) CALIBRATIONCa)(n) SURVEILLANCE REQUIRED
| |
| : 9. Turbine Stop Valve - Closure N.A. 1
| |
| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure Low N.A. 1
| |
| : 11. Reactor Mode Switch Shutdown Position N.A. N.A. 1, 2, 3, 4, 5
| |
| : 12. Manual Scram N.A. N.A. 1, 2, 3, 4, 5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
| |
| (b) The IRM and SRM channels shall determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 (c) Calibration includes verification that the OPRM Upscale trip auto-enable (not- ss) setpoint for APRM Simulated Thermal Power is ~ 29.5% and for recirculation drive flow is < 60%.
| |
| (d) The more frequent calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER ~25% of RATED THERMAL POWER.
| |
| ust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.
| |
| (e) CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.
| |
| (f) The LPRMs shall be calibrated at least once per 2000 effective full power hours (EFPH).
| |
| (g) The less frequent calibration includes the flow input function.
| |
| (h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
| |
| (i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| (j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.
| |
| (k) DELETED (1) Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1 until 12 hours after entering OPERATIONAL CONDITION 2.
| |
| (m) With THERMAL POWER ~ 25% of RATED THERMAL POWER.
| |
| (n) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| LIMERICK UNIT 1 3/4 3-8 Amendment No. ~,4+,~,ee,
| |
| -l-H,++/-+ ,-+/-4+/--,+7+ ,+%,201
| |
| | |
| TABLE 4.3.1.1-1 (Continued)
| |
| REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS (0) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.
| |
| (p) The instrument channel setpoint 11 be reset to a value that is within the as-left tolerance around the Trip Setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the Trip Setpoint are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in the associated Technical Specifications Bases.
| |
| LIMERICK - UNIT 1 3/4 3-8a Amendment No .201
| |
| | |
| INSTRUMENTATION 3/4.3.2. ISOLATION ACTUATION INSTRUMENTATION
| |
| *\_ITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip
| |
| *Setpoint column of Table 3.3.2.-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.
| |
| APPLICABIL ITY: As shown in Table 3.3.2-1.
| |
| ACTION:
| |
| a) With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint-adjusted consistent with the Trip Setpoint value.
| |
| b) With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirements for one trip system:
| |
| : 1. If placing the inoperable channel(s) in the tripped condition would cause an isolation, the inoperable channel(s) shall be restored to OPERABLE status within 6 hours. If this cannot be accomplished, the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken, or the channel shall be placed in the tripped condition.
| |
| or
| |
| : 2. If placing the inoperable channel(s) in the tripped condition would not cause an isolation, the inoperable channel(s) and/or that trip system shall be placed in the tripped condition within:
| |
| a) 12 hours for trip functions common* to RPS Instrumentation.
| |
| b) 24 hours for trip functions not common* to RPS Instrumentation.
| |
| I
| |
| * Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1.
| |
| LIMERICK - UNIT 1 3/4 3-9 Amendmeiit No. 53, 6X, 169
| |
| | |
| INSTRUMENTATION LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : c. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system** in the tripped condition within 1 hour and take the ACTION required by Table 3.3.2-1.
| |
| SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown in Table 4.3.2.1-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.2.1-1.
| |
| 4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operations of all channels shall be performed in accordance with the Surveillance Frequency Control Program. I 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit in accordance with the Surveillance Frequency Control Program. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times the frequency specified in accordance with the Surveillance Frequency Control Program, where N is the total number of redundant channels in a specific isolation trip system.
| |
| ** The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.
| |
| LIMERICK - UNIT I 3/4 3-10 Amendment No. .53, 4-1, 186
| |
| | |
| 4 C TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION
| |
| : 1. MAIN STEAM LINE ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low-Level 2 B 2 1, 2, 3 21
| |
| : 2) Low, Low, Low-Level 1 C 2 1, 2, 3 21
| |
| : b. DELETED DELETED DELETED DELETED DELETED
| |
| : c. Main Steam Line Pressure - Low P 2 1 22
| |
| : d. Main Steam Line Flow - High E 2/line 1, 2, 3 20
| |
| : e. Condenser Vacuum - Low Q 2 1, 2**, 3** 21
| |
| : f. Outboard MSIV Room Temperature - High F(f) 2 1, 2, 3 21
| |
| : 9. Turbine Enclosure - Main Steam Line Tunnel Temperature - High F(f) 14 1, 2, 3 21
| |
| : h. Manual Initiation NA 2 1, 2, 3 24
| |
| : 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level Low - Level 3 A 2 1, 2, 3 23
| |
| : b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High V 2 1, 2, 3 23
| |
| : c. Manual Initiation NA 1 1, 2, 3 24
| |
| | |
| ( C TABLE 3.3.2-1 (Continued)
| |
| C ISOLATION ACTUATION INSTRUMENTATION C3
| |
| ;2a MINIMUM APPLICABLE c' ISOLATION OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION Z 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. RWCS A Flow - High J 1 1, 2, 3 23
| |
| : b. RWCS Area Temperature - High J 6 1, 2, 3 23
| |
| : c. RWCS Area Ventilation A Temperature - High J 6 1, 2, 3 23
| |
| : d. SLCS Initiation Y(d) NA 1, 2, 3 23 w e. Reactor Vessel Water Level -
| |
| Low, Low.- Level 2 B 2 1, 2, 3 23
| |
| : f. Manual Initiation NA 1 1, 2, 3 24
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
| |
| : a. HPCI Steam Line A Pressure - High L 1 1, 2, 3 23
| |
| : b. HPCI Steam Supply Pressure - Low LA 2 1, 2, 3 23
| |
| : c. HPCI Turbine Exhaust Diaphragm Pressure - High L 2 1, 2, 3 23
| |
| : d. HPCI Equipment Room Temperature - High L 1 1, 2, 3 23
| |
| : e. HPCI Equipment Room A Temperature - High L 1 1, 2, 3 23
| |
| | |
| C 0 C *(
| |
| TABLE 3.3.2--1 (Continued)
| |
| I.-
| |
| ISOLATION ACTUAT][ON INSTRUMENTATION -
| |
| ai rr h
| |
| MINIMUM APPLICABLE X
| |
| ISOLATION OPERABLE CHANNELS OPERATIONAL I.- TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION ((Continued)
| |
| : f. HPCI Pipe Routing Area Temperature - High L 4 1, 2, 3 23
| |
| : g. Manual Initiation NAWe) 1/system 1, 2, 3 24' C
| |
| : h. HPCI Steam Line A Press Timer NA '1. 1, 2, 3 23 P"
| |
| : 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
| |
| : a. RCIC Steam Line A Pressure - High K 1 1, 2, 3 23 I
| |
| : b. RCIC Steam Supply Pressure - Low KA 2 1, 2, 3 23
| |
| : c. RCIC Turbine Exhaust Diaphragm Pressure - High K 2 1, 2, 3 23
| |
| : d. RCIC Equipment Room Temperature - High K 1 1, 2, 3 23
| |
| : e. RCIC Equipment Room A Temperature - High K 1 1, 2, 3 23
| |
| : f. RCIC Pipe Routing Area Temperature - High K 5 1, 2, 3 23 C-,
| |
| -f 9.
| |
| g Manual Initiation NA~e) 1/system 1, 2, 3 24 C).
| |
| 0c
| |
| : h. RCIC Steam Line tD
| |
| * A Pressure Timer NA *1 1, 2, 3 23
| |
| | |
| C C I-TABLE 3.3.2-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION C,
| |
| MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL SIGNAL(") PER TRIP SYSTEM (b) QCONDITION-
| |
| -e 6. PRIMARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level I) Low, Low - Level 2 B 2 1, 2, 3 20
| |
| : 2) Low, Low, Low - Level 1 C 2 1, 2, 3 20
| |
| : b. Drywell Pressure - High H 2 1, 2, 3 20
| |
| : c. North Stack Effluent
| |
| .Radiation - High (" U 1 1, 2, 3 23
| |
| : d. Deleted
| |
| : e. Reactor Enclosure Ventilation Exhaust Duct-Radiation - High S 2 1, 2, 3 23
| |
| : f. Deleted I
| |
| : 9. Deleted
| |
| : h. Drywell Pressure - High/
| |
| Reactor Pressure .- Low G 2/2 1, 2, 3 26
| |
| -2 3
| |
| : i. Primary Containment Instrument M 1 1, 2, 3 26 Gas Line to Drywell A Pressure - Low J. Manual Initiation NA 1 1, 2, 3 24 r(
| |
| .LA..
| |
| | |
| LLE 3.3.2-1 (Continued)
| |
| ISOLATICINACTUATION INSTRUMENTATION MINIMUM APPLICABLE ISOLATIO1qN OPERABLE CHANNELS OPERATIONAL SIGNALE ,'" PER TRIP SYSTEM (b)
| |
| TRIP FUNCTION CONDITION AUTIN
| |
| - 7. SECONDARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level Low, Low - Level 2 B 2 1, 2, 3 25
| |
| : b. Drywell Pressure - High H 2 1, 2, 3 25 c.l. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High R 2 *1 25
| |
| : 2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High R 2 25
| |
| : d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High S 2 1, 2, 3 25
| |
| : e. Deleted (D
| |
| * 3 f. Deleted AD e g. Reactor Enclosure Manual Initiation NA 1 1, 2, 3 24
| |
| : h. Refueling Area Manual Initiation NA 1 25 ba
| |
| .I' i
| |
| | |
| TABLE 3.3.2-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 - Be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24-hours.
| |
| ACTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| ACTION 22 - Be in at least STARTUP within 6 hours.
| |
| ACTION 23 - In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves are closed within 1 hour and declare the affected system inoperable. In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within 12 hours.
| |
| ACTION 24 - Restore the manual initiation function to OPERABLE status within 8 hours or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour.
| |
| ACTION 26 - Close the affected system isolation valves within 1 hour.
| |
| TABLE NOTATIONS
| |
| * Required when handling RECENTLY IRRADIATED FUEL in the secondary containment.
| |
| ** May be bypassed under administrative control, with all turbine stop valves closed.
| |
| # During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.
| |
| (a) DELETED (b) A channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1. In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours for required surveillance without placing the channel or trip system in the tripped condition.
| |
| LIMERICK - UNIT 1 3/4 3-16 Amendment No. ~.4G,-9-d,-e-9,-14i, ~.227
| |
| | |
| TABLE 3.3.2-1 (Continued)
| |
| TABLE NOTATIONS (c) Actuates secondary containment isolation valves. Signals B, H, S, and R also start the standby gas treatment system. I (d) RWCU system inlet outboard isolation valve closes on SLCS "B" initiation.
| |
| RWCU system inlet inboard isolation valve closes on SLCS "A" or SLCS "C" initiation.
| |
| (e)Manual initiation isolates the steam supply line outboard isolation valve and only-following manual or automatic initiation of the system.
| |
| (f) In the event of a loss of-ventilation the temperature - high setpoint may be raised by 500F for a period not to exceed 30 minutes-to permit restoration of the ventilation flow without a spurious trip. During the 30 minute period, an operator, or other qualified member of the technical staff, shall observe the temperature' indications continuously, so that, in the event of rapid increases in temperature, the main steam lines shall be manually isolated.
| |
| (g)Wide range accident monitor per Specification 3.3.7.5.
| |
| Amendment 28, 5a,112, 146 OCT 18 2Uo0 LIMERICK - UNIT I 3/4 3-17
| |
| | |
| TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
| |
| : 1. MAIN STEAM LINE ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low - Level 2 z - 38 inches* z - 45 inches
| |
| : 2) Low, Low, Low - Level 1 z - 129 inches* z - 136 inches
| |
| : b. DELETED DELETED DELETED
| |
| : c. Main Steam Line Pressure - Low z 840 psig 2 821 psig
| |
| : d. Main Steam Line Flow - High ~ 122.l psid ~ 123 psid
| |
| : e. Condenser Vacuum - Low 10.5 psia 210.1 psia/~ 10.9 psia
| |
| : f. Outboard MSIV Room Temperature - High ~ 192°F ~ 200°F
| |
| : g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High ~ 165°F ~ l 75°F
| |
| : h. Manual Initiation N.A. N.A.
| |
| : 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level Low - Level 3 z 12.5 inches* z 11.0 inches
| |
| : b. Reactor Vessel CRHR Cut-in Permissive) Pressure - High ~ 75 psig ~ 95 psig
| |
| : c. Manual Initiation N.A. N.A.
| |
| LIMERICK - UNIT 1 3/4 3-18 Amendment No. ~. &9-, ~. 222
| |
| | |
| TABLE 3.3.2-2 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
| |
| : 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. RWCS ~ Flow - High :.,:; 54.9 gpm :.,:; 65.2 gpm
| |
| : b. RWCS Area Temperature - High :.,:; 155°F or :.,:; 120°F** :.,:; 160°F or :.,:; 125°F**
| |
| : c. RWCS Area Ventilation
| |
| ~Temperature - High :.,:; 52°F or :.,:; 32°F** :.,:; 60°F or :.,:; 40°F**
| |
| : d. SLCS Initiation N.A. N.A.
| |
| : e. Reactor Vessel Water Level -
| |
| Low, Low, - Level 2 2 -38 inches
| |
| * 2 -45 inches
| |
| : f. Manual Initiation N.A. N.A.
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
| |
| : a. HPCI Steam Line~ Pressure - High :.,:; 974" H20 :.,:; 984" H20
| |
| : b. HPCI Steam Supply Pressure - Low 2 100 psig 2 90 psig
| |
| : c. HPCI Turbine Exhaust Diaphragm Pressure - High :.,:; 10 psig :.,:; 20 psig
| |
| : d. HPCI Equipment Room Temperature - High 180°F 2 177°F, :.,:; 191°F
| |
| : e. HPCI Equipment Room
| |
| ~Temperature - High :.,:; 104°F :.,:; 108.5°F
| |
| : f. HPCI Pipe Routing Area Temperature - High 180°F 2 177°F, :.,:; 191°F
| |
| : g. Manual Initiation N.A. N.A.
| |
| : h. HPCI Steam Line~ Pressure - Timer 3 :.,:; T:.,:; 12.5 seconds 2.5:.,:; T:.,:; 13 seconds LIMERICK - UNIT 1 3/4 3-19 Amendment No. JJ,&9,+GB,+&+,~.213
| |
| | |
| TABLE 3.3.2-2 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
| |
| : 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
| |
| : a. RCIC Steam Line 11 Pressure - High ~ 373" H20 ~ 381" H20
| |
| : b. RCIC Steam Supply Pressure - Low ~ 64.5 psig ~ 56.5 psig
| |
| : c. RCIC Turbine Exhaust Diaphragm Pressure - High ~ 10.0 psig ~ 20.0 psig
| |
| : d. RCIC Equipment Room Temperature - High 180°F ~ 161°F, ~ 191°F
| |
| : e. RCIC Equipment Room 11 Temperature - High ~ 109oF ~ 113. 5°F
| |
| : f. RCIC Pipe Routing Area Temperature - High 180oF ~ 16l°F, ~ 191°F
| |
| : g. Manual Initiation N.A. N.A.
| |
| : h. RCIC Steam Line 11 Pressure Timer 3 ~ t ~ 12.5 seconds 2.5 ~ t ~ 13 seconds LIMERICK - UNIT 1 3/4 3-20 Amendment No. JJ, ~.-+/--%, 213
| |
| | |
| C. C r-TABLE 3.3.2-2 (Continued)
| |
| C-,
| |
| ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE cTRIP FUNCTION TRIP SETPOINT VALUE
| |
| ' 6. PRIMARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1. Low, Low - Level 2 2 -38 inches* 2 -45 inches
| |
| : 2. Low, Low, Low - Level 1 2 -129 inches* a -136 inches
| |
| : b. Drywell Pressure - High 5 1.68 psig < 1.88 psig
| |
| : c. North Stack Effluent Radiation - High < 2.1 pCi/cc < 4.0 pCi/cc
| |
| : d. Deleted
| |
| : e. Reactor Enclosure Ventilation Exhaust Duct - Radiation - High 5 1.35 mR/h 5 1.5 mR/h
| |
| : f. Deleted I
| |
| : g. Deleted
| |
| : h. Drywell Pressure - High/ < 1.68 psig/
| |
| * 1.88 psig/
| |
| Reactor Pressure - Low 2 455 psig (decreasing) 2 435 psig (decreasing)
| |
| AD C i. Primary Containment Instrument 2 2.0 psi 2 1.9 psi M Gas to Drywell A Pressure - Low a
| |
| J. AdJ. Manual Initiation N.A. N.A.
| |
| C.C.
| |
| Cj-->
| |
| <73 DO
| |
| .. t-
| |
| | |
| Q' C C I--
| |
| zr, IABLE 3.3.2-2 (Continued)
| |
| C, ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE
| |
| -FUNCTIO TRIP SETPOINT VALUE
| |
| = 7.
| |
| -Il. SECONDARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low, Low - Level 2 2 -38 inches* 2-45 inches
| |
| : b. Drywell Pressure - High S 1.68 psig
| |
| * 1.88 psig c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High S 2.0 mR/h S 2.2 mR/h
| |
| : 2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S 2.0 mR/h S 2.2 mR/h W
| |
| 4I.' d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High S 1.35 mR/h S 1.5 mR/h
| |
| : e. Deleted
| |
| : f. Deleted (D
| |
| : g. Reactor Enclosure Manual tC-..
| |
| C, Initiation N.A. N.A.
| |
| a tD
| |
| : h. Refueling Area Manual Initiation N.A. N.A.
| |
| _ A Pa 20 cl A:1
| |
| * See Bases Figure B 3/4 3-1.
| |
| ** The low setpoints are for the RWCU Heat Exchanger Rooms; the high setpoints are for the pump rooms.
| |
| It
| |
| | |
| TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
| |
| : 1. MAIN STEAM LINE ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low - Level 2 N.A.
| |
| : 2) Low, Low, Low- Level 1 .::::1.0###*
| |
| : b. DELETED DELETED
| |
| : c. Main Steam Line Pressure - Low .::::1.0 ###*
| |
| : d. Main Steam Line Flow - High ~1.0 ###*
| |
| : e. Condenser Vacuum - Low N.A.
| |
| : f. Outboard MSIV Room Temperature - High N.A.
| |
| : g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
| |
| : h. Manual Initiation N.A.
| |
| : 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level Low - Level 3 N.A.
| |
| : b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High N.A.
| |
| : c. Manual Initiation N.A.
| |
| : 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. RWCS 6 Flow - High N.A. ##
| |
| : b. RWCS Area Temperature - High N.A.
| |
| : c. RWCS Area Ventilation 6 Temperature - High N.A.
| |
| : d. SLCS Initiation N.A.
| |
| : e. Reactor Vessel Water Level -
| |
| Low, Low - Level 2 N.A.
| |
| : f. Manual Initiation N.A.
| |
| LIMERICK - UNIT 1 3/4 3-23 Amendment No. ~.gg.~, 214
| |
| | |
| TABLE 3.3.2-3 (Continued)
| |
| ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
| |
| : a. HPCI Steam Line A Pressure - High N.A.
| |
| : b. HPCI Steam Supply Pressure - Low N.A.
| |
| : c. HPCI Turbine Exhaust Diaphragm Pressure - High N.A.
| |
| : d. HPCI Equipment Room Temperature - High N.A.
| |
| : e. HPCI Equipment Room
| |
| ^ Temperature - High N.A.
| |
| : f. HPCI Pipe Routing Area Temperature - High N.A.
| |
| : g. Manual Initiation N.A.
| |
| : 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
| |
| : a. RCIC Steam Line
| |
| & Pressure - High N.A.
| |
| : b. RCIC Steam Supply Pressure - Low N.A.
| |
| : c. RCIC Turbine Exhaust Diaphragm Pressure - High N.A.
| |
| : d. RCIC Equipment Room Temperature - High N.A.
| |
| : e. RCIC Equipment Room
| |
| & Temperature - High N.A.
| |
| : f. RCIC Pipe Routing Area Temperature - High N.A.
| |
| : 9. Manual Initiation N.A.
| |
| LIMERICK - UNIT 1 3/4 3-24 Amendment lb. 33-,132 DEC 1 4 1998
| |
| | |
| TABLE 3.3.2-3 (Continued)
| |
| ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME K) TRIP FUNCTION RESPONSE TIME (Seconds)#
| |
| : 6. PRIMARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level.
| |
| : 1) Low, Low - Level 2 N.A.
| |
| : 2) Low, Low, Low - Level I N.A. I'
| |
| : b. Drywell Pressure - High N.A.
| |
| : c. North Stack Effluent Radiation - High N.A.
| |
| : d. Deleted
| |
| : e. Reactor Enclosure Ventilation Exhaust Duct - Radiation - High N.A.
| |
| : f. Deleted
| |
| : 9. Deleted
| |
| : h. Drywell Pressure - High/
| |
| Reactor Pressure - Low N.A.
| |
| : i. Primary Containment Instrument Gas to Drywell A Pressure - Low N.A.
| |
| : j. Manual Initiation N.A.
| |
| : 7. SECONDARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level Low, Low - Level 2 N.A.
| |
| : b. Drywell Pressure - High N.A.
| |
| c.l. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High N.A.
| |
| : 2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High N.A.
| |
| : d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High N.A.
| |
| : e. Deleted LIMERICK - UNIT I 3/4 3-25 Amendment No. 2 412,132 DEd 1 4 '1998
| |
| | |
| TABLE 3.3.2-3 (Continued)
| |
| ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
| |
| : f. Deleted
| |
| : g. Reactor Enclosure Manual Initiation N.A.
| |
| : h. Refueling Area Manual Initiation N.A.
| |
| TABLE NOTATIONS (a) DELETED (b) DELETED
| |
| * Isolation system instrumentation response time for MSIV only. No diesel generator delays assumed for MSIVs.
| |
| DELETED Isolation system instrumentation response time specified for the Trip
| |
| .Function actuating each valve group shall be.added to the isolation time for the valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME I for each valve.
| |
| With 45 second time delay.
| |
| Sensor is eliminated from response time testing for the MSIV actuation logic circuits. Response time testing and conformance to the administrative limits for the remaining channel including trip unit and relay logic are required.
| |
| Amendment No. 6, 89, all, p, iA'4 C'T 1 2G00 LIMERICK - UNIT 1 3/4 3-26
| |
| | |
| TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOTIIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK(a) TEST(a) CALIBRATION(a) SURVEILLANCE REOUIRED I
| |
| : 1. MAIN STEAM LINE ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low, Level 2 1, 2, 3
| |
| : 2) Low, Low, Low - Level 1 1, 2, 3
| |
| : b. DELETED
| |
| : c. Main Steam Line Pressure - Low 1
| |
| : d. Main Steam Line Flow - High 1, 2, 3
| |
| : e. . Condenser Vacuum - Low 1, 2**, 3**
| |
| : f. Outboard MSIV Room Temperature - High 1, 2, 3
| |
| : g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High 1, 2, 3
| |
| : h. Manual Initiation N.A. N.A. 1. 2, 3 I
| |
| : 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level##
| |
| Low - Level 3 1, 2, 3 I
| |
| : b. Reactor Vessel (RHR Cut-In I
| |
| 1, 2, 3 Permissive) Pressure - High
| |
| : c. Manual Initiation N.A. N.A. 1, 2, 3 I LIMERICK - UNIT 1 3/4 3-27 Amendment No. 8,6 ,
| |
| 186
| |
| | |
| TABLE 4.3.2.1-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNE L CONDITIONS FOR WHICH TRIP FUNCTION CHECK(a) TEST(a) CALIBRAT ION(a) SURVEILLANCE REOUIREb
| |
| : 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. RWCS A Flow - High 1, 2, 3 I
| |
| : b. RWCS Area Temperature - High 1, 2, 3
| |
| : c. RWCS Area Ventilation A Temperature -'High 1, 2, 3
| |
| : d. SLCS Initiation N.A. N.A. 1, 2, 3
| |
| : e. Reactor Vessel Water Level Low, Low, - Level 2 1, 2, 3
| |
| : f. Manual Initiation N.A. N.A. 1, 2, 3
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
| |
| : a. HPCI Steam Line A Pressure - High 1, 2, 3
| |
| : b. HPCI Steam Supply Pressure, Low 1, 2, 3
| |
| : c. HPCI Turbine Exhaust Diaphragm Pressure - High 1, 2, 3
| |
| : d. HPCI Equipment Room Temperature - High 1, 2, 3
| |
| : e. HPCI Equipment Room A Temperature - High 1, 2, 3
| |
| : f. HPCI Pipe Routing Are'a Temperature - High 1, 2, 3
| |
| : g. Manual Initiation N.A. N.A. 1, 2, 3
| |
| : h. HPCI Steam Line A Pressure Timer N.A. 1, 2, 3 LIMERICK - UNIT 1 3/4 3-28 Amendment No. -, &9, 186
| |
| | |
| TABLE 4.3.2.1-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUI REMENTS REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK (a) TEST (a) CALIBRATION(a) SURVEILLANCE REQUIRED I
| |
| : 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
| |
| : a. RCIC Steam Line A Pressure - High 1, 2, 3
| |
| : b. RCIC Steam Supply Pressure - Low 1, 2, 3
| |
| : c. RCIC Turbine Exhaust Diaphragm Pressure - High 1, 2, 3
| |
| : d. RCIC Equipment Room Temperature - High 1, 2, 3
| |
| : e. RCIC Equipment Room A Temperature - High 1, 2, 3
| |
| : f. RCIC Pipe Routing Area Temperature - High 1, 2, 3
| |
| : g. Manual Initiation N.A. N.A. 1, 2, 3
| |
| : h. RCIC Steam Line A Pressure Timer N.A. 1, 2, 3 LIMERICK - UNIT 1 3/4 3-29 Amendment No. -34, 6-6,186
| |
| | |
| TABLE 4.3.2.1-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK(A). TESTW-L CALIBRATION(.U.). SURVEILLANCE REOUIRED
| |
| : 6. PRIMARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level 1, 2, 3 1) 2)
| |
| Low, Low - Level 2 Low, Low, Low - Level 1 1, 2, 3 II
| |
| : b. Drywell Pressure##t - High 1, 2, 3
| |
| : c. North Stack Effluent Radiation - High 1, 2, 3
| |
| : d. Deleted
| |
| : e. Reactor Enclosure Ventilation Exhaust Duct - Radiation - High 1, 2,3 I
| |
| : f. Deleted
| |
| : g. Deleted
| |
| : h. Drywell Pressure - High/
| |
| Reactor Pressure - Low 1, 2, 3
| |
| : i. Primary Containment Instrument Gas to Drywell A Pressure - Low N.A. 1, 2, 3 I 1, 2, 3 I
| |
| : j. Manual Initiation N.A. N.A.
| |
| LIMERICK - UNIT 1 3/4 3-30 Amendment No. 6, -53, 69, 4-l-9, 186
| |
| | |
| TABLE 4.3.2.1-1 (Continued)
| |
| ISO_LAIJON ACTUATION INSTRUMrnTA_llON SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK (a) TEST (a) CALIBRATION(a) SURVEILLANCE REQUIRED
| |
| : 7. SECONDARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level Low, Low - Level 2 1, 2, 3
| |
| : b. Drywel l Pressure## - High 1, 2, 3 c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High *#
| |
| : 2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High *#
| |
| : d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High 1, 2, 3
| |
| : e. Deleted
| |
| : f. Deleted
| |
| : g. Reactor Enclosure Manual Initiation N.A. N.A. 1, 2, 3
| |
| : h. Refueling Area Manual Initiation N.A. N.A. *
| |
| (a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| *Required when handling RECENTLY IRRADIATED FUEL in the secondary containment.
| |
| **When not administratively bypassed and/or when any turbine stop valve is open.
| |
| #During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.
| |
| ##These trip functions (2a, 6b, and 7b) are common to the RPS actuation trip function.
| |
| LIMERICK - UNIT 1 3/4 3-31 Amendment No . .iJ, 4G, w, @, g.g., ~ . ~ . ~ . 227
| |
| | |
| INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3.
| |
| APPLICABILITY: As shown in Table 3.3.3-1 ACTION:
| |
| : a. With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to Operable status with its trip setpoint adjusted consistent with the Trip Setpoint value.
| |
| : b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.
| |
| : c. With either ADS trip system subsystem inoperable, restore the inoperable trip system to OPERABLE status within:
| |
| : 1. 7 days, provided that the HPCI and RCIC systems are OPERABLE.
| |
| : 2. 72 hours.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 100 psig within the following 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown in Table 4.3.3.1-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.3.1-1.
| |
| 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.
| |
| 4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3 shall be demonstrated to be within the limit in accordance with the Surveillance Frequency Control Program. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times the frequency specified in the Surveillance Frequency Control Program where N is the total number of redundant channels in a specific ECCS trip system.
| |
| ,LIMERICK - UNIT 1 3/4 3-32 Amendment No. ;4, 186
| |
| | |
| TABLE 3.3.3-1 EMER_G_ENCY_CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRIP OPERATIONAL TRIP FUN CTI ON FUNCTION(al CONDITIONS ACTION
| |
| : 1. CORE SPRAY SYSTEM***
| |
| (b)
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 2/pump(bl 1, 2, 3 30
| |
| : b. Drywel l Pressure - High 2/pump 1, 2, 3 30 6(b)
| |
| C. Reactor Vessel Pressure - Low (Permissive) 1, 2, 3 31 z(e) 1, 2, 3
| |
| : d. Manual Initiation 33
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM***
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1, 2, 3 30
| |
| : b. Drywel l Pressure - High 2 1, 2, 3 30 C. Reactor Vessel Pressure - Low (Permissive) 2 1, 2, 3 31
| |
| : d. Injection Valve Differential Pressure-Low 1/valve 1, 2, 3 31 (Permissive)
| |
| : e. Manual Initiation 1 1, 2, 3 33
| |
| : 3. HIGH PRESSURE COOLANT INJECTION SYSTEM##
| |
| : a. Reactor Vessel Water LeltBJ - Low Low Level 2 4 1, 2, 3 34
| |
| : b. Drywel l Pressure - High 4 1, 2, 3 34 z(c)
| |
| C. Condensate Storage Tank Level - Low 1, 2, 3 35
| |
| : d. Suppression Pool Water Level - High 2(d) 1, 2, 3 35
| |
| : e. Reactor Vessel WaMJ Level - High, Level 8 4 1, 2, 3 31
| |
| : f. Manual Initiation 1/system 1, 2, 3 33 LIMERICK - UNIT 1 3/4 3-33 Amendment No. -bM-, 227
| |
| | |
| . I TABLE 3.3.3-1 (Continued) m
| |
| ' -4 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION so C-6-4 ,, -MINII MUM OPERABLE CHjkNNELS PER APPLICABLE 1
| |
| TRIP OPERATIONAL TRIP FUNCTION FUI NCTION a CONDITIONS ACTION
| |
| : 4. AUTOMATIC DEPRESSURIZATION SYSTEM#***
| |
| N. a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1, 2, 3 30
| |
| : b. Drywell Pressure - High 2 1, 2, 3 30 r-J c. ADS Timer 1 1, 2, 3 31
| |
| : d. Core Spray Pump Discharge Pressure - High (Permissive) 2 1, 2, 3 31
| |
| : e. RHR LPCI Mode Pump Discharge Pressure High (Permissive) 4 1, 2, 3 31
| |
| : f. Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1, 2, 3 31 (A)
| |
| : g. Manual Initiation 2 1, 2, 3 33
| |
| : h. ADS Drywell Pressure Bypass Timer. 2 1, 2, 3 31 4>1 MINIMUM APPLICABLE TOTAL NO. CHANNELS CHANNELS OPERATIONAL OF CHANNELS(f) TO TRIP OPERABLE CONDITIONS ACTION
| |
| : 5. LOSS OF POWER
| |
| : 1. 4.16 kV Emergency Bus Under-voltage (Loss of Voltage) 1/bus 1/bus 1/bus 1,2,3,4**,5** 36
| |
| : 2. 4.16 kV Emergency Bus Under-voltage (Degraded Voltage) 1/source/ 1/source/ 1/source/ 1,2,3,4**,5** 37 bus bus bus
| |
| .3 0-.
| |
| =3 0
| |
| CLA rta
| |
| ***The Minimum OPERABLE Channels Per Trip Function is per subsystem.
| |
| (7' ' , ' '( '' (.
| |
| | |
| TABLE 3.3.3-1 (Continued)
| |
| EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
| |
| (b) Also provides input to actuation logic for the associated emergency diesel generators.
| |
| (c) One trip system. Provides signal to HPCI pump suction valves only.
| |
| (d) On 1 out of 2 taken twice logic, provides a signal to trip the HPCI pump turbine only.
| |
| (e) The manual initiation push buttons start the respective core spray pump and diesel generator. The "A" and "B" logic manual push buttons also actuate an initiation permissive in the injection valve opening logic.
| |
| (f) A channel as used here is defined as the 127 bus relay for Item 1 and the 127, 127Y, and 127Z feeder relays with their associated time delay relays taken together for Item 2.
| |
| * DELETED
| |
| # Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
| |
| ** Required when ESF equipment is required to be OPERABLE.
| |
| ## Not required to be OPERABLE when reactor steam dome pressure is 1ess than or equal to 200 psig.
| |
| ### The injection functions of Drywell Pressure - High and Manual Initiation are not required to be OPERABLE with reactor steam dome pressure less than 550 psig.
| |
| LIMERICK - UNIT 1 3/4 3-35 Amendment No. -eJ,-l24, 227
| |
| | |
| TABLE 3.3.3-1 (Continued)
| |
| EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
| |
| : a. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours or declare the associated system inoperable.
| |
| : b. With more than one channel inoperable, declare the associated system inoperable.
| |
| ACTION 31 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ECCS inoperable within 24 hours.
| |
| ACTION 32 - DELETED ACTION 33 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 24 hours or declare the associated ECCS inoperable.
| |
| ACTION 34 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
| |
| : a. For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours or declare the HPCI system inoperable.
| |
| : b. With more than one channel inoperable, declare the HPCI system inoperable.
| |
| ACTION 35 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 24 hours or declare the HPCI system inoperable.
| |
| ACTION 36 - With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator and the associated offsite source breaker that is not supplying the bus inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.
| |
| LIMERICK - UNIT 1 3/4 3-36 Amendment No. ii, ~. -+/--e&, 227
| |
| | |
| TABLE 3.3.3-1 (Continued)
| |
| E1-'RGEN7Y CORE COSING SYST2' A_-TUATo:: -.. 7u_.!: - ATSON ACT IO'N S T.-T V£E7S ACTION 37 - With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable device in the bypassed condition subject to the following conditions:
| |
| Inoperable Device Condition 127-l1XOX 127Y-llXOX and 127Z-llXOX operable 127Y-1lXOX 127-l1XOX and 127Z-llXOX operable 127Z-llXOX 127-llXOX and 127Y-llXOX operable.
| |
| 127Z-llYOY operable for the other 3 breakers monitoring that source, offsite source grid voltage for that source is maintained at or above 230kV (for the 101 Safeguard Bus Source) or 525kV (for the 201 Safeguard Bus Source),
| |
| Load Tap Changer for that source is in service and in automatic operation, and the electrical buses and breaker alignments are maintained within bounds of approved plant procedures.
| |
| or, place the inoperable channel in the tripped condition within 1 hour and take the Action required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.
| |
| Operation may then continue until perfo mance of the next required CHANNEL FUNCTIONAL TEST.
| |
| LI.-1ERIi - U>NIT 1 3/4 3-36a Amendment No. 158 I AR 2 0 2WJ2
| |
| | |
| N
| |
| ) 0
| |
| -) 1 TABLE 3.3.3-2 I-
| |
| - -EMERGENCY
| |
| '; CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS
| |
| -II ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
| |
| : 1. CORE SPRAY SYSTEMs S T. I i
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 > -129 inches* > -136 inches I
| |
| : b. Drywell Pressure - High Z 1.68 psig < 1.88 psig
| |
| : c. Reactor Vessel Pressure - Low 5 455 psig,(decreasing) 5 435 1ssig, (decreasing)
| |
| : d. Manual Initiation N.A. N.A.
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 > -129 inches* > -136 inches
| |
| : b. Drywell Pressure - High < 1.68 psig < 1.88 psig
| |
| : c. Reactor Vessel Pressure - Low 5 455 psig,(decreasing) > 435 psig, (decreasing)
| |
| CkJ d. Injection Valve Differential Pressure - Low > 74 psid, (decreasing) > 64 psid and < 84 psid CAJ e. Manual Initiation N.A. N.A.,
| |
| : 3. HIGH PRESSURE COOLANT INJECTION SYSTEM
| |
| : a. Reactor Vessel Water Level - Low Low, Level 2 > -38 inches* > -45 inches I
| |
| : b. Drywell Pressure - High Z 1.68 psig < 1.88 psig
| |
| : c. Condensate Storage Tank Level - Low 5 167.8 inches** 5 164.3 inches
| |
| : d. 'Suppression Pool Water Level - High Z 24 feet 1.5 inches < 24 feet 3 inches
| |
| : e. Reactor Vessel Water Level - High, Level 8 < 54 inches < 60 inches
| |
| : f. Manual Initiation N.A. N.A.
| |
| : 4. AUTOMATIC DEPRESSURIZATION SYSTEM
| |
| , D> a. Reactor'Vessel Water Level - Low.Low Low, I.
| |
| aD "
| |
| Level 1 > -129 inches* > -136 inches CD b.-.Drywell Pressure - High < 1.68 psig < 1.88 psig 03
| |
| : c. ADS Timer < 105 seconds < 117 seconds
| |
| : d. ';Core Spray Pump Discharge Pressure - High *> 145 psig,(increasing) 5 125 psig, (increasing),
| |
| o £
| |
| : e. RHR LPCI Mode Pump Discharge Pressure-High > 125 psig,(increasing)- 5 115 psig, (increasing) 0 . f. ' Reactor Vessel Water Level-Low, Level 3 > 11.0 inches co > 12.5 inches
| |
| .9
| |
| : g. Manual Initiation N.A; N.A.
| |
| : h. ADS Drywell Pressure Bypass Timer < 420 seconds < 450 seconds W
| |
| *See Bases Figure B 3/4.3-1.
| |
| **Corresponds to 2.3 feet indicated.
| |
| 1
| |
| | |
| aJ
| |
| - % v l We;/m
| |
| >siid* . . 2esa)
| |
| . . . .. ; .TABLE 3.3.3-2 (Continued) . *,
| |
| -7 . EMERGENCY CORE- COOLING SYSTEM ACTUATION'INSTRUMENTATION SETPOINTS.! . ;
| |
| ' ' ',,ALLOWABLE-
| |
| * Cr TRIP FUNCTION.- TRIP SETPOINT' VALUE' 0*..
| |
| I* 5. 4LQSS OFPOWER ; RELAY t
| |
| : a. 4.16 kV Emergency Bus Undervoltage NA ,7NA-'
| |
| -!r; - ; , I , , , ,.
| |
| b.-' 4.16'kV Emergency Bus Undervoltage RELAY
| |
| ,(Degrade~dVotage) ,,,, ;, , 127-11XOX a. 4.16 kV Basis ,
| |
| - .. , . .' . and . 2905 +/- 115'volts '2905 1145'volts
| |
| - ' . .- ' ' - '102-11XOX b.; 120 V.Basis . t4 l fl2
| |
| '+'Aiunlte 4
| |
| : 0. L W wUra I ,,,.,_- ,,,,
| |
| : c. .1 second time <1.5'second time
| |
| 'elay delay 127Y-11XOXA* a. 4.16 kV Basis
| |
| .Wp
| |
| -and 3640 +/- 91 volts -'3640 +/- 182 volts.
| |
| 127Y-1-l1XOX b. ,120 V Basis-I .*.. '
| |
| ,''-104 +/-'3 volts 104 +/- 5.2 volts,',
| |
| I, I " '. C. <52 second time <60 second time B'elay delay' ,.
| |
| 127Z-11XOX a. 4.16 kV Basis
| |
| . .0
| |
| ---- and 162Y-IIXOX b.
| |
| 3910 +/- 11 volts 120 V Basis 3910 i 19 volts
| |
| . II 111.7 +/- 0.3 volts 111.7 +/- 0.5 volts
| |
| .. C. < 10 secondtime < 11 second-time 4:0 (0
| |
| S
| |
| . . elay Uelay I 127Z-l1XOX a. 4.16 kV Basis and 3910 + .11 volts 3910 +/- 19 volts 162Z- 11XOX b. 120 V Basis
| |
| -c.,-
| |
| 111.7 +/- 0.3 volts
| |
| <'61 second time 111.7_+/- 0.5 volts,
| |
| <;64'second time I
| |
| delay delay.
| |
| "AThis is an inverse time delay voltage relay. The voltages shown are the-,maximum that will not result in
| |
| * a trip. Some voltage conditions will result in decreased trip times. '
| |
| .. ,1).. 3
| |
| | |
| TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds)
| |
| : 1. CORE SPRAY SYSTEM s 27# I
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM s 40# I
| |
| : 3. AUTOMATIC DEPRESSURIZATION SYSTEM N.A.
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM s 60# I
| |
| : 5. LOSS OF POWER N.A.
| |
| # ECCS actuation instrumentation is eliminated from response time testing.
| |
| LIMERICK - UNIT 1 3/4 3-39 Amendment No. 44i,132 DEC 1 4 1998
| |
| | |
| TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK(a) TEST(a) CALIBRATION(a) SURVEILLANCE REQUIRED
| |
| : 1. CORE SPRAY SYSTEM
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low Low, Level 1 1, 2, 3
| |
| : b. Drywel l Pressure - High 1, 2, 3 C. Reactor Vessel Pressure - Low 1, 2, 3
| |
| : d. Manual Initiation N.A. N.A. 1, 2, 3
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low Low, Level 1 1, 2, 3
| |
| : b. Drywell Pressure - High 1, 2, 3 C. Reactor Vessel Pressure - Low 1, 2, 3
| |
| : d. Injection Valve Differential Pressure - Low (Permissive) 1, 2, 3
| |
| : e. Manual Initiation N.A. N.A. 1, 2, 3
| |
| : 3. HIGH PRESSURE COOLANT INJECTION SYSTEM***
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low, Level 2 1, 2, 3
| |
| : b. Drywell Pressure - High 1, 2, 3 C. Condensate Storage Tank Level -
| |
| Low 1, 2, 3
| |
| : d. Suppression Pool Water Level -
| |
| High 1, 2, 3
| |
| : e. Reactor Vessel Water Level -
| |
| High, Level 8 1, 2, 3
| |
| : f. Manual Initiation N.A. N.A. 1, 2, 3 LIMERICK - UNIT 1 3/4 3-40 Amendment No. BJ, -7+, ~ . 227
| |
| | |
| TABLE 4.3.3.1-1 (Continued)
| |
| EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHE..G.KW TEST( a) CALIBRATION(a) SURVEILLANCE REQUIRED
| |
| : 4. AUTOMATIC DEPRESSURIZATION SYSTEM#
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low Low, Level 1 1, 2, 3
| |
| : b. Drywel l Pressure - High 1, 2, 3 C. ADS Timer N.A. 1, 2, 3
| |
| : d. Core Spray Pump Discharge Pressure - High 1, 2, 3
| |
| : e. RHR LPCI Mode Pump Discharge Pressure - High 1, 2, 3
| |
| : f. Reactor Vessel Water Level - Low, Level 3 1, 2, 3
| |
| : g. Manual Initiation N.A. N.A. 1, 2, 3
| |
| : h. ADS Drywell Pressure Bypass Timer N. A. 1, 2, 3
| |
| : 5. LOSS OF POWER
| |
| : a. 4.16 kV Emergency Bus Under-voltage (Loss of Voltage)~ N.A. N.A. 1, 2, 3, 4**, 5**
| |
| : b. 4.16 kV Emergency Bus Under -
| |
| voltage (Degraded Voltage) 1, 2, 3, 4**, 5**
| |
| (a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| * DELETED
| |
| ** Required OPERABLE when ESF equipment is required to be OPERABLE.
| |
| *** Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.
| |
| # Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
| |
| ## Loss of Voltage Relay 127-llX is not field setable.
| |
| LIMERICK - UN IT 1 3/4 3-41 Amendment No. -&J, ~ . 227
| |
| | |
| INSTRUMENTATION 3/4.3.3.A REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC) INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.A The RPV Water Inventory Control (WIC) instrumentation channels shown in Table 3.3.3.A-l shall be OPERABLE.
| |
| APPLICABILITY: As shown in Table 3.3.3.A-l ACTION:
| |
| : a. With one or more channels inoperable in a trip system, take the ACTION referenced in Table 3.3.3.A-l for the trip system.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.1.A Each RPV Water Inventory Control (WIC) instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and LOGIC SYSTEM FUNCTIONAL TEST as shown in Table 4.3.3.A-l and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.3.A-l.
| |
| LIMERICK - UNIT 1 3/4 3-4la Amendment No. 227
| |
| | |
| TABLE 3.3.3.A-l RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRIP OPERATIONAL TRIP FUNCTION FUNCTION CONDITIONS ACTION
| |
| : 1. CORE SPRAY SYSTEM
| |
| : a. Reactor Vessel Pressure - Low (Permissive) 5(a) 4, 5 39
| |
| : b. Manual Initiation 2(a)(c) 4, 5 40
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
| |
| : a. Injection Valve Differential Pressure - Low (Permissive) 1/val ve(a) 4, 5 39
| |
| : b. Manual Initiation 1(a) 4, 5 40
| |
| : 3. RHR SYSTEM SHUTDQ_WN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low - Level 3 2 in one ( b) 38 trip system
| |
| : 4. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low, Low - Level 2 2 in one ( b) 38 trip system (a) Associated with an ECCS subsystem required to be OPERABLE by LCD 3.5.2, "REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC)."
| |
| (b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.
| |
| (c) The manual initiation push buttons start the respective core spray pump and diesel generator. The "A" and "B" logic manual push buttons also actuate an initiation permissive in the injection valve opening logic.
| |
| LIMERICK - UNIT 1 3/4 3-4lb Amendment No. 227
| |
| | |
| TABLE 3.3.3.A-1 (Continued)
| |
| RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION ACTION STATEMENTS ACTION 38 - Declare the associated trip system for the penetration flow path(s) incapable of automatic isolation and calculate DRAIN TIME.
| |
| ACTION 39 - Within 1 hour, place channel in trip. Otherwise, declare associated low pressure ECCS injection/spray subsystem inoperable.
| |
| ACTION 40 - Within 24 hours, restore channel to OPERABLE status. Otherwise, declare associated low pressure ECCS injection/spray subsystem inoperable.
| |
| LIMERICK - UNIT 1 3/4 3-41c Amendment No. 227
| |
| | |
| TABLE 3.3.3.A-2 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION VALUE
| |
| : 1. CORE SPRAY SYSTEM
| |
| : a. Reactor Vessel Pressure - Low (Permissive) ~ 435 psig (decreasing)
| |
| : b. Manual Initiation N.A.
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
| |
| : a. Injection Valve Differential Pressure - Low (Permissive) ::s 84 psid
| |
| : b. Manual Initiation N.A.
| |
| : 3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low - Level 3 ~ 11.0 inches
| |
| : 4. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low, Low - Level 2 ~ -45 inches LIMERICK - UNIT 1 3/4 3-4ld Amendment No. 227
| |
| | |
| TABLE 4.3.3.A-1 RPV WATER INVENTORY CONTROL (WIC) INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL LOGIC SYSTEM OPERATIONAL CHANNEL FUNCTIONAL FUNCTIONAL CONDITIONS FOR WHICH TRIP FUNCTION CHECK( a) TEST( a) TEST(a) SURVEILLANCE REQUIRED
| |
| : 1. CORE SPRAY SYSTEM
| |
| : a. Reactor Vessel Pressure - Low (Permissive) N.A. 4, 5
| |
| : b. Manual Initiation N.A. N.A. 4, 5
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
| |
| : a. Injection Valve Differential Pressure - Low (Permissive) N.A. 4, 5
| |
| : b. Manual Initiation N.A. N.A. 4, 5
| |
| : 3. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low - Level 3 N.A. ( b)
| |
| : 4. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low, Low - Level 2 N.A. ( b)
| |
| (a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| (b) When automatic isolation of the associated penetration flow path(s) is credited in calculating DRAIN TIME.
| |
| LIMERICK - UNIT 1 3/4 3-41e Amendment No. 227
| |
| | |
| INSTRUMENTATION 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1.
| |
| ACTION:
| |
| : a. With an ATWS recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel trip setpoint adjusted consistent with the Trip Setpoint value.
| |
| : b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel(s) in the tripped condition within 24 hours.
| |
| : c. With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:
| |
| : 1. If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure channel, place both inoperable channels in the tripped condition within 24 hours, or, if this action will initiate a pump trip, declare the trip system inoperable.
| |
| : 2. If the inoperable channels include two reactor vessel water level channels or two reactor vessel pressure channels, declare the trip system inoperable.
| |
| : d. With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours or be in at least STARTUP within the next 6 hours.
| |
| : e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour or be in at least STARTUP within the next 6 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.3.4.1.1 Each of the required ATWS recirculation pump trip system instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.
| |
| 4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 3-42 Amendment No. 0, -4L, 186
| |
| | |
| TABLE 3.3.4.1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER TRIP FUNCTION TRIP SYSTEM *
| |
| : 1. Reactor Vessel Water Level -
| |
| Low Low, Level 2 2
| |
| : 2. Reactor Vessel Pressure - High 2
| |
| * One channel may be placed in an inoperable status for up to 6 hours for I required surveillance provided the other channel is OPERABLE.
| |
| LIMERICK - UNIT 1 3/4 3-43 Amendment No. 70 j.:Od;I r, ~6 t 9
| |
| | |
| -n TABLE 3.3.4.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE
| |
| > TRIP FUNCTION SETPOINT VALUE
| |
| : 1. Reactor Vessel, Water Level -
| |
| Low Low, Level 2 2 -38 inches* 2 -45 inches
| |
| : 2. Reactor Vessel Pressure - High < 1149 psig < 1156 psig I
| |
| * See Bases Figure B3/4 3-1.
| |
| LIMERICK - UNIT 1 3/4 3-44 Amendment No. 106 OU AVLfadut -bySbAtp W -JAN 2 4 1996
| |
| _ XV kog oto I
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED I LIMERICK - UNIT 1 3/4 3-45 Amendment No. 740,186
| |
| | |
| INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION 3.3.4.2 The end-of-cycle recirculation pump trip CEOC-RPT) system instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2 3.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 29.5% of RATED THERMAL POWER.
| |
| ACTION:
| |
| : a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.
| |
| : b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channelCs) in the tripped condition within 12 hours.
| |
| : c. With the number of OPERABLE channels two or more less than uired by the Minimum OPERAB Channels r Trip System requirement for one trip system and:
| |
| : 1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 12 hours.
| |
| : 2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.
| |
| : d. With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours or take the ACTION required by Specification 3.2.3.
| |
| : e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or take the ACTION required by Specification 3.2.3.
| |
| LIMERICK - UNIT 1 3/4 3-46 Amendment No. +Q, 201
| |
| | |
| INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.4.2.1 Each of the required end-of-cycle recirculation pump trip system instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST, including trip system logic testing, and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.
| |
| 4.3.4.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.
| |
| 4.3.4.2.3 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table 3.3.4.2-3 shall be demonstrated to be within its limit in accordance with the Surveillance Frequency Control Program. Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are tested in accordance with the Surveillance Frequency Control Program. The measured time shall be added to the most recent breaker arc suppression time and the resulting END-OF-CYCLE-RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be verified to be within its limit.
| |
| 4.3.4.2.4 The time interval necessary for breaker arc suppression from energi-zation of the recirculation pump circuit breaker trip coil shall be measured in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 3-47 Amendment No. ;-14,186
| |
| | |
| TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM OPERABLE CHANNELS TRI P FUNCTION PER TRIP SYSTEM*
| |
| : 1. Turbine Stop Valve - Closure 2**
| |
| : 2. Turbine Control Valve-Fast Closure 2**
| |
| * A trip system may be placed in an inoperable status for up to 6 hours for required surveillance provided that the other trip system is OPERABLE.
| |
| ** This function shall be automatically bypassed when turbine first stage pressure is equivalent to THERMAL POWER LESS than 29.5% of RATED THERMAL POWER.
| |
| LIMERICK - UNIT 1 3/4 3-48 Amendment No. ~, 201
| |
| | |
| TABLE 3.3.4.2-2 END-OF-CYCLE RECIRCULATION PUMP TRIP SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE :
| |
| : 1. Turbine Stop Valve-Closure <-5% closed <,7% closed
| |
| : 2. Turbine Control Valve-Fast Closure-. > 500 psig '- > 465 psig j
| |
| LIMERICK - UNIT 1 3/4 3~-49
| |
| | |
| TABLE 3.3.4.2-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Milliseconds)
| |
| : 1. Turbine Stop Valve-Closure < 175
| |
| : 2. Turbine Control Valve-Fast Closure < 175
| |
| - I LIMERICK - UNIT 1 3/4 3-50
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED I LIMERICK - UNIT 1 3/4 3-51 Amendment No. -0,186
| |
| | |
| k INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumentation channels shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig.
| |
| ACTION:
| |
| : a. With a RCIC system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
| |
| : b. With one or more RCIC system actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.5-1.
| |
| SURVEILLANCE REQUIREMENTS 4.3.5.1 Each of the required RCIC system actuation instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.
| |
| CHANNEL CHECK and CHANNEL CALIBRATION are not required for manual initiation.
| |
| 4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 3-52 Amendment No. 4-1, 186
| |
| | |
| TABLE 3.3.5-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS FUNCTIONAL UNITS PER TRIP FUNCTION* ACTION
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low, Level 2 4# 50
| |
| : b. Reactor Vessel Water Level -
| |
| High, Level 8 4# 51
| |
| : c. Condensate Storage Tank Water Level - Low 2"'* 52
| |
| : d. Manual Initiation## 1/system*** 53
| |
| *A channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition provided all other channels monitoring that parameter are OPERABLE.
| |
| **One trip system with one-out-of-two logic.
| |
| ***One trip system with one channel.
| |
| #One trip system with one-out-of-two twice logic.
| |
| ##The injection function of Manual Initiation is not required to be OPERABLE with reactor steam dome pressure less than 550 psig.
| |
| LIMERICK - UNIT 1 3/4 3-53 Amendment No. 3-, 224
| |
| | |
| TABLE 3.3.5-1 (Continued)
| |
| REACTOR CORE ISOLATION COOLING SYSTEM ACTION STATEMENTS ACTION 50 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
| |
| : a. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours or declare the RCIC system inoperable.
| |
| : b. With more than one channel inoperable, declare the RCIC system inoperable.
| |
| ACTION 51 With the number of OPERABLE channels less than required by the minimum OPERABLE channels per Trip System requirement, declare the RCIC system inoperable within 24 hours.
| |
| ACTION 52 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, place at least one inoperable channel in the tripped condition within 24 hours or declare the RCIC system inoperable.
| |
| ACTION 53 - With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel to OPERABLE status within 24 hours or declare the I RCIC system inoperable.
| |
| LIMERICK - UNIT I 314 3-54 Amendment No. 53
| |
| . ', .. . , . V~ -Piro /II, /? 4/
| |
| | |
| TABLE 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNITS TRIP SETPOINT -VALUE
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low, Level 2 >-38 inches* >-45 inches
| |
| : b. Reactor Vessel Water Level -
| |
| High, Level 8 ; < 54 inches < 60 inches
| |
| : c. Condensate Storage Tank Level ^
| |
| Low > 135.8** inches > 132.3 inches
| |
| : d. Manual Initiation N.A. N.A.
| |
| *See Bases Figure B 3/4.3-1.
| |
| *"Corresponds to 2.3 feet indicated.
| |
| LIMERICK - UNIT 1 3/4 3-55 Amendment No. 33 I
| |
| I , d-v i\ ......;.;-.A
| |
| . I .. . .. 1/4; oCr 3019
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED I ,
| |
| LIMERICK - UNIT-1 3/4 3-56 Amendment No *.3, 186
| |
| | |
| INSTRUMENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.
| |
| APPLICABILITY: As shown in Table 3.3.6-1.
| |
| ACTION:
| |
| : a. With a control rod block instrumentation channel trip setpoint** less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
| |
| : b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.
| |
| SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE* by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown in Table 4.3.6-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.6-1.
| |
| * A channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition, provided at least one other operable channel in the same trip system is monitoring that parameter.
| |
| **The APRM Simulated Thermal Power - Upscale Functional Unit need not be declared inoperable upon entering single reactor recirculation loop operation provided that the flow-biased setpoints are adjusted within 6 hours per Specification 3.4.1.1.
| |
| LIMERICK - UNIT 1 3/4 3-57 Amendment No. 0, 43-1, 44-1, 186
| |
| | |
| Tj4 6-1I CONTROL ROD BLOCK INSTRUMENTATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION
| |
| : 1. ROD BLOCK MONITOR (
| |
| : a. Upscale 2 1* 60
| |
| : b. Inoperative 2 1* 60
| |
| : c. Downscale 2 1* 60
| |
| : 2. APRM
| |
| : a. Simulated Thermal Power - Upscale 3 1 61
| |
| : b. Inoperative 3 1, 2 61
| |
| : c. Neutron Flux - Downscale 3 1 61
| |
| : d. Simulated Thermal Power - Upscale (Setdown) 3 2 61
| |
| : e. Recirculation Flow - Upscale 3 1 61
| |
| : f. LPRM Low Count 3 1, 2 61
| |
| : 3. SOURCE RANGE MONITORS ***
| |
| : a. Detector not full in"b' . 3 2 61 2 5 61
| |
| : b. Upscale(c) 3 2 61 2 5 61
| |
| : c. Inoperative(") 3 2 61 2 5 61
| |
| : d. Downscale'd) 3 2 61 2 5 61
| |
| : 4. INTERMEDIATE RANGE MONITORS
| |
| : a. Detector not full in 6 2, 5** 61
| |
| : b. Upscale 6 2, 5** 61
| |
| : c. Inoperative 6 2, 5** 61
| |
| : d. Downscalee') 6 2, 5** 61
| |
| : 5. SCRAM DISCHARGE VOLUME
| |
| : a. Water Level-High 2 1, 2, 5** 62
| |
| : 6. DELETED DELETED DELETED DELETED
| |
| : 7. REACTOR MODE SWITCH SHUTDOWN POSITION 2 3, 4 63 LIMERICK - UNIT 1 3/4 3-58 Amendment No. 4, 41, -44, 177
| |
| | |
| TABLE 3.3.6-1 (Continued)
| |
| CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 Declare the affected RBM channel inoperable and take the ACTION I required by Specification 3.1.4.3.
| |
| ACTION 61 With the number of OPERABLE Channels:
| |
| : a. One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 12 hours or place the inoperable channel in the tripped condition.
| |
| : b. Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped.condition within one hour.
| |
| ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 12 hours.
| |
| ACTION 63 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block.
| |
| NOTES
| |
| * For OPERATIONAL CONDITION of Specification 3.1.4.3.
| |
| ** With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| * These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.
| |
| (a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of RATED THERMAL POWER.
| |
| (b) This function shall be automatically bypassed if detector count rate is
| |
| > 100 cps or the IRM channels are on range 3 or higher.
| |
| (c) This function is automatically bypassed when the associated IRM channels are on range 8 or higher.
| |
| (d) This function is automatically bypassed when the IRM channels are on range 3 or higher.
| |
| (e) This function is automatically bypassed when the IRM channels are on range 1.
| |
| (f) DELETED LIMERICK - UNIT 1 3/4 3-59 Amendment No. 4. X1, 00, 70, 141 APR 1 2 200D
| |
| | |
| TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRI P FUNCTI ON TRIP SETPOINT ALLOWABLE VALUE
| |
| : 1. ROD BLOCK M9NITOR
| |
| : a. Upscale a )
| |
| : 1) Low Tri p Setpoi nt (LTSP) * *
| |
| : 2) Intermediate Trip Setpoint (ITSP) * *
| |
| : 3) High Trip Setpoint (HTSP) * *
| |
| : b. Inoperative N/A N/A
| |
| : c. Downscale (DTSP) * *
| |
| : d. Power Range Setpo
| |
| : 1) Low Power Setpoint (LPSP) .1% RATED THERMAL POWER 28.4% RATED THERMAL POWER
| |
| : 2) Intermediate Power Setpoint (IPSP) 63.1% RATED THERMAL POWER 63.4% RATED THERMAL POWER
| |
| : 3) High Power Setpoint (HPSP) 83.1% RATED THERMAL POWER 83.4% RATED THERMAL
| |
| : 2. APRM
| |
| -a-.-Simul ated Thermal Power cale:
| |
| Two Recirculation ration ~ 0.65 W+ 54.3% and ~ 0.65 W+ 54.7% and
| |
| ~ 108.0% of RATED ~ 108.4% RATED THERMAL POWER THERMAL POWER Single Recirculation Loop Operation**** ~ 0.65 CW-7.6%) + 54.1% and ~ 0.65 CW-7.6%) + 54.5% and
| |
| ~ 108.0% of RATED ~ 108.4% of RATED THERMAL POWER THERMAL POWER
| |
| : b. Inoperative N.A. N.A.
| |
| : c. Neut ron Fl ux - Downsca I e ~ 3.2% of RATED THERMAL ~ 2.8% of RATED THERMAL POWER POWER
| |
| : d. Simulated Thermal Power - Upscale ~ 12.0% of THERMAL ~ 13.0% of RATED THERMAL (Setdown) POWER POWER
| |
| : e. Recirculation Flow - Upscale * *
| |
| : f. LPRM Low Count < 20 per channel < 20 per channel
| |
| < 3 per axial level < 3 per axi al 1 evel
| |
| : 3. SOURCE RANGE MONITORS
| |
| : a. Detector not full in N.A. N.A.
| |
| : b. Upscale ~ 1 x 10 5 cps ~ 1. 6 X lU cps J
| |
| : c. Inoperative N.A. N.A.
| |
| : d. Downsca 1e ~ 3 cps** ~ 1.8 cps**
| |
| LIMERICK UNIT 1 3/4 3 60 No. +,W,JG.,J+,
| |
| &&,+Ge,+/-4+/-,+/-++,201
| |
| | |
| TABLE 3.3.6- ntinued)
| |
| CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE
| |
| : 4. INTERMEDIATE RANGE MONITORS
| |
| : a. Detector not full in N.A. N.A.
| |
| : b. Upscale
| |
| * 108/125 divisions of
| |
| * 110/125 divisions of full scale full scale
| |
| : c. Inoperative N.A. N.A.
| |
| : d. Downscale 2 5/125 divisions of full > 3/125 divisions of full scale scale
| |
| : 5. SCRAM DISCHARGE VOLUME
| |
| : a. Water Level-High
| |
| * 257' 5 9/16" elevation***
| |
| * 257' 7 9/16" elevation
| |
| : a. Float Switch
| |
| : 6. DELETED DELETED DELETED
| |
| : 7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. N.A.
| |
| * Refer to the COLR for these setpoints.
| |
| ** May be reduced provided the Source Range Monitor has an observed count rate and signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.
| |
| *** Equivalent to 13 gallons/scram discharge volume.
| |
| **** The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 2 7.6%. For flows W < 7.6%, the (W-7.6%) term is set equal to zero.
| |
| (a) There are three upscale trip levels. Each is applicable only over its specified operating core thermal power range. All RBM trips are automatically bypassed below the low power setpoint (LPSP). The upscale LTSP is applied between the low power setpoint (LPSP) and the intermediate power setpoint (IPSP). The upscale ITSP is applied between the intermediate power setpoint and the high power setpoint (HPSP).
| |
| The HTSP is applied above the high power setpoint.
| |
| (b) Power range setpoints control enforcement of appropriate upscale trips over the proper core thermal power ranges. The power signal to the RBM is provided by the APRM.
| |
| LIMERICK - UNIT I 3/4 3-60a Amendment No. 3, 30, 34, 3X, 66, 4-4,177
| |
| | |
| a 3.0
| |
| *--Imma-!III-I 2.8 b
| |
| . I
| |
| .4 p9
| |
| *4 - . I 1--a--aa aa.L 2.
| |
| lIII na-i .aa-11111 1iij
| |
| _ ea _
| |
| 0-2.
| |
| I*I 11111 u
| |
| I-el I
| |
| : 1. .iAiaaaa~iiaa a.
| |
| .I
| |
| .aaaaaa.a
| |
| %n 1i
| |
| ;I.4 , a.
| |
| lu aaaaa-a a.
| |
| I-I'. 11111 111 1 11
| |
| : 0. aa aaaa I I 6 I I a I I I 2 6 10 14 is 22 26 30 SIGNL TO NOISE RATIO SRX COUNT RATE VERSUS SIGNAL TO NOISE RATIO
| |
| -N sIGM3.3.6-1 0 LLorick .- Uant I 314 3-60b Amndment No. 34 OEC 13 M9 I
| |
| | |
| TABLE 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK(h) TEST(h) CALIBRATION(a)(h) SURVEILLANCE REQUIRED
| |
| : 1. ROD BLOCK MONITOR
| |
| : a. Upscale N.A. (c) 1*
| |
| : b. Inoperative N.A. (c) N.A. 1*
| |
| : c. Downscale N.A. (c)
| |
| : 2. APRM
| |
| : a. Simulated Thermal Power-Upscale N.A. 1
| |
| : b. Inoperative N.A. N.A. 1, 2 C. Neutron Flux - Downscale N.A. 1
| |
| : d. Simulated Thermal Power -
| |
| Upscale (Setdown) ' N.A. 2
| |
| : e. Recirculation Flow - Upscale N.A. 1
| |
| : f. LPRM Low Count N.A. 1, 2
| |
| : 3. SOURCE RANGE MONITORS
| |
| : a. Detector not full in N.A. (e) N.A. 2, 5
| |
| : b. Upscale N.A. (e) 2, 5
| |
| : c. Inoperative N.A. (e) N.A. 2, 5
| |
| : d. Downscale N.A. (e) 2, 5
| |
| : 4. INTERMEDIATE RANGE MONITORS
| |
| : a. Detector not full in N.A. N.A. 2,
| |
| : b. Upscale N.A. 2,
| |
| : c. Inoperative N.A. N.A. 2,
| |
| : d. Downscale N.A. 2,
| |
| : 5. SCRAM DISCHARGE VOLUME
| |
| : a. Water Level - High N.A. 1, 2, 5** I I
| |
| : 6. DELETED
| |
| : 7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. (g) N.A. 3, 4 LIMERICK - UNIT 1 3/4 3-61 Amendment No. 44, 3, &6, 99, S474, 41-7-, 18 6
| |
| | |
| TABLE 4.3.6-1 (Continued)
| |
| CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
| |
| (b) Deleted.
| |
| (c) Includes reactor manual control multiplexing system input.
| |
| * For OPERATIONAL CONDITION of Specification 3.1.4.3.
| |
| ** With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| *** Deleted.
| |
| (d) Deleted. I (e) The provisions of Specification 4.0.4 are not applicable provided that the surveillance is performed within 12 hours after the IRMs are on Range 2 or below during a shutdown.
| |
| (f) Deleted. I (g) The provisions of Specification 4.0.4 are not applicable provided that the surveillance is performed within 1 hour after the Reactor Mode Switch has been placed in the shutdown position.
| |
| (h) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| LIMERICK - UNIT 1 3/4 3-62 Amendment No. 44, 66, 99, 186 44-1,
| |
| | |
| INSTRUMENTATION 3/4.3.7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1 The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE with their alarm/trip setpoints within the specified limits.
| |
| APPLICABILITY: As shown in Table 3.3.7.1-1.
| |
| ACTION:
| |
| : a. With a radiation monitoring instrumentation channel alarm/trip setpoint exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the limit within 4 hours or declare the channel inoperable.
| |
| : b. With one or more radiation monitoring channels inoperable, take the ACTION required by Table 3.3.7.1-1.
| |
| : c. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.7.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the conditions shown in Table 4.3.7.1-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.7.1-1.
| |
| LIMERICK - UNIT I 3/4 3-63 Amendment No. -14,186
| |
| | |
| TABLE 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP INSTRUMENTATION OPERABLE CONDITIONS SETPOINT ACTION
| |
| : 1. Main Control Room Normal 4 1,2,3, 1 x 10- 5 µCi /cc 70 Fresh Air Supply Radiation and
| |
| * Monitor
| |
| : 2. Area Monitors
| |
| : a. Criticality Monitors
| |
| : 1) Spent Fuel 2 (a) ~ 5 mR/h and :s;20mR/h<bJ 71 Storage Pool
| |
| : b. Control Room Direct 1 At All Times N.A. <bl 73 Radiation Monitor
| |
| : 3. Reactor Enclosure Cooling Water Radiation Monitor 1 At A11 Ti mes :s; 3 x Background<bl 72 LIMERICK - UNIT 1 3/4 3-64 Amendment No. 185
| |
| | |
| TABLE 3.3.7.1-1 (Continued)
| |
| RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS
| |
| *When RECENTLY IRRADIATED FUEL is being handled in the secondary containment with the vessel head removed and fuel in the vessel.
| |
| (a) With fuel in the spent fuel storage pool.
| |
| (b) Alarm only.
| |
| ACTION STATEMENTS ACTION 70 With one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7 days or, within the next 6 hours, initiate and maintain operation of the control room emergency filtration system in the radiation isolation mode of operation.
| |
| With two or more of the monitors inoperable, within one hour, initiate and maintain operation of the control room emergency filtration system in the radiation mode of operation.
| |
| ACTION 71 With one of the required monitor inoperable, assure a portable continuous monitor with the same alarm setpoint is OPERABLE in the vicinity of the installed monitor during any fuel movement.
| |
| If no fuel movement is being made, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours.
| |
| ACTION 72 With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once per 24 hours.
| |
| ACTION 73 With the required monitor inoperable, assure a portable alarming monitor is OPERABLE in the vicinity of the installed monitor or perform area surveys of the monitored area with portable monitor-ing instrumentation at least once per 24 hours.
| |
| LIMERICK - UNIT 1 3/4 3-65 Amendment No. ~ . 227
| |
| | |
| TABLE 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE INSTRUMENTATION CHECK(c) TEST (c) CALIBRATION(c) IS REQUIRED
| |
| : 1. Main Control Room Normal Fresh Air Supply Radiation Monitor 1, 2, 3, and *
| |
| : 2. Area Monitors
| |
| : a. Criticality Monitors
| |
| : 1) Spent Fuel Storage (a)
| |
| Pool
| |
| : b. Control Room Direct At All Times Radiation Monitor
| |
| : 3. Reactor Enclosure Cooling Water Radiation Monitor (b) At All Times LIMERICK - UNIT 1 3/4 3-66 Amendment No. .i-54, 186 47-,
| |
| | |
| TABLE 4.3.7.1-1 (Continued)
| |
| RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS
| |
| *When RECENTLY IRRADIATED FUEL is being handled in the secondary containment with the vessel head removed and fuel in the vessel.
| |
| (al With fuel in the spent fuel storage pool.
| |
| (bl The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
| |
| (cl Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| LIMERICK - UNIT 1 3/4 3-67 Amendment No. ~ . ~ . 227
| |
| | |
| Section 3.3.7.2 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM. TECHNICAL SPECIFICATIONS PAGES 3/4 3-69 THROUGH 3/4 3-72 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 3-68 Amendment No. 44, 4-, --5192
| |
| | |
| Section 3;3.7.3 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 3-74 THROUGH 3/4 3-75 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 3-73 Amendment No. /, 48 I
| |
| Qffb2ZLLC 9auaa i?' / f/
| |
| | |
| INSTRUMENTATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS LIMITING CONDITION FOR-OPERATION 3.3.7.4 The remote shutdown system instrumentation and controls shown in Table 3.3.7.4-1 shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| : a. With the number of OPERABLE remote shutdown system instrumentation channels less than required by Table 3.3.7.4-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. With the number of OPERABLE remote shutdown system controls less than required in Table 3.3.7.4-1, restore the inoperable control(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.3.7.4.1 Each of the above required remote shutdown monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK*
| |
| and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.
| |
| 4.3.7.4.2 Each of the above remote shutdown control switch(es) and control circuits shall be demonstrated OPERABLE by verifying its capability to perform its intended function(s) in accordance with the Surveillance Frequency Control Program. I
| |
| * Control is not required to be transferred to perform the CHANNEL CHECK.
| |
| I LIMERICK - UNIT I 3/4 3-76 Amendment No. ;4-, 46-9, 186
| |
| | |
| TABLE 3.3.7.4-1 3 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS clr
| |
| ;< -MINIMUM I INSTRUMENTS c INSTRUMENT OPERABLE
| |
| : 1. Reactor Vessel Pressure 1
| |
| : 2. Reactor Vessel Water Level 1
| |
| : 3. Safety/Relief Valve Position, 3 valves 1/valve
| |
| : 4. Suppression Chamber Water Level 1
| |
| : 5. Suppression Chamber Water Temperature 1
| |
| : 6. Drywell Pressure 1
| |
| : 7. Drywell Temperature 1 As 8. RHR System Flow 1
| |
| 'V 9. RHR Service Water Pump Discharge Pressure 1 i 10. RHR Heat Exchanger Service Water Outlet Pressure 1
| |
| : 11. RCIC System Flow 1
| |
| : 12. RCIC Turbine Speed 1
| |
| : 13. Emergency Service Water Pump Discharge Pressure 1 C 14. Condensate Storage Tank Level 1
| |
| : 15. RHR Heat Exchanger Bypass Valve (HV51-lF048A) Position Indication (O- 100%) 1
| |
| : 16. RCIC Turbine Tripped Indication 1
| |
| : 17. RCIC Turbine Bearing Oil Pressure Low Indication 1 A_ 18. RCIC LP Bearing Oil Temperature High Indication 1 S 19. RHR Heat Exchanger Discharge Line High Radiation Indication 1 to
| |
| | |
| TABLE 3.3.7.4-1 (Continued) I REMOTE SHUTDOWN SYSTEM CONTROLS RCIC SYSTEM HSS-49-191 Control-Transfer Switch HSS-49-192 Control-Transfer Switch HSS-49-193 Control-Transfer Switch HSS-49-195 Control-Transfer Switch HSS-49-196 Control-Transfer Switch HV-49-IF076 Control-Steam Line warmup bypass valve HV-49-IF060 Control-RCIC turb exhaust to suppression pool isolation HV-50-112 Control-Turb trip throttle valve HV-50-IF045 Control-Turbine steam supply valve HV-49-IF008 Control-Turbine steam line outboard isolation valve HV-49-1F007 Control-Turbine steam line inboard isolation valve HV-49-IF031 Control-RCIC pump suction from suppression pool HV-49-1F029 Control-RCIC pump suction from suppression pool HV-49-IFO1O Control-RCIC pump suction from condensate storage tank HV-49-1F019 Control-Minimum flow bypass to suppression pool HV-49-1F022 Control-Test return to condensate storage tank HV-50-1FO46 Control-RCIC turbine cooling water valve HV-49-1F012 Control-RCIC pump disch valve HV-49-lF013 Control-RCIC pump disch valve 10P220 Control-Vacuum tank condensate pump 10P219 Control-Barometric condenser vacuum pump HV-49-1F002 Control-Barometric condenser vacuum pump disch K-I1_
| |
| LIMERICK - UNIT 1 3/4 3-78 Amendment No. 33 OCT 3 0 1989 I
| |
| | |
| Table 3.3.7.4-1 (Continued)
| |
| RCIC SYSTEM (Continued)
| |
| HV-49-1F080 control-Vacuum breaker outboard isolation valve HV-49-lF084 - Control-Vacuum breaker inboard isolation valve FIC-49-lROOl Controller-RCIC discharge flow con trol E51-S45 RCIC Turbine Trip Bypass NUCLEAR BOILER SYSTEM -
| |
| HSS-41-191 Control -iranster switch aft ;. rr 1 PSV-41-1FO13A Control-Main steam line safety/rel ieI -valve Control-Main steam line s,fet PSV-41-lF013C Control-Miin steam line safety/rel ie1 fvalve I .. !.. , - I .
| |
| PSV-41-lFO13N Control-Main steam line sifiiydrel iel fvalve
| |
| -i ., ; :1..:
| |
| RHR SYSTEM HSS-51-192 .Control-Transferswitch HSS-51-193 Control-Transfer switch HSS 194 Control-Transfer switch HSS-51-195 Control-Transfer switch HSS-51-196 Control-Transfer switch HSS-51-197 Control-Transfer switch HSS-51-198 Control-Transfer switch HV-51-1FO09 Control-RHR pump swhutdownc ooling su:tion inboard isolation isolation HV-51-1FO08 Control-RHR shutdown cooling suction outboard HV-51'1FO06A Control-LA RHR loop shutdown cooling suction HV-51-lF0068 Control-1B RHR loop shutdown cooling suction HV-51-1FO04A Control-1A RHR pump suction 1AP202 Control-1A RHR pump LIMERICK - UNIT 1 3/4 3-79 f . .. .
| |
| | |
| I ZUlt TABLE 3.3.7.4-1 (Continued)
| |
| RHR SYSTEM (Continued)
| |
| HV-43-IF023A Control-Recirculation pump A suction valve HSS-43-191 Control-Transfer switch HV-51-lFO07A Control-lA RHR pump minimum flow bypass valve HV-51-1F048A Control-lA heat exchanger shell side bypass HV-51-IF015A Control-lA shutdown cooling injection valve HV-51-1FO16A Control-Reactor containment spray I
| |
| HV-51-1FO17A Control-lA RHR loop injection valve HV-51-1F024A Control-lA RHR loop test return HV-51-1F027A Control-Suppression pool sparger isolation HV-51-lFO47A Control-lA Heat exchanger shell side inlet HV-51-IF003A Control-IA Heat exchanger shell side outlet HV-51-1F049 Control-RHR Discharge to radwaste outboard isolation HV-51-125A Control-lA/1C test return line to suppression pool RHR SERVICE WATER SYSTEM HSS-12-015A-2 Control-Spray pond/cooling tower select HSS-12-OlSC-2 Control-Spray pond/cooling tower select HSS-12-016A-2 Control-Spray/bypass select HSS-12-016C-2 Control-Spray/bypass select LIMERICK - UNIT 1 3/4 3-80 Amendment No. Z4, B3, 74 AUG 2 3 1994
| |
| | |
| Table 3.3.7.4-1 (Continued)
| |
| RHR SERVICE WATER SYSTEM (Continued)
| |
| HSS-12-094 Control-Transfer switch
| |
| ' HSS-12-093 Control-Transfer switch *
| |
| ' HV-51-1F014A OAP506 Control-lA RHR heat exchanger.tube-side inlet Control-RHR Service Water pump I
| |
| HV-51-1F068A Control-lA RHR Heat exchanger tube side outl et' EMERGENCY SERVIICE WATER SYSTEM
| |
| -OAP548; Control-A emergency service water pump..
| |
| HV-11-O1lA Control-A emergency service water disch to RHR service water-HSS-11-091 Control-Transfer-switch HSS-11-092 Control-Transfer switch
| |
| .' HSS-11-093 Control-Transfer switch The following valves of the ESW and.RHRSW systems are'actuated by signals from the transfer'switches:
| |
| HV-12-005-. ESW and RHRSW pumps wetwell intertie gate
| |
| .HV-11-015A- , ESW loop A discharge to RHRSW loop B HV-12-017A..' ESW and RHRSW cooling tower return cross-tie
| |
| . STANDBY AC POWER SUPPLY 152-,11509/CSR - 101-D11 Safeguard SWGR feeder bkr. -
| |
| 152-11609/CSR 101-012 Safeguard.SWGR feeder bkr. -
| |
| 152-11709/CSR 101-D13 Safeguard.SWGR feeder bkr. .
| |
| '152-11502/CSR- 201-D11 Safeguard SWGR feeder bkr. '
| |
| 152-11602/CSR ' 201-D12 Safeguard SWGR feeder bkr.
| |
| - 152-11702/CSR
| |
| * 201-013 Safeguard SWGR feeder bkr.'
| |
| 152-11505/CSR D114 Safeguard LC XFMR breaker
| |
| ,. , }/. ' . ' : . .
| |
| :-4
| |
| - LIMERICK - UNIT 1 43/4 3-81 .AmendmentNo .33 . .. I OCT 30 1989 .;
| |
| | |
| I pr Table 3.3.7.4-1 (Continued) A\
| |
| STANDBY AC POWER SUPPLY (Continued) 152-11605/CSR 0124 Safeguard LC XFMR breaker 152-11705/CSR D134 Safeguard LC XFMR breaker 143-115/CS Transfer switch 143-116/CS Transfer switch 143-117/CS Transfer switch LIMERICK - UNIT 1 3/4 3-82
| |
| .',I .. , :A .
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 3-83 Amendment No.186
| |
| | |
| INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.5 The accident monitoring instrumentation channels shown in Table 3.3.7.5-1 shall be OPERABLE.
| |
| APPLICABILITY: As shown in Table 3.3.7.5-1.
| |
| ACTION:
| |
| With one or more accident monitoring instrumentation channels inoperable, take the ACTION required by Table 3.3.7.5-1.
| |
| SURVEILLANCE REQUIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.7.5-1.
| |
| LIMERICK - UNIT 1 3/4 3-84 Amendment No.186
| |
| | |
| TABLE 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION MINIMUM APPLICABLE REQUIRED NUMBER CHANNELS OPERATIONAL INSTRUMENT OF CHANNELS OPERABLE CONDITIONS ACTION
| |
| : 1. Reactor Vessel Pressure 2 1 1,2 80
| |
| : 2. Reactor Vessel Water Level 2 1 1,2 80
| |
| : 3. Suppression Chamber Water Level 2 1 1,2 80
| |
| : 4. Suppression Chamber Water Temperature 8, 6 locations 6, 1,2 80 1/location
| |
| : 5. Del eted
| |
| : 6. Drywell Pressure 2 1 1,2 80
| |
| : 7. Deleted
| |
| : 8. Deleted
| |
| : 9. Deleted
| |
| : 10. Deleted
| |
| : 11. Primary Containment Post-LOCA Radiation Monitors 4 2 1,2,3 81
| |
| : 12. North Stack Wide Range Accident Monitor**- 3* 3* 1,2,3 81 13 Neutron Flux 2 1 1,2 80 LIMERICK - UNIT 1 3/4 3-85 Amendment No. 2-4, 4-5-1, 4-73, 4-7 -191
| |
| | |
| Table 3.3.7.5-1 (Continued)
| |
| ACCIDENT MONITORING INSTRUMENTATION TABLE NOTATIONS
| |
| *Three noble gas detectors with overlapping ranges (10' to 10', 104 to 10, 10' to 105 sLCi/cc).
| |
| **High range noble gas monitor.
| |
| ACTION STATEMENTS ACTION 80 -
| |
| : a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| ACTION 81 - With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitor-ing the appropriate parameters within 72 hours, and
| |
| : a. Either restore the inoperable channel(s) to OPERABLE status within 7 days of the event, or
| |
| : b. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
| |
| ACTION 82 - DELETED LIMERICK - UNIT 1 3/4 3-86 Amendment Mo. 4-54, 173
| |
| | |
| TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK(a) CALIBRATION(a)
| |
| : 1. Reactor Vessel Pressure
| |
| : 2. Reactor Vessel Water Level
| |
| : 3. Suppression Chamber Water Level
| |
| : 4. Suppression Chamber Water Temperature
| |
| : 5. Deleted
| |
| : 6. Primary Containment Pressure
| |
| : 7. Deleted
| |
| : 8. Deleted
| |
| : 9. Deleted
| |
| : 10. Deleted
| |
| : 11. Primary Containment Post LOCA Radiation Monitors **
| |
| : 12. North Stack Wide Range Accident Monitor***
| |
| : 13. Neutron Flux (a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| **CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source.
| |
| ***High range noble gas monitors.
| |
| LIMERICK - UNIT 1 3/4 3-87 Amendment No. 4-1-6, 4-7-3, --7-a, 4-46191
| |
| | |
| INSTRUMENTATION SOURCE RANGE MONITORS LIMITING CONDITION FOR OPERATION 3.3.7.6 At least the following source range monitor channels shall be OPERABLE:
| |
| : a. In OPERATIONAL CONDITION 2*, three.
| |
| : b. In OPERATIONAL CONDITION 3 and 4, two.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 2*, 3, and 4.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 2* with one of the above required source range monitor channels inoperable, restore at least three source range monitor channels to OPERABLE status within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. In OPERATIONAL CONDITION 3 or 4 with one or more of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour.
| |
| SURVEILLANCE REOUIREMENTS 4.3.7.6 Each of the above required source range monitor channels shall be demonstrated OPERABLE by:
| |
| : a. Performance of a:
| |
| : 1. CHANNEL CHECK in accordance with the Surveillance Frequency Control Program:
| |
| a) In CONDITION 2*, AND b) In CONDITION 3 or 4.
| |
| : 2. CHANNEL CALIBRATION** in accordance with the Surveillance Frequency Control Program.
| |
| : b. Performance of a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.
| |
| : c. Verifying, prior to withdrawal of control rods, that the SRM count rate is at least 3.0 cps*** with the detector fully inserted.
| |
| *With IRM's on range 2 or below.
| |
| **Neutron detectors may be excluded from CHANNEL CALIBRATION.
| |
| ***May be reduced, provided the source range monitor has an observed count rate and signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.
| |
| LIMERICK - UNIT 1 3/4 3-88 Amendment No. 34, -4, 9986
| |
| | |
| INSTRUMENTATION Section 3/4.3.7.7 THE INFORMATION FROM THIS TECHNICAL SPECIFICATION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM)
| |
| LIMERICK - UNIT 1 3/4 3-89 Amendment No. 11, 117 JUN 11 1996
| |
| | |
| INSTRUMENTATION CHLORINE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8.1 Two independent chlorine detection system subsystems shall be OPERABLE with their alarm and trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 0.5 ppm APPLICABILITY: All OPERATIONAL CONDITIONS.
| |
| ACTION:
| |
| : a. With one chlorine detection subsystem inoperable, restore the inoperable detection system to OPERABLE status within 7 days or, within the next 6 hours, initiate and maintain operation of at least one control room emergency filtration system subsystem in the chlorine isolation mode of operation.
| |
| : b. With both chlorine detection subsystems inoperable, within 1 hour initiate and maintain operation of at least one control room emer-gency filtration system subsystem in the chlorine isolation mode of operation.
| |
| SURVEILLANCE REQUIREMENTS 4.3.7.8.1 Each of the above required chlorine detection system subsystems shall be demonstrated OPERABLE by performance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 3-90 Amendment No. 4-ý, -, ;4, 186
| |
| | |
| INSTRUMENTATION TOXIC GAS DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8.2 Three independent toxic gas detection system subsystems shall be OPERABLE with their alarm setpoints adjusted to actuate at a toxic gas concen-tration of less than or equal to:
| |
| MONITOR SET POINT CHEMICAL (ppm)
| |
| Ammonia 25 Ethylene Oxide 50 Formaldehyde 5 Vinyl Chloride 10 Phosgene 0.4 APPLICABILITY: All OPERATIONAL CONDITIONS.
| |
| ACTION:
| |
| : a. With one toxic gas detection subsystem inoperable, place the inoperable subsystem in the tripped condition within 24 hours.
| |
| : b. With two toxic gas detection system subsystems inoperable, place one inoperable subsystem in the tripped condition within 1 hour, restore one inoperable detection subsystem to OPERABLE status within 7 days, or initiate and maintain operation of at least one control room emergency filtration subsystem in the chlorine isolation mode of operation.
| |
| : c. With three toxic gas detection subsystems inoperable, within 1 hour initiate and maintain operation of at least one control room emergency filtration subsystem in the chlorine isolation mode of operation.
| |
| SURVEILLANCE REOUIREMENTS 4.3.7.8.2 Each of the above required toxic gas detection system subsystems shall be demonstrated OPERABLE by performance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 3-91 Amendment No. 44, .4-1, 84,186
| |
| | |
| INSTRUMENTATION Section 3/4.7.9 (Deleted) I THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM) FIRE PROTECTION SECTION. TECHNICAL SPECIFICATIONS PAGES 3/4 3-92 THROUGH 3/4 3-96 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 3-92 Amendment No. 6q,104 NO'V 2 0 1993
| |
| | |
| INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 3-97 Amendment No. 44, At,. 153 SEP 1 2001
| |
| | |
| Section 3:3.7.1 (Deleted) I1 THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCH. TECHNICAL SPECIFICATIONS-PAGES 3/4 3-99 THROUGH 3/4 3-102 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| I Amendment No. /948 LIMERICK - UNIT 1 3/4 3-98 AI I I al OfAaJlJjJ a J9q9
| |
| | |
| Section 3/4.3.7.12 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM AND THE TRM. TECHNICAL SPECIFICATIONS PAGES 3/4 3-104 THROUGH 3/4 3-108 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 3-103 Amendment No. 48, -+J-1., ~ , ~ , 228
| |
| | |
| INTENTIONALLY LEFT BLANK LIMERICK - UNIT 1 3/4 3-109 Amendment No.48 I
| |
| ,o~tmaL qaatLaI c?, / qq9 I
| |
| | |
| Section 3/4.3.8 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM.
| |
| TECHNICAL SPECIFICATIONS PAGE 3/4 3-111 HAS BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 3-110 Amendment No. 4-9-9, 192
| |
| | |
| INSTRUMENTATION 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The feedwater/main turbine trip system actuation instrumentation channels shown in the Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9-2.
| |
| APPLICABILITY: As shown in Table 3.3.9-1.
| |
| ACTION:
| |
| : a. With a feedwater/main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.9-2, declare the channel inoper-able and either place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip set-point adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable.
| |
| : b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours.
| |
| : c. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE status within 72 hours or be in at least STARTUP within the next 6 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.3.9.1 Each of the required feedwater/main turbine trip system actuation instrumentation channels shall be demonstrated OPERABLE* by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.
| |
| 4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed in accordance with the Surveillance Frequency Control Program.
| |
| * A channel may be placed in an inoperable status for up to 6 hours for required surveillance without placing the trip system in the tripped condition.
| |
| LIMERICK - UNIT 1 3/4 3-112 Amendment No. W4., *-4, 186
| |
| | |
| TABLE 3.3.9-1 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE APPLICABLE CHANNELS PER OPERATIONAL TRIP FUNCTION TRIP SYSTEM CONDITIONS
| |
| : 1. Reactor Vessel Water Level-.High, Level 8 4 1* I
| |
| * With Thermal Power greater than or equal to 25% of Rated Thermal Power. I LIMERICK - UNIT I 3/4 3-113 Amendment No. 91 JUN 13 1995
| |
| | |
| TABLE 3.3.9-2 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
| |
| : 1. Reactor Vessel Water Level-High, Level 8 < 54 inches* < 55.5 inches
| |
| *See Bases Figure B 3/4.3-1 LIMERICK - UNIT 1 3/4 3-114
| |
| | |
| {
| |
| INFORMATION ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 3-115 Amendment No. - 94, 186
| |
| | |
| 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.
| |
| APPLICABI LITY: OPERATIONAL CONDITIONS 1* and 2*.
| |
| ACTION:
| |
| : a. With one reactor coolant system recirculation loop not in operation:
| |
| : 1. Within 4 hours:
| |
| : a. Place the recirculation flow control system in the Local Manual mode, and
| |
| : b. Reduce THERMAL POWER to ~ 74.9% of RATED THERMAL POWER, and,
| |
| : c. Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and
| |
| : d. Verify that the differential temperature requirements of Surveillance Requirement 4.4.1.1.5 are met if THERMAL POWER is ~ 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is ~ 50% of rated loop flow, or suspend the THERMAL POWER or recirculation loop flow increase.
| |
| *See Special Test Exception 3.10.4.
| |
| LIMERICK - UNIT 1 3/4 4-1 Amendment No. JG,~,~,~,201
| |
| | |
| REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)
| |
| )ACTION: (Continued)
| |
| : 2. Within 6 hours:
| |
| Reduce the Average Power Range Monitor (APRM) Simulated Thermal Power -
| |
| Upscale Scram and Rod Block Trip Setpoints and Allowable Values, to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, or declare the associated channel(s) inoperable and take the actions required by the referenced specifications.
| |
| : 3. Otherwise be in at least HOT SHUTDOWN within the next 12 hours.
| |
| : b. With no reactor coolant system recirculation loops in operation, initiate measures to place the unit in at least HOT SHUTDOWN within the next 12 hours.
| |
| LIMERICK - UNIT 1 3/4 4-lta Amendment No. 3X, 46, 444, -69,177
| |
| | |
| REACTOR COOLANT SYSTEM 4.4.1.1.1 DELETED 4.4.1.1.2 DELETED 4.4.1.1.3 DELETED 4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, in accordance with the Surveillance Frequency Control Program, verify that:
| |
| : a. Reactor THERMAL POWER is ~ 74.9% of RATED THERMAL POWER,
| |
| : b. The recirculation flow control system is in the Local Manual mode, and
| |
| : c. The speed of the operating recirculation pump is ~ 90% of rated pump speed.
| |
| 4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is ~ 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is ~ 50% of rated loop flow.
| |
| : a. ~ 145°F between reactor vessel steam space coolant and bottom head drain line coolant,
| |
| : b. ~ 50°F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and
| |
| : c. ~ 50°F between the reactor coolant within the loop not in operation and the operating loop.
| |
| The differential temperature requirements of Specification 4.4.1.1.5b. and
| |
| : c. do not apply when the loop not in operation is isolated from the reactor pressure vessel.
| |
| LIMERICK - UNIT 1 3/4 4-2 Amendment No. J,JQ,++/-,+&,++,~,
| |
| ~,+/-++,~, 201
| |
| | |
| CONTENTS OF THIS PAGE HAVE BEEN DELETED LIMERICK - UNIT I 3/4 4-3 Amendment No. 4O, 406,177
| |
| | |
| REACTOR COOLANT SYSTEM
| |
| ~IET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours.
| |
| SlJRVEILLANCE REOUIREMENTS 4.4.1.2 All jet pumps shall be demonstrated OPERABLE as follows:
| |
| : a. During two recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program while greater than 25% of RATED THERMAL POWER by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when both recirculation loop indicated flows are in compliance with Specification 3.4.1.3.
| |
| : 1. The indicated recirculation loop flow differs by more than 10%
| |
| from the established pump speed-loop flow characteristics.
| |
| : 2. The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.
| |
| : 3. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from the established patterns by more than 10%.
| |
| LIMERICK - UNIT 1 3/4 4-4 Amendment No. JG, ~,
| |
| 196
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)
| |
| : b. During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that no two of the follOWing conditions occur:
| |
| : 1. The indicated recirculation loop flow in the operating loop differs by more than 10% from the established pump speed-loop flow characteristics.
| |
| : 2. The indicated total core flow differs by more than 10% from the established total core flow value derived from single recirculation loop flow measurements.
| |
| : 3. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established single recirculation loop patterns by more than 10%.
| |
| : c. The provisions of Specification 4.0.4 are not applicable provided that this surveillance is performed within 24 hours after exceeding 25% of RATED THERMAL POWER and upon entering single recirculation loop operation.
| |
| LIMERICK - UNIT 1 3/4 4-4a Amendment No. JG, +/-&&,
| |
| 196
| |
| | |
| REACTOR COOLANT SYSTEM RECIRCULATION PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation loop flow mismatch shall be maintained within:
| |
| : a. 5% of each other with core flow greater than or equal to 70% of rated core flow.
| |
| : b. 10% of each other with core flow less than 70% of rated core flow.
| |
| APPLICABIIT.JY: OPERATIONAL CONDITIONS 1* and 2* during two recirculation loop operation.
| |
| ACTION:
| |
| With the recirculation loop flows different by more than the specified limits, either:
| |
| : a. Restore the recirculation loop flows to within the specified limit within 2 hours, or
| |
| : b. Shutdown one of the recirculation loops within the next 8 hours and take the ACTION required by Specification 3.4.1.1.
| |
| SURVEILLANCE REOUIREMENTS 4.4.1.3 Recirculation loop flow mismatch shall be verified to be within the limits in accordance with the Surveillance Frequency Control Program.
| |
| *See Special Test Exception 3.10.4.
| |
| LIMERICK - UNIT 1 3/4 4-5 Amendment No. W,186 (GorrecteO
| |
| | |
| 'IZZ-.
| |
| REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal to 1451F, and:
| |
| : a. When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to 501F, or
| |
| : b. When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recircula-tion loops is less than or equal to 500 F and the operating loop flow rate is less than or equal to 50% of rated loop flow.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.
| |
| ACTION: /
| |
| With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop.
| |
| SURVEILLANCE REQUIREMENTS 4.4.1.4 The temperature differentials and flow rate shall be determined to be within the limits within 15 minutes prior to startup of an idle recirculation loop.
| |
| LIMERICK - UNIT 1 3/4 4-6
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY/RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 12 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function lift settings:*#
| |
| 4 safety/relief valves @ 1170 psig +/-3%
| |
| 5 safety/relief valves @ 1180 psig +/-3%
| |
| 5 safety/relief valves @ 1190 psig +/-3%
| |
| APPLICABILIT Y: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With the safety valve function of one or more of the above required safety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| : b. DELETED C. DELETED SURVEILLANCE REQUIREMENTS 4.4.2.1 DELETED 4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previouslyset pressure tested and stored in accordance with manufacturer's recommendations in accordance with the Surveillance Frequency Control Program, and they shall be rotated such that all 14 safety relief valves are removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations in accordance with the Surveillance Frequency Control Program. All safety valves will be recertification tested to meet a +/-1% tolerance prior to returning the valves to service.
| |
| * The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
| |
| # Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling.
| |
| LIMERICK - UNIT 1 3/4 4-7 Amendment No. 6O,7,-O,40,7,4-57,-7-9, 186
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3.4.3.1 The following reactor coolant leakage detection systems shall be OPERABLE:
| |
| : a. The primary containment atmosphere gaseous radioactivity monitoring system,
| |
| : b. The drywell sump monitoring system,
| |
| : c. The drywell unit coolers condensate flow rate monitoring system, and
| |
| : d. The primary containment pressure and temperature monitoring system.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.*
| |
| * The primary containment gaseous radioactivity monitor is not required to be operable until Operational Condition 2.
| |
| ACTIONS:
| |
| A. With the primary containment atmosphere gaseous radioactivity monitoring system inoperable, analyze grab samples of primary containment atmosphere at least once per 12 hours AND restore primary containment atmosphere gaseous radioactivity monitoring system to OPERABLE status within 30 days.
| |
| B. With the drywell sump monitoring system inoperable, restore the drywell sump monitoring system to OPERABLE status within 30 days AND increase monitoring frequency of drywell unit cooler condensate flow rate (SR 4.4.3.2.1.c) to once every 8 hours.
| |
| C. With the drywell unit coolers condensate flow rate monitoring system inoperable, AND the primary containment atmosphere gaseous radioactivity monitoring system OPERABLE, perform a channel check of the primary containment atmosphere gaseous radioactivity monitoring system (SR 4.4.3.1.a) once per 8 hours.
| |
| D. With the primary containment pressure and temperature monitoring system inoperable, restore the primary containment pressure and temperature monitoring system to OPERABLE status within 30 days. NOTE: All other Tech Spec Limiting Conditions For Operation and Surveillance Requirements associated with the primary containment pressure/temperature monitoring system still apply. Affected Tech Spec Sectjons include: 3/4.3.7.5. 4.4.3.2.1, 3/4.6.1.6. and 3/4.6.1.7.
| |
| E. With the primary containment atmosphere gaseous radioactivity monitoring system inoperable AND the drywell unit coolers condensate flow rate monitoring system inoperable, restore the primary containment atmosphere gaseous radioactivity monitoring system to OPERABLE status within 30 days OR restore the drywell unit coolers condensate flow rate monitoring system to OPERABLE status within 30 days.
| |
| With the primary containment atmosphere gaseous radioactivity monitoring system inoperable, analyze grab samples of primary containment atmosphere at least once per 12 hours.
| |
| LIMERICK - UNIT 1 3/4 4-8
| |
| | |
| REACTOR COOLANT SYSTEM F. With the drywell floor drain sump monitoring system inoperable AND the drywell unit coolers condensate flow rate monitoring system inoperable analyze grab samples of the primary containment atmosphere once per 12 hours, AND monitor Reactor Coolant System leakage by administrative means once per 12 hours AND restore either the drywell floor drain sump monitoring system to OPERABLE status within 7 days restore the drywell unit coolers condensate flow rate monitoring system to OPERABLE status within 7 days.
| |
| G. With any other two or more leak detection systems inoperable other than ACTIONS E and F above OR with required Actions and associated Completion Time of ACTIONS A, B, C, D, E or F not met, be in HOT SHUTDOWN within 12 hours AND in COLD SHUTDOWN within the next 24 hours.
| |
| 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated operable by:
| |
| : a. Perform a CHANNEL CHECK of the primary containment atmosphere gaseous radioactivity monitoring system in accordance with the Surveillance Frequency Control Program.
| |
| : b. Perform a CHANNEL FUNCTIONAL TEST of required leakage detection instrumentation in accordance with the Surveillance Frequency Control Program.
| |
| This does not apply to containment pressure and temperature monitoring system.
| |
| : c. Perform a CHANNEL CALIBRATION of required leakage detection instrumentation in accordance with the Surveillance Frequency Control Program. This does not apply to containment pressure and temperature monitoring system.
| |
| : d. Monitor primary containment pressures AND primary containment temperature in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 4-8a Amendment No. ~, ~,205
| |
| | |
| REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FQR OPERATIO_
| |
| 3.4.3.2 Reactor coolant system leakage shall be limited to:
| |
| : a. No PRESSURE BOUNDARY LEAKAGE.
| |
| : b. 5 gpm UNIDENTIFIED LEAKAGE.
| |
| : c. 30 gpm total leakage.
| |
| : d. 25 gpm total leakage averaged over any 24-hour period.
| |
| : e. 1 gpm leakage at a reactor coolant system pressure of 950 +/-10 psig from any reactor coolant system pressure isolation valve.** I
| |
| : f. 2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| : b. With any reactor coolant system leakage greater than the limits in b, c and/or d above, reduce the leakage rate to within the limits within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : c. With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least one other closed manual, deactivated automatic, or check* valves, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : d. With one or more of the high/low pressure interface valve leakage pressure monitors inoperable, restore the inoperable monitor(s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours; restore the inoperable monitor(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : e. With any reactor coolant system leakage greater than the limit in f above, identify the source of leakage within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| * Which have been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.
| |
| ** Pressure isolation valve leakage is not included in any other allowable operational leakage specified in Section 3.4.3.2.
| |
| LIMERICK - UNIT 1 3/14 4-9 Amendment No. 2.8, 49, 4-72, 182
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
| |
| : a. Monitoring the primary containment atmospheric gaseous radioactivity in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage),
| |
| *b. Monitoring the drywell floor drain sump and drywell equipment drain tank flow rate in accordance with the Surveillance Frequency Control Program, I
| |
| : c. Monitoring the drywell unit coolers condensate flow rate in accordance with I the Surveillance Frequency Control Program, I
| |
| : d. Monitoring the primary containment pressure in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage),
| |
| : e. Monitoring the reactor vessel head flange leak detection system in accordance with the Surveillance Frequency Control Program, and
| |
| : f. Monitoring the primary containment temperature in accordance with the I Surveillance Frequency Control Program (not a means of quantifying leakage). I 4.4.3.2.2 Each reactor coolant system pressure isolation valve shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:
| |
| : a. In accordance with the Surveillance Frequency Control Program, and
| |
| : b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.
| |
| The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.
| |
| 4.4.3.2.3 The high/low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints set less than the specified allowable values by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies specified in the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 4-10 Amendment No. -3, 4-9, -74P -4'9 186
| |
| | |
| TABLE 3.4.3.2-1 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM)
| |
| LIMERICK - UNIT 1 3/4 4-11 Amendment No. 44, 182
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.4 (DELETED)
| |
| SPECIFICATIONS THE INFORMATION FROM THIS TECHNICAL TECHNICAL REQUIREMENTS SECTION HAS BEEN RELOCATED TO THE PAGES 3/4 4-13 MANUAL (TRM). TECHNICAL SPECIFICATIONS OMITTED.
| |
| AND 3/4 4-14 HAVE BEEN INTENTIONALLY K>
| |
| K-'
| |
| Amendment No. 169,172 3/4 4-12 LIMERICK - UNIT 1 t
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:
| |
| : a. Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT I-131.
| |
| : b. (Deleted)
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 1, 2, or 3 with the specific activity of the primary coolant;
| |
| : 1. Greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcuries per gram, DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours. The provisions of Specification 3.0.4.c are applicable.
| |
| : 2. (Deleted)
| |
| : b. In OPERATIONAL CONDITION 1, 2, 3, or 4, with the specific activity K) /of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131, perform the sampling and analysis requirements of Item 4.a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.
| |
| : c. In OPERATIONAL CONDITION 1 or 2, with:
| |
| : 1. THERMAL POWER changed by more than 15% of RATED THERMAL POWER in 1 hour, or
| |
| : 2. The off-gas level, at the SJAE, increased by more than 10,000 microcuries per second in 1 hour during steady-state operation at release rates less than 75,000 microcuries per second, or
| |
| : 3. The off-gas level, at the SJAE, increased by more than 15% in 1 hour during steady-state operation at release rates greater than 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4.b of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.
| |
| LIMERICK - UNIT 1 3/4 4-15 Amendment No. pO, 4BE9- 174
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to I
| |
| be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.
| |
| LIMERICK - UNIT 1 3/4 4-16 Amendment No. 20 I MAY 1 9 1989
| |
| | |
| TABLE 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM OPERATIONAL CONDITIONS TYPE OF MEASUREMENT SAMPLE AND ANALYSIS IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANAIYSTS IS REOIITRFD
| |
| : 1. (Deleted)
| |
| : 2. Isotopic Analysis for DOSE In accordance with the 1 EQUIVALENT 1-131 Concentration Surveillance Frequency Control Program
| |
| : 3. (Deleted)
| |
| : 4. Isotopic Analysis for Iodine a) At least once per 4 hours, 1"*, 2**, 3**, 4**
| |
| whenever the specific activity exceeds a limit, as required by ACTION b.
| |
| b) At least one-sample, between 1, 2 2 and 6 hours following the change in THERMAL POWER or off-gas level, as required by ACTION c.
| |
| : 5. Isotopic Analysis of an Off- In accordance with the 1 gas Sample Including Quantitative Surveillance Frequency Measurements for at least Xe-133, Control Program Xe-135, and Kr-88
| |
| **Until the specific activity of the primary coolant system is restored to within its limits.
| |
| LIMERICK - UNIT 1 3/4 4-17 Amendment No. ., I--4,186
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A for hydrostatic or leak testing; (2) curve B for heatup by non-nuclear means, cool-down following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with:
| |
| : a. A maximum heatup of 100°F in any 1-hour period,
| |
| : b. A maximum cooldown of 100°F in any 1-hour period,
| |
| : c. A maximum temperature change of less than or equal to 20°F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, and
| |
| : d. The reactor vessel flange and head flange temperature greater than or equal to 80°F when reactor vessel head bolting studs are under tension.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 curve A, B, or C as applicable, in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 4-18 Amendment No. 436, 406, 4-4r,186
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curve C within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and in accordance with the Surveillance Frequency Control Program during system heatup.
| |
| 4.4.6.1.3 DELETED 4.4.6.1.4 DELETED 4.4.6.1.5 The reactor vessel flangeand head flange temperature shall be verified to be greater than or equal to 80°F:
| |
| : a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:
| |
| : 1.
| |
| * 100°F, in accordance with the Surveillance Frequency Control Program.
| |
| : 2. g 90°F, in accordance with the Surveillance Frequency Control Program.
| |
| : b. Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during tensioning of the reactor vessel head bolting studs.
| |
| LIMERICK - UNIT 1 3/4 4-19 Amendment No. 2-9, 36, 1-2-6, 4-6-;, 186
| |
| | |
| A-- A B 1400 l 1300 1200 1100 0m C 1000 w
| |
| 0-0 900
| |
| -i w
| |
| Up 800 cn w
| |
| ;A, B, C - CORE BELTLINE AFTER o 700 'ASSUMED 690F SHIFT FROM AN C-)
| |
| * INITIAL PLATE RTNDT OF 200F fit M 600 jA - SYSTEM HYDROTEST z
| |
| !WITH FUEL INTHE VESSEL
| |
| , 500 BH-NON-NUCLEAR Iti X HEATUP/COOLDOWN LIMIT Un 400 C - NUCLEAR (CORE CRITICAL)
| |
| LIMIT 300
| |
| _ CURVES A, B, C ARE VALID UP 200 jTO 32 EFPY OF OPERATION I
| |
| I'CURVE A 22 IS VALID UP TO 100 0
| |
| 0 25 50 I _ .-.-.-
| |
| -22 EFPY OF OPERATION 75 100 125 150 175 200 225 250 275 300 325 350 375 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (GF)
| |
| MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE FIGURE 3.4.6.1-1 LIMERICK - UNIT 1 3/4 4-20 Amendment No. 36, 4-06, 4A-5,455, 163 JAN 0 2 -e3
| |
| | |
| (( C LIMERICK - UNIT 1 3/4 4 -21 Amendment No. 446, 455,167
| |
| | |
| REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1053 psig.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.
| |
| ACTION:
| |
| With the reactor steam dome pressure exceeding 1053 psig, reduce the pressure to less than 1053 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1053 psig in accordance with the Surveillance Frequency Control Program. I
| |
| *Not applicable during anticipated transients.
| |
| LIMERICK - UNIT I 3/4 4-22 Amendment No. 4-06,186
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal to 5 seconds.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| With one or more MSIVs inoperable: I
| |
| : a. Maintain at least one MSIV OPERABLE in each affected main steam line that I is open and within 8 hours, either:
| |
| : 1. Restore the inoperable valve(s) to OPERABLE status, or I
| |
| : 2. Isolate the affected main steam line by use of a deactivated MSIV in I the closed position.
| |
| : b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD I SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to Specification 4.0.5.
| |
| LIMERICK - UNIT 1 3/4 4-23 Amendment No.169
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.8 (DELETED)
| |
| PAGE INTENTIONALLY LEFT BLANK LIMERICK - UNIT 1 3/4 4-24 Amendment No. ++/-,199
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.9 RESIDUAL HEAT REMOVAL HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.1 Two (2) independent RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one (1) RHR shutdown cooling subsystem shall be in operation. * ** ***
| |
| Each independent RHR shutdown cooling subsystem shall consist of at least:
| |
| : a. One OPERABLE RHR pump, and
| |
| : b. One OPERABLE RHR heat exchanger, not common to the two (2) independent subsystems.
| |
| APPLICABILITY: OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint.
| |
| ACTION:
| |
| : a. With less than the above required independent RHR shutdown cooling subsystems OPERABLE, immediately initiate corrective action to return the required independent subsystems to OPERABLE status as soon as possible.
| |
| Within 1 hour and at least once per 24 hours thereafter, verify the availability of at least one alternate method capable of decay heat removal for each inoperable independent RHR shutdown cooling subsystem. Be in at least COLD SHUTDOWN within 24 hours.****
| |
| : b. With no independent RHR shutdown cooling subsystem in operation, immediately initiate corrective action to return at least one (1) independent subsystem to operation as soon as possible. Within 1 hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.
| |
| SURVEILLANCE REQUIREMENTS 4.4.9.1.1 At least one independent RHR shutdown cooling subsystem or alternate method shall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.
| |
| 4.4.9.1.2 Verify RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.*****
| |
| *One independent RHR shutdown cooling subsystem may be inoperable for up to 2 hours for surveillance testing provided the other independent subsystem is OPERABLE and in operation.
| |
| **The shutdown cooling pump may be removed from operation for up to 2 hours per 8-hour period provided the other independent subsystem is OPERABLE.
| |
| ***The independent RHR shutdown cooling subsystem may be removed from operation during hydrostatic testing.
| |
| ****Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
| |
| *****Not required to be performed until 12 hours after reactor steam dome pressure is less I than the RHR cut-in permissive setpoint.
| |
| LIMERICK - UNIT 1 3/4 4-25 Amendment No. -9+,-+/--+/-9,~. 216
| |
| | |
| REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.2 Two (2) RHR shutdown cooling subsystems shall be OPERABLE, and with no recirculation pump in operation, at least one (1) RHR shutdown cooling subsystem shall be in operation. * ** ***
| |
| APPLICABILITY: OPERATIONAL CONDITION 4.
| |
| ACTION: #
| |
| : a. With one (1) or two (2) RHR shutdown cooling subsystems inoperable:
| |
| : 1. Within one (1) hour, and once per 24 hours thereafter, verify an alternate method of decay heat removal is available for each inoperable RHR shutdown cooling subsystem.
| |
| : b. With no RHR shutdown cooling subsystems in operation and no recirculation pump in operation:
| |
| : 1. Within one (1) hour from discovery of no reactor coolant circulation, and once per 12 hours thereafter, verify reactor coolant circulating by an alternate method; and
| |
| : 2. Once per hour monitor reactor coolant temperature and pressure.
| |
| SURVEILLANCE REQUIREMENTS 4.4.9.2.l At least one (1) RHR shutdown cooling subsystem or recirculation pump is operating or an alternate method shall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.
| |
| 4.4.9.2.2 Verify RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.
| |
| * Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up two (2) hours per eight (8) hour period.
| |
| ** One (1) RHR shutdown cooling subsystem may be inoperable for up to two (2) hours for the performance of Surveillances.
| |
| *** The shutdown cooling subsystem may be removed from operation during hydrostatic testing.
| |
| # Separate Action entry is allowed for each shutdown cooling subsystem.
| |
| LIMERICK - UNIT 1 3/4 4-26 Amendment No. f.,-+/--+/--9-.~. 216
| |
| | |
| 3/4.5 EMERGENCY CORE COOLING SYSTEMS
| |
| .3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems-shall be OPERABLE with:
| |
| : a. The core spray system (CSS) consisting of two subsystems with each subsystem comprised of:
| |
| : 1. Two OPERABLE CSS. pumps, and
| |
| : 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water through the'spray sparger to the reactor vessel.
| |
| *b. The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of four subsystems with each subsystem comprised of:
| |
| : 1. One OPERABLE LPCI pump, and
| |
| : 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
| |
| : c. The high pressure coolant injection (HPCI) system consisting of:
| |
| : 1. One OPERABLE HPCI pump, and
| |
| : 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
| |
| : d. The automatic depressurization system (ADS) with at least five OPERABLE ADS valves.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1, 2* ** #, and 3* **
| |
| *The HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.
| |
| **The ADS is not required to be OPERABLE when the reactor steam dome pressure is less that or equal to 100 psig.
| |
| #See Special Test Exception 3.10.6.
| |
| ##Two LPCI subsystems of the RHR system may be inoperable in that they are aligned in the shutdown cooling mode when reactor vessel pressure is less than the RHR Shutdown cooling permissive setpoint.
| |
| LIMERICK - UNIT 1 3/4 5-1 Amendment No. 8-,
| |
| 9-, 4-34, 192
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS
| |
| *IMITING CONDITION FOR OPERATION (Continued) 11T9.N:
| |
| : a. For the core spray'system:
| |
| : 1. With one CSS subsystem inoperable, provided that at least two LPCI subsystems are OPERABLE, restore the inoperable CSS subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 2. With both CSS subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| : b. For the LPCI system:
| |
| : 1. With one LPCI subsystem inoperable, provided that at least one CSS subsystem is OPERABLE, restore the inoperable LPCI pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 2. With one RHR cross-tie valve (HV-51-182 A or B) open, or power not removed from one closed RHR cross-tie valve operator, close the open valve and/or remove power from the closed valves operator within 72 hours, or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| : 3. With no RHR cross-tie valves (HV-51-182 A, B) closed, or power not removed from both closed RHR cross-tie valve operators, or with one RHR cross-tie valve open and power not removed from the other RHR cross-tie valve operator, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| : 4. With two LPCI subsystems inoperable, provided that at least one CSS subsystem is OPERABLE, restore at least three LPCI subsystems to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 5. W4th three LPCI subsystems inoperable, provided that both CSS subsystems are OPERABLE, restore at least two LPCI subsystems to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 6. With all four LPCI subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.*
| |
| *Whenever both .shutdown cooling subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
| |
| IMERICK - UNIT 1 3/4 5-2 Amendment No. 66, 94,131 OCT 2 3 1993
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : c. For the HPCI system:
| |
| : 1. With the HPCI system inoperable, provided the CSS, the LPCI system, the ADS and the RCIC system are OPERABLE, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to~ 200 psig within the following 24 hours.
| |
| : 2. With the HPCI system inoperable, and one CSS subsystem, and/or LPCI subsystem inoperable, and provided at least one CSS subsystem, three LPCI subsystems, and ADS are operable, restore the HPCI to OPERABLE within 8 hours, or be in HOT SHUTDOWN in the next 12 hours, and in COLD SHUTDOWN in the next 24 hours.
| |
| : 3. Specification 3.0.4.b is not applicable to HPCI.
| |
| : d. For the ADS:
| |
| : 1. With one of the above required ADS valves inoperable, provided the HPCI system, the CSS and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to~ 100 psig within the next 24 hours.
| |
| : 2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours and reduce reactor steam dome pressure to~ 100 psig within the next 24 hours.
| |
| : e. With a CSS and/or LPCI header 6P instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours or determine the ECCS header 6P locally at least once per 12 hours; otherwise, declare the associated CSS and/or LPCI, as applicable, inoperable.
| |
| : f. DELETED LIMERICK UNIT 1 3/4 5-3 Amendment No. JJ, 94, +69, 211
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.l The emergency core cooling systems shall be demonstrated OPERABLE by:
| |
| : a. In accordance with the Surveillance Frequency Control Program:
| |
| : 1. For the CSS, the LPCI system, and the HPCI system:
| |
| a) Verifying locations susceptible to gas accumulation are sufficiently filled with water.
| |
| b) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct*
| |
| position.***
| |
| : 2. For the LPCI system, verifying that both LPCI system subsystem cross-tie valves CHV-51-182 A, B) are closed with power removed from the valve operators.
| |
| : 3. For the HPCI system, verifying that the HPCI pump flow controller is in the correct position.
| |
| : 4. For the CSS and LPCI system, performance of a CHANNEL FUNCTIONAL TEST of the injection header ~P instrumentation.
| |
| : b. Verifying that, when tested pursuant to Specification 4.0.5:
| |
| : 1. Each CSS pump in each subsystem develops a flow of at least 3175 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of ~ 105 psid plus head and line losses.
| |
| : 2. Each LPCI pump in each subsystem develops a flow of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of
| |
| ~ 20 psid plus head and line losses.
| |
| : 3. The HPCI pump develops a flow of at least 5600 gpm against a test line pressure which corresponds to a reactor vessel pressure of 1040 psig plus head and line losses when steam is being supplied to the turbine at 1040, +13, -120 psig.**
| |
| : c. In accordance with the Surveillance Frequency Control Program:
| |
| : 1. For the CSS, the LPCI system, and the HPCI system, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.
| |
| * Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
| |
| ** The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than 200 psig within the following 72 hours.
| |
| *** Not required to be met for system vent flow paths opened under administrative control.
| |
| LIMERICK - UNIT 1 3/4 5-4 Amendment No . .&9,++/-.~.~. 216
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
| |
| : 2. For the HPCI system, verifying that:
| |
| a) The system develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of Ž200 psig plus head and line losses, when steam is being supplied to the turbine at 200 + 15, - 0 psig.**
| |
| b) The suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber water level - high signal.
| |
| : 3. Performing a CHANNEL CALIBRATION of the CSS, LPCI, and HPCI system discharge line "keep filled" alarm instrumentation.
| |
| : 4. Performing a CHANNEL CALIBRATION of the CSS header AP instru-mentation and verifying the setpoint to be 5 the allowable value of 4.4 psid.
| |
| : 5. Performing a CHANNEL CALIBRATION of the LPCI header AP instru-mentation and verifying the setpoint to be 5 the allowable value of 3.0 psid.
| |
| : d. For the ADS:
| |
| : 1. In accordance-with the Surveillance Frequency Control Program, verify ADS accumulator gas supply header pressure is Ž90 psig.
| |
| : 2. In accordance with the Surveillance Frequency Control Program:
| |
| a) Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
| |
| b) Verify that when tested pursuant to Specification 4.0.5 that each ADS valve is capable of being opened.
| |
| c) DELETED
| |
| ** The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test. If HPCI OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than 200 psig within the following 72 hours.
| |
| LIMERICK - UNIT 1 3/4 5-5 Amendment No. 2-9, -33; 74-9Q, 4-6-5,186
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL (WIC)
| |
| LIMITING CONDITION FOR OPERATION 3.5.2 DRAIN TIME of RPV water inventory to the top of active fuel (TAF) shall be~ 36 hours At least one of the following shall be OPERABLE:
| |
| : a. Core spray system (CSS) subsystem comprised of:
| |
| : 1. Two OPERABLE CSS pumps, and
| |
| : 2. An OPERABLE flow path capable of taking suction from at least one of the following water sources and transferring the water through the spray sparger to the reactor vessel:
| |
| a) From the suppression chamber, or b) When the suppression chamber water level is less than the limit or is drained, from the condensate storage tank containing at least 135,000 available gallons of water, equivalent to a level of 29 feet.
| |
| : b. Low pressure coolant injection (LPCI) system subsystem comprised of:
| |
| : 1. One OPERABLE LPCI pump, and
| |
| : 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.**
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5.
| |
| ACTION:
| |
| : a. With none of the above required subsystems OPERABLE, immediately suspend CORE ALTERATIONS. Restore at least one subsystem to OPERABLE status within 4 hours. Otherwise, initiate action to establish a method of water injection capable of operating without offsite electrical power.
| |
| : b. DELETED
| |
| **One LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.
| |
| LIMERICK - UNIT 1 3/4 5-6 Amendment No. ~.227
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION:
| |
| : c. With DRAIN TIME less than 36 hours and greater than or equal to 8 hours, within 4 hours:
| |
| : 1. Verify SECONDARY CONTAINMENT INTEGRITY is capable of being established in less than the DRAIN TIME,
| |
| : 2. Verify each secondary containment penetration flow path is capable of being isolated in less than the DRAIN TIME, and
| |
| : 3. Verify one standby gas treatment subsystem is capable of being placed in operation in less than the DRAIN TIME.
| |
| : d. With DRAIN TIME less than 8 hours, immediately:
| |
| : 1. Initiate action to establish an additional method of water injection with water sources capable of maintaining RPV water level greater than TAF for greater than or equal to 36 hours,***
| |
| : 2. Initiate action to establish SECONDARY CONTAINMENT INTEGRITY,
| |
| : 3. Initiate action to isolate each secondary containment penetration flow path or verify it can be manually isolated from the control room, and
| |
| : 4. Initiate action to verify one standby gas treatment subsystem is capable of being placed in operation.
| |
| : e. With required ACTION and associated allowed outage time for ACTIONs c. or d. not met, or DRAIN TIME less than 1 hour, immediately initiate action to restore DRAIN TIME to greater than or equal to 36 hours.
| |
| ***The required injection/spray subsystem or an additional method of water injection shall be capable of operating without offsite electrical power.
| |
| LIMERICK - UNIT 1 3/4 5-6a Amendment No. 227 .
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2.1 Verify DRAIN TIME is greater than or equal to 36 hours in accordance with the Surveillance Frequency Control Program.
| |
| 4.5.2.2 Verify, for a required LPCI subsystem, the suppression chamber water level is greater than or equal to 16 feet O inches in accordance with the Surveillance Frequency Control Program.
| |
| 4.5.2.3 Verify, for a required CSS subsystem, that the suppression chamber water level is greater than or equal to 16 feet O inches or the condensate storage tank water level is greater than or equal to 29 feet O inches in accordance with the Surveillance Frequency Control Program.
| |
| 4.5.2.4 Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.
| |
| 4.5.2.5 Verify, for the required ECCS injection/spray subsystem, each manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position in accordance with the Surveillance Frequency Control Program.#A 4.5.2.6 Operate the required ECCS injection/spray subsystem through the recirculation line for greater than or equal to 10 minutes in accordance with the Surveillance Frequency Control Program.
| |
| 4.5.2.7 Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal in accordance with the Surveillance Frequency Control Program.
| |
| 4.5.2.8 Verify the required ECCS injection/spray subsystem actuates on a manual initiation signal in accordance with the Surveillance Frequency Control Program.##
| |
| *DELETED.
| |
| #Not required to be met for system vent flow paths open under administrative control.
| |
| AExcept that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
| |
| ##Vessel injection/spray may be excluded.
| |
| LIMERICK - UNIT 1 3/4 5-7 Amendment No.~.~. 227
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION 3.5.3 The suppression chamber shall be OPERABLE:
| |
| : a. In OPERATIONAL CONDITIONS 1, 2, and 3 with a contained water volume of at least 122,120 ft 3 , equivalent to a level of 22'0".
| |
| : b. DELETED APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 1, 2 or 3 with the suppression chamber water level less than the above limit, restore the water level to within the limit within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. DELETED LIMERICK - UNIT 1 3/4 5-8 Amendment No. 227
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying the water level to be greater than or equal to, as applicable:
| |
| : a. 22'0" in accordance with the Surveillance Frequency Control Program.
| |
| : b. DELETED 4.5.3.2 DELETED LIMERICK - UNIT 1 3/4 5-9 Amendment No.~. 227
| |
| | |
| 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING.CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3.
| |
| ACTION:
| |
| Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:
| |
| : a. After each closing of each penetration subject to Type B testing, except the primary containment air locks, if opened following Type A or B test, by leak rate testing in accordance with the Primary Containment Leakage Rate Testing Program.
| |
| : b. In accordance with the Surveillance Frequency Control Program by verifying that all primary containment penetrations** not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except for valves that are opened under administrative control as permitted by Specification 3.6.3.
| |
| : c. By verifying the primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.
| |
| : d. By verifying the suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
| |
| * See Special Test Exception 3.10.1
| |
| **Except valves, blind flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often than once per 92 days.
| |
| LIMERICK - UNIT 1 3/4 6-1 Amendment No. 4--8, 446,186
| |
| | |
| CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:
| |
| : a. An overall integrated leakage rate (Type A Test) in accordance with the Primary Containment Leakage Rate Testing Program.
| |
| : b. A combined leakage rate in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted.
| |
| : c. *Less than or equal to 100 scf per hour through any one main steam isolation valve not to exceed 200 scf per hour for all four main steam lines, when tested at Pt. 22.0 psig.
| |
| : d. A combined leakage rate of less than or equal to 1 gpm times the total number of containment isolation valves in hydrostatically tested lines which penetrate the primary containment, when tested at 1.10 Pa, 48.4 psig.
| |
| APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1.
| |
| ACTION:
| |
| With:
| |
| : a. The measured overall integrated primary containment leakage rate (Type A Test) exceeding the leakage rate specified in the Primary Containment Leakage Rate Testing Program, or
| |
| : b. The measured combined leakage rate exceeding the leakage rate specified in the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, or
| |
| : c. The measured leakage rate exceeding 100 scf per hour through any one main steam isolation valve, or exceeding 200 scf per hour for all four main steam lines, or
| |
| : d. The measured combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment exceeding 1 gpm times the total number of such valves, restore:
| |
| : a. The overall integrated leakage rate(s) (Type A Test) to be in accordance with the Primary Containment Leakage Rate Testing Program, and
| |
| *Exemption to Appendix J of 10 CFR Part 50. Amendment No. 3A97, 148, 146 ITMPOTrv
| |
| ,,-. '.\.W -- IIWTT vats1I. 1 la4A A/' A-V-L OCT 1 4 !J
| |
| | |
| CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : b. The combined leakage rate to be in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type B and C tests, except for: main steam line isolation valves*, valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has been granted, and
| |
| : c. The leakage rate to *100 scf per hour for any main steam isolation valve I that exceeds 100 scf per hour, and restore the combined maximum pathway leakage to *200 scf per hour, and
| |
| : d. The combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing the reactor coolant system temperature above 200°F.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated to be in accordance with the Primary Containment Leakage Rate Testing Program, or approved exemptions, for the following:
| |
| : a. Type A Test
| |
| : b. Type B and C Tests (including air locks)
| |
| : c. Main Steam Line Isolation Valves
| |
| : d. Hydrostatically tested Containment Isolation Valves
| |
| * Exemption to Appendix "J" to 10 CFR Part 50.
| |
| LIMERICK - UNIT I L3/4 6-3 Amendment No. 6*, 4-P, 44-8, -446,185
| |
| | |
| THIS PAGE IS INTENTIONALLY LEFT BLANK LIMERICK - UNIT I 3/4 6-4 Amendment No. Z, 3, 71, $H, 118 JAN 2 4 1997
| |
| | |
| CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK LIMITING CONDITION FOR OPERATION 3.6.1.3 The primary containment air lock shall be OPERABLE with:
| |
| : a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and
| |
| : b. An overall air lock leakage rate in accordance with the Primary Containment Leakage Rate Testing Program.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3.
| |
| ACTION:
| |
| : a. With one primary containment air lock door inoperable:
| |
| : 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed.
| |
| : 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.
| |
| : 3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With the primary containment air lock inoperable, except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| *See Special Test Exception 3.10.1.
| |
| LIMERICK - UNIT 1 3/4 6-5 Amendment No. 33, 4-18,169
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 The primary containment air lock shall be demonstrated OPERABLE:
| |
| : a. By verifying the seal leakage rate is in accordance with the the Primary Containment Leakage Rate Testing Program.
| |
| : b. By conducting an overall air lock leakage test in accordance with the Primary Containment Leakage Rate Testing Program.
| |
| : c. In accordance with the Surveillance Frequency Control Program by I verifying that only one door'in the air lock can be opened at a time.***
| |
| * Except that the airlock doors need not be opened to verify interlock OPERA-BILITY when the primary containment is inerted, provided that the airlock doors' interlock is tested within 8 hours after the primary containment has been deinerted and provided the shield door to the airlock is maintained locked closed.
| |
| LIMERICK - UNIT 1 3/4 6-6 Amendment No. 4-18, 186
| |
| | |
| CONTAINMENT SYSTEMS MSIV LEAKAGE ALTERNATE DRAIN PATHWAY LIMIJING CONDITION FOR OPERAJION 3.6.1.4 The MSIV Leakage Alternate Drain Pathway shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| With the MSIV Leakage Alternate Drain Pathway inoperable, restore the pathway to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURYEILLANCE REQUIREMENTS 4.6.1.4 The MSIV Leakage Alternate Drain Pathway shall be demonstrated OPERABLE:
| |
| : a. In accordance with Specification 4.0.5, by cycling each motor operated valve, required to be repositioned, through at least one complete cycle of full travel.
| |
| LIMERICK - UNIT 1 3/4 6-7 Amendment No. ~. 191,225
| |
| | |
| CONTAINMENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.5 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.5.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.5.1 The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment, including the liner plate, shall be determined by a visual inspection of those surfaces. This inspection shall be performed in accordance with the Primary Containment Leakage Rate Testing Program.
| |
| 4.6.1.5.2 DELETED LIMERICK - UNIT 1 3/4 6-8 Amendment No. t+&, 211
| |
| | |
| CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.6 Drywell and suppression chamber internal pressure shall be maintained between -1.0 and +2.0 psig.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| With the drywell and/or suppression chamber internal pressure outside of the specified limits, restore the internal pressure to within the limit within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.6 The drywell and suppression chamber internal pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program. I LIMERICK - UNIT 1 3/4 6-9 Amendment No. 2-9, 186
| |
| | |
| CONTAINMENT SYSTEMS DRYWELL AVERAGE AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.7 Drywell average air temperature shall not exceed 145°F.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| With the drywell average air temperature greater than 145°F, reduce the average air temperature to within the limit within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.6.1.7 The drywell average air temperature shall be the volumetric average of the temperatures at the following locations and shall be determined to be within the limit in accordance with the Surveillance Frequency Control Program: I Approximate Number of Elevation Installed Sensors*
| |
| : a. 330' 3
| |
| : b. 320' 3
| |
| : c. 260' 3
| |
| : d. 248' 6
| |
| *At least one reading from each elevation is required for a volumetric average calculation.
| |
| LIMERICK - UNIT I 3/4 6-10 Amendment No. 2-9, 4.5-9, 186
| |
| | |
| CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.8 The drywell and suppression chamber purge system may be in operation with the supply and exhaust isolation valves in one supply line and one exhaust line open for inerting, deinerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.
| |
| APPLICABIITJY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With a drywell and/or suppression chamber purge supply and/or exhaust isolation valve open, except as permitted above, close the valve(s) within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.6.1.8 In accordance with the Surveillance Frequency Control Program, verify I each primary containment purge valve [18" or 24"] is closed.*, **
| |
| * Only required to be met in OPERATIONAL CONDITIONS 1, 2 and 3.
| |
| ** Not required to be met when the primary containment purge valves are open for inerting, deinerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require these valves to be open.
| |
| LIMERICK - UNIT 1 3/4 6-11 Amendment No. 441, 444,186
| |
| | |
| CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION 3.6.2.1 The suppression chamber shall be OPERABLE with:
| |
| : a. The pool water:
| |
| : 1. Volume* between 122,120 ft3 and 134,600 ft3, equivalent to a level between 22' 0" and 24' 3", and a
| |
| : 2. Maximum average temperature of 950 F except that the maximum average temperature may be permitted to increase to:
| |
| a) 1050 F during testing which adds heat to the suppression chamber.
| |
| b) 110F with THERMAL POWER less than or equal to 1% of RATED THERMAL POWER.
| |
| c) 1201F with the main steam line isolation valves closed following a scram.
| |
| : b. Drywell-to-suppression chamber bypass leakage less than or equal to 10% of the acceptable A/4X design value of 0.0500 ft2.
| |
| : c. At least eight suppression pool water temperature instrumentation indicators, one in each of the eight locations.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With the suppression chamber water level outside the above limits, restore the water level to within the limits within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With the suppression chamber average water temperature greater than 950F, restore the average temperature to less than or equal to 950 F within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, except, as permitted above:
| |
| : 1. With the suppression chamber average water temperature greater than 1050 F during testing which adds heat to the suppression chamber, stop all testing which adds heat to the suppression chamber and restore the average temperature to less than 950 F within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 2. With the suppression chamber average water temperature greater than:
| |
| a) 950F for more than 24 hours and THERMAL POWER greater than 1% of RATED THERMAL POWER, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| b) 110 0F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode.
| |
| *Includes the volume inside the pedestal.
| |
| LIMERICK - UNIT 1 3/4 6-12 Amendment No. 29 l JUN 2 2 1989
| |
| | |
| CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : 3. With the suppression chamber average water temperature greater than 120 0 F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours.
| |
| : c. With only one suppression chamber water level indicator OPERABLE and/or with less than eight suppression pool water temperature indicators, one in each of the eight locations OPERABLE, restore the inoperable indicator(s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at least once per 12 hours.
| |
| : d. With no suppression chamber water level indicators OPERABLE and/or with less than seven suppression pool water temperature indicators covering at least seven locations OPERABLE, restore at least one water level indicator and at least seven water temperature indicators to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : e. With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200 0 F.
| |
| SURVEILLANCE REQUIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:
| |
| : a. By verifying the suppression chamber water volume to be within the limits in accordance with the Surveillance Frequency Control Program.
| |
| : b. In accordance with the Surveillance Frequency Control Program by verifying the suppression chamber average water temperature to be less than or equal to 95°F, except:
| |
| : 1. At least once per 5 minutes during testing which adds heat to the suppression chamber, by verifying the suppression chamber average water temperature less than or equal to 105 0 F.
| |
| : 2. At least once per hour when suppression chamber average water temperature is greater than or equal to 95 0 F, by verifying:
| |
| a) Suppression chamber average water temperature to be less than or equal to 110 0 F, and b) THERMAL POWER to be less than or equal to 1% of RATED THERMAL POWER 12 hours after suppression chamber average water temperature has exceeded 95 0 F for more than 24 hours.
| |
| : 3. At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 95°F by verifying suppression chamber average water temperature less than or equal to 120 0 F.
| |
| LIMERICK - UNIT I 3/4 6-13 Amendment No.186
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
| |
| : c. By verifying at least 8 suppression pool water temperature indicators in at least 8 locations, OPERABLE by performance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies specified in the Surveillance Frequency Control Program with the temperature alarm setpoint for:
| |
| : 1. High water temperature:
| |
| a) First setpoint
| |
| * 95°F b) Second setpoint
| |
| * 105°F c) Third setpoint : 110 F d) Fourth setpoint
| |
| * 120'F
| |
| : d. By verifying at least two suppression chamber water level indicators OPERABLE by performance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies specified in the Surveillance Frequency Control Program with the water level alarm setpoint for high water level
| |
| * 24'1-1/2".
| |
| : e. Drywell-to-suppression chamber bypass leak tests shall be conducted to coincide with the Type A test at an initial differential pressure of 4 psi and verifying that the A/IFk calculated from the measured leakage is within the specified limit. If any drywell-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed.
| |
| : f. By conducting a leakage test on the drywell-to-suppression chamber vacuum breakers at a differential pressure of at least 4.0 psi and verifying that the total leakage area A/.F- contributed by all vacuum breakers is less than or equal to 24% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.e is not conducted.
| |
| LIMERICK -UNIT 1 3/4 6-14 Amendment No. 48,47-0,;4-,40-8,*98, 186
| |
| | |
| CONTAINMENT SYSTEMS SUPPRESSION POOL SPRAY LIMITING CONDITION FOR OPERATION 3.6.2.2 The suppression pool spray mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:
| |
| : a. One OPERABLE RHR pump, and
| |
| : b. An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and the suppression pool spray sparger(s).
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one suppression pool spray loop inoperable, restore the inoperable loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With both suppression pool spray loops inoperable, restore at least one loop to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN* within the following 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.2.2 The suppression pool spray mode of the RHR system shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
| |
| : b. By verifying that each of the required RHR pumps develops a flow of at least 500 gpm on recirculation flow through the RHR heat exchanger and the suppression pool spray sparger when tested pursuant to Speci-fication 4.0.5.
| |
| : c. By verifying RHR suppression pool spray subsystem locations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.
| |
| *Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
| |
| LIMERICK - UNIT 1 3/4 6-15 Amendment No. .
| |
| ~ .J-J.+/-..~, 216
| |
| | |
| CONTAINMENT SYSTEMS SUPPRESSION POOL COOLING LIMITING CONDITION FOR OPERATION 3.6.2.3 The suppression pool cooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:
| |
| : a. One OPERABLE RHR pump, and
| |
| : b. An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one suppression pool cooling loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours** or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With both suppression pool cooling loops inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN* within the next 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.2.3 The suppression pool cooling mode of the RHR system shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
| |
| : b. By verifying that each of the required RHR pumps develops a flow of at least 10,000 gpm on recirculation flow through the flow path including the RHR heat exchanger and its associated closed bypass valve, the suppression pool and the full flow test line when tested pursuant to Specification 4.0.5.
| |
| : c. By verifying RHR suppression pool cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.
| |
| *Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
| |
| **During the extended 7-day Allowed Outage Time (AOT) specified by TS LCO 3.7.1.1, Action a.3.a) or a.3.b) to allow for RHRSW subsystem piping repairs, the 72-hour AOT for one inoperable suppression pool cooling loop may also be extended to 7 days for the same 7-day period.
| |
| LIMERICK - UNIT 1 3/4 6-16 Amendment No . .e+,~,ge.J..J+,+ge,~,216
| |
| | |
| CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each primary containment isolation valve and each instrumentation lin.e excess flow check valve shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one or more of the primary containment isolation valves inoperable,**
| |
| maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours either:
| |
| : 1. Restore the inoperable valve(s) to OPERABLE status, or
| |
| : 2. Isolate each affected penetration by use of at least one de-activated automatic valve secured in the isolated position,* or
| |
| : 3. Isolate each affected penetration by use of at least one closed manual valve or blind flange.*
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and, in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With one or more of the instrumentation line excess flow check valves inoperable, operation may continue and the provisions of Specification 3.0.3 are not applicable provided that within 4 hours either:
| |
| : 1. The inoperable valve is returned to OPERABLE status, or
| |
| : 2. The instrument line is isolated and the associated instrument is declared inoperable.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in.COLD SHUTDOWN within the following 24 hours.
| |
| : c. With one or more scram discharge volume vent or drain valves inoperable, perform the applicable actions specified in Specification 3.1.3.1.
| |
| * Isolation valves closed to satisfy these requirements may be reopened on an intermittent basis under administrative control.
| |
| ** Except for the scram discharge volume vent and drain valves.
| |
| LIMERICK - UNIT 1 3/4 6-17 Amendment No. *,-!-4-68,4-64,192
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
| |
| 4.6.3.2 Each primary containment automatic isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.
| |
| 4.6.3.3 The isolation time of each primary containment power operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
| |
| 4.6.3.4 A representative sample of reactor instrumentation line excess flow check valves shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program, such that each valve is tested in accordance with the Surveillance Frequency Control Program by verifying that the valve checks flow.*
| |
| 4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying the continuity of the explosive charge.
| |
| : b. In accordance with the Surveillance Frequency Control Program by removing the explosive squib from the explosive valve, such that each explosive squib in each explosive valve will be tested in accordance with the Surveillance Frequency Control Program, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life and/or operating life, as applicable.
| |
| *The reactor vessel head seal leak detection line (penetration 29A) excess flow check valve is not required to be tested pursuant to this requirement.
| |
| LIMERICK - UNIT 1 3/4 6-18 Amendment No. ,9,,,-!46,48, 186
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| | |
| TABLE 3.6.3-1 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM), PCIV SECTION.
| |
| TECHNICAL SPECIFICATION PAGES 3/4 6-19 THROUGH 3/4 6-43a HAVE BEEN INTENTIONALLY OMITTED.
| |
| Amendment No. X, 33,99, 146 LIMERICK - UNIT 1 3/4 6-19 OCT 1 3
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| | |
| CONTAINMENT SYSTEMS 3/4.6.4 VACUUM RELIEF SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Three pairs of suppression chamber - drywell vacuum breakers shall be OPERABLE and all suppression chamber - drywell vacuum breakers shall be closed.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one or more vacuum breakers in one of the three required pairs of suppression chamber - drywell vacuum breaker pairs inoperable for opening but known to be closed, restore at least one inoperable pair of vacuum breakers to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With one suppression chamber - drywell vacuum breaker open, verify the other vacuum breaker in the pair to be closed within 2 hours; restore the open vacuum breaker to the closed position within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : c. With one position indicator of any suppression chamber - drywell vacuum breaker inoperable:
| |
| : 1. Verify the other vacuum breaker in the pair to be closed within 2 hours and at least once per 15 days thereafter, or
| |
| : 2. Verify the vacuum breaker(s) with the inoperable position indicator to be closed by conducting a test which demonstrates that the AP is maintained at greater than or equal to 0.7 psi for one hour without makeup within 24 hours and at least once per 15 days thereafter.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next .12 hours and in COLD SHUTDOWN within the following 24 hours.
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| LIMERICK - UNIT 1 3/4 6-44 Amendment No. 46 0o; i 1u
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| | |
| CONTAINMENT SYSTEMS SURVETLLANCE REOUITREMENTS 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be:
| |
| : a. Verified closed in accordance with the Surveillance Frequency Control Program.
| |
| : b. Demonstrated OPERABLE:
| |
| : 1. In accordance with the Surveillance Frequency Control Program and within 2 hours after any discharge of steam to the suppression chamber from the safety/relief valves, by cycling each vacuum breaker through at least one complete cycle of full travel.
| |
| : 2. In accordance with the Surveillance Frequency Control Program by verifying both position indicators OPERABLE by observing expected valve movement during the cycling test.
| |
| : 3. In accordance with the Surveillance Frequency Control Program by:
| |
| a) Verifying each valve's opening setpoint, from the closed position, to be 0.5 psid +/- 5%, and b) Verifying both position indicators OPERABLE by performance of a CHANNEL CALIBRATION.
| |
| c) Verifying that each outboard valve's position indicator is capable of detecting disk displacement Ž0.050", and each inboard valve's position indicator is capable of detecting disk displacement ;0.120".
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| LIMERICK - UNIT 1 3/4 6-45 Amendment No. W4,186
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| | |
| CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be maintained.
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| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| Without REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY, restore REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
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| SURVEILLANCE REQUIREMENTS 4.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be demon-strated by:
| |
| : a. Verifying in accordance with the Surveillance Frequency Control Program that the pressure within the reactor enclosure secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.*
| |
| : b. Verifying in accordance with the Surveillance Frequency Control Program that:
| |
| : 1. All reactor enclosure secondary containment equipment hatches and blowout panels are closed and sealed.
| |
| : 2. At least one door in each access to the reactor enclosure secondary containment is closed, except when the access opening is being used for entry and exit.
| |
| : 3. All reactor enclosure secondary containment penetrations not capable of being closed by OPERABLE secondary containment auto-matic isolation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers/valves secured in position.
| |
| : c. In accordance with the Surveillance Frequency Control Program:
| |
| : 1. Verifying that one standby gas treatment subsystem will. draw down the reactor enclosure secondary containment to greater than or equal to 0.25 inch of vacuum water gauge in less than or equal to 916 seconds with the reactor enclosure recirc system in operation and
| |
| : 2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.25 inch of vacuum water gauge in the reactor enclosure secondary containment at a flow rate not exceeding 2500 cfm with wind speeds of~ 7.0 mph as measured on the wind instrument on Tower 1, elevation 30' or, if that instrument is unavailable, Tower 2, elevation 159'.
| |
| *Not required to be met for 4 hours if analysis demonstrates one standby gas treatment subsystem is capable of establishing the required secondary containment vacuum.
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| LIMERICK - UNIT 1 3/4 6-46 Amendment No. g,+l-,-+/--Oe,+2-2:,-l-86,-2-2-0, 229
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| | |
| CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REFUELING AREA SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained.
| |
| APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment.
| |
| ACTION:
| |
| Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:
| |
| : a. Verifying in accordance with the Surveillance Frequency Control Program that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.*
| |
| : b. Verifying in accordance with the Surveillance Frequency Control Program that:
| |
| : 1. All refueling area secondary containment equipment hatches and blowout panels are closed and sealed.
| |
| : 2. At least one door in each access to the refueling area secondary containment is closed, except when the access opening is being used for entry and exit.
| |
| : 3. All refueling area secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic iso-lation dampers/valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers/valves secured in position.
| |
| : c. In accordance with the Surveillance Frequency Control Program:
| |
| Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal to 0.25 inch of vacuum water gauge in the refueling area secondary containment at a flow rate not exceeding 764 cfm.
| |
| *Not required to be met for 4 hours if analysis demonstrates one standby gas treatment subsystem is capable of establishing the required secondary containment vacuum.
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| LIMERICK - UNIT 1 3/4 6-47 Amendment No. ~.+l-.~.~.~.~.
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| 229
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| | |
| CONTAINMENT SYSTEMS REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.1 The reactor enclosure secondary containment ventilation system auto-matic isolation valves shall be OPERABLE:
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| With one or more of the reactor secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours either:
| |
| : a. Restore the inoperable valves to OPERABLE status, or
| |
| : b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or
| |
| : c. Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.
| |
| Otherwise, in OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REOUTREMENTS 4.6.5.2.1 Each reactor enclosure secondary containment ventilation system automatic isolation valve shall be demonstrated OPERABLE:
| |
| : a. Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
| |
| : b. In accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.
| |
| : c. By verifying the isolation time to be within its limit in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 6-48 Amendment No. 24, 49, -74, 45, 186
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| | |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED.TO THE TRM.
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| LIMERICK - UNIT 1 3/4 6-49 Amendment No. Q-4, 4-4-&,192
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| | |
| CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.
| |
| APPLICABILITY: When RECENTLY IRRADIATED FUEL is being handled in the secondary containment.
| |
| ACTION:
| |
| With one or more of the refueling area secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours either:
| |
| : a. Restore the inoperable valves to OPERABLE status, or
| |
| : b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or
| |
| : c. Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.
| |
| Otherwise, suspend handling of RECENTLY IRRADIATED FUEL in the refueling area secondary containment. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shall be demonstrated OPERABLE:
| |
| : a. Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
| |
| : b. In accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.
| |
| : c. By verifying the isolation time to be within its limit in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT 1 3/4 6-50 Amendment No. B,40,+l-,-l-Q..e.,+/-&5-, ~ . 227
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| | |
| THE INFORMATIONFROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM.
| |
| LIMERICK - UNIT I 3/4 6-51 Amendment No. 2*-, 192 4-g-E,
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| | |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM.
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| LIMERICK - UNIT 1 3/4 6-51a Amendment No. 2-4, 41O-,192
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| | |
| CONTAINMENT SYSTEMS STANDBY GAS TREATMENT SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and when (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE ALTE RA TI ON S .
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 1, 2, or 3:
| |
| : 1. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. When (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE ALTERATIONS:
| |
| : 1. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or suspend handling of irradiated fuel in the secondary containment and CORE ALTERATIONS. The provisions of Specification 3.0.3 are not applicable.
| |
| : 2. With both standby gas treatment subsystems inoperable, if in progress, suspend handling of irradiated fuel in the secondary containment and CORE ALTERATIONS. The provisions of Specification 3.0.3. are not applicable.
| |
| SURVEILLANCE REOlJIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.
| |
| LIMERICK - UNIT 1 3/4 6-52 Amendment No. ~.4-G,~.~. -&G-G, 227
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| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:
| |
| : 1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%
| |
| and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 5764 cfm + 10%.
| |
| : 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 0.5%
| |
| when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C (86 0 F), at a relative humidity of 70% and at a face velocity of 66 fpm.
| |
| : 3. Verify that when the fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III) when tested in accordance with ANSI N510-1980.
| |
| : 4. Verify that the pressure drop across the refueling area to SGTS prefilter is less than 0.25 inches water gage while operating at a flow rate of 2400 cfm +/- 10%.
| |
| : c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 0.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C (86 0 F), at a relative humidity of 70% and at a face velocity of 66 fpm.
| |
| : d. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 9.1 inches water gauge while operating the filter train at a flow rate of 8400 cfm +/- 10%.
| |
| LIMERICK - UNIT 1 3/4 6-53 Amendment No. 6, -33, W4, 4-12, 444, 186
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| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 2. Verifying that the fan starts and isolation valves necessary to draw a suction from the refueling area or the reactor enclosure recirculation discharge open on each of the following test signals:
| |
| a) Manual initiation from the control room, and b) Simulated automatic initiation signal.
| |
| : 3. Verifying that the temperature differential across each heater is > 15*F when tested in accordance with ANSI N510-1980.
| |
| : e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 5764 cfm i 10%. I
| |
| : f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05%
| |
| in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 5764 cfm i 10%. I
| |
| : 9. After any major system alteration:
| |
| : 1. Verify that when the SGTS fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III).
| |
| : 2. Verify that one standby gas treatment subsystem will drawdown reactor enclosure Zone I secondary containment to greater than or equal to 0.25 inch of vacuum water gauge in less than or equal to 916 seconds with the reactor enclosure recirculation I system in operation and the adjacent reactor enclosure and refueling area zones are in their isolation modes.
| |
| LIMERICK - UNIT I 3/4 6-54 Amendment No. 40, 122 0,I1 1997
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| | |
| CONTAINMENT SYSTEMS REACTOR ENCLOSURE RECIRCULATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.4 Two independent reactor enclosure recirculation subsystems shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one reactor enclosure recirculation subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With both reactor enclosure recirculation subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hors.
| |
| SURVEILLANCE REOUTREMENTS 4.6.5.4 Each reactor enclosure recirculation subsystem shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates properly.
| |
| : b. In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:
| |
| : 1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%
| |
| and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 60,000 cfm +/- 10%.
| |
| : 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 2.5%
| |
| when tested in accordance with ATM D3803-1989 at a temperature of 30 0 C (86 0 F) and a relative humidity of 70%.
| |
| : 3. Verifying a subsystem flow rate of 60,000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1980.
| |
| LIMERICK - UNIT 1 3/4 6-55 Amendment No. W4-, 1-44, 186
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows, the methyl iodide penetration of less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C (86 0 F) and a relative humidity of 70%.
| |
| : d. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the filter train at a flow rate of 60,000 cfm +/- 10%, verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.
| |
| : 2. Verifying that the filter train starts and the isolation valves which take suction on and return to the reactor enclosure open on each of the following test signals:
| |
| : a. Manual initiation from the control room, and
| |
| : b. Simulated automatic initiation signal.
| |
| : e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the-inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 60,000 cfm +/- 10%.
| |
| : f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow rate of 60,000 cfm +/- 10%.
| |
| LIMERICK - UNIT 1 3/4 6-56 Amendment No. 4-1-, ;4, 44, 186
| |
| | |
| CONTAINMENT SYSTEMS 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL PRIMARY'CONTAINMENT HYDROGEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.6.1 DELETED LIMERICK - UNIT 1 3/4 6-57 Amendment No. X8, -74,173
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| | |
| CONTAINMENT SYSTEMS DRYWFLL HYDROGEN MTXTNG SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.2 Four independent drywell unit cooler hydrogen mixing subsystems (IAV212, IBV212, IGV212, 1HV212) shall be OPERABLE with each subsystem consist-ing of one unit cooler fan.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| With one drywell unit cooler hydrogen mixing subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.6.6.2 Each drywell unit cooler hydrogen mixing subsystem shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by: I
| |
| : a. Starting the system from the control room, and
| |
| : b. Verifying that the system operates for at least 15 minutes.
| |
| LIMERICK - UNIT 1 3/4 6-58 Amendment No. 186
| |
| | |
| CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.6.6.3 The drywell and suppression chamber atmosphere oxygen concentration shall be less than 4% by volume.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1*, during the time period:
| |
| : a. Within 24 hours** after THERMAL POWER is greater than 15% of RATED THERMAL POWER, following startup, to
| |
| : b. Within 24 hours** prior to reducing THERMAL POWER to less than 15% of RATED THERMAL POWER, preliminary to a scheduled reactor shutdown.
| |
| ACTION:
| |
| With the drywell and/or suppression chamber oxygen concentration exceeding the limit, restore the oxygen concentration to within the limit within 24 hours or be in at least STARTUP within the next 8 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.6.6.3 The drywell and suppression chamber oxygen concentration shall be verified to be within the limit within 24 hours after THERMAL POWER is greater than 15% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program thereafter. I
| |
| *See Special Test Exception 3.10.5.
| |
| **Specification 3.6.1.8 is applicable during this 24 hour period.
| |
| LIMERICK - UNIT 1 3/4 6-59 Amendment No'. 186
| |
| | |
| - COMMON SYSTEM 3.7.1.1 At least the following independent residual heat removal service water (RHRSW) system subsystems, with each subsystem comprised of:
| |
| : a. Two OPERABLE RHRSW pumps, and
| |
| : b. An OPERABLE flow path capable of taking suction from the RHR service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water through one Unit 1 RHR heat exchanger, shall be OPERABLE:
| |
| : a. In OPERABLE CONDITIONS 1, 2, and 3, two subsystems.
| |
| : b. In OPERABLE CONDITIONS 4 and 5, the subsystem(s) associated with systems and components required OPERABLE by Specification 3.4.9.2, 3.9.11.1, and 3.9.11.2.
| |
| APPLICABI LITY; OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 1, 2, or 3:
| |
| : 1. With one RHRSW pump inoperable, restore the inoperable pump to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 2. With one RHRSW pump in each subsystem inoperable, restore at least one of the inoperable RHRSW pumps to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 3. With one RHRSW subsystem otherwise inoperable, restore the inoperable subsystem to OPERABLE status with at least one OPERABLE RHRSW pump within 72 hours, unless otherwise specified in a) or b) be1ow**, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| a) When the 'A' RHRSW subsystem is inoperable to allow for repairs of the 'A' RHRSW subsystem piping, with Limerick Generating Station Unit 2 shutdown, reactor vessel head removed and reactor cavity flooded, the 72-hour Allowed Outage Time may be extended to 7 days once every other calendar year with the following compensatory measures established:
| |
| ** Only one of these two Actions, either a.3.a) or a.3.b), may be entered on Unit 1 in a calendar year. However, if either Unit 2 TS LCD 3.7.1.1, Action a.3.a) or a.3.b) has previously been entered in the calendar year, then Unit 1 Action a.3.a) or a.3.b) may not be entered during that same calendar year.
| |
| LIMERICK - UNIT 1 3/4 7-1 Amendment No. %8,g& , 203
| |
| | |
| 3/4.7 PLANT SYSTEMS ACTION: (Continued)
| |
| : 1) The following systems and subsystems will be protected in accordance with applicable station procedures:
| |
| * 'B' RHRSW subsystem
| |
| * ' B' ESW loop
| |
| * 'B' and 'D' RHR subsystems
| |
| * 012, D14. 022. and D24 4kV buses and emergency diesel generators
| |
| * Division 2 and Division 4 Safeguard DC, and
| |
| : 2) The 'A' and 'B' loop of ESW return flow shall be aligned to the operable 'B' RHRSW return header only.
| |
| The ESW return valves to the 'B' RHRSW return header (i.e., HV-ll 015A and HV-11-015B) will be administratively controlled in the open position and de-energi prior to entering the extended AOT. The ESW return valves to the 'A' RHRSW return header (i .e., HV 11-011A and HV 11 011B) wi 11 be administratively controlled in the closed position and de energized as part of the work boundary.
| |
| b) When the 'B' RHRSW subsystem is inoperable to allow for repairs of the 'B' RHRSW subsystem piping, with Limerick Generating Station Unit 2 shutdown. reactor vessel head removed and reactor cavity flooded, the 72-hour Allowed Outage Time may be extended to 7 days once every other calendar year with the following compensatory measures established:
| |
| : 1) The following systems and subsystems will be protected in accordance with applicable station procedures:
| |
| * 'A' RHRSW subsystem
| |
| * ' A' ESW loop
| |
| * 'A' and 'C' RHR subsystems
| |
| * 011, D13, 021, and 023 4kV buses and emergency diesel generators
| |
| * Division 1 and Division 3 Safeguard DC, and
| |
| : 2) The 'A' and 'B' loop of ESW return flow shall be aligned to the operable 'A' RHRSW return header only.
| |
| The ESW return valves to the 'A' RHRSW return header (i.e .* HV 11 OllA and HV-U-OllB) will be administratively controlled in the open position and de-energized prior to entering the extended AOT. The ESW return valves to the 'B' RHRSW return header (i .e., HV-ll-015A and HV 11-015B) will be administratively controlled in the closed position and de energized as part of the work boundary.
| |
| : 4. With both RHRSW subsystems otherwise inoperable, restore at least one subsystem to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN* within the following 24 hours.
| |
| *Whenever both RHRSW subsystems are inoperable, if unable to attain COLD SHUTDOWN as requi by the ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
| |
| LIMERICK UNIT 1 3/4 7-1a Amendment No. ~,ge,~. 203 I
| |
| | |
| PLANTLSYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : 5. With two RHRSW pump/diesel generator pairs* inoperable, restore at least one inoperable RHRSW pump/diesel generator pair* to OPERABLE status within 30 days, or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 6. With three RHRSW pump/diesel generator pairs* inoperable, restore at least one inoperable RHRSW pump/diesel generator pair* to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 7. With four RHRSW pump/diesel generator pairs* inoperable, restore at least one inoperable RHRSW pump/diesel generator pair* to OPERABLE status within 8 hours, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. In OPERATIONAL CONDITION 3 or 4 with the RHRSW subsystem(s), which is associated with an RHR loop required OPERABLE by Specification 3.4.9.1 or 3.4.9.2, inoperable, declare the associated RHR loop inoperable and take the ACTION required by Specification 3.4.9.1 or 3.4.9.2, as applicable.
| |
| : c. In OPERATIONAL CONDITION 5 with the RHRSW subsystem(s), which is associated with an RHR loop required OPERABLE by Specification 3.9.11.1 or 3.9.11.2, inoperable, declare the associated RHR system inoperable and take the ACTION required by Specification 3.9.11.1 or 3.9.11.2, as applicable.
| |
| SURVEILLANCE RFOITREMFNTS 4.7.1.1 At least the above required residual heat removal service water system subsystem(s) shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying that each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
| |
| *A RHRSW pump/diesel generator pair consists of a RHRSW pump and its associated diesel generator. If either a RHRSW pump or its associated diesel generator becomes inoperable, then the RHRSW pump/diesel generator pair is inoperable.
| |
| LIMERICK - UNIT 1 3/4 7-2 Amendment No. 2-4, 49, 4-64, 186
| |
| | |
| PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM - COMMON SYSTEM 3.7.1.2 At least the following independent emergency service water system loops, with each loop comprised of:
| |
| : a. Two OPERABLE emergency service water pumps, and
| |
| : b. An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water to the associated Unit 1 and common safety related equipment, s hall be OPERABLE:
| |
| : a. In OPERATIONAL CONDITIONS 1, 2, and 3, two loops.
| |
| : b. In OPERATIONAL CONDITIONS 4, 5, and *, one loop.
| |
| APPLICABI LITY; OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and
| |
| * ACTION:
| |
| : a. In OPERATION CONDITION 1, 2, or 3:
| |
| : 1. With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 45 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 2. With one emergency service water pump in each loop inoperable.
| |
| restore at least one inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 3. With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable loop inoperab1e**. restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours# or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| *When handling irradiated fuel in the secondary containment.
| |
| **The diesel generators may be aligned to the OPERABLE emergency service water system loop provided confirmatory flow testing has been performed. Those diesel generators no aligned to the OPERABLE emergency service water system loop shall declared inoperable and the actions of 3.8.1.1 taken.
| |
| # During the extended 7-day Allowed Outage Time (AOT) specified by TS LCO 3.7.1.1, Action a.3.a) or a.3.b) to allow for RHRSW subsystem piping repairs, the 72 hour AOT for one inoperable emergency service water system loop may also be extended to 7 days for the same 7-day period.
| |
| LIMERICK UNIT 1 3/4 7 3 Amendment No. ,4{,},~,-l-M, 203
| |
| | |
| PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : 4. With three ESW pump/diesel generator pairs** inoperable, restore at least.one inoperable ESW pump/diesel generator pair** to OPERABLE status within 72 hours, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within.the following.24 hours.
| |
| : 5. With four ESW pump/diesel generator pairs** inoperable, restore at. least one inoperable ESW pump/diesel generator pair** to OPERABLE status within 8 hours, or be in at least. HQT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. In OPERATIONALCONDITION 4 or 5:
| |
| : 1. With only one emergency service water pump and its associated flowpath OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours or declare the associated safety related equipment inoperable and take the ACTION required by Specifications 3.5.2 and 3.8.1.2.
| |
| : c. In OPERATIONAL CONDITION *
| |
| .1. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at. least one flow path to OPERABLE status within 72 hours or verify
| |
| .--- - - ..- adequate cool i ng--rema-i ns- -avai-l ab-l e -for-the---di-esel - generators -... .
| |
| required to be OPERABLE or declare the associated diesel generator(s) inoperable and take the ACTION required by Specification 3.8.1.2. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REOUIREMENT 4.7.1.2 At least .the above required emergency service water system loop(s) shall be demonstrated. OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) that is not locked, sealed, or otherwise secured in position, is in its correct position.
| |
| : b. In accordance with the Surveillance Frequency Control Program by verifying that:
| |
| : 1. Each automatic valve actuates to its correct position on its appropriate ESW pump start signal.
| |
| : 2. Each pump starts automatically when its associated diesel generator starts.
| |
| *When handling irradiated fuel in the secondary containment.
| |
| **An ESW pump/diesel generator pair consists of an ESW pump and its associated diesel generator. If either an ESW pump or its associated diesel generator becomes inoperable, then the ESW pump/diesel generator pair is inoperable.
| |
| 92 LIMERICK - UNIT 1 3/4 7-4 Amendment No. 2-T,40,4,;ý,4-9,1
| |
| | |
| PLANT SYSTEMS ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION ....
| |
| 3.7.1.3 The spray pond shall be OPERABLE with:
| |
| : a. A minimum pond water level at or above elevation 250' 10" Mean Sea Level, and
| |
| : b. A pond water temperature of less than or equal to 88°F.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and
| |
| * ACTION:
| |
| With the requirements of the above specification not satisfied:
| |
| : a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| : b. In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.
| |
| : c. In OPERATIONAL CONDITION *, declare the emergency service water system inoperable and take the ACTION required by Specification 3.7.1.2. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE:
| |
| : a. By verifying the pond water level to be greater than its limit in accordance with the Surveillance Frequency Control Program.
| |
| : b. By verifying the water surface temperature (within the upper two feet of the surface) to be less than or equal to 88°F:
| |
| : 1. in accordance with the Surveillance Frequency Control Program when the spray pond temperature is greater than or equal to 80'F; and
| |
| : 2. in accordance with the Surveillance Frequency Control Program when the spray pond temperature is greater than or equal to 85°F; and
| |
| : 3. in accordance with the Surveillance Frequency Control Program when the spray pond temperature is greater than 320F.
| |
| : c. By verifying all piping above the frost line is drained:
| |
| : 1. within one (1) hour after being used when ambient air temperature is below 40°F; or
| |
| : 2. when ambient air temperature falls below 40°F if the piping has not been previously drained.
| |
| *When handling irradiated fuel in the secondary containment.
| |
| LIMERICK - UNIT 1 3/4 7-5 Amendment No. 2-5, 40, 40,186
| |
| | |
| PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE.
| |
| NOTE: The main control room envelope (CRE) boundary may be opened intermittently under administrative control APPLICABILITY: All OPERATIONAL CONDITIONS and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 1, 2, or 3:
| |
| : 1. With one control room emergency fresh air supply subsystem inoperable for reasons other than Condition a.2, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 2. With one or more control room emergency fresh air supply subsystems inoperable due to an inoperable CRE boundary,
| |
| : a. Initiate action to implement mitigating actions immediately or be in HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours; and
| |
| : b. Within 24 hours, verify mitigating actions ensure CRE occupant exposures to radiological and chemical hazards will not exceed limits and actions to mitigate exposure to smoke hazards are taken or be in HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours; and
| |
| : c. Restore CRE boundary to operable status within 90 days or be in HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. In OPERATIONAL CONDITION 4, 5, or when RECENTLY IRRADIATED FUEL is being handled in the secondary containment:
| |
| : 1. With one control room emergency fresh air supply subsystems inoperable for reasons other than Condition b.3, restore the inoperable subsystem to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.
| |
| : 2. With both control room emergency fresh air supply subsystems inoperable for reasons other than Condition b.3, immediately suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment. The provisions of Specification 3.0.3 are not applicable.
| |
| LIMERICK - UNIT 1 3/4 7-6 Amendment No. 4-G,+--+/--,~.~. ~ . 227
| |
| | |
| PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : 3. With one or more control room emergency fresh air subsystems inoperable due to an inoperable CRE boundary, immediately suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment.
| |
| The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.7.2.1 Each control room emergency fresh air supply subsystem shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying the control room air temperature to be less than or equal to 85°F effective temperature.
| |
| : b. In accordance with the Surveillance Frequency Control Program on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.
| |
| : c. In accordance with the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:
| |
| : 1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm +/- 10%.
| |
| : 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 2.5%
| |
| when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.
| |
| : 3. Verifying a subsystem flow rate of 3000 cfm +/- 10% during subsystem operation when tested in accordance with ANSI N510-1980.
| |
| : d. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration of less than 2.5%
| |
| when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.
| |
| LIMERICK - UNIT 1 3/4 7-7 Amendment No. ~.4-G,+l-,-+/--44,~.
| |
| -+/--&&, J..gg, 227
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
| |
| : e. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the subsystem at a flow rate of 3000 cfm +/- 10%; verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.
| |
| : 2. Verifying that on each of the below chlorine isolation mode actuation test signals, the subsystem automatically switches to the chlorine isolation mode of operation and the isolation valves close within 5 seconds:
| |
| a) Outside air intake high chlorine, and b) Manual initiation from the control room.
| |
| : 3. Verifying that on each of the below radiation isolation mode actuation test signals, the subsystem automatically switches to the radiation isolation mode of operation:
| |
| a) Outside air intake high radiation, and b) Manual initiation from control room.
| |
| : f. -After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetra-tion and bypass leakage testing acceptance criteria of less than 0.05%
| |
| in accordance with ANSI N510-1980 while operating the system at a flow rate of 3000 cfm +/- 10%.
| |
| : q. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 3000 cfm +/- 10%.
| |
| 4.7.2.2 The control room envelope boundary shall be demonstrated OPERABLE:
| |
| : a. At a frequency in accordance with the Control Room Envelope Habitability Program by performance of control room envelope unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.
| |
| LIMERICK - UNIT 1 3/4 7-8 Amendment No. B,4O,.-4,-1-&* ,
| |
| 4-9&6,188
| |
| | |
| PLANT SYSTEMS 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig.
| |
| ACTION:
| |
| : a. With the RCIC system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours.
| |
| : b. DELETED
| |
| : c. Specification 3.0.4.b is not applicable to RCIC.
| |
| SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Verifying locations susceptible to gas accumulation are sufficiently filled with water.
| |
| : 2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.**
| |
| : 3. Verifying that the pump flow controller is in the correct position.
| |
| : b. In accordance with the Surveillance Frequency Control Program by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1040 + 13, - 120 psig.*
| |
| * The prov1s1ons of Specification 4.0.4 are not applicable, provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam pressure to less than 150 psig within the following 72 hours.
| |
| ** Not required to be met for system vent flow paths opened under administrative control.
| |
| LIMERICK - UNIT 1 3/4 7-9 Amendment No. -&9,-+/--Ge,-+/-6-9,~.-&+/--+/-. 216
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : c. In accordance with the Surveillance Frequency Control Program by: I
| |
| : 1. Performing a system functional test which includes simulated automatic actuation and restart and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded.
| |
| : 2. Verifying that the system will develop a flow of greater than or equal to 600 gpm in the test flow path when steam is supplied to the turbine at a pressure of 150 + 15, - 0 psig.*
| |
| : 3. Verifying that the suction for the RCIC system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank water level-low signal.
| |
| : 4. Performing a CHANNEL CALIBRATION of the RCIC system discharge line "keep filled" level alarm instrumentation.
| |
| *The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the tests. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam pressure to less than 150 psig within the following 72 hours.
| |
| LIMERICK - UNIT 1 3/4 7-10 Amendment No. 2-9, .74,186
| |
| | |
| PLANT SYSTEMS 3/4.7.4 SNUBBERS LIMITIN~ CQNDITION FOR. OPERATION 3.7.4 All snubbers shall be OPERABLE.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. OPERATIONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERABLE in those OPERATIONAL CONDITIONS.
| |
| ACTION:
| |
| With one or more snubbers inoperable on any system, within 72 hours replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per the Snubber Program on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.
| |
| SURVEILLANCE REQUIREMENTS 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of the Snubber Program.
| |
| LIMERICK - UNIT 1 3/4 7-11 Amendment No. -51:, 2 2 3
| |
| | |
| PLANT SYSTEMS TABLE 4.7.4.1 (Deleted)
| |
| TEOiNICAL SPECIFICATION PAGES 3/4 7-lla THROUGH 3/4 7-llb HAVE BEEN INTENTIONALLY OMITTED LIMERICK - UNIT 1 3/4 7-lla Amendment No. 5+/-, 223
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REOUIREME&TS (Continued)
| |
| THE INFORMATION FROM TECHNICAL SPECIFICATIONS SECTION 4.7.4.d HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM) SNUBBERS SECTION. TECHNICAL SPEOFICATIONS PAGES 3/4 7-13 THROUGH 7-16 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 7-12 Amendment No. :l, 2 2 3
| |
| | |
| PLANT SYSTEMS 3/4.7.5 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.5 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With a sealed source having removable contamination in excess of the above limit, withdraw the sealed source from use and either:
| |
| : 1. Decontaminate and repair the sealed source, or
| |
| : 2. Dispose of the sealed source in accordance with Commission Regulations.
| |
| : b. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REOUIREMENTS 4.7.5.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:
| |
| : a. The licensee, or
| |
| : b. Other persons specifically authorized by the Commission or an Agreement State.
| |
| The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample.
| |
| 4.7.5.2 Test Frequencies - Each category of sealed sources, excluding startup sources and fission detectors previously subjected to core flux, shall be tested at the frequency described below.
| |
| : a. Sources in use - In accordance with the Surveillance Frequency Control Program for all sealed sources containing radioactive material:
| |
| : 1. With a half-life greater than 30 days, excluding Hydrogen 3, and
| |
| : 2. In any form other than gas.
| |
| LIMERICK - UNIT I 3/4 7-17 Amendment No. *--, 186
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.
| |
| : c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.
| |
| 4.7.5.3 DELETED LIMERICK - UNIT 1 3/4 7-18 Amendment No. 211
| |
| | |
| PLANT SYSTEMS Section 3/4.7.6 through 3/4.7.7 (Deleted) I THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAVE BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRH) FIRE PROTECTION SECTION. TECHNICAL SPECIFICATIONS PAGES 3/4 7-19 THROUGH 3/4 7-32 HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 7-19 Amendment No. 14, 104
| |
| `O'V 2 0 19i
| |
| | |
| PLANTLSYSTEMS 3/4.7.8 MAIN TURBINE BYPASS SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 The main turbine bypass system shall be OPERABLE as determined by the number of operable main turbine bypass valves being greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
| |
| ACTION: With the main turbine bypass system inoperable, restore the system to OPERABLE status within 1 hour or take the ACTION required by Specification 3.2.3.c.
| |
| SURVEILLANCE REOUIREMENTS 4.7.8 The main turbine bypass system shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program:
| |
| : a. By cycling each turbine bypass valve through at least one complete cycle of full travel,
| |
| : b. By performing a system functional test which includes simulated automatic actuation, and by verifying that each automatic valve actuates to its correct position, and
| |
| : c. By determining TURBINE BYPASS SYSTEM RESPONSE TIME to be less than or equal to the value specified in the CORE OPERATING LIMITS REPORT.
| |
| LIMERICK - UNIT I 3/4 7-33 Amendment No. .&, ;4,186
| |
| | |
| 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. SOURCES - OPERATING 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
| |
| : a. Two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system, and
| |
| : b. Four separate and independent diesel generators, each with:
| |
| : 1. A separate day tank containing a minimum of 250 gallons of fuel,
| |
| : 2. A separate fuel storage system containing a minimum of ,500 gallons of fuel, and
| |
| : 3. A separate fuel transfer pump.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.
| |
| sources by performing Surveillance Requirement 4.8.1.1.1.a within 24 hours and at least once per 7 days thereafter. If the diesel generator became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining operable diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at a time, within 24 hours, unless the absence of any potential common-mode failure for the remaining diesel generators is determined. Restore the inoperable diesel generator to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. See also ACTION e.
| |
| : b. With two diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.
| |
| sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter. If either of the diesel generators became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining di generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at a time, within 8 hours, unless the absence of any potential common-mode failure for the remaining diesel generators is determined. Restore at least one of the inoperable diesel generators to OPERABLE status within 72 hours* or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. See also ACTION e.
| |
| * During the extended 7-day Allowed Outage Time CAOT) specified by TS LCO 3.7.1.1, Action a.3.a) or a.3.b) to allow for RHRSW subsystem piping repairs, the 72-hour AOT for two inoperable diesel generators may also be extended to 7 days for the same 7 day period.
| |
| LIMERICK - UNIT 1 3/4 8 1 Amendment No. ~.~,~,+9J,203
| |
| | |
| 3/4.8 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : c. With three diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within I hour and at least once per 8 hours thereafter; and perform Surveillance Requirement 4.8.1..1.2.a.4 for the remaining diesel generator, within 1 hour. Restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. See also ACTION e.
| |
| : d. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter.
| |
| If the diesel generator became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at a time, within 8 hours, unless the absence of any potential common-mode failure for the remaining diesel generators is determined. Restore at least two offsite circuits to OPERABLE status within 72 hours from the time of initial lossor be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. See also ACTION e.
| |
| LIMERICK - UNIT 1 3/4 8-1a Amendment No. 42, 4-, 189
| |
| | |
| ELECTRICAL POWER SYSTEMS ACTION: (Continued)
| |
| : e. In addition to the ACTIONS above:
| |
| : 1. For two train systems, with one or more diesel generators of the above required A.C. electrical power sources inoperable, verify within 2 hours and at least once per 12 hours thereafter that at least one of the required two train system subsystem, train, components, and devices is OPERABLE and its associated diesel generator is OPERABLE. Otherwise, restore either the inoperable diesel generator or the inoperable system subsystem to an OPERABLE status within 72 hours* or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : 2. For the LPCI systems, with two or more diesel generators of the above required A.C. electrical power sources inoperable, verify within 2 hours and at least once per 12 hours thereafter that at least two of the required LPCI system subsystems, trains, components, and devices are OPERABLE and its associated di generator is OPERABLE. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| This ACTION does not apply for those systems covered in Specifications 3.7.1.1. and 3.7.1.2.
| |
| * During the extended 7-day Allowed Outage me (AOT) specified by TS LCO 3.7.1.1, Action a.3.a) or a.3.b) to allow for RHRSW subsystem piping repairs, the 72-hour AOT may also be extended to 7 days for the same 7-day period.
| |
| LIMERICK UNIT 1 3/4 8 2 Amendment No. ~,4G, 203
| |
| | |
| FIFCTRTCAI POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : f. With one offsite circuit of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter. Restore at least two offsite circuits to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and COLD SHUTDOWN within the following 24 hours.
| |
| : g. With two of the above required offsite circuits inoperable, restore at least one of the inoperable offsite circuits to OPERABLE status within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours. With only one offsite circuit restored to OPERABLE status, restore at least two offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : h. With one offsite circuit and two diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance, Requirements 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter. If either of the diesel generators became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at a time, within 8 hours, unless the absence of any potential common-mode failure for the remaining diesel generators is determined. Restore at least one of the above required inoperable A.C. sources to OPERABLE status withi~n 12 hours or be in a at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. Restore at least two offsite circuits and at least three of the above required diesel generators to OPERABLE status within 72 hours from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. See also ACTION e.
| |
| : i. Specification 3.0.4.b is not applicable to diesel generators.
| |
| LIMERICK - UNIT 1 3/4 8-2a Amendment No. 4-, 4-64,189
| |
| | |
| F1 FflTRTC.AI PnWFR SYSTFMSý
| |
| .SURVEILLANCE REOUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:
| |
| : a. Determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability, and.
| |
| : b. Demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by transferring, manually and automatically, unit power supply from the normal circuit to the alternate circuit.
| |
| 4.8.1.1.2 Each of the above required diesel generators shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program on a STAGGERED TEST BASIS by:
| |
| : 1. Verifying the fuel level in the day fuel tank.
| |
| : 2. Verifying the fuel level in the fuel storage tank.
| |
| : 3. Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day fuel tank.
| |
| : 4. Verify that the diesel can start* and gradually accelerate to synchronous speed with generator voltage and frequency at 4280 +/- 120 volts and 60 +/- 1.2 HZ.
| |
| : 5. Verify diesel is synchronized, gradually loaded* to an indicated 2700-2800 KW** and operates with this load for at least 60 minutes.
| |
| : 6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
| |
| : 7. Verifying the pressure in all diesel generator air start receivers to be greater than or equal to 225 psig.
| |
| *This test shall be conducted in accordance with the manufacturer's recommendations regarding engine pre-lube and.warmup procedures, and as applicable regarding loading and shutdown recommendations.
| |
| **This band is meant as guidance to avoid routine overloading of the engine.
| |
| Loads in excess of this band for special testing under direct monitoring by the manufacturer or momentary variations due to changing bus loads shall not invalidate the test.
| |
| LIMERICK - UNIT I 3/4 8-3 Amendment No. *,-7,4-O,48.,189
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. By removing accumulated water:
| |
| : 1) From the day tank in accordance with the Surveillance Frequency Control Program and after each occasion when the diesel is operated for greater than 1 hour, and
| |
| : 2) From the storage tank in accordance with the Surveillance Frequency Control Program.
| |
| : c. By sampling new fuel oil in accordance with ASTM D4057-81 prior to addition to the storage tanks and:
| |
| : 1) By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has:
| |
| a) An API Gravity of within 0.3 degrees at 60'F or a specific gravity of within 0.0016 at 60/60 0 F, when compared to the supplier's certificate or an absolute specific gravity at 60/60'F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity at 60'F of greater than or equal to 27 degrees but less than or equal to 39 degrees.
| |
| b) A kinematic viscosity at 40 0 C of greater than or equal to 1.9 centistokes, but less, than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification.
| |
| c) A flash point equal to or greater than 125°F, and d) A clear and bright appearance with proper color when tested in accordance with ASTM D4176-82.
| |
| : 2) By verifyihg within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.
| |
| : d. In accordance with the Surveillance Frequency Control Program by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-78, and verifying that total particulate contamination is less than 10 mg/liter when checked in accordance with ASTM D2276-78, Method A, except that the filters specified in ASTM D2276-78, Sections 5.1.6 and 5.1.7, may have a nominal pore size of up to three (3) microns.
| |
| : e. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Deleted
| |
| : 2. Verifying each diesel generator's capability to reject a load of greater than or equal to that of its single largest post-accident load, and:
| |
| I a) Following load rejection, the frequency is
| |
| * 66.5 Hz; b) Within 1.8 seconds following the load rejection, voltage is 4285
| |
| +/- 420 volts, and frequency is 60 +/- 1.2 Hz; and c) After steady-state conditions are reached, voltage is maintained at 4280 +/- 120 volts.
| |
| LIMERICK - UNIT 1 3/4 8-4 Amendment No. 21-,4,4-44,-1-24-,
| |
| 4,4--3,4-,8, 189
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
| |
| : 3. Verifying the diesel generator capability to reject a load of 2850 kW without tripping. The generator voltage shall not exceed 4784 volts during and following the load rejection.
| |
| : 4. Simulating a loss-of-offsite power by itself, and:
| |
| a) Verifying deenergization of the emergency busses and load shedding from the emergency busses.
| |
| b) Verifying the diesel generator starts* on the auto-start signal, energizes the emergency busses within 10 seconds, energizes the auto-connected loads through the individual load timers and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4280 +/- 120 volts and 60 +/- 1.2 Hz during this test.
| |
| : 5. Verifying that on an ECCS actuation test signal, without loss-of-offsite power, the diesel generator starts* on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall reach 4280 +/- 120 volts and 60 +/- 1.2 Hz within 10 seconds after the auto-start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test.
| |
| : 6. Simulating a loss-of-offsite power in conjunction with an ECCS actuation test signal, and:
| |
| a) Verifying deenergization of the emergency busses and load shedding from the emergency busses.
| |
| b) Verifying the diesel generator starts* on the auto-start signal, energizes the emergency busses within 10 seconds, energizes the auto-connected shutdown loads through the individual load timers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4280 +/- 120 volts and 60 +/- 1.2 Hz during this test.
| |
| : 7. Verifying that all automatic diesel generator trips, except engine overspeed and generator differential over-current are automatically bypassed upon an ECCS actuation signal.
| |
| * This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warm-up procedures, and as applicable regarding loading and shutdown recommendations.
| |
| LIMERICK - UNIT I 3/4 8-5 Amendment No. -, W-4, 4-04*,186
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 8. a) Verifying the diesel generator operates* for at least 24 hours.
| |
| During the first 2 hours of this test, the diesel generator shall be loaded to an indicated 2950-3050 kW** and during the remaining 22 hours of this test, the diesel generator shall be loaded to an indicated 2700-2800 kW**.
| |
| b) Verifying that, within 5 minutes of shutting down the diesel generator after the diesel generator has operated* for at least 2 hours at an indicated 2700-2800 kW**, the diesel generator starts*.
| |
| The generator voltage and frequency shall reach 4280 +/- 120 volts and 60 +/- 1.2 Hz within 10 seconds after the start signal.
| |
| : 9. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 3100 kW.
| |
| : 10. Verifying the diesel generator's capability to:
| |
| a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
| |
| : 11. Verifying that with the diesel generator operating in a test mode and connected to its bus, a simulated ECCS actuation signal overrides the test mode by (1) returning the diesel generator to standby operation, and (2) automatically energizes the emergency loads with offsite power.
| |
| : 12. Verifying that the automatic load sequence timers are OPERABLE with the interval between each load block within +/- 10% of its design interval.
| |
| * This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warmup procedures, and as applicable regarding loading and shutdown recommendations.
| |
| ** This band is meant as guidance to avoid routine overloading of the engine.
| |
| Loads in excess of this band for special testing under direct monitoring by the manufacturer or momentary variations due to changing bus loads shall not invalidate the test.
| |
| LIMERICK - UNIT 1 3/4 8-6 Amendment No. 3, -74, 4-a, 4-04, 4-36, 186
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
| |
| : 13. Verifying that the following diesel generator lockout I features prevent diesel generator starting only when required:
| |
| a) Control Room Switch In Pull-To-Lock (With Local/Remote Switch in Remote) b) Local/Remote Switch in Local c) Emergency Stop
| |
| : f. In accordance with the Surveillance Frequency Control Program or I after any modifications which could affect diesel generator interdependence by starting* all four diesel generators simultaneously, during shutdown, and verifying that all four diesel generators accelerate to at least 882 rpm in less than or equal to 10 seconds.
| |
| : g. In accordance with the Surveillance Frequency Control Program by: I
| |
| : 1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite or equivalent solution, and
| |
| : 2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code in accordance with ASME Code Section XI Article IWD-5000.
| |
| *This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warmup procedures, and as applicable regarding loading and shutdown recommendations.
| |
| LIMERICK - UNIT 1 3/4 8- 7 Amendment No. 4, ;74, 186
| |
| | |
| -L-CT-,CA T(- E SY!STEMS1F
| |
| - -A SURVEILLANCE REOUIREMENTS (Continued)
| |
| : h. In accordance with the Surveillance Frequency Control Program the diesel generator shall be started* and verified to accelerate to synchronous speed in less than or equal to 10 seconds. The generator voltage and frequency shall reach 4280 +/- 120 volts and 60 +/- 1.2 Hz within 10 seconds after the start signal. The diesel generator shall be started for this test by using one of the following signals:
| |
| a) Manual***
| |
| b) Simulated loss-of-offsite power by itself.
| |
| c) Simulated loss-of-offsite power in conjunction with an ECCS actuation test signal.
| |
| d) An ECCS actuation test signal by itself.
| |
| The generator shall be manually synchronized to its appropriate emergency bus, loaded to an indicated 2700-2800 KW** and operate for at least 60 minutes. This test, if it is perfori,,ed so iL coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5, may also serve to concurrently meet those requirements as well.
| |
| 4.8.1.1.3 Deleted
| |
| *This test shall be conducted in accordance with the manufacturer's
| |
| ... J.mm)i ... - .. i.. ...1/4 .. prellube and .,rmup pr e, , and as applicable regarding loading and shutdown recommendations.
| |
| **This band is meant as guidance to avoid routine overloading of the engine. Loads in excess of this band for special testingkunder dire'ct monitoring by the manufacturer or momentary variations due to changing bus loads shall not invalidate the test.
| |
| ***If diesel generator started manually from the control room, 10 seconds after the automatic prelube period.
| |
| `MýýDTýV - INTT I 3/4 8- 7a Amendment No. -a, , go,189
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 8-8 Amendment No. -32, , 189
| |
| | |
| ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
| |
| : a. One circuit between the offsite transmission network and the onsite Class lE distribution system, and
| |
| : b. Two diesel generators each with:
| |
| : 1. A day fuel tank containing a minimum of 250 gallons of fuel.
| |
| : 2. A fuel storage system containing a minimum of 33,500 gallons of fuel.
| |
| : 3. A fuel transfer pump.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and*
| |
| ACTION:
| |
| : a. With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment, and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet above the reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.
| |
| : b. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1 and 4.8.1.1.2.
| |
| *When handling irradiated fuel in the secondary containment.
| |
| LIMERICK - UNIT 1 3/4 8-9 Amendment No. Ji, -+/---9--i, -+/---9-J, 227
| |
| | |
| i ELECTRICAL PG.iER SYSTEMS 3/4.8.2 D.C. SOURCES D.C. SOURCES - 0PERATi'JG LIMTTING CONDITION FnP OPPRATtON 3.8.2.1 As a minimum, the follow.ing D.C. electrical power sources shall be OPERABLE:
| |
| : a. Division 1, Consistinq of:
| |
| : 1. 125-Volt Battery lAI (IAlD101).
| |
| : 2. 125-Volt Battery 1A2 (1A20101).
| |
| : 3. 125-Volt Battery Charger IBCA1 (QA1D103).
| |
| : 4. 125-Volt Battery Charger 1BCA2 (1A2D103).
| |
| : b. Division 2, Consisting of:
| |
| : 1. 125-Volt Battery 1B1 (iB0l101).
| |
| : 2. 125-Volt Battery 1B2 (1B2D101).
| |
| : 3. 125-Volt Battery Charger 1BCBI (1B1D103).
| |
| : 4. 125-Volt Battery Charger 1BCB2 (1B2D103).
| |
| : c. Division 3. Consisting of:
| |
| : 1. 125-Volt Battery 1C (lCD101).
| |
| : 2. 125-Volt Battery Charger 1BCC (lCD103).
| |
| : d. Division 4, Consisting of:
| |
| : 1. 125-Volt Battery ID (lDD101).
| |
| : 2. 125-Volt Battery Charger lBCD (lDD103).
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one or two battery chargers on one division inoperable:
| |
| : 1. Restore battery termir'al vo tage to greater than or equal to the minimum established float voltaqe within 2 hours,
| |
| : 2. Verify associated Division 1 or 2 float current < 2 amps, or Division 3 or 4 float current < I amp awithin 18 hours and once per 12 hours thereafter, and
| |
| : 3. Restore battery charger(s) to OPERABLE status within 7 days.
| |
| : b. With one or more batteries inoperable due to:
| |
| : 1. One or two batteries on one division with one or more battery cells float voltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hours for affected battery(s) and restore affected cell(s) voltage
| |
| > 2.07 volts within 24 hours.
| |
| : 2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4 with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours for affected battery(s) ard restore battery float current to within iimits within 18 hours.
| |
| LIMERICK - UNIT 1 3/4 8-10 Amendment No. 164
| |
| | |
| 3.
| |
| ELECTPICAL POQ'JER SYSTEMS ITMJTTTG CONDITION FOR OPERATION ACTION: (Continued)
| |
| : 3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if electrolyte level was below the top of the plates restore electrolyte level to above top of plates within 8 hours and verify no evidence of leakage(*) within 12 hours. In all cases, restore electrolyte level to greater than or equal to minimum established design limits within 31 days.
| |
| : 4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restore battery pilot cell temperature to greater than or equal to minimum established design limits within 12 hours.
| |
| : 5. Batteries in more than one division affected, restore battery parameters for all batteries in all but one division to within limits within 2 hours.
| |
| : 6. (i) Any battery having both (Action b.1) one or more battery cells float voltage < 2.07 volts and (Action b.2) float current not within limits. and/or (ii) Any battery not meeting any Action b.1 through b.5, Restore the battery parameters to within limits within 2 hours.
| |
| : c. With any battery(ies) on one division of the above required D.C. electrical power sources inoperable for reasons other than Action b., restore the inoperable division battery to OPERABLE status within 2 hours.
| |
| Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| (*i Contrary to the provisions of Specification 3.0.2, if electrolyte level was below the top of the plates, the verification that there is no evidence of leakage is required to be completed regardless of when electrolyte level is restored.
| |
| L11MERICK - UNIT I 3/4 8-10a LAmendment No. 1641
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS 4.8.2.1 Each of the above required division batteries and chargers shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by verifying that:
| |
| : 1. Each Division 1 and 2 battery float current is s 2 amps, and Division 3 and 4 battery float current is s 1 amp when battery terminal voltage is greater than or equal to the minimum established float voltage of 4.8.2.1.a.2, and
| |
| : 2. Total battery terminal voltage for each 125-volt battery is greater than or equal to the minimum established float voltage.
| |
| : b. In accordance with the Surveillance Frequency Control Program by verifying that:
| |
| : 1. Each battery pilot cell voltage is > 2.07 volts,
| |
| : 2. Each battery connected cell electrolyte level is greater than or equal to minimum established design limits, and
| |
| : 3. The electrolyte temperature of each pilot cell is greater than or equal to minimum established design limits.
| |
| : c. In accordance with the Surveillance Frequency Control Program by verifying that each battery connected cell voltage is > 2.07 volts.
| |
| : d. In accordance with the Surveillance Frequency Control Program by verifying that:
| |
| : 1. The battery chargers will supply the currents listed below at greater than or equal to the minimum established float voltage for at least 4 hours.
| |
| Charger Current (Amperes) 1BCA1 300 1BCA2 300 1BCB1 300 IBCB2 300 IBCC 75 1BCD 75
| |
| : 2. The battery capacity is adequate to supply and maintain in OPERABLE status the required emergency loads for the design duty cycle when subjected to a battery service test.
| |
| LIMERICK - UNIT 1 3/4 8-11 Amendment No. *4, 92, 464286 Corrpr-tMd bv lpttepr dited Jun*' 10 1995
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
| |
| : e. In accordance with the Surveillance Frequency Control Program by verifying I that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test or modified performance discharge test. The modified performance discharge test may be performed in lieu of the battery service test (Specification 4.8.2.1.d.2).
| |
| : f. Performance discharge tests or modified performance discharge tests of battery capacity shall be given as follows:
| |
| : 1. In accordance with the Surveillance Frequency Control Program when:
| |
| (a) The battery shows degradation or (b) The battery has reached 85% of expected life with battery capacity < 100% of manufacturer's rating, and
| |
| : 2. In accordance with the Surveillance Frequency Control Program when I battery has reached 85% of expected life with battery capacity 100% of manufacturer's rating.
| |
| LIMERICK - UNIT 1 3/4 8-12 Amendment No. W74, 92, 4-64,186 Corrccted by letter dated june 19, 1995
| |
| | |
| TABLE 4.8.2.1-1 (DELETED) I THE INFOPRAMTIDiN ON THIS PAGE HAS BEEN DELETED I
| |
| I I
| |
| I LIMERICK - UNIT I 3/4 8- 13 Amendment No. 4-a4, 164
| |
| | |
| i I
| |
| ELC-~A~-DP, .. SV STr.- IS D.C. SOURCES - SHU1TDOWN LrNI-JTIltj C0,,D4DTTON FOR OPFRATiOQ; 3.8.2.2 As a minimum, two of the follo-wing four divisions of the D.C.
| |
| electric& power sources system shall be OPERABLE wJth:
| |
| : a. Division 1, Consistirc
| |
| : 1. 125-Volt Battery 1A1 M1AID101).
| |
| : 2. 125-Volt Battery 1A2 (142D101).
| |
| : 3. 125-Volt Battery Charger IBCA1 (HA1DI03).
| |
| : 4. 125-Volt Battery Charger 1BCA2 (lA2D103).
| |
| : b. Division 2, Consisting of:
| |
| : 1. 125-Volt Battery IBI (IBlDI01).
| |
| : 2. 125-Volt Battery 1B2 (IB2D101).
| |
| : 3. 125-Volt Batterv Charger IBCB1 (1B1D1O3).
| |
| : 4. 125-Volt Battery Charger 1BCB2 (IB2D103).
| |
| : c. Division 3, Consisting of:
| |
| : 1. 125-Volt Battery iC (ICD101).
| |
| : 2. 125-Volt Battery Charger 1BCC (ICD103).
| |
| : d. Division 4, Consisting of:
| |
| : 1. 125-Volt Battery ID (UDD101).
| |
| : 2. 125-Volt Battery Charger 1BCD C1DD103).
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and *.
| |
| ACTION:
| |
| : a. With one or two required battery chargers on one required division inoperable:
| |
| : 1. Restore battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours,
| |
| : 2. Verifv associated Division 1 or 2 float current < 2 amps, or Division 3 or 4 float current < 1 amp within 18 hours and once per 12 hours thereafter. and
| |
| : 3. Restore battery charger(s) to OPERABLE status within 7 days.
| |
| : b. With one or more required batteries inoperable due to:
| |
| : 1. One or two batteries on one division with one or more battery cells float voltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hours for affected battery(s) and restore affected cell(s) voltage > 2.07 volts within 24 hours.
| |
| *When handling irradiated fuel in the secondary containment.
| |
| LIMERICK - UNIT 1 3/4 8-14 Amendment No. 164
| |
| | |
| ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : 2. Division 1 or 2 with float current> 2 amps, or with Division 3 or 4 with float current> 1 amp, perform 4.8.2.1.a.2 within 2 hours for affected battery(s) and restore battery float current to within limits within 18 hours.
| |
| : 3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if electrolyte level was below the top of the plates restore electrolyte level to above top of plates within 8 hours and verify no evidence of leakage(*) within 12 hours. In all cases, restore electrolyte level to greater than or equal to minimum established design limits within 31 days.
| |
| : 4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restore battery pilot cell temperature to greater than or equal to minimum established design limits within 12 hours.
| |
| : 5. Batteries in more than one division affected, restore battery parameters for all batteries in one division to within limits within 2 hours.
| |
| : 6. (i) Any battery having both (Action b.l) one or more battery cells float voltage< 2.07 volts and (Action b.2) float current not within limits, and/or (ii) Any battery not meeting any Action b.l through b.5, Restore the battery parameters to within limits within 2 hours.
| |
| : c. 1. With the requirements of Action a. and/or Action b. not met, or
| |
| : 2. With less than two divisions of the above required D.C. electrical power sources OPERABLE for reasons other than Actions a. and/orb.,
| |
| Suspend CORE ALTERATIONS and handling of irradiated fuel in the secondary containment.
| |
| : d. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.8.2.2 At least the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.
| |
| (*) Contrary to the prov1s1ons of Specification 3.0.2, if electrolyte level was below the top of the plates, the verification that there is no evidence of leakage is required to be completed regardless of when electrolyte level is restored.
| |
| LIMERICK - UNIT 1 3/4 8-14a Amendment No. +/--e-4-, 227
| |
| | |
| ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following power distribution system divisions shall be energized:
| |
| : a. A.C. power distribution:
| |
| : 1. Unit I Division 1, Consisting of:
| |
| a) 4160-VAC Bus: D1I (lOA115) b) 480-VAC Load Center: D114 (1OB201) c) 480-VAC Motor Control Centers: D114-R-C1 (1OB219)
| |
| D114-R-C (lOB213)
| |
| D114-R-G (10B211)
| |
| D114-R-GI (10B215)
| |
| D114-D-G (1OB515) d) 120-VAC Distribution Panels: IOY101 lOY206
| |
| : 2. Unit I Division 2, Consisting of:
| |
| a) 4160-VAC Bus: D12 (10A116) b) 480-VAC Load Center: D124 (10B202) c) 480-VAC Motor Control Centers: D124-R-C1 (1OB220)
| |
| D124-R-C (1OB214)
| |
| D124-R-G (10B212)
| |
| D124-R-GI (1OB216)
| |
| D124-D-G (10B516) d) 120-VAC Distribution Panels: 10Y102 10Y207
| |
| : 3. Unit 1 Division 3, Consisting of:
| |
| a) 4160-VAC Bus: D13 (10A117) b) 480-VAC Load Center: D134 (10B203) c) 480-VAC Motor Control Centers: D134-R-H1 (10B221)
| |
| D134-R-H (1OB217)
| |
| D134-R-E (1OB223)
| |
| D134-C-B (00B131)
| |
| D134-D-G (1OB517) d) 120-VAC Distribution Panels: I0Y103 10Y163
| |
| : 4. Unit I Division 4, Consisting of:
| |
| a) 4160-VAC Bus: D14 (IOA118) b) 480-VAC Load Center: D144 (1OB204)
| |
| Amendment No. 24,131 LIMERICK - UNIT I 3/4 8-15 OCT 2 3 1998
| |
| | |
| ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c) 480-VAC Motor Control Centers: 0144-R-G (10B222)
| |
| D144-R-H (10B218) 0144-R-E (10B224)
| |
| D144-C-B (OOB132) 0144-0-G (10B518) d) 120-VAC Distribution Panels: l0Y104 10Y164
| |
| : 5. Unit 2 and Common Division 1, Consisting of:
| |
| a) 4160-VAC bus: D21 (20A115) b) 480-VAC load center: D214 (20B201) c) 480-VAC motor control centers: D114-S-L (OOB519)
| |
| D214-R-C (20B213)
| |
| D214-0-G (20B515) d) 120-VAC distribution panels: 01Y501 20Y101 20Y206
| |
| : 6. Unit 2 and Common Division 2, Consisting of:
| |
| a) 4160-VAC bus: D22 (20A116).
| |
| b) 480-VAC load center: D224 (20B202) c) 480-VAC motor control centers: D124-5-L (OOB520)
| |
| D224-D-G (20B516) d) 120-VAC distribution panels: 02Y501 20Y102 20Y207
| |
| : 7. Unit 2 and Common Division 3, Consisting of:
| |
| a) 4160-VAC bus: 023 (20A117) b) 480-VAC load center: D234 (20B203) c) 480-VAC motor control centers: 0234-S-L (OOB521)
| |
| D234-D-G (20B517) d) 120- VAC distribution panels: 03Y501 20Y103 20Y163 8V.t.{--Unit 2 and Common Division 4, Consisting of:
| |
| a) 4160-VAC bus: D24 (20A118) b) 480-VAC load center: D224 (20B204) c) 480-VAC motor control centers: 0244-S-L (OOB522)
| |
| D244-D-G (20B518) d) 120-VAC distribution panels: 04Y501 20Y104 20Y164 LIMERICK - UNIT 1 3/4 8-16 Amendment No. 24 JUN 15 1989
| |
| | |
| ELECTRICAL POWER SYSTEMS LHJITING CONDITION FOR OPERATION (Continued)
| |
| : b. D.C. Power Distribution Panels
| |
| : 1. Unit 1 Division 1, Consisting of:
| |
| a) 250-V DC Fuse Box: 1FA (IAD105) b) 250-V DC Motor Control Center: 1DA (10D201) I c) 125-V DC Distribution Panels: 1PPA1 (IAD102)
| |
| IPPA2 (IAD501) 1PPA3 (IAD162)
| |
| : 2. Unit 1 Division 2, Consisting of:
| |
| a) 250-V DC Fuse Box: 1FB (IBDI05) b) 250-V DC Motor Control Centers: 1DB-i (1OD202) 1DB-2 (1OD203) c) 125-V DC Distribution Panels: 1PPBI (IBD102) 1PPB2 (IBD501)
| |
| IPPB3 (1BD162)
| |
| : 3. Unit 1 Division 3, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FC (lCD105) b) 125-V DC Distribution Panels: 1PPC1 (ICD102)
| |
| IPPC2 (1CD501)
| |
| IPPC3 (ICD162)
| |
| : 4. Unit 1 Division 4, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FD (IDD105) b) 125-V DC Distribution Panels: 1PPD1 (1DD102)
| |
| IPPD2 (IDD501) 1PPD3 (IDD162)
| |
| : 5. Unit 2 and Common Division 1, Consisting of:
| |
| a) 250-V DC Fuse Box: 2FA (2AD105) b) 125-V DC Distribution Panels: 2PPAI (2AD102) 2PPA2 (2AD501)
| |
| : 6. Unit 2 and Common Division 2, Consisting of:
| |
| a) 250-V DC Fuse Box: 2FB (2BD105) b) 125-V DC Distribution Panels: 2PPBI (2BD102) 2PPB2 (2BD501)
| |
| : 7. Unit 2 and Common Division 3, Consisting of:
| |
| a) 125-V DC Fuse Box: 2FC (2CD105) b) 125-V DC Distribution Panels: 2PPCI (2CD102) I 2PPC2 (2CD501)
| |
| : 8. Unit 2 and Common Division 4, Consisting of:
| |
| a) 125-V DC Fuse Box: 2FD (2DD105) I b) 125-V DC Distribution Panels: 2PPD1 (2DD102) 2PPD2 (2DD501)
| |
| K.>
| |
| LIMERICK - UNIT 1 3/4 8-16a Amendment No. iZ, 139 M1AR 1 Q 2cm
| |
| | |
| ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one of the above required Unit 1 A.C. distribution system divisions not energized, reenergize the division within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. With one of the above required Unit 1 D.C. distribution system divisions not energized, reenergize the division within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : c. With any of the above required Unit 2 and common AC and/or DC distribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for that system.
| |
| SURVEILLANCE REOUIREMENTS 4.8.3.1 Each of the above required power distribution system divisions shall be determined energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and voltage on the busses/MCCs/panels.
| |
| LIMERICK - UNIT 1 3/4 8-17 Amendment No. 2-4,186
| |
| | |
| ELECTRICAL POWER SYSTEMS DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, 2 of the 4 divisions of the power distribution system shall be energized with:
| |
| : a. A.C. power distribution:
| |
| : 1. Unit 1 Division 1, Consisting of:
| |
| a) 4160-VAC Bus: 011 (lOAllS) b) 480-VAC Load Center: 0114 (10B201) c) 480-VAC Motor Control Centers: D114-R-C1 (10B219)
| |
| D114-R-C (10B213) 0114-R-G (10B211) 0114-R-G1 (108215)
| |
| D114-D-G (10B515) d) 120-VAC Distribution Panels: lOY101 lOY206
| |
| : 2. Unit 1 Division 2, Consisting of:
| |
| a) 4160-VAC Bus: D12 (lOA116) b) 480-VAC Load Center: 0124 (10B202) c) 480-VAC Motor Control Centers: D124-R-C1 (10B220)
| |
| D124-R-C (108214)
| |
| D124-R-G (1OB212) 0124-R-G1 (10B216)
| |
| D124-D-G (10B516) d) 120-VAC Distribution Panels: lOY102 lOY207
| |
| : 3. Unit 1 Division 3, Consisting of:
| |
| a) 4160-VAC Bus: D13 (10A117) b) 480-VAC Load Center: 0134 (10B203) c) 480-VAC Motor Control Centers: 0134-R-H1 (10B221)
| |
| D134-R-H (10B217)
| |
| D134-R-E. (10B223)
| |
| I 0134-C-B (008131) 0134-0-G (10B517) d) 120-VAC Distribution Panels: lOY103 lOY163
| |
| : 4. Unit 1 Division 4, Consisting of:
| |
| a) b)
| |
| 4160-VAC Bus:
| |
| 480-VAC Load Center:
| |
| D14 (10A118) 0144 (108204)
| |
| I LIMERICK - UNIT 1 3/4 8-18 Amendment No. 24 JUN 1 5 1989 P 1&b 0kht (- (S Scawgio
| |
| | |
| ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c) 480-VAC Motor Control Centers: D144-R-G (10B222)
| |
| D144-R-H (10B218)
| |
| D144-R-E (10B224)
| |
| D144-C-B (OOB132)
| |
| D144-D-G (10B518) I d) 120-VAC Distribution Panels: 10Y104 10Y164 I
| |
| : 5. Unit 2 and Common Division 1, Consisting of:
| |
| a) 4160-VAC Bus: D21 (20A115) b) 480-VAC Load Center: 0214 (208201) c) 480-VAC Motor Control Centers: D114-S-L (OOB519)
| |
| D214-R-C (20B213)
| |
| D214-D-G (20B515) d) 120-VAC Distribution Panels: 01Y501 20Y101 20Y206
| |
| : 6. Unit 2 and Common Division 2, Consisting of:
| |
| : a. 4160-VAC Bus: D22 (20A116).
| |
| : b. 480-VAC Load Center: D224 (20B202)
| |
| : c. 480-VAC Motor Control Centers: D124-5-L (OOB520)
| |
| D224-D-G (20B516)
| |
| : d. 120-VAC Distribution Panels: 02Y501 20Y102 20Y207
| |
| : 7. Unit 2 and Common Divison 3, Consisting of:
| |
| : a. 4160-VAC Bus: D23 (20A117)
| |
| : b. 480-VAC Load Center: D234 (20B203)
| |
| : c. 480-VAC Motor Control Centers: D234-S-L (OOB521)
| |
| D234-D-G (20B517)
| |
| : d. 120-VAC Distribution Panels: 03Y501 20Y103 20Y163
| |
| : 8. Unit 2 and Common Divison 4, Consisting of:
| |
| : a. 4160-VAC Bus: D24 (20A118)
| |
| : b. 480-VAC Load Center: D224 (20B204)
| |
| : c. 480-VAC Motor Control Centers: D244-S-L (OOB522)
| |
| D244-D-G (20B518)
| |
| : d. 120-VAC Distribution Panels: 04Y501 20Y104 20Y164 LIMERICK - UNIT 1 3/4 8-18a Amendment No. 24 I
| |
| JUN 1 5 1989 d ~lvi& 6ft O-L it kL 7
| |
| | |
| ELECTRICAL POWER SYSTEMS ITMTTIN rONDITTTON FOR OPFRATION (CnntintiedI
| |
| : b. D.C. Power Distribution:
| |
| : 1. Unit I Division 1, Consisting of:
| |
| a) 250-V DC Fuse Box: 1FA (IAD105) b) 250-V DC Motor Control Center: 1DA (IOD201)
| |
| C) 125-V DC Distribution Panels: 1PPAI (IAD102) 1PPA2 (1AD5OI) 1PPA3 (IAD162)
| |
| : 2. Unit 1 Division 2, Consisting of:
| |
| a) 250-V DC Fuse Box: 1FB (lBDI05) b) 250-V DC Motor Control Centers: 1DB-i (10D202) 1DB-2 (1OD203) c) 125-V DC Distribution Panels: IPPBI (lBD102) 1PPB2 (1BD501) 1PPB3 (IBDI 62)
| |
| : 3. Unit 1 Division 3, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FC (lCD105) b) 125-V DC Distribution Panels: IPPCI (lCD102)
| |
| IPPC2 (ICD5O1)
| |
| "1PPC3 (lCD162)
| |
| : 4. Unit I Division 4, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FD (lDD105) b) 125-V DC Distribution Panels: IPPDI (lDD102) 1PPD2 (1DD501) 1PPD3 (lDD162)
| |
| : 5. Unit 2 and Common Division 1, Consisting of:
| |
| a) 250-V DC Fuse Box: 2FA (2AD105) b) 125-V DC Distribution Panels: 2PPA1 (2AD102) 2PPA2 (2AD501)
| |
| : 6. Unit 2 and Common Division 2, Consisting of:
| |
| a) 250-V DC Fuse Box: 2FB (2BD105) b) 125-V DC Distribution Panels: 2PPB1 (2BD102) 2PPB2 (2BD501)
| |
| : 7. Unit 2 and Common Division 3, Consisting of:
| |
| a) 125-V DC Fuse Box: 2FC (2CD105) I b) 125-V DC Distribution Panels: 2PPC1 (2CD102) 2PPC2 (2CD501)
| |
| : 8. Unit 2 and Common Division 4, Consisting of:
| |
| a) 125-V DC Fuse Box: 2FD (2DD105) I b) 125-V DC Distribution Panels: 2PPD1 (2DD102) 2PPD2 (2DD501)
| |
| LIMERICK - UNIT 1 3/4 8-19 Amendment No. i4, 139 MAR 1 4 2000
| |
| | |
| ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and*
| |
| ACTION:
| |
| : a. With less than two divisions of the above required Unit 1 A.C. dis-tribution systems energized, suspend CORE ALTERATIONS and handling of irradiated fuel in the secondary containment.
| |
| : b. With less than two divisions of the above required Unit 1 D.C. dis-tribution systems energized, suspend CORE ALTERATIONS and handling of irradiated fuel in the secondary containment.
| |
| : c. With any of the above required Unit 2 and common AC and/or DC distribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for that system.
| |
| : d. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.8.3.2 At least the above required power distribution system divisions shall be determined energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and voltage on the busses/MCCs/panels.
| |
| *When handling irradiated fuel in the secondary containment.
| |
| LIMERICK - UNIT 1 3/4 8-20 Amendment No . .;?4- l&e-,
| |
| 227
| |
| | |
| .Section 3/4.8.4.1 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS BEEN RELOCATED TO THE TRM. TECHNICAL SPECIFICATIONS PAGES 3/4 8-21 THROUGH 3/4 8-26 OF THE SECTION HAVE BEEN INTENTIONALLY.OMITTED.
| |
| LIMERICK - UNIT 1 3/4 8-21 Amendment No. 3-2,4Q4,-7,-94,192
| |
| | |
| PAG' INTENTIONALLY LEFT BLANK LIMER CK - UNIT 1 3/4 8 27 Amendment No. ++/-, +&, 209
| |
| | |
| ELECTRICAL POWER SYSTEMS REACTOR PROTECTION SYSTEM ELECTRICAL POWER MONITORING LIMITING CONDITION FOR OPERATION 3.8.4.3 Two reactor protection system (RPS) electric power monitoring channels for each inservice RPS Inverter or alternate power supply shall be OPERABLE.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With one RPS electric power monitoring channel for an inservice RPS Inverter or alternate power supply inoperable, restore the inoperable power monitoring channel to OPERABLE status within 72 hours or remove the associated RPS Inverter or alternate power supply from service.
| |
| : b. With both RPS electric power monitoring channels for an inservice RPS Inverter or alternate power supply inoperable, restore at least one electric power monitoring channel to OPERABLE status within 24 hours or remove the associated RPS Inverter or alternate power supply from service.
| |
| SURVEILLANCE REOUIREMENTS 4.8.4.3 The above specified RPS electric power monitoring channels shall be determined OPERABLE:
| |
| : a. By performance of a CHANNEL FUNCTIONAL TEST each time the plant is in COLD SHUTDOWN for a period of more than 24 hours, unless performed in the previous 6 months.
| |
| : b. In accordance with the Surveillance Frequency Control Program by demonstrating the OPERABILITY of overvoltage, undervoliage, and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic, and output circuit breakers and verifying the following Allowable Values.
| |
| : 1. Overvoltage S 127.6 VAC,
| |
| : 2. Undervoltage Ž 110.7 VAC,
| |
| : 3. Underfrequency Z 57.05 Hz.
| |
| LIMERICK - UNIT 1 3/4 8-28 Amendment No. ;4, ;-9, 4-34, 186
| |
| | |
| 3.4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH IiMLTINM CDNDIThQN EORItAlNO R
| |
| 3.9.1 The reactor mode switch shall be OPERABLE and locked in the Shutdown or Refuel position. When the reactor mode switch is locked in the Refuel position:
| |
| : a. The Refuel position one-rod-out interlock shall be OPERABLE.
| |
| : b. The following Refuel position interlocks shall be OPERABLE:
| |
| : 1. All rods in.
| |
| : 2. Refuel Platform (over-core) position.
| |
| : 3. Refuel Platform hoists fuel-loaded.
| |
| : 4. Service Platform hoist fuel-loaded (with Service Platform installed).
| |
| APPLICABILITY: OPERATIONAL CONDITION 5* **, OPERATIONAL CONDITIONS 3 AND 4 when the reactor mode switch is in the Refuel position.
| |
| ACTION:
| |
| : a. With the reactor mode switch not locked in the Shutdown or Refuel position as specified, suspend CORE ALTERATIONS and lock the reactor mode switch in the Shutdown or Refuel position.
| |
| : b. With the one-rod-out interlock inoperable, verify all control rods are fully inserted and disable withdraw capabilities of all control rods
| |
| * or lock the reactor mode switch in the Shutdown position.
| |
| : c. With any of the above required Refuel Platform Refuel position interlocks inoperable, take one of the ACTIONS listed below, or suspend CORE ALTERATIONS.
| |
| : 1. Verify control rods are fully inserted and disable withdraw capabilities of all control rods***, or
| |
| : 2. Verify Refuel Platform is not over-core (limit switches not reached) and disable Refuel Platform travel over-core, or
| |
| : 3. Verify that no Refuel Platform hoist is loaded and disable all Refuel Platform hoists from picking up (grappling) a load.
| |
| : d. With the Service Platform installed over the vessel and any of the above required Service Platform Refuel position interlocks inoperable, take one of the ACTIONS listed below, or suspend CORE ALTERATIONS.
| |
| : 1. Verify all control rods are fully inserted and disable withdraw capabilities of all control rods***, or
| |
| : 2. Verify Service Platform hoist is not loaded and disable Service Platform hoist from picking up (grappling) a load.
| |
| * See Special Test Exceptions 3.10.1 and 3.10.3.
| |
| ** The reactor shall be maintained in OPERATIONAL CONDITION 5 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
| |
| ** Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| LIMERICK - UNIT 1 3/4 9-1 Amendment No. -it?7 149 APR 0 5 20-3
| |
| | |
| REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.1.1 The reactor mode switch shall be verified to be locked in the Shutdown or Refuel position as specified, in accordance with the Surveillance Frequency Control Program.
| |
| 4.9.1.2 Each of the above required reactor mode switch Refuel position interlocks* shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program during control rod withdrawal or CORE ALTERATIONS, as applicable.
| |
| 4.9.1.3 Each of the above required reactor mode switch Refuel position interlocks* that is affected shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST prior to resuming control rod withdrawal or CORE ALTERATIONS, as applicable, following repair, maintenance or replacement of any component that could affect the Refuel position interlock.
| |
| *The reactor mode switch may be placed in the Run or Startup/Hot Standby position to test the switch interlock functions provided that all control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.
| |
| LIMERICK - UNIT 1 3/4 9-2 Amendment No. 4--, 186
| |
| | |
| REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least two source range monitor (SRM) channels* shall be OPERABLE and inserted to the normal operating level with:
| |
| : a. Continuous visual indication in the control room,
| |
| : b. At least one with audible alarm in the control room,
| |
| : c. One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant, and
| |
| : d. Unless adequate shutdown margin has been demonstrated, the shorting links shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn.**
| |
| APPLICABILITY: OPERATIONAL CONDITION 5.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and insert all insertable control rods.
| |
| SURVEILLANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:
| |
| : a. In accordance with the Surveillance Frequency Control Program: I
| |
| : 1. Performance of a CHANNEL CHECK,
| |
| : 2. Verifying the detectors are inserted to the normal operating level, and
| |
| : 3. During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and another is located in an adjacent quadrant.
| |
| *These channels are not required when sixteen or fewer fuel assemblies, ad-jacent to the SRMs, are in the core. The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is per-missible as long as these special detectors are connected to the normal SRM circuits.
| |
| **Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| LIMERICK - UNIT*1 3/4 9-3 Amendment No. 4,186
| |
| | |
| REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. Performance of a CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.
| |
| : c. Verifying that the channel count rate is at least 3.0 cps:*
| |
| : 1. Prior to control rod withdrawal,
| |
| : 2. Prior to and in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS, and,
| |
| : 3. In accordance with the Surveillance Frequency Control Program.
| |
| : d. Verifying, within 8 hours prior to and in accordance with the Surveillance Frequency Control Program, that the RPS circuitry "shorting links" have been removed during:
| |
| : 1. The time any control rod is withdrawn**, unless adequate shutdown margin has been demonstrated, or
| |
| : 2. Shutdown margin demonstrations.
| |
| *May be reduced, provided the source range monitor has an observed count rate and signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.
| |
| These channels are not required when sixteen or fewer fuel assemblies, adja-cent to the SRMs, are in the core.
| |
| **Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| LIMERICK - UNIT 1 3/4 9-4 Amendment No. 4, 34, g, 4&g-F, 186
| |
| | |
| REFUELING OPERATIONS 3/4.9.3 CONTROL ROD POSITION LIMITING CONDITION FOR OPERAT-ION 3.9.3 All control rods shall be inserted.*
| |
| APPLICABILITY: OPERATIONAL CONDITION 5, during CORE ALTERATIONS.**
| |
| ACTION:
| |
| With all control rods not inserted, suspend all other CORE ALTERATIONS, except that one control rod may be withdrawn under control of the reactor mode switch Refuel position one-rod-out interlock.
| |
| SURVEILLANCE REOUIREMENTS 4.9.3 All control rods shall be verified to be inserted, except as above specified, in accordance with the Surveillance Frequency Control Program. I
| |
| *Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.
| |
| **See Special Test Exception 3.10.3.
| |
| LIMERICK - UNIT 1 3/4 9-5 Amendment No. 404,186
| |
| | |
| REFUELING OPERATIONS 3/4.9.4 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.4 The reactor shall be subcritical for at least 24 hours.
| |
| APPLICABILITY: OPERATIONAL CONDITION 5, during movement of irradiated fuel in the reactor pressure vessel.
| |
| ACTION:
| |
| With the reactor subcritical for less than 24 hours, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel.
| |
| SURVEILLANCE REQUIREMENTS 4.9.4 The reactor shall be determined to have been subcritical for at least 24 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.
| |
| LIMERICK - UNIT 1 3/4 9-6
| |
| | |
| REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communication shall be maintained between the control room and refueling floor personnel.
| |
| APPLICABILITY: OPERATIONAL CONDITION 5, during CORE ALTERATIONS.*
| |
| ACTION:
| |
| When direct communication between the control room and refueling floor personnel cannot be maintained, immediately suspend CORE ALTERATIONS.*
| |
| SURVEILLANCE REOUIREMENTS 4.9.5 Direct communication between the control room and refueling floor personnel shall be demonstrated in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS.* I
| |
| *Except movement of control rods with their normal drive system.
| |
| LIMERICK - UNIT 1 3/4 9-7 Amendment No. 14-9, 186
| |
| | |
| REFUELING OPERATIONS 3/4.9.6 REFUELING PLATFORM LIMITING CONDITION FOR OPERATION 3.9.6 The refueling platform shall be OPERABLE and used for handling fuel assemblies or control rods within the reactor pressure vessel.
| |
| APPLICABILITY: During handling of fuel assemblies or control rods within the reactor pressure vessel.
| |
| ACTION:
| |
| With the requirements for refueling platform OPERABILITY not satisfied, suspend use of any inoperable refueling platform equipment from operations involving the handling of control rods and fuel assemblies within the reactor pressure vessel after placing the load in a safe condition.
| |
| SURVEILLANCE REQUIREMENTS 4.9.6.1 The refueling platform main hoist used for handling of fuel assemblies I within the reactor pressure vessel shall be demonstrated OPERABLE within 7 days prior to the start of such operations by:
| |
| : a. Demonstrating operation of the overload cutoff on the main hoist when the load exceeds 1150 +/- 50 pounds.
| |
| : b. Demonstrating operation of the hoist loaded control rod block interlock on the main hoist when the load exceeds 485 +/- 50 pounds.
| |
| : c. Demonstrating operation of the redundant loaded interlock on the main hoist when the load exceeds 550 + 0, - 115 pounds.
| |
| : d. Demonstrating operation of the uptravel interlock.when uptravel brings the top of the active fuel to not less than 8 feet 0 inches below the I normal water level.
| |
| LIMERICK - UNIT 1 3/4 9-8 Amendment No.
| |
| AUG 1 6 1990 43 I
| |
| | |
| REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)-
| |
| 4.9.6.2 The refueling platform frame-mounted auxiliary hoist used for handling of control rods within the reactor pressure vessel shall be demon-strated OPERABLE within 7 days prior to.the use of such equipment by:
| |
| : a. Demonstrating operation of the overload cutoff on the frame mounted hoist when the load exceeds 500 +/- 50 pounds.
| |
| : b. Demonstrating operation of the uptravel mechanical stop on the frame I mounted hoist when uptravel brings the top of a control rod to not less than 6 feet 6 inches below the normal fuel storage pool water level.
| |
| 4.9.6.3 The refueling platform monorail mounted auxiliary,-hoist used for handling of control rods.within the-reactor pressure vesselishallibe demonstra-ted OPERABLE within 7 days prior to the use of such equipment by:
| |
| : a. Demonstrating operation of the overload cutoff on the monorail hoist when the load exceeds 500 +/- 50 pounds.
| |
| : b. Demonstrating operation of the uptravel mechanical stop on the monorail hoist when uptravel brings the top of a control rod to not less than 6 feet 6 inches below the normal fuel storage pool water level.
| |
| I I
| |
| LIMERICK - UNIT 2 3/4 9-9 Amendment No. 43 I . . .
| |
| AUG i 6 1990'
| |
| | |
| REFUELING OPERATIONS k
| |
| 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 1200 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage pool racks.
| |
| APPLICABILITY: With fuel assemblies in the spent fuel storage pool racks.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REOUIREMENTS 4.9.7 Crane interlocks which prevent crane travel over fuel assemblies in the spent fuel storage pool racks shall be demonstrated OPERABLE within 7 days prior to and in accordance with the Surveillance Frequency Control Program during I crane operation.
| |
| LIMERICK - UNIT I 3/4 9-10 Amendment No.186
| |
| | |
| REFUELING OPERATIONS 3/4.9.8 WATER LEVEL - REACTOR VESSEL
| |
| .LIMITINGC-ONDIT-ION FOR OPERATION 3.9.8 At least 22 feet of water shall be maintained over the top of the reactor pressure vessel flange.
| |
| APPLICABILITY: During handling of fuel assemblies or control rods within the reactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, suspend all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel after placing all fuel assemblies and control rods in a safe condition.
| |
| SURVEILLANCE REOUIREMENTS 4.9.8 The reactor vessel water level shall be determined to be at least its minimum required depth in accordance with the Surveillance Frequency Control Program during handling of fuel assemblies or control rods within the reactor pressure vessel.
| |
| LIMERICK - UNIT I 3/4 9-11 Amendment No. 409, 186
| |
| | |
| REFUELING OPERATIONS 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.9 At least 22 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.
| |
| APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel storage pool.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REOUIREMENTS 4.9.9 The water level in the spent fuel storage pool shall be determined to be at least at its minimum required depth in accordance with the Surveillance Frequency Control Program. I LIMERICK - UNIT 1 3/4 9-12 Amendment No. 186
| |
| | |
| REFUELING OPERATIONS 3/4.9.10 CONTROL ROD REMOVAL SINGLE CONTROL ROD'REMOVAL LIMITING CONDITION'FOR OPERATION 3.9.10.1 One control rod and/or the associated control rod drive mechanism may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is-fully inserted in the core.
| |
| : a. The reactor mode switch is OPERABLE and locked in the Shutdown position or in the Refuel position per Table 1.2 and-Specification 3.9.1.
| |
| : b. The source range monitors (SRM) are OPERABLE per Specification 3.9.2.
| |
| : c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, except that the control rod selected to be removed;
| |
| : 1. May be assumed to be the highest worth control rod required to be assumed to be fully withdrawn by the SHUTDOWN MARGIN test, and
| |
| : 2. Need not be assumed to be immovable or untrippable.
| |
| : d. All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
| |
| : e. All other control rods are inserted.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, suspend removal of the control rod and/or associated control rod drive mechanism from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.
| |
| LIMERICK - UNIT 1 3/4 9-13
| |
| | |
| REFUELING OPERATIONS SURVEILLANCF REOUIREMENTS 4.9.10.1 Within 4 hours prior to the start of removal of a control rod and/or the associated control rod drive mechanism from the core and/or reactor pressure vessel and in accordance with the Surveillance Frequency Control Program thereafter until a control rod and associated control rod drive mechanism are reinstalled and the control rod is inserted in the core, verify that:
| |
| : a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position with the "one rod out" Refuel position interlock OPERABLE per Specification 3.9.1.
| |
| : b. The SRM channels are OPERABLE per Specification 3.9.2.
| |
| : c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied per Specification 3.9.10.1c.
| |
| : d. All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
| |
| : e. All other control rods are inserted.
| |
| LIMERICK - UNIT 1 3/4 9-14 Amendment NoI86
| |
| | |
| REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.2 Any number of control rods and/or control rod drive'mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core.
| |
| : a. The reactor mode switch is OPERABLE and locked in'the Shutdown position or in the Refuel position per Specification 3.9.1, except that the Refuel position "one-rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after'the fuel'assemblies have been removed as specified below.
| |
| : b. The source range monitors'(SRM) are OPERABLE per Specification 3.9.2.
| |
| : c. The SHUTDOWN'MARGIN requirements'of Specification-3.1.1 are satisfied.
| |
| : d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.
| |
| : e. The four fuel assemblies'surrounding each control rod or control rod
| |
| -drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
| |
| APPLICABILITY: OPERATIONAL CONDITION 5.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, suspend removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.
| |
| LIMERICK - UNIT 1 3/4 9-15
| |
| | |
| REFUELING OPERATIONS SURVEILLANCE REOUIREMENTS 4.9.10.2.1 Within 4 hours prior to the start of removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and in accordance with the Surveillance Frequency Control Program thereafter until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core, verify that:
| |
| : a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position per Specification 3.9.1.
| |
| : b. The SRM channels are OPERABLE per Specification 3.9.2.
| |
| : c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
| |
| : d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.
| |
| : e. The four fuel assemblies surrounding each control rod and/or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
| |
| 4.9.10.2.2 Following replacement of all control rods and/or control rod drive mechanisms removed in accordance with this specification, perform a functional test of the "one-rod-out" Refuel position interlock, if this function had been bypassed.
| |
| LIMERICK - UNIT 1 3/4 9-16 Amendment No. 186
| |
| | |
| REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.1 One (1) RHR shutdown cooling subsystem shall be OPERABLE and in operation.
| |
| * APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet above the top of the reactor pressure vessel flange.
| |
| ACTION:
| |
| : a. With the required RHR shutdown cooling subsystem inoperable:
| |
| : 1. Within one (1) hour, and once per 24 hours thereafter, verify an alternate method of decay heat removal is available.
| |
| : b. With the required action and associated completion time of Action "a" above not met.
| |
| : 1. Immediately suspend loading of irradiated fuel assemblies into the reactor pressure vessel; and
| |
| : 2. Immediately initiate action to restore REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY to OPERABLE status; and
| |
| : 3. Immediately initiate action to restore one (1) Standby Gas Treatment subsystem to OPERABLE status; and
| |
| : 4. Immediately initiate action to restore isolation capability in each required Refueling Floor secondary containment penetration flow path not isolated.
| |
| : c. With no RHR shutdown cooling subsystem in operation:
| |
| : 1. Within one (1) hour from discovery of no reactor coolant circulation, and once per 12 hours thereafter, verify reactor coolant circulation by an alternate method; and
| |
| : 2. Once per hour monitor reactor coolant temperature.
| |
| SURVEILLANCE REQUIREMENTS 4.9.11.1.l At least one (1) RHR shutdown cooling subsystem, or an alternate method, shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.
| |
| 4.9.11.1.2 Verify required RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.
| |
| * The required RHR shutdown cooling subsystem may be removed from operation for up to two (2) hours per eight (8) hour period.
| |
| LIMERICK - UNIT 1 3/4 9-17 Amendment No. f.,++/--g,~,216
| |
| | |
| REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.2 Two (2) RHR shutdown cooling subsystems shall be OPERABLE, and one (1)
| |
| RHR shutdown cooling subsystem shall be in operation.
| |
| * APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is less than 22 feet above the top of the reactor pressure vessel flange.
| |
| ACTION:
| |
| : a. With one ( 1) or two ( 2) required RHR shutdown cooling subsystems inoperable:
| |
| : 1. Within one (1) hour, and once per 24 hours thereafter, verify an alternate method of decay heat removal is available for each inoperable required RHR shutdown cooling subsystem.
| |
| : b. With the required action and associated completion time of Action "a" above not met:
| |
| : 1. Immediately initiate action to restore REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY to OPERABLE status; and
| |
| : 2. Immediately initiate action to restore one (1) Standby Gas Treatment subsystem to OPERABLE status; and
| |
| : 3. Immediately initiate action to restore isolation capability in each required Refueling Floor secondary containment penetration flow path not isolated.
| |
| : c. With no RHR shutdown cooling subsystem in operation:
| |
| : 1. Within one (1) hour from discovery of no reactor coolant circulation, and once per 12 hours thereafter, verify reactor coolant circulation by an alternate method; and
| |
| : 2. Once per hour monitor reactor coolant temperature.
| |
| SURVEILLANCE REQUIREMENTS 4.9.11.2.1 At least one (1) RHR shutdown cooling subsystem, or an alternate method, shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.
| |
| 4.9.11.2.2 Verify RHR shutdown cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water in accordance with the Surveillance Frequency Control Program.
| |
| * The required operating shutdown cooling subsystem may be removed from operation for up to two (2) hours per eight (8) hour period.
| |
| LIMERICK - UNIT 1 314 9-18 Amendment No. -9+,-+/--+/--9,~,216
| |
| | |
| */.14-1 -PFr.TAI TFST HUCPTMONS T
| |
| 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.10.1 The provisions of Specifications 3.6.1.1, 3.6.1.3, and 3.9.1 and Table 1.2 may be suspended to permit the reactor pressure vessel closure head and the drywell head to be removed and the primary containment air lock doors to be open when the reactor mode switch is in the Startup position during low power PHYSICS TESTS with THERMAL POWER less than 1% of RATED THERMAL POWER and reactor coolant temperature less than 200 0 F.
| |
| APPLICABILITY: OPERATIONAL CONDITION 2, during low power PHYSICS TESTS.
| |
| ACTION:
| |
| With THERMAL POWER greater than or equal to 1% of RATED THERMAL POWER or with the reactor coolant temperature greater than or equal to 200 0 F, immediately place the reactor mode switch in the Shutdown position.
| |
| SURVEILLANCE REOUIREMENTS 4.10.1 The THERMAL POWER and reactor coolant temperature shall be verified to be within the limits in accordance with the Surveillance Frequency Control Program during low power PHYSICS TESTS. I LIMERICK - UNIT 1 3/4 10-1 Amendment No. 186
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod worth minimizer (RWM) per Specification 3.1.4.1 may be suspended for the following tests provided that control rod movement prescribed for this testing is verified I by a second licensed operator or other technically qualified member of the unit technical staff present at the reactor console:
| |
| : a. Shutdown margin demonstration, Specification 4.1.1.
| |
| : b. Control rod scram, Specification 4.1.3.2.
| |
| : c. Control rod friction measurements.
| |
| : d. Startup Test Program I APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER. I ACTION:
| |
| With the requirements of the above specifications not satisfied, verify that the RWM is OPERABLE per Specification 3.1.4.1.
| |
| SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed by the RWM are bypassed, verify; I
| |
| : a. That movement of control rods is blocked or limited to the approved control rod withdrawal sequence during scram and friction tests. I
| |
| : b. That movement of control rods during shutdown margin demonstrations is limited to the prescribed sequence per Specification 3.10.3.
| |
| : c. Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit technical staff.
| |
| LIMERICK - UNIT 1 3/4 10-2 Amendment No. 17 I MAR 2 2 1989
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS LIMITING CONDITION FOR OPERATION 3.10.3 The provisions of Specification 3.9.1, Specification 3.9.3, and Table 1.2 may be suspended to permit the reactor mode switch to be in the Startup position and to allow more than one control rod to be withdrawn for shutdown margin demonstration, provided that at least the following requirements are satisfied.
| |
| : a. The source range monitors are OPERABLE with the RPS circuitry "shorting links" removed per Specification 3.9.2.
| |
| : b. The rod worth minimizer is OPERABLE per Specification 3.1.4.1 and is programmed for the shutdown margin demonstration, or conformance with the shutdown margin demonstration procedure is verified by a second licensed operator or other technically qualified member of the unit technical staff.
| |
| : c. The "continuous rod withdrawal" control shall not be used during out-of-sequence movement of the control rods.
| |
| : d. No other CORE ALTERATIONS are in progress.
| |
| APPLICABILITY: OPERATIONAL CONDITION 5, during shutdown margin demonstrations.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown or Refuel position.
| |
| SURVEILLANCE REOUTREMENTS 4.10.3 Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during the performance of a shutdown margin demonstration, verify that;
| |
| : a. The source range monitors are OPERABLE per Specification 3.9.2,
| |
| : b. The rod worth minimizer is OPERABLE with the required program per Specification 3.1.4.1 or a second licensed operator or other techni-cally qualified member of the unit technical staff is present and verifies compliance with the shutdown margin demonstration procedures, and
| |
| : c. No other CORE ALTERATIONS are in progress.
| |
| LIMERICK - UNIT 1 3/4 10-3 Amendment No. 186
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.4 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.4.1.1 and 3.4.1.3 that recirculation loops be in operation may be suspended for up to 24 hours for the performance of:
| |
| : a. PHYSICS TESTS, provided that THERMAL POWER does not exceed 5% of RATED THERMAL POWER, or
| |
| : b. The Startup Test Program.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, during PHYSICS TESTS and the Startup Test Program.
| |
| ACTION:
| |
| : a. With the above specified time limit exceeded, insert all control rods.
| |
| : b. With the above specified THERMAL POWER limit exceeded during PHYSICS TESTS, immediately place the reactor mode switch in the Shutdown position.
| |
| SURVEILLANCE REOUIREMENTS 4.10.4.1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS and the Startup Test Program.
| |
| 4.10.4.2 THERMAL POWER shall be determined to be less than 5% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS.
| |
| LIMERICK -UNIT 1 3/4 10-4 Amendment No. 186
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.5 OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.10.5 The provisions of Specification 3.6.6.3 may be suspended during the performance of the Startup Test Program until either the required 100% of RATED THERMAL POWER trip tests have been completed or the reactor has operated for 120 Effective Full Power Days.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, be in at least STARTUP within 6 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.10.5 The Effective Full Power Days of operation shall be verified to be less than 120, by calculation, in accordance with the Surveillance Frequency Control Program during the Startup Test Program. I LIMERICK - UNIT 1 3/4 10-5 Amendment No. 186
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.6 TRAINING STARTUPS LIMITING CONDITION FOR OPERATION 3.10.6 The provisions of Specification 3.5.1 may be suspended to permit one RHR subsystem to be aligned in the shutdown cooling mode during training startups provided that the reactor vessel is not pressurized, THERMAL POWER is less than or equal to 1% of RATED THERMAL POWER and reactor coolant temperature is less than 200 0 F.
| |
| APPLICABILITY: OPERATIONAL CONDITION 2, during training startups.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position.
| |
| SURVEILLANCE REOUIREMENTS 4.10.6 The reactor vessel shall be verified to be unpressurized and the THERMAL POWER and reactor coolant temperature shall be verified to be within the limits in accordance with the Surveillance Frequency Control Program during I training startups.
| |
| LIMERICK - UNIT 1 3/4 10-6 Amendment No.186
| |
| | |
| SPECIAL TEST EXCEPTIONS
| |
| ; ,3/4.10.7 RESERVED-CURRENTLY NOT USED LIMERICK - UNIT 1 3/4 10-7 Amendment No. 133 JAN 1 2 1999
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING LIMITING CONDITIONS FOR OPERATION 3.10.8 When conducting inservice leak or hydrostatic testing, the average reactor coolant temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased to 212F, and operation considered not to be in OPERATIONAL CONDITION 3, to allow performance of an inservice leak or hydrostatic test provided the following OPERATIONAL CONDITION 3 Specifications are met:
| |
| : a. 3.3.2 ISOLATION ACTUATION INSTRUMENTATION, Functions 7.a, 7.c.1, 7.c.2 and 7.d of Table 3.3.2-1;
| |
| : b. 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY;
| |
| : c. 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY;
| |
| : d. 3.6.5.2.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES;
| |
| : e. 3.6.5.2.2 REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES; and
| |
| : f. 3.6.5.3 STANDBY GAS TREATMENT SYSTEM.
| |
| APPLICABILITY: OPERATIONAL CONDITION 4, with average reactor coolant temperature greater than 200°F and less than or equal to 212°F.
| |
| ACTION:
| |
| With the requirements of the above Specifications not satisfied:
| |
| X 1. Immediately enter the applicable (OPERATIONAL CONDITION 3) action for the affected Specification; or
| |
| : 2. Immediately suspend activities that could increase the average reactor coolant temperature or pressure and reduce the average reactor coolant temperature to 2007 or less within 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.10.8 Verify applicable OPERATIONAL CONDITION 3 surveillances for the Specifications listed in 3.10.8 are met.
| |
| I
| |
| ; , . - -:-.- .- - . .. . ... ~ ~ ` ~ % .': , - - - -
| |
| LIMERICK - UNIT 1 3/4 10-8 Amendment No. 133 I JAN 1 2 1299
| |
| | |
| Section 3/4 11.1.1 through 3/4 11.1.4 (Deleted)
| |
| THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 11-2 THROUGH 3/4 11-6 OF THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 11-1 Amendment No. , 48 ciQ ,/q/
| |
| I limit
| |
| | |
| RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in any outside temporary tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With the quantity of radioactive material in any of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit and describe the events leading to this condition in the next Annual Radioactive Effluent Release Report.
| |
| : b. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a repre-sentative sample of the tank's contents in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank.
| |
| LIMERICK - UNIT 1 3/4 11-7 Amendment No. 44, .-7, 186
| |
| | |
| I Mau-g Section 3/4 11.2.1 through Section 3/4 11.2.4 (Deleted)
| |
| THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE OCN. TECHNICAL SPECIFICATIONS PAGES 3/4 11-9 THROUGH 3/4 11-14 OF THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED.
| |
| ' i' I
| |
| LIMERICK - UNIT 1 3/4 11-8 Amendment No. ;48, 4 1 0t qua , 1qq I I
| |
| | |
| Section 3/4.11.2.5 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM.
| |
| LIMERICK - UNIT 1 3/4 11-15 Amendment No . .U, 22
| |
| | |
| RADIOACTIVE EFFLUENTS MAIN CONDENSER UMUHlli .C.ONUlllilll EORilPERAllQN 3.11.2.6 The rate of the sum of the activities of the noble gases Kr-85m, Kr-87, Kr-88, Xe-133, Xe-135, and Xe-138 measured at the recombiner after-condenser discharge shall be limited to less than or equal to 330 millicuries/second.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3*.
| |
| ACT ION:
| |
| With the rate of the sum of the activities of the specified noble gases at the recombiner after-condenser discharge exceeding 330 millicuries/second, restore the gross radioactivity rate to within its limit within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours.
| |
| S.Ufill IlWtiCL.RE.ilU lREME.fil.S.
| |
| 4.11.2.6.1 Relocated to the ODCM.
| |
| 4.11.2.6.2 The rate of the sum of the activities of the specified noble gases from the recombiner after-condenser discharge shall be determined to be within the limits of Specification 3.11.2.6 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the recombiner after condenser discharge:
| |
| : a. In accordance with the Surveillance Frequency Control Program.
| |
| : b. Within 4 hours following a noted increase of greater than 50%, after factoring out increases due to changes in THERMAL POWER level or air in-leakage, in the nominal steady-state fission gas release from the primary cool ant.
| |
| : c. The provisions of Specification 4.0.4 are not applicable.
| |
| *When the main condenser air ejector is in operation.
| |
| LIMERICK - UNIT 1 3/4 11-16 Amendment No. ~ . 228
| |
| | |
| Section 3/4 11-2.7 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL -.
| |
| SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM. ---;.I LIMERICK - UNIT 1 3/4 .11 Amendment No. )'48
| |
| .i:
| |
| ,.^,,,k, K' ,e;,";, I],
| |
| J~~9MuvL (&4, /I4/
| |
| \s.-VI ,
| |
| . * .I
| |
| | |
| Section 3/4 11.3 and Section 3/4 11.4 (Deleted) t..
| |
| ' /
| |
| THEINFORMATION FROMTHESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN -
| |
| RELOCATED TO THE PCP OR ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 11-19 THROUGH 3/4 11-20 OF THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 11-18 . Amendment No. .I:48 I
| |
| * --A~\\" ,1 X, . t. X: . A APtuLt tCPw~tLu . , jJqq
| |
| | |
| Section 3/4.12 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 12-2 THROUGH 3/4 12-14 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 12-1 Amendment No;48 0 cDAL4u a cz, 1q9q I 41-111-
| |
| | |
| SECTION 5.0 DESIGN FEATURES
| |
| | |
| 5.0 DESIGN FEATURES
| |
| . . I 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.3,1-1.
| |
| LOW POPULATION ZONE.
| |
| 5.1.2 The low po~pulation zone shall be as shown in Figure 5.1.2-1.-
| |
| MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS.AND LIQUID EFFLUENTS '
| |
| 5.1.3 Informationlregarding radioactive gaseous'and liquid effluents, which will allow identification of'structures' and release points as well as'definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible'to MEMBER OF THE PUBLIC, shall be as shown in Figures 5.1.3-la and 5.1.3-lb.
| |
| I I . . .
| |
| 5.1.4 (Deleted)''
| |
| 5_ rnNTATNMFNT -
| |
| CONFIGURATION '
| |
| 5.2.1 The primary containment is a steel lined reinforced concrete structure v<_,J consisting of a drywell and suppression chamber.:Thedrywell is a-steel-lined reinforced concrete-vessel in a shape'of a-truncated.cone on'top of a'water.filled suppression chamber and is separated by a diaphragm slab and connected -to the suppression chamber through a series of*.downcomer.vents. The drywell'has a maximum free.a'ir volume of 243,580'cubic feet at-a minimum suppression pool level of 22 feet.- 'The suppression chamber has a-maximum air region of'159,540 cubic feet and a minimum water region-of 122,120 cubic feet.
| |
| DESIGN TEMPERATURE AND PRESSURE '
| |
| 5.2.2 The primary containment is designed and shall be maintained for:
| |
| : a. Maximum Internal pressure 55 pslg.- :
| |
| : b. Maximum internal-temperature: drywellI34 0°F.
| |
| suppression pool220 F.-
| |
| : c. Maximum external to internal differential pressure5psid.
| |
| : d. Maximum floor differential pressure: 30 psid, downward.
| |
| 20 psid, upward.
| |
| LIMERICK - UNIT 1 Amendment No. 33,48 I 4"MLL qua cQ, i qq/
| |
| | |
| I FIGURE 5.1.1-1 EXCLUSION AREA LIMERICK - UNIT 1. 5-2
| |
| \\% ,; ,". nJ.9t} ! t~t-,f X .,ji
| |
| | |
| FIGURE 5.1.2-1 LOW POPULATION ZONE LIMERICK - UNIT 1 5-3
| |
| | |
| I
| |
| ' K i )-
| |
| FIGURE 5.1.3-la MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS LIMERICK - UNIT 1 5-4
| |
| | |
| li, IVATC coo~o~tt.-
| |
| IAoj ao.ce 1* m.eSY
| |
| ?Uw~tke IbCLSUU w66 POINTS iL A? 5 s
| |
| OASEO US RELEASE N, ?wo ..
| |
| e
| |
| ^C w~sn SO.
| |
| cowaUTH ueat TACK)
| |
| *.AT. C O,.O U 331.832
| |
| *
| |
| * 2.n...
| |
| e "^ aeess #0 *U45L546.
| |
| BOlSL 34 POIN*OF a n POINT RELEASE OF asI" at~ lotVA4150U ' OUTH TAK RELEASE CDOSLUNG wA'tS vaPoR 0.691 Two To051it 2\2*46 2.064.1141
| |
| \\ S LI VATfot 144 VIA
| |
| %,~~U
| |
| * 2.0023l 105654LIVATSOUt led
| |
| * INSIGNIFICANT MI50066 SUeLsWSJSI aAoWAST.
| |
| IUcLOswaS A l A"
| |
| -f Ads n ben T!
| |
| *- 4 UAZ FIGURE 5.1.3-lb MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS
| |
| '\\ .
| |
| LIMERICK - UNIT'll - i- v 55
| |
| | |
| I THE FIGURE ON THIS PAGE HAS BEEN RELOCATED TO THE ODCM.
| |
| LIMERICK - UNIT 1 5-6 Amendment No.48 JI' 4#ZMwL 9 g. v,.
| |
| | |
| DESIGN FEATURES SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of three distinct isolatable zones.
| |
| Zones I and II are the Unit 1 and Unit 2 reactor enclosures respectively.
| |
| Zone III is the common refueling area. Each zone has an independent normal ventilation system which is capable of providing secondary containment zone isolation as required.
| |
| Each reactor enclosure (Zone I or II) completely encloses and provides secondary containment for its corresponding primary containment and reactor auxiliary or service equipment, and has a minimum free volume of 1,800,000 cubic feet.
| |
| The common refueling area (Zone III) completely encloses and provides secondary containment for thte refueling servicing equipment and spent fuel storage facilities for Units 1 and 2, and has a minimum free volume of 2,200,000 cubic feet.
| |
| 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall consist of not more than 764 fuel assemblies and shall be limited to those fuel assemblies which have been analyzed with NRC approved codes and methods and have been shown to comply with all Safety Design Bases in the Final Safety Analysis Report (FSAR).
| |
| CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform-shaped control rod assemblies.
| |
| 5.4 REACTOR COOLANT SYSTEM I
| |
| DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
| |
| : a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, LIMERICK - UNIT 1 5-7 Amendment No. 19 APR 2 4 1989
| |
| | |
| ULilbN ttAIUKLb
| |
| - DESIGN ... PRESSURE AND TEMPERATURE (Continued)
| |
| : b. For a pressure of:
| |
| : 1. 1250 psig on the suction side of the recirculation pump.
| |
| : 2. 1500 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.
| |
| : 3. 1500 psig from the discharge shutoff valve to the Jet pumps.
| |
| : c. For a temperature of 5750F.
| |
| VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal steam dome saturation temperature of 552 0F. I 5.5 FUEL STORAGE CRITICALITY 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:
| |
| : a. A keff equivalent to less than or equal to 0.95 when flooded with unborated water, including all calculational uncertainties and biases as described in Section 9.1.2 of the FSAR.
| |
| : b. A nominal center-to-center distance between fuel assemblies placed in the storage racks of greater than or equal to 6.244 inches.
| |
| 5.5.1.2 The keff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.
| |
| DRAINAGE 5.5.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 346'0".
| |
| CAPACITY 5.5.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 4117 fuel assemblies.
| |
| 5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.6.1 The components identified in Table 5.6.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.6.1-1.
| |
| LIMERICK - UNIT 1 5-8 Amendment No. 7i,$i,106 1998
| |
| .1?i 9 ,4
| |
| - 7 V &" "MY)PUnVnW--' PUOIP--' -
| |
| --L-- or&
| |
| | |
| C:.,
| |
| I-
| |
| '-q m TABLE 5.6.1-1
| |
| '-4 n- COMPONENT CYCLIC OR TRANSIENT LIMITS a
| |
| 0-4 CYCLIC OR DESIGN CYCLE
| |
| -I4 COMPONENT TRANSIENT LIMIT OR TRANSIENT
| |
| '-a Reactor 120 heatup and cooldown cycles 700 F to 5600 F to 700 F 80 step change cycles Loss of feedwater heaters 180 reactor trip cycles 100% to 0% of RATED THERMAL POWER 130 hydrostatic pressure and Pressurized to > 930 and leak tests < 1250 psig C.
| |
| ED
| |
| | |
| SECTION 6.0 ADMINISTRATIVE CONTROLS
| |
| | |
| 6.0 ADMINISTRATIVE CONTROLS
| |
| ..- 6.1 RESPONSIBILITY
| |
| .1.1 The Plant Manager shall be responsible for overall unit operation and 4I.
| |
| all delegate in writing the succession to this responsibility during his absence.
| |
| 6.1.2 The Shift Manager, or during his absence from the control room, a designated individual shall be responsible for the control room command function.
| |
| A management directive to this-effect, signed by the Vice President, Limerick Generating Station shall be reissued to all station personnel on an annual basis.
| |
| 6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
| |
| : a. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organizational positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Limerick Quality Assurance Program.
| |
| : b. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
| |
| : c. The Vice President, Limerick Generating Station shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.
| |
| : d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
| |
| LIMERICK - UNIT 1 6-1 Amendment No. 10, 3, 96 JUL 18 l995
| |
| | |
| ADMINISTRATIVE_CONTROLS 6.2.2 UNIT STAFF
| |
| : a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1;
| |
| : b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2, or 3, at least one licensed Senior Operator shall be in the control room;
| |
| : c. A Health Physics Technician* shall be on site when fuel is in the reactor;
| |
| : d. ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
| |
| : e. (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.
| |
| : f. (Deleted)
| |
| * The Health Physics Technician position may be less than the mlnlmum requirements for a period of time not to exceed 2 hours, in order to accommodate unexpectsd absence, provided immediate action is taken to fill the required position.
| |
| LIMERICK - UNIT 1 6-2 Amendment No . .w, ~, 9-6, Ht4, H9,198
| |
| | |
| ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)
| |
| : g. The individual filling the position of Operations Manager as defined by ANSI/ANS-3.1-1978 or another Manager in Operations shall hold a Senior Reactor Operator License.
| |
| LIMERICK - UNIT 1 6-3 Amendment No. +/-G, ~, £, -BG, %,
| |
| ~, 198
| |
| | |
| This page is intentionally left blank.
| |
| LIMERICK - UNIT 1 6-4 Amendent No. AT,35 DEC 1 9 19
| |
| | |
| TABLE 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMMON CONTROL ROOM WITH UNIT 2 IN CONDITION 4 OR 5 OR DEFUELED k
| |
| -'POSITION NUMBER OF INDIVIDUALS REmUIRED TO FILL POSITION CONDITION 1, 2, or 3 CONDITION 4 OR 5 SM 1* 1*
| |
| SRO 2* 2*
| |
| RO 2 1 NLO 2 2**
| |
| STA 1*** None WITH UNIT 2 IN CONDITION 1, 2, OR 3 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITION 1, 2, or 3 CONDITION 4 or 5 SM 1* 1*
| |
| SRO 2* 2*
| |
| RO 2** 1 NLO 2** 1 STA 1*,*** None TABLE NOTATIONS
| |
| * Individual(s) may fill the same position on Unit 2.
| |
| One of the two required individuals may fill the same position on Unit 2.
| |
| UA*The STA position may be filled by an on-shift SM or SRO provided the individual meets the 1985 NRC Policy Statement on Engineering Expertise on Shift.
| |
| SM - Shift Manager with a Senior Operator license on Unit 1.
| |
| SRO - Individual with a Senior Operator license on Unit 1.
| |
| RO - Individual with an Operator license on Unit 1.
| |
| NLO - Non-licensed operator properly qualified to support the unit to which assigned.
| |
| STA - Shift Technical Advisor Except for the Shift Manager (SM), the shift crew composition may be one less than the minimum requirements of Table 6.2.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew compo-sition to within the minimum requirements of Table 6.2.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an upcoming shift crewman being late or absent.
| |
| During any absence of the Shift Manager (SM) from the control room while the unit I
| |
| *is in OPERATIONAL CONDITION 1, 2, or 3, an individual with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Manager (SM) from the control room while the unit is in OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command function.
| |
| LIMERICK - UNIT I 6-5 Amendment No 7$,73,00,96 WUL 1 8 iaa5
| |
| | |
| ADMINISTRATIVE,,LO.NIB.OLS 6.2.3 DELETED. The information from this section is located in the UFSAR.
| |
| ~.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.l The Shift Technical Advisor shall provide advisory technical support to Shift Supervision in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. The Shift Technical Advisor shall meet the qualifications specified by the 1985 NRC Policy Statement on Engineering Expertise on Shift.
| |
| 6.3 UNIT STAFF QUALIFICATIONS 6.3.l Each member of the unit staff shall meet or exceed the m1n1mum qualifications referenced for comparable positions as specified in the Exelon Quality Assurance Topical Report.
| |
| LIMERICK - UNIT 1 6-6 Amendment No. +-G, ~. hl,, M, -%, &W, 231 b8~Fet-&G-by-1=-ett~f'--4a-t-f.4 O&t-Gbe~-l+,-+9~§
| |
| | |
| ADMINISTRATIVE CONTROLS 6.4 DELETED L.5 DELETED THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR X MERICK - UNIT 1 6-7 Amendment No. 4-0, 3X, i5-5,86, 4-3-9, 44-6, 46, 176 Co irrectiomn I tv d*,4 10.'-7/LQr
| |
| | |
| ADMINISTRATIVE CONTROLS THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT 1 6-8 Amendment No. 40, 4-7, 404,, 1.31., -5.4,176
| |
| | |
| ADMINISTRATIVM CONTROLS THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT I 6-9 Amendment No. -1,A, 96, 4314, 176
| |
| | |
| AnMTNISTRATIVF CONTROLS THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT 1 6-10 Amendment No. 69. 176 . _
| |
| ,eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
| |
| | |
| THIS PAGE IS INTENTIONALLY LEFT BLANK.
| |
| LIMERICK - UNIT 1 6-11 Amendment No. 10,H,96
| |
| 'JUL 1 8 1995
| |
| | |
| ADMINISTRATIVE CONTROLS THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT 1 6-12 Amendment No. 40, 45, 4X-, -., 4-54,176
| |
| | |
| ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
| |
| : a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and
| |
| : b. Each REPORTABLE EVENT shall be reviewed by the PORC and submitted to the NRB, Plant Manager and the Vice President, Limerick Generating Station.
| |
| C7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
| |
| : a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour. The Vice President, Limerick Generating Station, Plant Manager, and the NRB shall be notified within 24 hours.
| |
| : b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the NRB. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon unit components, systems, or structures, and (3) corrective action taken to prevent recurrence.
| |
| : c. The Safety Limit Violation Report shall be submitted to the Commission, the NRB, Plant Manager, and the Vice President, Limerick Generating Station, within the 14 days of the violation.
| |
| LIMERICK - UNIT I 6 -12a Amendment No. 47, 96, 47P,176
| |
| | |
| ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued)
| |
| : d. Critical operation of the unit shall not be resumed until authorized by the Commission.
| |
| 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
| |
| : a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.
| |
| : b. The applicable procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737.
| |
| : c. Refueling operations.
| |
| : d. Surveillance and test activities of safety-related equipment.
| |
| : e. Security Plan implementation.
| |
| : f. Emergency Plan implementation.
| |
| : g. Fire Protection Program implementation.
| |
| : h. PROCESS CONTROL PROGRAM implementation.
| |
| : i. OFFSITE DOSE CALCULATION MANUAL implementation.
| |
| : j. Quality Assurance Program for effluent and environmental monitoring, using the guidance of Regulatory Guide 4.15, February 1979.
| |
| 6.8.2 The information from Section 6.8.2 has been relocated to QATR.
| |
| 6.8.3 The information from Section 6.8.3 has been relocated to QATR.
| |
| LIMERICK - UNIT 1 6-13 Amendment No. -4,4-, A_, 49, 176
| |
| | |
| ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4 The following programs shall be established, implemented, and maintained:
| |
| : a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the core spray, high pressure coolant injection, reactor core isolation cooling, residual heat removal, post-accident sampling system (until such time as a modification eliminates the PASS system as a potential leakage path), safeguard piping fill system, control rod drive scram discharge system, and containment air monitor systems. The program shall include the following:
| |
| : 1. Preventive maintenance and periodic visual inspection requirements, and
| |
| : 2. Integrated leak test requirements for each system at refueling cycle intervals or less.
| |
| : b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
| |
| : 1. Training of personnel,
| |
| : 2. Procedures for monitoring, and
| |
| : 3. Provisions for maintenance of sampling and analysis equipment.
| |
| : c. Deleted.
| |
| LIMERICK - UNIT 1 6-14 Amendment No. .2, 1'66
| |
| | |
| ?ROCEDURES AND PROGRAMS (Continued)
| |
| : d. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by oper at i no or ocedur es , and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
| |
| : 1) l imi t at i ons on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the (lDCM,
| |
| : 2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 times the concentration values in 10 CFR Part 20, Appendix B, Table 2, Column 2, J) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,
| |
| : 4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,
| |
| : 5) Determination of cumu~ative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.
| |
| Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days,
| |
| : 6) Limitations on the operability and use of the liqUid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,
| |
| : 7) Limitat:ons on the dose rate resu'ting from radioactive material released in gaseous effluents from the site to areas at or beyond the SIn BOUNDARY shall be limited to the following:
| |
| : a. For noble gases: less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin, and
| |
| : b. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: less than or equal to 1500 mrem/yr to any organ.
| |
| LIMERICK - UNIT 1 6-14a Amendment No. 197
| |
| | |
| t ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
| |
| : 8) Limitations on the annual quarterly'air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, -
| |
| : 9) Limitations on the annual and quarterly doses to a'MEMBER OF THE PUBLIC from Iodine-131, Iodine-133,'tritium, and all radionuclides-in particulate form with half-live'sgr'eater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part.50,
| |
| : 10) Limitations on venting and purging of the Mark II containment through the Standby Gas Treatment System to maintain releases as 'low as reasonably achievable, and
| |
| : 11) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity-and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
| |
| : e. Meteorological Monitoring Program A program shall be provided to provide meteorological information in the environs of the plant. The program shall provide sufficient meteorological data for estimating potential radiation'doses to the public.
| |
| The program shall (1) be contained in the ODCM, (2) conform to the guidance of Regulatory Guide 1.23, "Safety..Guide 23 - Onsite Meteorlogical Pro'gram", and (3) include limitations on the opera-bility of meteorological monitoring instrumentation including surveillance tests in accordance with the-'methodology in the ODCM.
| |
| : f. Radiological Environmental Monitoring Program A-program shall be p`rovided'to monitor the radiation and radionuclides in the enrvirons'of the plant. The'program shall provide (1) repre-sentative measurements of'radioactivity.in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling off environmental exposure pathways.' The program shall (1) be'contain'd in the 00CM, (2) con-form to the guidance of'Appendix I to 10 CFR Part 50, and (3) include the following:
| |
| : 1) Monitoring,.sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,-'
| |
| : 2) A Land Use Census to ensure that changes in the use of areas at and-beyond the SITE BOUNDARY are identified and that modifications to-the monitoringprogram are made if required by the results.of this census,''and
| |
| : 3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and'accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
| |
| LIMERICK - UNIT 1 6-14b Amendment No.48
| |
| | |
| ADMINISTRATIVE CONTROLS PROEDURES AND PROGRAMS. (Continued)
| |
| : g. Primary Containment Leakaqe Rate Testinq Program A program shal. be established to implement.the leakage rate testing of the containment as required. by 10 CFR 50.54 (o) and .10.CFR 50, Appendix J, Option B, as modified by approved exemptions.. This program shall be -in accordance with the guidelines contained *in Regulatory.Guide 1.163_
| |
| "Performance-Based Containment Leakage Test program," dated September 1995, as modified by the following. exception to NEI. 94-01,. Rev. 0, "Industry Guideline for Implementing Performance-Based Optlion of 10 CFR 50,
| |
| .Appendix J":
| |
| : a. Section 9.2.3: The first Type A test performed after:May 15, 1998 shall be performed no later-than May 15, 2013.
| |
| The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is .44.0 psig.
| |
| The maximum, allowable primary containment leakage rate, L,, at P,, shall be 0.5% of primary containment air weight per day.
| |
| Leakage rate acceptance criteria are:
| |
| : a. Primary Containment leakage rate acceptance criterion is less than or equal to 1.0 L. During the first unit startup following testing in accordance, with this program, the leakage rate acceptance criteria are less than or. equal to 0.60 L, for the Type B and Type C tests and less than or equal to 0.75 L, for Type A tests;
| |
| : b. Air lock testing acceptance criteria are:
| |
| : 1) Overall airlock leakage rate is less than or equal to 0.05 L, when tested. at greater than or equal to Pa.
| |
| : 2) Seal leakage rate is less than or equal to 5 scf per hour when the gap between the door seals is pressurized to 10 psig.
| |
| The provisions 'of Specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage. RateTesting Program.
| |
| The provisions of Specification 4.0.3 are applicable to the tests described in the Primary Containment Leakage Rate Testing'Program.
| |
| : h. Technical Specifications (TS) Bases Control Proaram This program provides a means for processing changes to the Bases of these Technical Specifications.'
| |
| : a. Changes to the Bases of the TS shall be made Under appropriate administrative controls and reviews.
| |
| : b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
| |
| A change in the TS incorporated in the license; or A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
| |
| : c. The Bases Control Program shall contain-provisions to ensure that the Bases are maintained consistent with the UFSAR.
| |
| : d. Proposed changes that meet the criteria of b. above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
| |
| LIMERICK - UNIT 1 6-14c Amendment No. 44ý9, 1-622,190
| |
| | |
| bDMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
| |
| : i. Battery Monitoring and Maintenance Program This Program provides for restoration and maintenance, based on the recorrunendations of IEEE Standard 450, "IEEE Recorrunended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries For Stationary Applications,n of the following:
| |
| : a. Actions to restore battery cells wi~h float voltage < 2.13 volts, and
| |
| : b. Actions to equalize and test battery cells that have been discovered with electrolyte level below the minimum established design limit.
| |
| : j. Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
| |
| : a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
| |
| : b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NE! 04-10, "Risk-Informed Method for Control of Surveillance Frequencies,"
| |
| Revision 0.
| |
| : c. The provisions of Surveillance Requirements 4. 0. 2 and 4. 0. 3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
| |
| : k. Snubber Program This program conforms to the examination, testing, and service life monitoring for dynamic restraints (snubbers) in accordance with 10 CFR
| |
| : 50. 55a inservice inspection ( ISI) requirements for supports. 'l'he program shall be in accordance with the following:
| |
| : a. This program shall meet 10 CFR 50. 55a ISI requirements fer supports.
| |
| : b. The program shall meet the requirements for ISI of supports set forth in subsequent editions of the Code of Record and addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code and the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) that a.!:'e incorporated by reference in 10 CFR 50.55a, subject to its conditions, and subject to Comrnission approval.
| |
| LIMERICK - UNIT 1 6-14d Amendment No. ~' ~' 223
| |
| | |
| ADMINISTE8.ll~£ ..C.ONIROLS PROCEDURES AND PROGRAMS (Continued)
| |
| : c. The program shal1, as al1owed by 10 CFR 50.55a, meet Subsection ISTA, "Genera1 Requirements," and Subsection ISTD, "Preservice and lnservice Examination and Testing of Dynamic Restraints (Snubbersl in Light-Water Reactor Nuc1ear Power Plants," in lieu of Section XI of the ASME B&PV Code ISi requirements for snubbers, or meet authorized alternatives pursuant to 10 CFR 50.55a.
| |
| : d. The 120-month program updates sha11 be made in accordance with 10 CFR 50.55a subject to the limitations and conditions listed therein.
| |
| : l. Explosive Gas Monitoring Program This program provides contro1s for potentia1ly explosive gas mixtures contained downstream of the off-gas recombiners.
| |
| The program shall include:
| |
| : a. The limit for the concentration of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limit is maintained. This limit shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);
| |
| The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the Explosive Gas Monitoring Program surveillance frequencies.
| |
| LIMERICK - UNIT l 6-14e Amendment No. ~ , 22;,
| |
| | |
| ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.
| |
| STARTUP REPORT
| |
| : 6. 9. 1. 1 Deleted
| |
| : 6. 9. 1. 2 Deleted 6.9.1.3 Deleted ANNUAL REPORTS*
| |
| 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year unless otherwise noted.
| |
| 6.9.1.5 Reports required on an annual basis shall include:
| |
| : a. Deleted
| |
| *A single submittal may be made for a multiple unit station.
| |
| LIMERICK - UNIT 1 6-15 Amendment No. ~.++S, 211
| |
| | |
| ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued)
| |
| : b. (Deleted)
| |
| : c. (Deleted)
| |
| : d. (Deleted)
| |
| MONTHLY OPERATING REPORTS*
| |
| 6.9.1.6 Deleted ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*
| |
| 6.9.1.7 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpreta-tions, analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
| |
| *A single submittal may be made for a multiple unit station.
| |
| LIMERICK - UNIT 1 6-16 Amendment No. ~.4G,4S,~,+&9.
| |
| -+/-+§., 211
| |
| | |
| ADMINISTRATIVE CONTROLS ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT*
| |
| 6.9.1.8 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The I report shall include a summary. of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
| |
| The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1. I
| |
| *A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.
| |
| LMI __UNT 1. 7o.
| |
| 6.ent 6, 44.,72, 135 LIMERICK - UNIT I 6-17 M~AY 2 4 1999
| |
| | |
| INTENTIONALLY LEFT BLANK I,
| |
| I LIMERICK - UNIT 1 6 Amendment No.Y'48 I
| |
| _JLUL~tZLa Q1AAfiJt&. Of fqq
| |
| | |
| CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:
| |
| : a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,
| |
| : b. MAPFAC(P) and MAP (F) factors for Specification 3.2.1,
| |
| : c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,
| |
| : d. The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,
| |
| : e. LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,
| |
| : f. The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6,
| |
| : g. The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,
| |
| : h. The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBOA) setpoints for Specification 2.2.1,
| |
| : i. The minimum required number of operable main turbine bypass valves for Specification 3.7.8 and the TURBINE BYPASS SYSTEM RESPONSE TIME for Specification 4.7.8.c.
| |
| 6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
| |
| : a. NEDE-24011 P A "General Electric Standard Application for Reactor Fuel" (Latest approved revision),*
| |
| : b. NEOO 32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
| |
| 6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.
| |
| 6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
| |
| SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
| |
| * For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98) to support License Amendment No. (127).
| |
| LIMERICK UNIT 1 6 18a Amendment No. J+, ,~,~,+/-++ 200
| |
| | |
| ADMINISTRATIVE CONTROLS 6.10 DELETED THE INFORMATION FROM SECTION 6.10 HAS BEEN RELOCATED TO THE QATR.
| |
| LIMERICK - UNIT 1 6-19 Amendment No. 47,176
| |
| | |
| ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared con-sistent with the requirements of 10 CFR Part 20 and shall be approved, main-tained, and adhered to for all operations involving personnel radiation exposure.
| |
| 6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:
| |
| 6.12.1 High Radiation Areas with dose rates (deep dose equivalent) greater than 0.1 rem/hr and not exceeding 1.0 rem/hour (at 30 centimeters from the radiation sources or from any surface penetrated by the radiation):
| |
| : a. Each accessible entryway to such an area shall be barricaded and conspicuously posted as a High Radiation Area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
| |
| LIMERICK - UNIT I 6-20 Amendment No. 48, 434, 176
| |
| | |
| ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)
| |
| : b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes radiation protection instructions, job coverage and monitoring requirements. Radiological information (i.e., dose rates) is included on the hadiation surveys associated with the RWP or equivalent.
| |
| : c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.
| |
| : d. Each individual or group entering such an area shall be provided with or accompanied by one or more of the following:
| |
| : 1. A radiation monitoring device that continuously displays radiation dose rates in the area ("radiation monitoring and indicating device"), OR
| |
| : 2. A radiation monitoring device with the capability to display accumulated dose and which continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached ("alarming dosimeter"), OR
| |
| : 3. A radiation monitoring device with the capability to display accumulated dose and which continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, OR
| |
| : 4. A direct reading dosimeter AND:
| |
| a) A health physics qualified individual (i.e.,
| |
| qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for controlling personnel radiation exposure within the area, OR b) Be under the surveillance, as specified in the RWP or equivalent, by means of closed circuit television, of a health physics qualified individual (i.e., qualified in radiation protection procedures),
| |
| responsible for controlling personnel radiation exposure in the area.
| |
| : e. Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been established and entry personnel are knowledgeable of them.
| |
| LIMERICK - UNIT 1 6-20a Mendment No. 13'
| |
| | |
| ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) 6.12.2 In addition to the requirements of Section 6.12.1, High U Radiation Areas with dose rates (deep dose equivalent) greater than 1.0 rem/hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation), but less than 500 rad/hr (at 1 meter from the radiation source or from any surface penetrated by the radiation source) accessible .to personnel shall be controlled as follows:
| |
| : a. Each accessible entryway to such an area shall be con-spicously posted as a High Radiation Area and shall be provided with a locked door, gate, or guard that prevents unauthorized entry, and in addition:
| |
| : 1. All such door and gate keys shall be maintained under the administrative control of radiation protection supervision.
| |
| : 2. Doors and gates shall remain locked or guarded except during periods of personnel or equipment entry or exit.
| |
| : b. Such individual areas that are within a larger area, such as containment, that is controlled as a High Radiation Area, where no enclosure exists for purpose of locking and where no enclosure can reasonable be constructed around the individual area need not be controlled by a locked door or gate, but shall be barricaded and conspicuously posted as a High Radiation Area, and a conspicuous, clearly visible flashing light shall be activated at the area as a warning device.
| |
| : c. Each individual entering such an area shall be provided with or accompanied by one or more of the following:
| |
| : 1. A dose rate survey meter and a radiation monitoring device with the capability to display accumulated dose and an integrating alarm setpoint, OR
| |
| : 2. A radiation monitoring device with the capability to display accumulated dose and which continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached ("alarming dosimeter"), OR
| |
| : 3. A radiation monitoring device with the capability to display accumulated dose and which continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area AND with the means to communicate with the individuals in the area, OR
| |
| : 4. A direct reading dosimeter AND:
| |
| LIMERICK - UNIT 1 6-21 kAnchent Nb. 48, 135 JAAY 24 S'9s
| |
| | |
| ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) a) A health physics qualified individual (i.e.,
| |
| qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for controlling personnel radiation exposure within the area, OR b) Be under the surveillance, as specified in the RWP or equivalent, by means of closed circuit television, of a health physics qualified individual (i.e., qualified in radiation protection procedures),
| |
| responsible for controlling personnel radiation exposure in the area, and with the means to communicate with the individuals in the area.
| |
| 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 Changes to the PCP:
| |
| : a. Shall be documented with the following information:
| |
| : 1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and LIMERICK - UNIT 1 6-21a Amendment No. 445, 176
| |
| | |
| ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (Continued)
| |
| : 2. A determination that the change did not reduce the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
| |
| : b. Shall become effective upon review and acceptance by the PORC and approval of the Plant Manager.
| |
| 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 Changes to the ODCM:
| |
| : a. Shall be documented with the following information:
| |
| : 1. Sufficient information to support the change together with the appropriate analyses or. evaluations justifying the change(,s) and
| |
| : 2. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
| |
| : b. Shall become effective upon review and acceptance by the RORC and the approval of the Plant Manager.
| |
| : c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report .in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.
| |
| 6.15 (Deleted) -INFORMATION FROM THIS SECTION RELOCATED TO THE OCM.
| |
| 6.16 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE h'abitability is maintained such that, with an OPERABLE Control Room Emergency Fresh Air Supply (CREFAS) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protectfon is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions withoutpersonnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
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| a.-__ Thhe-de.fini-t.i-on .o*__the --CRE--and the--CRE boundary.
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| .... ..... . Req iemen.ts for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
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| LIMERICK - UNIT I 6-22 Amendment No. 49,49,-74,4-6,-71-9-7, 188
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| ADMINISTRATIVE CONTROLS CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (Continued)
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| : c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.l and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
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| : d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREFAS, operating at the flow rate required by SR 4.7.2.1.c.l, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.
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| : e. The quantitative limits on unfiltered air inleakage into the CRE.
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| These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
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| : f. The provisions of Specification 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
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| 6.17 SAFETY FUNCTION DETERMINATION PROGRAM (SFDP)
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| This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into Specification 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system ACTIONS. This program implements the requirements of Specification 3.0.6. The SFDP shall contain the following:
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| : a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected,
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| : b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists,
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| : c. Provisions to ensure that an inoperable supported system's Allowed Outage Time is not inappropriately extended as a result of multiple support system inoperabilities, and LIMERICK - UNIT l 6-23 Amendment No. -+/-gg, 219
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| ADMINISTRATIVE CONTROLS 6.17 SAFETY FUNCTION DETERMINATION PROGRAM (SFDP) (Continued)
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| : d. Other appropriate limitations and remedial or compensatory actions.
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| A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
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| : a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable,
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| : b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable, or
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| : c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
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| The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate ACTIONs of the Limiting Condition for Operation in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate ACTIONs to enter are those of the support system.
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| LIMERICK - UNIT 1 6-24 Amendment No. 219 I
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| APPENDIX B TO FACILITY OPERATING LICENSE NO. NPF-39 LIMERICK GENERATING STATION UNIT 1 I EXELON GENERATION COMPANY, LLC DOCKET NO. 50-352 I ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL)
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| LIMERICK - UNIT 1 Amendment No. 4-34, 144, 180,
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| LIMERICK GENERATING STATION UNIT 1 ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL)
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| TABLE OF CONTENTS Section ts 1.0 Objectives of the Environmental Protection Plan ............. 1-1 2.0 Environmental Protection Issues ............................. 2-1 2.1 Aquatic Issues .2-1 2.2 Terrestrial Issues .2-2 2.3 Noise Issues .2-2 3.0 Consistency Requirements .3-1 3.1 Plant Design and Operation .3-1 3.2 Reporting Related to the NPDES Permit and State Certifications............................................ 3-2 3.3 Changes Required for Compliance with Other Environmental Regulations............................................... 3-3 4.0 Environmental Conditions .4-1 4.1 Unusual or Important Environmental Events .4-1 4.2 Environmental Monitoring.................................... 4-1 5.0 Administrative Procedures .5-1 5.1 Review and Audit .5-1 5.2 Records Retention. 5-1 5.3 Changes in Environmental Protection Plan. 51 5.4 Plant Reporting Requirements .5-2
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| 1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of nonradiological environmental values during operation of the nuclear Facility. The principal objectives of the EPP are as follows:
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| :1) Verify that the facility is operated in an environmentally acceptable manner, as established by the Final Environmental Statement- Operating Licensing Stage (FES-OL) and other NRC environmental impact assessments.
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| (2) Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.
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| (3) Keep NRC informed of the environmental effects of facility operation and of actions taken to control those effects.
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| Environmental concerns identified in the FES-OL which relate to water quality matters are regulated by way of the licensee's NPDES permit.
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| 1 -1 Amendment No.180,
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| 2.0 Environmental Protection Issues In the FES-OL dated April, 1984, the staff considered the environmental impacts associated with the operation of the two unit Limerick Generating Station. Certain environmental issues were identified which required study or license conditions to resolve environmental concerns and to assure adequate protection of the environment.
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| 2.1 Aquatic Issues (1) During operation, the station blowdown temperature will exceed the maximum permissible temperatures set by the applicable water quality standards. However, the affected area of the Schuylkill River is expected to be smaller than the maximum area permitted by the Delaware River Basin Commission. (FES Section 5.3.2.2)
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| (2) The water quality of the station discharge, after initial mixing with the Schuylkill River, is predicted to, at times, exceed the applicable quality criteria for some constituents, based on source water maximum constituent concentrations. These exceedances are expected for con-stituents whose maximum river concentrations also exceed the applicable criteria. (FES Section 5.3.2.3) 2-1
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| (3) Chlorination of station cooling waters for condenser and cooling tower biofouling control may result in some adverse impacts to Schuylkill River biota in the vicinity of the station discharge (FES Section 5.3.2.3)
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| (4) Operation of the Point Pleasant Diversion will alter the hydrology, aquatic habitats, and water quality of the headwater section of the East Branch of Perkiomen Creek but the diversion waters are expected to provide beneficial dilution of waste loads entering the stream in its middle and lower reaches. (FES Sections 5.3.2.3 and 5.2.2)
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| (5) The supplemental cooling water withdrawal from Perkiomen Creek using state-of-the-art technology will result in localized effects from entrainment of fish larvae. (FES Section 5.5.2) 2.2 Terrestrial Issues No specific terrestrial issues were identified by the NRC staff in the FES-OL.
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| 2.3 Noise Issues (1) Tones from the Point Pleasant pumphouse transformers are predicted to be audible and may cause annoyance at a nearby residence. Noise monitor-2-2 Amendment No.180,
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| ing and, if necessary, mitigative measures to make the tones inaudible have been mandated by the ASLB. (FES Sections 5.12.1 and 5.14.4.1)
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| (2) Noise from transformers and pumps in the Bradshaw Reservoir pumphouse may be audible at nearby residences. The licensee has committed to ambient and operational noise level monitoring and implementation of identified mitigative measures, if necessary, to reduce noise levels below those likely to cause annoyance and complaints. (FES Sections 5.12.2 and 5.14.4.2)
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| (3) Offsite noise levels in the vicinity of the Limerick site during station operation are not expected to be high enough above ambient levels to annoy nearby residents. But because of uncertainties in the assessment, a confirmatory noise monitoring program and implementa-tion of mitigative measures, if necessary, will be undertaken. (FES Sections 5.12.3 and 5.14.4.3)
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| NRC requirements with regard to noise issues are specified in Section 4.3 of this EPP.
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| 2-3
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| 3.0 Consistency Requirements 3.1 Plant Design and Operation The licensee may make changes in station design or operation or perform tests or experiments affecting the environment provided such activities do not involve an unreviewed environmental question and do not involve a change in the EPP*. Changes in station design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this EPP. Activities governed by Section 3.3 are not subject to the requirements of this Section.
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| Before engaging in additional construction or operational activities which may significantly affect the environment, the licensee shall prepare and record an environmental evaluation of such activity. Activities are excluded from this requirement if all measurable nonradiological environ-mental effects are confined to the on-site areas previously disturbed during site preparation and plant construction. When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activity and obtain prior NRC approval. When such activity involves a change in the EPP, such activity and change to the EPP may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3 of this EPP.
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| * This provision does not relieve the licensee of the requirements of 10 CFR 50.59.
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| 3-1
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| A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns: (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the FES-OL, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) a significant change in effluents or power level; or (3) a matter, not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact.
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| The licensee shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this Sub-section. These records shall include written evaluations which provide bases for the determination that the change, test, or experiment does not involve an unreviewed environmental question or constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0.
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| 3.2 Reporting Related to t.he NPDES Permit and State Certification Changes to, or renewals of, the NPDES Permits or the State certification shall be reported to the NRC wii:hin 30 days following the date the change or renewal is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.
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| 3-2 Amendment No. 180
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| 3.3 Changes Required for Compliance with Other Environmental Regulations Changes in plant design or operation and performance of tests or experiments that are either regulated or mandated by Federal, State, and local environmental regulations are not subject to the requirements of Section 3.1. However, if any environmental impacts of a change are not evaluated under other Federal, State, or local environmental regulations, then those impacts are subject to the requirements of Section 3.1.
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| 3-3 Amendment No.180
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| 4.0 Environmental Conditicns 4.1. Unusual or Important Environmental Events Any' occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded *and reported to the NRC within 24 hours followed by a written report per Subsection 5.4.2. If an event is reportable under 10 CFR 50.72, then a duplicate immediate report under this subsection is not required. However, a follow-up written report is required in accordance with Subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973, fish kills, increase in nuisance organisms or conditions, and unanticipated or emergency discharge of waste water or chemical substances.
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| No routine monitoring programs are required to implement this condition.
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| 4.2 Environmental Monitoring 4.2.1 Aquatic Monitoring The certifications and permits required under the Clean Water Act provide mechanisms for protecting water quality and, indirectly, aquatic biota.
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| The NRC will rely on the decisions made by the Commonwealth of Pennsylvania, under the authority of the Clean Water Act, for any requirements for aquatic monitoring 4 -1 Amendment No.180
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| 4.2.2 Terrestrial Monitoring No terrestrial monitoring is required.
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| 4.2.3 Maintenance of Transmission Line Corridors The use of herbicides within the Limerick Generating Station transmission line corridors (Limerick to Cromby, Cromby to Plymouth Meeting, Cromby to North Wales, and Limerick to Whitpain) shall conform to the approved use of selected herbicides as registered by the Environmental Protection Agency and approved by Commonwealth authorities and applied as directed on the pesticide label.
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| 4-2 Amendment No.180
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| 4.2.4 Noise Monitoring All initial Environmental Noise Assessments have been completed.
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| The information in Subsections 4.2.4.1, 4.2.4.2, 4.2.4.3, and 4.2.4.4 has been deleted.
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| Pages 4-4, 4-5, 4-6 and 4-7 have been removed from this Section.
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| 4-3 Amendment No.180
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| 5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the EPP.
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| The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and made available for inspection.
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| 5.2 Records Retention Records associated with this Environmental Protection Plan shall be made and retained in a manner convenient for review and inspection. These records shall be made available to NRC on request.
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| Records of modifications to station structures, systems and components determined to potentially affect the continued protection of the environ-ment shall be retained until the date of termination of the operating license.
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| All other records relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
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| 5-1I Amendment No. 180
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| 5.3 Changes in Environmental Protection Plan Requests for changes in the EPP shall include an assessment of the environmental impact of the proposed change and a supporting justification.
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| implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the EPP.
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| 5.4 Plant Reporting Requirements 5.4.1 Deleted 5 -2 Amendment No.180
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| 5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of an Unusual or Important Environmental event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact, and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective 5-3 Amendment No.180
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| action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.
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| Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided with a copy of such report at the same time it is submitted to the other agency.
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| 5-4
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| APPENDIXC ADDITIONAL CONDITIONS OPERATING LICENSE NO. NPF-39 Exelon Generation Company, LLC shall comply with the following conditions on the schedule noted below:
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| Amendment No. Additional Conditions 230 Exelon is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e.,
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| seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 License Amendment No. 230 dated July 31, 2018.
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| Exelon will complete the implementation items listed in Attachment 2 of Exelon letter to NRC dated April 23, 2018 prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.
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| Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
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| Renewed License No. NPF-39 Amendment No.~. ~ . 447, 484, 230}}
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