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| {{#Wiki_filter:License No. NPF-39 Limerick Generating Station, Unit No. 1 Docket No. 50-352 Issued by the U.S. Nuclear Regulatory Commiulssion Office of Nuclear Reactor Regulation | | {{#Wiki_filter:}} |
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| LIMERICK GENERATING STATION UNIT 1 OPERATING LICENSE NPF-39 PAGE REVISION LIST (Generated by Exelon - Not part of Operating License)
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| Page/Attachment Amendment No.
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| 1 thru 2 Revised by letter dated October 28, 2004 1 thru 4 147 5 131 6 Deleted 7 thru 7a 147 8 181 I 9 128 Dated August 8, 1985 Dated August 8, 1985 License Amendment Change List (Generated by Exelon)
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| Page Dated December 19, 1994 LIMERICK - UNIT 1 - A - Amendment No. 181
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| EXELON GENERATION COMPANY, LLC LIMERICK GENERATING STATION, UNIT 1 DOCKET NO. 50-352 ADMINISTRATIVE LICENSE CHANGE TO FACILITY OPERATING LICENSE License No. NPF-39 The Nuclear Regulatory Commission (the Commission) has found that:
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| A. Consistent with Order EA-03-086, the Order requiring compliance with the revised design basis threat, (DBT Order) issued on April 29, 2003, conforming administrative changes to Facility Operating Licenses (FOLs) are required to ensure implementation of DBT Order requirements. Therefore, an administrative license change to FOL No. NPF-39 is being made to incorporate the reference to the revised Physical Security Plan, Safeguards Contingency Plan, and Training and Qualification Plan required by the DBT Order. These changes comply with the standards and requirements of the Atomic Energy Act (the Act) of 1954, as amended and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this administrative license change 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; D. The issuance of this administrative license change will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this administrative license change is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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| Enclosure 1
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| : 2. Accordingly, the FOL is changed as indicated in the Attachment.
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| : 3. This administrative license change is effective as of its date of issuance and shall be implemented on or before October 29, 2004.
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| FOR THE NUCLEAR REGULATORY COMMISSION IRA/
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| John A. Nakoski, Chief Security Plan Review Team Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
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| ==Attachment:==
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| Changes to the FOL Date of Issuance: October 28, 2004
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| UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO 50-352 LIMERICK GENERATING STATION, UNIT 1 FACILITY OPERATING LICENSE License No. NPF-39 The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
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| A. The application for 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 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 actMties authorized by this 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 license will not be inimical to the common defense and security or the health and safety of the public; Amendment No. 4a1-, 147
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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 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 license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70.
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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 Facility Operating License NPF-39 hereby issued to the Exelon Generation Company (the licensee), to read as follows:
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| A. This 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 license; (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; Amendment No. 431-, 147
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| (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 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 3458 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 license shall be completed as specified. Attachment 1 is hereby incorporated into this 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. 186, are hereby incorporated into this license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
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| Amendment No. 62, 16,4 31,, 147
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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 ifthose changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
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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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| Amendment No. 404, 4-1-, 147
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| The Information contained on FOL pages 5 and 6 were intentionally omitted.
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| Amendment No. 131 NOVY 5 1998
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| (16) Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 147, are hereby incorporated into this license. 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.
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| 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 Amendment No. -148, 42-8, 431, 147
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| - 7a -
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| from the trust shall be made if the trustee receives prior written notice of objection from the NRC.
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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 Commission's 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, license and the requirements of the Order approving the transfer, and consistent with the safety evaluation supporting the Order.
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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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| Amendment No. 147
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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. 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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| 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). The combined set of plans', which contain Safeguards Information protected under 10,CFR 73.21, is entitled: "Limerick Generating Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 0," submitted by letter dated October 21, 2004.
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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 accord-ance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
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| The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
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| 9 Amendment No. ,- 131, 181 Revised by letter dated October 28, 2004 Revised by lcttcr dated Deccmbcr 10, .2001
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| H. This ficanse is effective as of the date of issuance and shall expire at midnight on October 26, 2024.
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| FOR THE NUCEAR REGULATORY COMMISSION original signed by Darrell Eisenhut for Harold R. Denton, 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 (NUREG-1 149)
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| : 3. Appendix B - Environmental Protection Plan
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| : 4. Appendix C - Additional Conditions Date of Issuance: August 8, 1985 Amendment No. 128 JUN 1 5 199
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| ATTACHMENT 1 8387 2 0 0 3 7 33 To-NPF 39 38851 25590 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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| AU6 8 195
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| ... ) 3872003733 ATTACHMENT 2 To NPF-39 388512559C 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 inaividual 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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| AUG6 8INS
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| UNIT 1 OPERATING LICENSE NPF - 39 LICENSE AMENDMENT CHANGE LIST (Generated by PECo from NRC Issued Amendments)
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| License Section 2.C.(2)
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| The License and the Technical Specifications contained in Appendix A have been revised through the Amendment number on the cover sheet of the License. The Technical Specifications as contained in Appendix A are hereby incorporated into this license. PECO Energy shall operate the facility in accordance with Technical Specifications and the Environmental Protection Plan.
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| License Sections 2.C(10),(11)
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| License Amendment No. 13 issued January 10, 1989 The license is amended by deleting license conditions 2.C.10 and 2.C.11.
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| License Section 2.E License Amendment No. 9 issued August 3, 1988 The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical, guard training and qualification, and safeguards contingency plans including amendments and pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:
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| "Limerick Generating Station, Units 1 and 2, Physical Security Plan," with revisions submitted through December 7, 1987; "Limerick Generating Station, Units 1 and 2, Plant Security Personnel Training and Qualification Plan," with revisions submitted through October 1 1985; and "Limerick Generating Station, Units 1 and 2, Safeguards Contingency Plan," with revisions submitted through November 15, 1986. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
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| LIMERICK - UNIT 1 - i- 12/19/94
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| Technical Specifications Limerick Generating Station, Unit No. 1 Docket No. SO-=S2 Appendix "A" to License No. NPF- 39 I2umd by the U.S. Nuci~. ReuLd*ry c ommfl Rc Otfc @1 Nuciaar Reacl~r Regulation
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| % "
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 66 ii 66 iii 66 iv 33 v 37 vi 66 via Deleted vii 186 viii 186 ix 186 x 186 xi 182 xii 146 xiii 173 xiv 104 xv 93 xvi 133 xvii 48 xviii 69 xix 153 xx 33 xxi 104 xxii 133 xxiii 48 xxiv 48 xxv Original Issue xxvi 176 xxvii 176 xxviii 48 Section 1.0 Definitions 1-1 Original Issue 1-2 185 1-3 66 1-4 73 1-5 146 1-6 185 1-7 185 1-8 52 1-9 71 1-10 149 LIMERICK - UNIT 1 -A-
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| Section 2.0 Safety Limits and Limiting Safety System Settings 2-1 183 2-2 Original Issue 2-3 141 2-4 177 Bases for Section 2.0 B 2-1 183 B 2-2 7 B 2-3 7 B 2-4 7 B 2-5 33 B 2-6 177 B 2-7 177 B 2-7a 177 B 2-8 89 B 2-9 Original Issue B 2-10 177 Sections 3.0 and 4.0 Limiting Conditions for Operation and Surveillance Requirements 3/4 0-1 169 3/4 0-2 169 3/4 0-3 171 3/4 1-1 Original Issue 3/4 1-2 29 3/41-3 178 3/4 1-4 186 3/4 1-5 186 3/4 1-6 186 3/4 1-7 Original Issue 3/4 1-8 169 3/4 1-9 143 3/4 1-10 186 3/4 1-11 169 3/4 1-12 124 3/4 1-13 169 3/4 1-14 186 3/4 1-15 Original Issue 3/4 1-16 17 3/4 1-17 17 LIMERICK - UNIT 1 -B -
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 3/4 1-18 186 3/4 1-19 186 3/4 1-20 186 3/4 1-21 22 3/4 1-22 22 3/4 2-1 186 3/4 2-2 37 3/4 2-3 thru 3/4 2-6c Deleted 3/4 2-7 66 3/4 2-8 66 3/4 2-9 186 3/4 2-10 37 3/4 2-1Oa thru 3/4 2-1Oc Deleted 3/42-11 Deleted 3/4 2-12 186 3/4 3-1 177 3/4 3-1a 186 3/4 3-2 177 3/4 3-3 89 3/4 3-4 177 3/4 3-5 177 3/4 3-6 177 3/4 3-7 186 3/4 3-8 186 3/4 3-9 169 3/4 3-10 186 3/43-11 89 3/4 3-12 Original Issue 3/4 3-13 33 3/4 3-14 112 3/4 3-15 112 3/4 3-16 185 3/4 3-17 146 3/4 3-18 106 3/4 3-19 161 3/4 3-20 106 3/4 3-21 112 3/4 3-22 112 3/4 3-23 132 3/4 3-24 132 3/4 3-25 132 3/4 3-26 146 3/4 3-27 186 3/4 3-28 186 3/4 3-29 186 LIMERICK - UNIT 1 -C-
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 3/4 3-30 186 3/4 3-31 186 3/4 3-32 186 3/4 3-33 Original Issue 3/4 3-34 33 3/4 3-35 53 3/4 3-36 158 3/4 3-36a 158 3/4 3-37 33 3/4 3-38 18 3/4 3-39 132 3/4 3-40 186 3/4 3-41 186 3/4 3-42 186 3/4 3-43 70 3/4 3-44 106 3/4 3-45 186 I 3/4 3-46 70 I 3/4 3-47 186 3/4 3-48 70 3/4 3-49 Original Issue 3/4 3-50 Original Issue 3/4 3-51 186 3/4 3-52 186 3/4 3-53 53 3/4 3-54 53 3/4 3-55 33 3/4 3-56 186 3/4 3-57 186 3/4 3-58 177 3/4 3-59 141 3/4 3-60 177 3/4 3-60a 177 3/4 3-60b 34 3/4 3-61 186 3/4 3-62 186 3/4 3-63 186 3/4 3-64 185 3/4 3-65 185 3/4 3-66 186 3/4 3-67 186 3/4 3-68 75 3/4 3-69 thru 3/4 3-72 Deleted 3/4 3-73 48 LIMERICK - UNIT 1
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 3/4 3-74 thru 3/4 3-75 Deleted 3/4 3-76 186 3/4 3-77 45 3/4 3-78 33 3/4 3-79 Original Issue 3/4 3-80 74 3/4 3-81 33 3/4 3-82 Original Issue 3/4 3-83 186 3/4 3-84 186 3/4 3-85 179 3/4 3-86 173 3/4 3-87 186 3/4 3-88 186 3/4 3-89 117 3/4 3-90 186 3/4 3-91 186 3/4 3-92 104 3/4 3-92a Deleted 3/4 3-93 Deleted 3/4 3-94 Deleted 3/4 3-95 Deleted 3/4 3-96 Deleted 3/4 3-97 153 3/4 3-98 48 3/4 3-99 thru 3/4 3-102 Deleted 3/4 3-103 186 3/4 3-104 48 3/4 3-105 48 3/4 3-106 48 3/4 3-107 186 3/4 3-108 186 3/4 3-109 48 3/43-110 100 3/43-111 Deleted 3/43-112 186 3/43-113 91 3/43-114 Original Issue 3/43-115 186 3/44-1 177 3/4 4-1a 177 3/4 4-2 186 3/4 4-3 177 LIMERICK - UNIT 1
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 3/4 4-4 186 3/4 4-4a 186 3/4 4-5 186 3/4 4-6 Original Issue 3/4 4-7 186 3/4 4-8 169 3/4 4-8a 186 3/4 4-9 182 3/44-10 186 3/44-11 182 3/4 4-12 174 3/4 4-13 Deleted 3/4 4-14 Deleted 3/4 4-15 174 3/4 4-16 20 3/4 4-17 186 3/4 4-18 186 3/4 4-19 186 3/4 4-20 163 3/4 4-21 167 3/4 4-22 186 3/4 4-23 169 3/4 4-24 11 3/4 4-25 186 3/4 4-26 186 3/4 5-1 131 3/4 5-2 131 3/4 5-3 169 3/4 5-4 186 3/4 5-5 186 3/4 5-6 95 3/4 5-7 186 3/4 5-8 Original Issue 3/4 5-9 186 3/4 6-1 186 3/4 6-2 146 3/4 6-3 185 3/4 6-4 118 3/4 6-5 169 3/4 6-6 186 3/4 6-7 107 3/4 6-8 118 3/4 6-9 186 LIMERICK - UNIT 1 -F
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 3/4 6-10 186 3/46-11 186 3/4 6-12 29 3/4 6-13 186 3/4 6-14 186 3/4 6-15 186 3/4 6-16 186 3/4 6-17 169 3/4 6-18 186 3/4 6-19 146 3/4 6-20 thru 3/4 6-43a Deleted 3/4 6-44 46 3/4 6-45 186 3/4 6-46 186 3/4 6-47 186 3/4 6-48 186 3/4 6-49 105 3/4 6-50 186 3/4 6-51 105 3/4 6-51a 105:
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| 3/4 6-52 186 3/4 6-53 186 3/4 6-54 122 3/4 6-55 186 3/4 6-56 186 3/4 6-57 173 3/4 6-58 186 3/4 6-59 186 3/4 7-1 131 3/4 7-2 186 3/4 7-3 131 3/4 7-4 186 3/4 7-5 186 3/4 7-6 186 3/4 7-7 186 3/4 7-8 Original Issue 3/4 7-9 3/4 7-10 3/47-11 186 186 51 I
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| 3/4 7-11 a 51 3/4 7-11b 51 3/4 7-12 51 3/4 7-13 81 LIMERICK - UNIT 1
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 3/4 7-14 54 3/4 7-15 54 3/4 7-16 54 3/4 7-17 186 3/4 7-18 Original Issue 3/4 7-19 104 3/4 7-20 Deleted 3/4 7-21 Deleted 3/4 7-22 Deleted 3/4 7-23 Deleted 3/4 7-24 Deleted 3/4 7-25 Deleted 3/4 7-26 Deleted 3/4 7-27 Deleted 3/4 7-28 Deleted 3/4 7-29 Deleted 3/4 7-30 Deleted 3/4 7-31 Deleted 3/4 7-33 186 I 3/48-1 40 3/4 8-1a 40 3/4 8-2 40 3/4 8-2a 169 3/4 8-3 186 3/4 8-4 186 3/4 8-5 186 3/4 8-6 186 3/4 8-7 186 3/4 8-7a 186 3/4 8-8 186 3/4 8-9 32 3/4 8-10 164 3/4 8-1Oa 164 3/48-11 186 3/4 8-12 186 3/4 8-13 164 3/4 8-14 164 3/4 8-14a 164 3/4 8-15 131 3/4 8-16 24 3/4 8-16a 139 3/4 8-17 186 I 3/4 8-18 24 3/4 8-1Ba 24 3/4 8-19 139
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| -H -
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| LIMERICK - UNIT 1
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| 3/4 8-20 186 3/4 8-21 93 3/4 8-22 Deleted 3/4 8-23 Deleted 3/4 8-24 Deleted 3/4 8-25 Deleted 3/4 8-26 Deleted 3/4 8-27 186 3/4 8-28 186 3/49-1 149 3/4 9-2 186 3/4 9-3 186 3/4 9-4 186 3/4 9-5 186 3/4 9-6 Original Issue 3/4 9-7 186 3/4 9-8 43 3/4 9-9 43 3/4 9-10 186 3/49-11 186 3/4 9-12 186 3/4 9-13 Original Issue 3/4 9-14 186 3/4 9-15 Original Issue 3/4 9-16 186 3/4 9-17 186 3/4 9-18 186 3/4 10-1 186 3/4 10-2 17 3/4 10-3 186 3/4 10-4 186 3/4 10-5 186 3/4 10-6 186 3/4 10-7 133 3/4 10-8 133 3/4 11-1 48 3/4 11-2 thru 3/4 11-6 Deleted 3/4 11-7 186 I 3/4 11-8 48 3/4 11-9 thru 3/4 11-1 4 Deleted 3/4 11-15 11 3/4 11-16 186 3/4 11-17 48 3/4 11-18 48 3/4 11-19 thru 3/4 11 -120 Deleted 3/4 12-1 48 3/4 12-2 thru 3/4 12-1 4 Deleted LIMERICK - UNIT 1 - I-
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| Bases for Sections 3.0 and 4.0 B 3/4 0-1 16 B 3/4 0-2 B 3/4 0-3 169 169 B 3/4 0-3a 169 B 3/4 0-3b 169 B 3/4 0-3c 169 B 3/4 0-4 162 B 3/4 0-5 169 B 3/4 0-6 171 B 3/4 1-1 Original Issue B 3/4 1-2 168 B 3/4 1-2a 178 B 3/4 1-3 186 B 3/4 1-4 185 B 3/4 1-5 186 B 3/4 2-1 66 B 3/4 2-2 66 B 3/4 2-3 7 B 3/4 2-4 ECR LG 99-01138 B 3/4 2-5 66 B 3/4 3-1 186 B 3/4 3-1 a 177 B 3/4 3-1b 177 B 3/4 3-1 c 177 B 3/4 3-id 177 B 3/4 3-1 e 186 B 3/4 3-2 186 B 3/4 3-3 186 B 3/4 3-3a 158 B 3/4 3-4 186 B 3/4 3-5 186 B 3/4 3-5a 151 B 3/4 3-6 186 B 3/4 3-7 177 B 3/4 3-8 Original Issue B 3/4 3-9 141 B 3/4 4-1 177 B 3/4 4-2 186 B 3/4 4-3 140 B 3/4 4-3a 140 B 3/4 4-3b 169 B 3/4 4-3c 186 B 3/4 4-3d 186 B 3/4 4-3e 182 LIMERICK - UNIT 1
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| B 3/4 4-4 174 B 3/4 4-5 167 B 3/4 4-6 171 B 3/4 4-6a 119 B 3/4 4-7 145 B 3/4 4-8 106 B 3/4 5-1 ECR LG 00-00177 B 3/4 5-2 186 B 3/4 5-2a 152 B 3/4 6-1 118 B 3/4 6-2 186 B 3/4 6-3 106 B 3/4 6-3a 68 B 3/4 6-4 186 I B 3/4 6-4a 148 B 3/4 6-5 186 B 3/4 6-6 173 B 3/4 7-1 27 B 3/4 7-1 a 185 B 3/4 7-2 51 B 3/4 7-3 81 B 3/4 7-4 104 B 3/4 7-5 52 B 3/4 8-1 164 B 3/4 8-la 169 B 3/4 8-1 b 164 B 3/4 8-ic 164 B 3/4 8-2 186 B 3/4 8-2a 186 B 3/4 8-2b 164 B 3/4 8-3 134 B 3/4 9-1 4 B 3/4 9-2 119 B 3/4 9-2a ECR LG 01-00386 B 3/4 10-1 133 B 3/4 10-2 167 B3/4 11-1 48 B 3/4 11-2 48 B 3/4 11-3 48 B 3/4 11-4 48 B 3/4 11-5 48 B 3/4 12-1 48 B 3/4 12-2 Deleted
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| -K -
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| LIMERICK - UNIT 1
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| LIMERICK GENERATING STATION UNIT 1 TECHNICAL SPECIFICATION PAGE REVISION LIST Index Amendment Nos.
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| Section 5.0 Design Features 5-1 48 5-2 Original Issue 5-3 Original Issue 5-4 Original Issue 5-5 Original Issue 5-6 48 5-7 19 5-8 106 5-9 Original Issue Section 6.0 Administrative Controls 6-1 96 6-2 139 6-3 162 6-4 35 6-5 96 6-6 96 6-7 176 6-8 176 6-9 176 6-10 176 6-11 96 6-12 176 6-12a 176 6-13 176 6-14 166 6-14a 48 6-14b 48 6-14c 162 6-14d 186 6-15 175 6-16 175 6-17 135 6-18 48 6-18a 177 6-19 176 6-20 176 6-20a 135 6-21 135 6-21a 176 6-22 176 LIMERICK - UNIT 1
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| INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACTION ..................................................... 1-1 1.2 AVERAGE PLANAR EXPOSURE .................................... 1-1 1.3 AVERAGF PLANAR LINEAR HEAT GENERATION RATE ................. 1-1 1.4 CHANNEL CALIBRATION ......................................... 1-1 1.5 CHANNEL CHECK .............................................. 1-1 1.6 CHANNEL FUNCTIONALTEST ................................... . 1-1 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 1-131 ...................................... 1-2 1.9a DOWNSCALE TRIP SETPOINT (DTSP) ............................. 1-2 1.10 E-AVERAGE DISINTEGRATION ENERGY ............................ 1-2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ......... 1-2 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 IDENTIFIED 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. $3, 37, 66 FEB 10 1994
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| INDEX DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.20a LOW (POWER) TRIP SETPOINT (LTSP) ........................ 1-4 1.21 (DELETED) ................................................... 1-4 1.22 MEMBER(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 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 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY ............ 1-6 1.36 REPORTABLE EVENT ........................................... 1-7 1.37 ROD DENSITY ................................................ 1-7 1.38 SHUTDOWN MARGIN ............................................ 1-7 1.39 SITE BOUNDARY .............................................. 1-7 1.40 (DELETED) .................................................. 1-7 1.41 SOURCE CHECK ............................................... 1-7 LIMERICK - UNIT 1 ii Amendment No. 3, $8, *2, FEB 10 !994 66
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| INDEX DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.42 STAGGERED TEST BASIS ....................................... 1-8 1.43 THERMAL POWER ............................................. 1-8 1.43A TURBTNF BYPASS SYSTEM RESPONSE TIME ........................ 1-8 1.44 UNIDENTIFIED LEAKAGE ....................................... 1-8 1.45 UNRESTRICTED AREA .......................................... 1-8 1.46 VENTILATION EXHAUST TREATMENT SYSTEM ....................... 1-8 1.47 VENTING .................................................. -8 Table 1.1, Surveillance Frequency Notation ........................... 1-9 Table 1.2, Operational Conditions .................................... 1-10 LIMERICK - UNIT I iii Amendment No. 66 FEB 10 1994
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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 I iv Amendment No.' %, 33 OCT 3 0 798q
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY ............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 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 Control Rod Operability ............................ ...... 3/4 1-3 Control Rod Maximum Scram Insertion Times ................ 3/4 1-6 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 Control Rod Drive Coupling ............................... 3/4 1-11 Control Rod Position Inaication .......................... 3/4 1-13 Control Rod Drive Housing Support ........................ 3/4 1-15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth 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 I LIMERICK'- UNIT 1 V Amendment No.,,2f, 37 MAY 1 5 1990
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| 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,.... 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 LIMERICK UNIT - 1 vi Amendment No. 7, 19, 32, V7, 66 FEB 1 0 !'%i-4
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 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
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| ,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 3/4.3.4 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 40. _(3, 186
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUTREMENTS 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 I 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 I NSTRUMENTATION ............................. 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 LIMERICK - UNIT 1 viii Amendment No. -34, 186
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 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, -75, 4--4, 444-L,186
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (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 Offgas Monitoring Instrumentation ............................... 3/4 3-103 Table 3.3.7.12-1 Offgas Monitoring Instrumentation ................ .... 3/4 3-104 Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements ................ ..... 3/4 3-107 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 .
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| 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 I 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops ............................................. 3/4 4-1 LIMERICK - UNIT I x Amendment No. 48, 4-40, 4-347, 44-3, 186
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 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 I 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 STRUCTURAL INTEGRITY ........................................ 3/4 4-24 LIMERICK - UNIT 1 xi Amendment No. 4-T-7, I-7.4, 1-7-7, 182
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown .......................................... 3/4 4-25 Col d 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 ECCS - SHUTDOWN -. 5-6 3.............
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| 3/4 3/4.5.3 SUPPRESSION CHAMBER ........ ............. .... 3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY. CONTAINMENT.
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| Primary Containment Integrity....... 3/4-6-1 PrImary Containment Leakage ........................ 3/4 6-2 Prima*. Containment Air Lock...................... 3/4 6-5 MSIV Leakage Alternate Drain Pathway................ 3/4 6-7 Primary Containment Structural Integrity ............ 3/4 6-8 Drywell 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
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| -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 OCT 18 ZJO0 LIMERICK - UNIT I x1i Amendment No. *3, W4, 146
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS 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
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| - Drywell 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 xiii Amendment No. 2-, 49, 4-QF, 173
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| a INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 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 Diesel Generator Test Schedule .............................. 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 1 xiv Amendment No. 33,49,52,104 DEC 2 0
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PA' ELECTRICAL POWER SYSTEMS (Continued)
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| Table 4.8.2.1-1 Battery Surveillance Requirements .................... 3/4 8-13 D.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 Motor-Operated Valves Thermal Overl oad Protection .... 3/4 8-27 Reactor Protection System Electric Power Monitoring... 3/4 8-28 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH ................................... 3/4 9-1 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 LIMERICK - UNIT I XV Amendment No. 93 Jut 0 7 1995
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 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 DEMONSTRATIONS ........................ 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 ..................................... 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 Liquid Holdup Tanks ................................... 3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS 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. 17, $$, 0, 133 FEB 1 1 1999
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| INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE RADIOACTIVE EFFLUENTS (Continued)
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| Explosive Gas Mixture .................................. 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
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| - " 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 I xvii Amendment No, 48 JAN 6 2 1921
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| INDEX BASES, SECTION PAGE 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 3/4.3.4 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. 7, 2, g9,69 Ap 9?R !lq4
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| INDEX BASES SECTION PAGE 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) ................................................. B 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 Offgas Monitoring Instrumentation ......................... 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. 48, Q, 69, .7-5, 4-04, 7, 4-34, 153
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| INDEX BASES SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY ......................... B 3/4 4-4 3/4.4.6 PRESSURE/TEMPERATURE LIMITS.......................... 8 3/4 4-r Bases Table B 3/4.4.6-1 Reactor Vessel Toughness ................. B 314 4-7 Bases Figure 8 3/4.4.6-1 Fast Neutron Fluence (E>1 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 STRUCTURAL INTEGRITY ............ .................. ,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 and 3/4.5.2 ECCS - OPERATING and SHUTDOW ............ B 3/4 5-1 3/4.5.3 SUPPRESSION CHAMBER. ............................... B 3/4 5-2 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 I MSIV Leakage Control System .................... B 3/4 6-1 Primary Containment Structural Integrity ........... B 3/4 6-2 Drywall and Suppression Chamber Internal Pressure ......................................... B 3/4 6-2 Drywell Average Air Temperature .................... B 3/4 6-2 Drywall and Suppression Chamber Purge System ....... B 3/4 6-2 3/4.6.2 OEPRESSURIZATION SYSTEMS ........................... S 3/4 i-3 LIMERICK - UNIT I xx Amendment No. 3 OCT 3 0 W3U
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| INDEX BASES SECTION 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 7-1a 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM ............ B 3/4 7-1a 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 3/4.9 REFUELING 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 Dlic 2 oU LIMERICK - UNIT 1 xxi Amendment NIo.27,40,52, 104
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| IflDEX BASES 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 3j4.9.9 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 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY....................... B3/4 10-1 3/4.10.2 ROD WORTH MINIMIZER................................ 8B3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS ...................... 3/4 10-1 3/4.10.4 RECIRCULATION LOOPS................................ B3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION............................... 8B3/4 10-1 3/4.10.6 TRAINING STARTUPS.................................. B3/4 10-1 3/4.10.7 RESERVED - CURRENTLY NOT USED....................... B3/4 10-1 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING ................ 3/4 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 3/411-2 Liquid Holdup Tanks ............................ B 3/411-2 3/4.11.2 GASEOUS EFFLUENTS (Deleted)...................................... 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) ...................................... 83/4 11-4 LIMERICK - UNIT 1 xxi i Amendment No.- 17/, 1/i, 133 FEB 1 1 1999
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| INDEX BASES SECTION PAGE RADIOACTIVE EFFLUENTS (Continued)
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| Explosive Gas Mixture .............................. 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 L-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 I xxiii Amendment No.48 JAN 02 1951
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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 5.1.3-1a 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 JAN 02 1991
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| INDEX 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 LIMERICK - UNIT 1 xxv 0 a It
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| INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY ..................................................... 6-1 6.2 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 DELETED ............................................................ 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 LIMERICK - UNIT 1 xxvi Amendment No. , -- , -96, 4-06, 176
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| INDEX ADMINISTRATIVE 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 Semiannual 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 I xxvii Amendment No. 31, 4-, 96, 176
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| bh ADMINISTRATIVE CONTROLS SECTION PAGE 6.23 PROCESS CONTROL PROGRAM (PCP) .............................. 6-21 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) ..................... 6-22 6.15 (Deleted) .................................................. 6-22 I
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| LIMERICK - UNIT I xxvii i Amendment No.48 JAN 02 1991 I
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| SECTION 1.0 DEFINITIONS
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| INTENTIONALLY LEFT BLANK 1.0 DEFINITIONS The 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 remdial measures required under designated conditions.
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| AVERAGE PLAjR 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 nuber 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 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 byI observation. This determination shill include, where possible, comparison of the channel indication and/or status.oAth 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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| -AU& 0tES
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| DEFINITIONS 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 1-131 1.9 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131, microcuries per gram, which alone would produce the same inhalation committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 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 (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.
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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 1-2 Amendment No. ;, 46, 44;, 4-74, 185
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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 +hat the entire response time is measured.
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| 1.13 (Deleted) 1.14 (Deleted)
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| FREOUENCY 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 (HTSP) 1.15a 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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| INTERMEDIATE (POWER) TRIP SETPOINT (ITSP) 1.16a The intermediate power trip setpoint associated with the Rod Block Monitor (RBM) 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 (LHGR) 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. 66 SEB 10 134
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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(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
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| This category does include persons who use portions of the site for recrea-tional, occupational, or other purposes not associated with the plant.
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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 standar-:
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| 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 methodolocy 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 RadiologicaTEnvironmental 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 1 1-4 Amendment No. 4$, 6, 73 AUC 1.0 !994
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| DEFINITIONS 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 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:
| |
| : a. All primary containment penetrations required to be closed during accident conditions are either:
| |
| : 1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or
| |
| : 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except for valves that are opened under administrative control as permitted by Specification 3.6.3.
| |
| b% All primary containment equipment hatches are closed and sealed.
| |
| : c. The primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.
| |
| : d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.
| |
| : e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
| |
| : f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or O-rings, is OPERABLE.
| |
| 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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| OGI 16LG LIMERICK - UNIT I 1-5 Amendment No. 48, 66, 146
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| | |
| QEFINITIONS 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 3458 MWt.
| |
| REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:
| |
| : a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:
| |
| : 1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or
| |
| : 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.
| |
| : b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.
| |
| : c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
| |
| : d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.
| |
| : e. At least one door in each access to the reactor enclosure secondary containment is closed.
| |
| : f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
| |
| : g. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.1a.
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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.
| |
| REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:
| |
| : a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:
| |
| LIMERICK - UNIT I 1-6 Amendment No. 33, &G, 40-5, 06, 185
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| | |
| DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)
| |
| : 1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or
| |
| : 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.
| |
| : b. All refueling floor secondary containment hatches and blowout panels are closed and sealed.
| |
| : c. The standby gas treatment system is in compliance with the requirements of specification 3.6.5.3.
| |
| : d. At least one door in each access to the refueling floor secondary containment is closed.
| |
| : e. The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.
| |
| : f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.
| |
| REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
| |
| ROD DENSITY 1.38 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.
| |
| SHUTDOWN MARGIN 1.39 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68°F; and xenon free.
| |
| SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-1a.
| |
| 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.
| |
| LIMERICK - UNIT I 1-7 Amendment No. 48, 66, 4--5, 185
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| | |
| DEFINITIONS STAGGERED TEST BASIS 1.42 A STAGGERED TEST BASIS shall consist of:
| |
| : a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.
| |
| : b. The testing of one system, subsystem, train, or other designated conent at the beginning of each subinterval.
| |
| THERMAL POWER 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
| |
| 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.
| |
| The response time may be measured by any series of sequential, overlapping, or total steps such that the entire response ttme is measured.
| |
| UNIDENTIFIED LEAKAGE 1.44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
| |
| UNRESTRICTED AREA 1.45 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, comercial, institutional, and/or recreational purposes.
| |
| VENTILATION EXHAUST TREATMENT SYSTEM 1.46 A VENTILATION EXHAUST TREATMENT SYSTEM s hall be any system designed and installed to reduce gaseous radiolodine 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 strean 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.
| |
| VENTING 1.47 IENTING 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 not provided or required during VENTING. Vent, used in system names, does not Imply a VENTING process.
| |
| LIMERICK - UNIT 1 1-8 Amendment No. 52 OCT 24 1991
| |
| | |
| DEFINITIONS TABLE 1.1 SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours.
| |
| D At least once er 24 hours.
| |
| w At least once er 7 days.
| |
| M At least once er 31 days.
| |
| 0 At least once er 92 days.
| |
| SA At least once er 184 days.
| |
| A At least once er 366 days.
| |
| E At least once er 18 months (550 days).
| |
| R (Refueling Interval) At least once er 24 months (731 days).
| |
| S/U Prior to each eactor startup.
| |
| P Prior to each r adioactive release.
| |
| N.A. Not applicable.
| |
| LIMERICK - UNIT 1 1-9 Amendment No. 71
| |
| | |
| 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# *** > 200OF
| |
| : 4. COLD SHUTDOWN Shutdown# ## ***
| |
| * 200°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 I 1-10 Amendment No. 48, 4-3-3, 149
| |
| | |
| SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS MEB 8 198
| |
| | |
| INTENTIONALLY LEFT BLANK 2.0 SAFETY LIMITS AND LIMITING 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 785 psig 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 785 psig 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 1.07 for two recirculation loop operation and shall not be less than 1.09 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
| |
| ACTION:
| |
| With MCPR less than 1.07 for two recirculation loop operation or less than 1.09 for single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig 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.
| |
| 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.
| |
| LIMERICK - UNIT 1 2-1 Amendment No. -, 0, 4-144-2, 4-56, 4-74-0, 183
| |
| | |
| 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: OPERATIONAL CONDITIONS 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.
| |
| 06 LIMERICK-- UNIT 1 2-2 AbQ 08 1QA%
| |
| | |
| 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.
| |
| API 12U LIMERICK - UNIT 1 2-3 Amendment No. 141 S.
| |
| | |
| 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 scale
| |
| : 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
| |
| _ 0.66 W + 62.8% and < 0.66 W + 63.3% and
| |
| < 116.6% of RATED < 117.0% of RATED THERMAL POWER THERMAL POWER
| |
| - Single Recirculation Loop Operation*** < 0.66 (W-7.6%) + 62.8% and _<0.66 (W-7.6%) + 63.3% and
| |
| _<116.6% of RATED !5117.0% of RATED THERMAL POWER THERMAL POWER
| |
| : c. Neutron Flux - Upscale 118.3% of RATED 118.7% of RATED THERMAL POWER THERMAL POWER
| |
| : d. Inoperative N.A. N.A.
| |
| : e. 2-Out-Of-4 Voter N.A. N.A.
| |
| : f. OPRM Upscale N.A.
| |
| : 3. Reactor Vessel Steam Dome Pressure - High 1096 psig ! 1103 psig
| |
| : 4. Reactor Vessel Water Level - Low, Level 3 12.5 inches above instrument 11.0 inches above zero* instrument zero
| |
| : 5. Main Steam Line Isolation Valve - Closure
| |
| * 8% closed 12% closed
| |
| : 6. DELETED DELETED DELETED
| |
| : 7. Drywell Pressure - High 1.68 psig
| |
| * 1.88 psig
| |
| : 8. Scram Discharge Volume Water Level - High
| |
| : a. Level Transmitter 260' 9 5/8" elevation** *261' 5 5/8" elevation
| |
| : b. Float Switch 260' 9 5/8" elevation**
| |
| * 261' 5 5/8" el evati on
| |
| : 9. Turbine Stop Valve - Closure
| |
| * 5% closed *7% closed
| |
| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low > 500 psig > 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 Ž 7.6%. For flows W < 7.6%, the (W-7.6%) term is set equal to zero.
| |
| * See COLR for OPRM period based detection algorithm trip setpoints. OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power Ž 30% and recirculation drive flow < 60%.
| |
| LIMERICK - ' -TT 1 2-4 Amendment No. ;-,-0, &6, 99, -!46, VJ- 177
| |
| | |
| BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
| |
| !W6 81385
| |
| | |
| INTENTIONALLY LEFT BLANK NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.
| |
| JW6 ~8
| |
| | |
| INTENTIONALLY LEFT BLANK 2.1 SAFETY LIMITS BASES
| |
| | |
| ==2.0 INTRODUCTION==
| |
| | |
| The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07 for two recirculation loop operation and 1.09 for single recirculation loop operation. MCPR greater than 1.07 for two recirculation loop operation and 1.09 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incre-mentally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.
| |
| While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a signi-ficant departure from the condition intended by design for planned operation.
| |
| 2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the (GEXL) correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10' lb/h, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 101 lb/h. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thi.s, a THEPMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
| |
| LIMERICK - UNIT I B 2-1 Amendment No. -, 0, -144, -124, 1-56 fCR n00 00209,En q 0r10055, 1-7-0, 183
| |
| | |
| SAFETY LIMITS BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
| |
| The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to I
| |
| calculate critical power. Calculation of the Safety Limit MCPR is described in Reference 1.
| |
| I
| |
| | |
| ==Reference:==
| |
| | |
| I. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (latest approved revision).
| |
| LIMERICK - UNIT I B 2-2 Amendment No.7 II AUGUST 14, 1987
| |
| | |
| 3875101020 LEFT INTENTIONALLY BLANK LIMERICK - UNIT 1 B 2-3 Amendment No. 7 I AUG 14 1987
| |
| | |
| 3875101020 LEFT INTENTIONALLY BLANK UMERICK - IMIT I B 2-4 Aeenanent Ho. 7 AUG 1 4 198?
| |
| | |
| 3901000402 SAFETY LIMITS BASES 2.1.3 REACTOR COOLANT SYSTEM PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code 1968 Edition, including Addenda through Summer 1969, which permits a maximum pres-sure transient of 110%, 1375 psig, of design pressure 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor vessel steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor coolant system is designed to the ASME Boiler and Pressure Vessel Code, 1977 Edition, including Addenda through Summer 1977 for the reactor recirculation piping, which permits a maximum pressure transient I
| |
| of 110%, 1375 psig of design pressure, 1250 psig for suction piping and 1500 psig for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the ASME Boiler and Pressure Vessel Code Section III, Class I.
| |
| 2.1.4 REACTOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is shutdown, consideration must be given to water level requirements due to the effect of decay heat. If thewater level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.
| |
| LIMERICK - UNIT 1 B 2-5 OCT 3 0 1989 Amendment No. 33 I
| |
| | |
| 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their'Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Set point but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
| |
| : 1. Intermediate Range Monitor, Neutron Flux High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.
| |
| The most significant source of reactivity changes during the power increase is due to control rod withdrawal. In order to ensure that the IRM rovides the required protection, a range of rod withdrawal accidents have een analyzed. The results of these analyses are in Section 15.4 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal/gm.
| |
| Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
| |
| : 2. Average Power Range Monitor The APRM system is divided into four APRM channels and four 2-Out-Of -4 Voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. All four voters will trip (full scram) when any two unbypassed APRM channels exceed their trip setpoints.
| |
| APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently from OPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unbypassed APRM channels will result in a full trip in each of the four voter channels. Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full trip from each of the four voter channels.
| |
| For operation at low pressure and low flow during STARTUP, the APRM Neutron Flux-U pscale (Setdown) scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-ture coefficients are small and control rod patterns are constrained by the RWM. Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.
| |
| LIMERICK - UNIT I B 2-6 Amendment No. 4--7, 4-;, 177
| |
| | |
| LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
| |
| Average Power Range Monitor (Continued)
| |
| Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant-amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.
| |
| The 15% Neutron Flux - Upscale (Setdown) trip remains active until the mode switch is placed in the Run position.
| |
| The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux - Upscale setpoint; i.e.,
| |
| for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Simulated Thermal Power - Upscale setpoint, a time constant of 6 +/- 0.6 seconds is introduced into the flow-biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow-biased setpoint as shown in Table 2.2.1-1.
| |
| A reduced Trip Setpoint and Allowable Value is provided for the Simulated Thermal Power - Upscale Function, applicable when the plant is operating in Single Loop Operation (SLO) per LCO 3.4.1.1. In SLO, the drive flow values (W) used in the Trip Setpoint and Allowable Value equations is reduced by 7.6%. The 7.6% value is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop. The Trip Setpoint and Allowable Value thus maintain thermal margins essentially unchanged from those for two-loop operation.
| |
| The Trip Setpoint and Allowable Value equations for single loop operation are only valid for flows down to W = 7.6%. The Trip Setpoint and Allowable Value do not go below 62.8% and 63.3% RATED THERMAL POWER, respectively. This is acceptable because back flow in the inactive recirculation loop is only an issue with drive flows of approximately 40% or greater (Reference 1).
| |
| The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.
| |
| The APRM channels also include an Oscillation Power Range Monitor (OPRM) Upscale Function. The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR Safety Limit due to anticipated thermal-hydraulic power oscillations. The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms.
| |
| References 2, 3 and 4 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.
| |
| LIMERICK - UNIT 1 B 2-7 Amendment No. 66, -4A4, 177
| |
| | |
| LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
| |
| Average Power Range Monitor (Continued)
| |
| The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is Ž 30% and recirculation drive flow is < 60%
| |
| as indicated by APRM measured recirculation drive flow. (NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow. Refer to TS Bases 3/4.3.1 for further discussion concerning the recirculation drive flow/core flow relationship.) This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. See Reference 5 for additional discussion of OPRM Upscale trip enable region limits. These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Upscale trip enabled region. The APRM Simulated Thermal Power auto-enable setpoint has 1% deadband while the drive flow setpoint has a 2% deadband. The deadband for these setpoints is established so that it 'increases the enabled region.
| |
| An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints.
| |
| One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel.
| |
| There are four "sets" of OPRM related setpoints or adjustment parameters:
| |
| a) OPRM trip auto-enable setpoints for APRM Simulated Thermal Power (30%) and recirculation drive flow (60%); b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude based algorithm (ABA) setpoints.
| |
| The first set, the OPRM auto-enable region setpoints, are treated as nominal setpoints with no additional margins added as discussed in Reference 5.
| |
| The settings, 30% APRM Simulated Thermal Power and 60% recirculation drive flow, are defined (limit values) in a note to Table 2.2.1-1. The second set, the OPRM PBDA trip setpoints, are established in accordance with methodologies defined in Reference 4, and are documented in the COLR. There are no allowable values for these setpoints. The third set, the OPRM PBDA "tuning" parameters, are established or adjusted in accordance with and controlled by station procedures.
| |
| The fourth set, the GRA and ABA setpoints, in accordance with References 2 and 3, are established as nominal values only, and controlled by station procedures.
| |
| : 3. Reactor Vessel Steam Dome Pressure-Hiqh High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power/flow conditions when the turbine stop valve and control fast closure trips are bypassed. For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
| |
| LIMERICK - UNIT 1 B 2-7a Amendment No. 66, *A1-], 177 1
| |
| | |
| INTENTIONALLY LEFT BLANK LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
| |
| : 4. Reactor Vessel Water Level-Low -
| |
| The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease. The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
| |
| : 5. Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIVs are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature, and low steam line pressure. I The MSIVs closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal/hydraulic Safety Limits.
| |
| : 6. DELETED I
| |
| : 7. Drywell Pressure-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment. The trip setting was selected as low as possible without causing spurious trips.
| |
| FED 1 6 W LIMERICK - WIT 1 B 2-8 Amendment No. 89
| |
| | |
| LIMITING SAFETY SYSTEM SETTING BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
| |
| : 8. Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water
| |
| -at pressures below 65 psig, control rod insertion would be hindered.. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are. tripped. The trip setpoint for each scram discharge volume is equivalent toa contained volume of 25.45 gallons of water.
| |
| : 9. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With.'a trip setting of 5% of valve closure from full open, the resultant increase in heat flux is such that adequatethermal margins are maintained during the worst design basis transient.
| |
| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low The turbine control valve fast closure-trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection with or without coincident fail-ure of. the turbine bypass valves. The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the Reactor Protection System. This trip setting, a faster closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve. Relevant transient analyses are discussed 'in Section 15.2.2 of the Final Safety Analysis Report.
| |
| : 11. Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.
| |
| : 12. Manual Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
| |
| LIMERICK - UNIT 1 B 2-9
| |
| | |
| LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
| |
| | |
| ==REFERENCES:==
| |
| : 1. NEDC-31300, "Single-Loop Operation Analysis for Limerick Generating Station, Unit 1," August 1986.
| |
| : 2. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
| |
| : 3. NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
| |
| : 4. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
| |
| : 5. BWROG Letter 96113, K. P. Donovan (BWROG) to L. E. Phillips (NRC),
| |
| "Guidelines for Stability Option III 'Enable Region' (TAC M92882),"
| |
| September 17, 1996.
| |
| LIMERICK - UNIT 1 B 2-10 Amendment No. 177 1
| |
| | |
| SECTIONS 3.0 and 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Avg SES
| |
| | |
| PAGE INTENTIONALLY LEFT BLANK 3/4.0 APPLICABILITY LIMITING CONDITION FOR 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.
| |
| 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. 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 Specifications.
| |
| 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 I 3/4 0-1 Amendment No. 4-4, 169
| |
| | |
| APPLICABILITY SURVEILLANCE 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. 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 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, & 3 components shall be applicable as follows:
| |
| : a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a.
| |
| : b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
| |
| LIMERICK - UNIT I 3/4 0-2 Amendment No. 4*, -38, 42-B, 4-62, 169
| |
| | |
| APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)
| |
| ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days
| |
| : c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.
| |
| : d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
| |
| : e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
| |
| : f. The Inservice Inspecti-on ISI) Program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule, methods and personnel, and sample expansion included in the Generic Letter, or in accordance with alternate measures approved by the NRC staff. Details for implementation of these requirements are included as augmented inspection requirements in the ISI Program.
| |
| LIMERICK - UNIT 1 3/4 0-3 Amendment No. 14, 4-9, 4-2-5, 171
| |
| | |
| INTENTIONALLY LEFT BLANK 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% Ak/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.
| |
| LIMERICK - UNIT 1 3/4 1-1
| |
| | |
| REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall not exceed 1% Ak/k.
| |
| APPLICABILITY: OPERATIONAL CONDITION 1 and 2.
| |
| ACTION:
| |
| With the reactivity equivalence difference exceeding 1% Ak/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.
| |
| SURVEILLANCE REQUIREMENTS 4.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted ROD DENSITY shall be verified to be less than or equal to 1% Ak/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. 29
| |
| | |
| 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 I from all other inoperable withdrawn control rods by at least I 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 1 3/4 1-3 Amendment No. 4-68, 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 REQUIREMENTS 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:
| |
| : a. Verifying each valve to be open,* and
| |
| : b. Cycling each valve through at least one complete cycle of full travel.
| |
| * These valves may be closed intermittently 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 1 3/4 1-4 Amendment No. 14, 4-68, 4-6-9,186
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (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 aftl6r 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 1 3/4 1-5 Amendment No. --0, Z-7, -98, 41-6, 1-7-,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 I 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:
| |
| 1.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. 9, 4-6-9186
| |
| | |
| 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 7Inser-Fully Withdrawn tion Time (Seconds) 45 0.43 39
| |
| * 0.86 25 1.93 05 3.49 APPLICABILITY: OPERATIONAL CONDITIONS I 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 I 3/4 1-7 JW6 B LM
| |
| | |
| REACTIVITY CONTROL SYSTEMS FOHR CONTRnl ROD GROI1P SWRAM INSFRTION TIMFS 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.
| |
| SURVEILLANCE REOUIREMENTS 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 a1400 psig or by inserting at least one withdrawn control rod at least one notch. If no control rod drive pump is operating and:
| |
| I) If reactor pressure is A900 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:
| |
| 1I Electrically, or 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:
| |
| al 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 2 2 Z200
| |
| | |
| 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.
| |
| LIMERICK - UNIT I 3/4 1-10 Amendment No. 3-4, 39, 7-7, 4-43, 4-&9, 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 except as in 3.1.3.6.c or 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 directional control valves** either:
| |
| a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.
| |
| : c. For control rod 50-27, for the remainder of Unit 1 Cycle 7, if coup-ling can not be established the uncoupled rod may be withdrawn when rated thermal power exceeds 10% only if all the following conditions are satisfied:
| |
| : 1) The uncoupled control rod may not be withdrawn past notch position 46, and
| |
| : 2) No other uncoupled control rod is withdrawn.
| |
| * 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 1-11 Amendment No. 44, 4-24, 169
| |
| | |
| REACTIVITY CONTROL SYSTEMS 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, I C. Following maintenance on or modification to the control rod or control rod drive system which could have affected the control rod drive coupling integrity, and
| |
| : d. When repositioning the uncoupled control rod per Specification 3.1-3.6.d the uncoupled rod's position shall be verified to have followed the control rod drive by neutron instrumentation (LPRII or TIP). If the control blade can not be verified to have followed the drive out to its final position, then the rod shall be completely inserted and the control rod directional valves disarmed as stated in 3.1.3.6-.a.2.
| |
| JAN 1 6 1 LIMERICK - UNIT I 3/4 1-12 Aamxbmt No. 12A
| |
| | |
| 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 I 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
| |
| | |
| ,D0,7jU0j OJU REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE HOUSING SUPPORT LIM"TING CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be in place.
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
| |
| ACTION:
| |
| With the control rod drive housing support not in place, be in at least NOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 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.
| |
| LIMERICK - UNIT 1 3/4 1-15 AUG 8 1985
| |
| | |
| 3893087830 REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS 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 withdraVial 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. 11, 17 MAR 2 2 15g9 I
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.4.1 The 4%M 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 RWN automatic initia-tion when reducing T._'-RMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.
| |
| : b. In OPERATIONAL 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-sequence control 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 control 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 LIMERICK - UNIT I 3/4 1-17 Amendment No. 17 MAR 22 1989
| |
| | |
| RFACTIVITY 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 REOUIREMENTS 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. 46, 186
| |
| | |
| REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE and consist of a minimum of the following:
| |
| : a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,
| |
| : b. In OPERATIONAL CONDITION 3, 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.
| |
| SURVEILLANCE REQUIREMENTS 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 pentaborate 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 I 3/4 1-19 Amendment No. 9,66,&,U5, 186
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Verifying the continuity of the explosive charge.
| |
| : 2. Determining by chemical analysis and calculation* that the available weight of Boron-lO is greater than or equal to 185 Ibs; 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:
| |
| C x E x 1. Ž1 13% wt. 29 atom % 86 gpm where C = Sodium pentaborate 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.
| |
| : 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 pentaborate 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. -9,6-1,f4,-,4-O6,1-5, 186
| |
| | |
| 3890020620 80 13.8 75 I
| |
| 70 OOPERATING RANGEL Li~
| |
| C a
| |
| La
| |
| =
| |
| 65 C
| |
| 1, La C-S La I-60 m OPERATING LIMIT-r l I I I
| |
| * F I I ei*. ,lII'.
| |
| 55 50 I I1 11 12 13 14 15 CONCENTRATION, % BY WEIGHT SODIUM PENTABORATE SOLUTION TEMPERATURE/CONCENTRATION REQUIREMENTS FIGURE 3.1.5-1 LIMERICK - UNIT 1 3/4 1-21 JUN 8 19f Amendment No. 22 I
| |
| | |
| 3890020620 THIS PAGE LEFT BLANK INTENTIONALLY LIMERICK - UNIT I 3/4 1-22 Amendment No. 22 JUN 8198
| |
| | |
| 3/42 POWER flISTRIBUTION 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: OPERATIONA'L 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 value
| |
| : 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.
| |
| I
| |
| : d. The provisions of Specification 4.0.4 are not applicable.
| |
| LIMERICK - UNIT I 3/4 2-1 Amendment No. -, 30, 66, 6 186
| |
| | |
| Figures on pages 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 ONITTED.
| |
| LIMERICK - UNIT 1 3/4 2-2 Amendmnt No. 37 MAY is w I
| |
| | |
| POWER DISTRIBUTION LIMITS Section 3/4.2.2 (DELETED)
| |
| INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED Amendment No. 7, 30, 66 FEB 10 1994 LIMERICK - UNIT 3/4 2-7
| |
| -
| |
| | |
| POWER DISTRIBUTION LIMIIS 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:
| |
| T= (rave TA -B) 7B where:
| |
| TA - 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, TB = 0.672 + 1.65 N1 I n t Ni
| |
| )1/2(o.016),
| |
| i-I n ]*Niri 7ave- i-n I Ni jil!
| |
| n - number of surveillance tests performed to date in cycle, Ni - number of active control rods measured in the V"tsurveillance test, Ti - average scram time to notch 39 of all rods measured in the ih surveillance test, and N, - 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.
| |
| LIMERICK - UNIT 1 3/4 2-8 Amendment No. 17, V, $2, 66 FEB 10 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, operationmay 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 I 3/4 279 Amendment No. 44, 4-9, 3-7, -, 66, , 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 UNITS REPORT.
| |
| Technical Specifications pages 3/4 2-20a thru 3/4 2-21 have been INTENTIONALLY OHITTM.
| |
| LIPERICK; UNIT 1 3/4 2-10 Amendwnt No. 2. 1,, $1, 37 vIBv 15 IM0 I
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| | |
| POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITIONFOR 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.
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| APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
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| ACTION:
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| 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.
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| SURVEILLANCE REQUIREMENTS 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
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| : 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.
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| I
| |
| : d. The provisions of Specification 4.0.4 are not applicable.
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| LIMERICK - UNIT 1 3/4 2- 12 Amendment No. ;, -, 186
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| 3901038720 THIS PAGE INTENTIONALLY LEFT BLANK
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| | |
| 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.
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| APPLICABILITY: As shown in Table 3.3.1-1.
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| ACTION:
| |
| Note: Separate condition entry is allowed for each channel.-
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| : 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.
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| : b. With the number of OPERABLE channels in either tip 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.***
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| : 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.**.***
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| : d. If 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.
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| *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.
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| **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.
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| ***A channel or trip system which has been placed in the tripped condition to satisfy Action b. or c. may be returned to the untripped condition under administrative control for up to two hours solely to perform testing required to demonstrate its operability or the operability of other equipment provided Action a. continues to be satisfied.
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| LIMERICK - UNIT 1 3/4 3-1 Amendment No. 63, .7;4, 1, 177
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| 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 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.
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| 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.
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| 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.
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| LIMERICK - UNIT 1 3/4 3-1a Amendment No. 4,4-7-7, 186
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| INTENTIONALLY LEFT BLANK TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION
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| : 1. Intermediate Range Monitorstb):
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| : a. Neutron Flux - High 2 3 1 3(i), 4(i) 3 2 5(i) 3(d) 3
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| : b. Inoperative 2 3 1 3(0), 4(i) 3 2 5(i) 3(d) 3
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| : 2. Average Power Range Monitor(e':
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| : a. Neutron Flux - Upscale (Setdown) 2 3(m) 1
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| : b. Simulated Thermal Power - Upscale 1 3(m) 4
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| : c. Neutron Flux - Upscale 1 3(m) 4
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| : d. Inoperative 1, 2 3(m) 1
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| : e. 2-Out-Of-4 Voter 1, 2 2 1
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| : f. OPRM Upscale 1(o)(p) 3(m) 10
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| : 3. Reactor Vessel Steam Dome Pressure - High 1, 2(f) 2 1
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| : 4. Reactor Vessel Water Level - Low, Level 3 1, 2 2 1
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| : 5. Main Steam Line Isolation Valve-Closure 1(g) 1/valve 4 LIMERICK - UNIT 1 3/4 3-2 Amendment No. 2-8, 4-4, --44, 44-9, 177
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| | |
| TABLL ..,.,-1 (Continued)
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| REACTOR PROTECTION SYSTEM INWTRIJMFNTATION APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION
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| : 6. DELETED DELETED DELETED DELETED
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| : 7. Drywel l Pressure - High 1, 2(h) 2 I
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| : 8. Scram Discharge Volume Water Level - High
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| : a. Level Transmitter 1, 2 2 1 5(i) 2 3
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| : b. Float Switch 1, 2 2 1 5(i) 2 3
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| : 9. Turbine Stop Valve - Closure 1(j) 4(k) 6
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| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 1(j) 2(k) 6
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| : 11. Reactor Mode Switch Shutdown Position 1, 2 2 1 3, 4 2 7 5 2 3
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| : 12. Manual Scram 1, 2 2 1 3, 4 2 8 5 2 9 LIMERICK UNIT I 3/4 3-3 Amendment No. 89
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| | |
| TABLE 3.3.1-1 (Continued)
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| REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 Be in at least HOT SHUTDOWN within 12 hours.
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| ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the SHUTDOWN position within 1 hour.
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| ACTION 3 Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within 1 hour.
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| ACTION 4 Be in at least STARTUP within 6 hours.
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| 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.
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| 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.
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| ACTION 7 - Verify all insertable control rods to be inserted within 1 hour.
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| ACTION 8 - Lock the reactor mode switch in the Shutdown position within 1 hour.
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| 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.
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| ACTION 10 - a. If 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, OR
| |
| : b. Reduce THERMAL POWER to < 25% RATED THERMAL POWER within 4 hours.
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| Unanticipated characteristic of the instability detection algorithm or equipment that renders all OPRM channels inoperable at once.
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| LIMERICK - UNIT I 3/4 3-4 Amendment No. -44, 4A19, 177
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| | |
| TABLE 3.3.1-1 (Continued)
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| 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.
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| (b) This function shall be automatically bypassed when the reactor mode switch is in the Run position.
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| (c) DELETED (d) The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems. There'fore, 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.
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| (g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
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| (h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.
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| (i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
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| (j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.
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| Ck) Also actuates the EOC-RPT system.
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| (1) DELETED (m) Each APRM channel provides inputs to both trip systems.
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| (n) DELETED (o) With THERMAL POWER e 25% RATED THERMAL POWER. The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is Ž 30% and recirculation drive flow is < 60%. The OPRM trip output may be automatically bypassed when APRM Simulated Thermal Power is < 30% or recirculation drive flow is Ž 60%.
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| (p) A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for an OPRM channel to be OPERABLE.
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| LIMERICK - UNIT 1 3/4 3-5 Amendment No. 44-, -53, 444r, 177
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| | |
| TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT (Seconds)
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| : 1. Intermediate Range Monitors:
| |
| : a. Neutron Flux - High N.A.
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| : b. Inoperative N.A.
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| : 2. Average Power Range Monitor*:
| |
| : a. Neutron Flux - Upscale (Setdown) N.A.
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| : b. Simulated Thermal Power - Upscale N.A.
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| : c. Neutron Flux - Upscale N.A.
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| : d. Inoperative N.A.
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| : e. 2-Out-Of-4 Voter *0.05*
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| : 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
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| : a. Level Transmitter N.A.
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| : b. Float Switch N.A.
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| : 9. Turbine Stop Valve - Closure *0.06
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| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 0.08**
| |
| : 11. Reactor Mode Switch Shutdown Position N.A.
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| : 12. Manual Scram N.A.
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| * 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 APRM outputs are considered to be separate channels, so N = 8. Testing of OPRM and APRM outputs shall alternate.
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| ** Measured from start of turbine control valve fast closure.
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| # 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.
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| LIMERIr'' . UNIT I 3/4 3-r Amendment No. 89, 4-39, 4-4-, 17'
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| | |
| TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK(n) TEST(n) CALIBRATION(a)(n) SURVEILLANCE REQUIRED
| |
| : 1. Intermediate Range Mon itors:
| |
| : a. Neutron Flux - High (b) 2 3(0) 40i) 5(i)
| |
| (j)
| |
| : b. Inoperative N.A. (j) N.A. 2, 3(i), 4(0), 5(0)
| |
| : 2. Average Power Range Monitor(f):
| |
| : a. Neutron Flux - Upscale (Setdown) (b) (1) 2
| |
| : b. Simulated Thermal Power - Upscale (e) (d), (g) 1
| |
| : c. Neutron Flux - Upscale (d)
| |
| : d. Inoperative N.A. N.A. 1, 2
| |
| : e. 2-Out-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. Drywell Pressure - High 1, 2
| |
| : 8. Scram Discharge Volume Water Level - High
| |
| : a. Level Transmitter 1, 2, 5Ci)
| |
| : b. Float Switch N.A. 1, 2, 5Ci)
| |
| LIMERICK - UNIT 1 3/4 3-7 Amendment No. 4-1, 8-3, 89, 64, --7-,
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| 186
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| | |
| TABLE 4.3.1.1-1 (Continued)
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| REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REGUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK(n) TEST(n) CALIBRATION(a)(n) SURVEILLANCE REQUIRED
| |
| : 9. Turbine Stop Valve - Closure N.A.
| |
| : 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A.
| |
| : 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 be 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 days.
| |
| (c) Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRM Simulated Thermal Power is > 30% 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.
| |
| Adjust 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.
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| (f) The LPRMs shall be calibrated at least once per 1000 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.
| |
| Wk) DELETED (I) 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. 2-9, 4-i, 3, 6, 4-143, 4-1-7, 4-34-, -44, -7,186
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| INSTRUMENTATION 3/4.3.2. ISOLATION ACTUATION INSTRUMENTATION LIMITING 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.
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| APPLICABILITY: 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.
| |
| Trip functions common to RPS Actuation Instrumentation are shown in Table 4.3.2.1-1.
| |
| LIMERICK - UNIT I 3/4 3-9 Amendment No. 3, 69, 169
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| | |
| 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. -, ;4-,186
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| | |
| TABLE 3.3.2-1 I-
| |
| -I ni ISOLATION ACTUATION INSTRUMENTATION
| |
| '-4 MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL C TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION
| |
| -4 1. MAIN STEAM LINE ISOLATION I-.
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low-Level 2 B 2 1, 2, 3 21
| |
| : 2) Low, Low, Low-Level I C 2 1, 2, 3 21
| |
| : b. DELETED DELETED DELETED DELETED DELETED
| |
| : c. Main Steam Line Pressure - Low P 2 I 22
| |
| : d. Main Steam Line 4:h Flow - High E 2/1 ine 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
| |
| : g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High F(f) 14 1, 2, 3 21 CO-rt h. Manual Initiation NA 2 1, 2,3 24 "I 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
| |
| | |
| TABLE 3.3.2-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION I- MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION C=
| |
| : 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. RWCS A Flow - High 1 is 2, 3 23'
| |
| : b. RWCS Area Temperature - High 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
| |
| : e. Reactor Vessel Water Level -
| |
| Low, Low - Level 2 B 2 1, 2, 3 23 Manual Initiation NA 1 1, 2, 3
| |
| : f. 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 i 1, 2, 3 23 Pressure - High L 2
| |
| : 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,
| |
| w
| |
| | |
| TABLE 3.3., ?-I (Continued)
| |
| ISOLATWFTUAM N INSTRUMENTATION I-ISOLATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION 4. SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACT ION
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION (Continued)
| |
| : f. HPCI Pipe Routing Area Temperature - High L 4 1, 2, 3 23
| |
| : g. Manual Initiation NA(e) 1/system 1, 2, 3 24
| |
| : h. HPCI Steam Line A Press Timer HA 1 1, 2, 3 23
| |
| : 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION I-
| |
| : a. RCIC Steam Line A Pressure - High K 1 1, 2, 3 23 I
| |
| : b. RCIC Steam Supply Pressure - Low KA 2 is 2, 3 23 02
| |
| : c. RCIC Turbine Exhaust Diaphragm Pressure - High K 2 1, 2, 3 .23
| |
| : d. RCIC Equipment Room 1 Temperature - High K 1, 2, 3 23
| |
| : e. RCIC Equipment Room 1 A Temperature - High K 1, 2, 3 23-
| |
| : f. RCIC Pipe Routing Area
| |
| * g Temperature - High K 5 1, 2, 3 23 I
| |
| NA(e) I/system is*2, 3 24
| |
| : g. Manual Initiation'
| |
| : h. RCIC Steam Line 23 A Pressure Timer NA 1 1, 2. 3
| |
| | |
| f-TABLE (Continued) m ISOLATION ACTUATION INSTRUMENTATION MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL STRIP FUNCTION PER TRIP SYSTEM 1b) CONDITION SIGNAL'.@ ACTION
| |
| -4 6. PRIMARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low - Level 2 B 2 3 20 1, 2, 1,
| |
| 2, 3 20
| |
| : 2) Low, Low, Low - Level 1 C 2
| |
| : b. Drywell Pressure - High H 2 1, 2, 3 20
| |
| : c. North Stack Effluent Radiation - High (' W I 1, 2, 3 23
| |
| : d. Deleted
| |
| : e. Reactor Enclosure Ventilation Exhaust Duct-Radiation - High S 2 1, 2, 3 23
| |
| : f. Deleted I
| |
| : g. Deleted
| |
| : h. Drywell Pressure - High/
| |
| Reactor Pressure - Low G 2/2 1, 2, 3 26 C+Cc i. Primary Containment Instrument M 1 1, 2, 3 26 Gas Line to Drywell A Pressure - Low NJ J. Manual Initiation NA I 1, 2, 3 24
| |
| | |
| I- TAjBLE 3.3.2-1 (Continued)
| |
| ISOLATICON ACTUATION INSTRUMENTATION
| |
| ~ISOLATIO MINIMUM' APPLICABLE OPERABLE OPERATIONAL TRIP FUNCTION IN SIGNAL'*' .,(0) PER TRIP CHANNELS SYSTEM (b)
| |
| CONDITION
| |
| -' 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 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.
| |
| f.
| |
| Deleted Deleted I
| |
| : g. Reactor Enclosure Manual Initiation NA 1, 2, 3 24
| |
| : h. Refueling Area Manual Initiation NA 25 r"
| |
| cc ba r\3
| |
| | |
| 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 hoursand 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 (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
| |
| ** 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 I 3/4 3-16 Amendment No. 2-3, 40, -53, W8, 4.46, 185
| |
| | |
| TABLE 3.3.2-] (Continued)
| |
| TABLE NOTATIONS Signals B, H, S, and R also (c) Actuates secondary containment isolation valves.
| |
| 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 50'F 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.
| |
| OC= 18 =
| |
| LIMERICK - UNIT 1 3/4 3-17 Amendment a8, r&,4-I-, 146
| |
| | |
| I TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUNENTAT[ON SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT -VALUE 1.. MAIN STEAM LINE ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low - Level 2 - 38 Inches* - 45 inches
| |
| : 2) Low, Low, Low - Level I - 129 inches* - 136 inches
| |
| : b. DELETED DELETED DELETED
| |
| : c. Main Steam Line Pressure - Low 756 pstg k 736 psig
| |
| : d. Main Steam Line 122.1 pstd 5 123 psid Flow - High 10.5 psia :10.1 psia/: 10.9 psia I
| |
| : e. Condenser Vacuum - Low
| |
| : f. Outboard MSIV Room Temperature - High :5192*F 200OF
| |
| : g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High 5 165OF S175OF
| |
| : h. Manual Initiation N.A. N.A.
| |
| : 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level Low - Level 3 12.5 inches* 11.0 Inches
| |
| : b. Reactor Vessel (RHR Cut-in Permissive) Pressure - High 75 psig S 95 psig N.A. N.A.
| |
| : c. Manual Initiation
| |
| | |
| TABLE 3.3.2-2 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
| |
| : 3. REACTOR WATER CLEANUP.SYSTEM ISOLATION
| |
| : a. RWCS A Flow - High < 54.9 gpm < 65.2 gpm
| |
| : b. RWCS Area Temperature - High < 155 0 F or
| |
| * 1320F** < 160°F or _<137 0 F**
| |
| : c. RWCS Area Ventilation A Temperature - High < 520 F or _<320 F** < 600F or < 40 0 F**
| |
| : d. SLCS Initiation N.A. N.A.
| |
| : e. Reactor Vessel Water Level -
| |
| Low, Low, - Level 2 > -38 inches * > -45 inches
| |
| : f. Manual Initiation N.A. N.A.
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
| |
| : a. HPCI Steam Line A Pressure - High < 974" H20 < 984" H20
| |
| : b. HPCI Steam Supply Pressure - Low >_100 psig > 90 psig
| |
| : c. HPCI Turbine Exhaust Diaphragm Pressure - High < 10 psig _ 20 psig
| |
| : d. HPCI Equipment Room Temperature - High 2250F > 2180F, < 247oF
| |
| : e. HPCI Equipment Room A Temperature - High
| |
| * 126 0 F < 130.5°F
| |
| : f. HPCI Pipe Routing Area Temperature - High 1750F > 165 0 F, 200OF gý. Manual Initiation N.A. N.A.
| |
| : h. HPCI Steam Line A Pressure - Timer 3 < T < 12.5 seconds 2.5 < T < 13 seconds LIMERICK - UNIT I 3/4 3-19 Amendment No. 3-3, 8-, 4-0-6,161
| |
| | |
| TABLE 3.3.2-2 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT - ALUE
| |
| : 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
| |
| : a. RCIC Steam Line A Pressure - High
| |
| : b. RCIC Steam Supply Pressure - Low 313" 1120
| |
| Ž 64.5 psig 381" H2 0 I
| |
| > 56.5 psig
| |
| : c. RCIC Turbine Exhaust Diaphragm Pressure - High : 10.0 psig : 20.0 psig
| |
| : d. RCIC Equipment Room Temperature - High 2050F >-198 0F, :5 2270F
| |
| : e. RCIC Equipment Room A Temperature - High 109 0 F 113.5 0 F CAJ
| |
| : f. RCIC Pipe Routing Area Temperature - High 175°F > 165 0 F, < 200OF 0
| |
| : g. Manual Initiation N.A. N.A.
| |
| : h. RCIC Steam Line A Pressure Timer 3 5 r - 12.5 seconds 2.5 < T < 13 seconds 0.
| |
| C+
| |
| CcI
| |
| | |
| r'-
| |
| TABLE 3.3.2-2 (Continued) rrl ISOLATION ACTUATION INSTRUMENTATION SETPOINTS TRIP FUNCTION ALLOWABLE TRIP SETPOINT VALUE S6. PRIMARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1. Low, Low - Level 2 -38 inches* -45 inches
| |
| : 2. Low, Low, Low - Level I a -129 inches* -136 inches
| |
| : b. Drywell Pressure - High 5 1.68 pslg 1.88 psig
| |
| : c. North Stack Effluent Radiation - High 5 2.1 pCi/cc : 4.0 pCI/cc
| |
| : d. Deleted
| |
| , e. Reactor Enclosure Ventilation Exhaust Duct - Radiation - High 1.35 1 mR/h 51.5 mR/h
| |
| : f. Deleted I
| |
| : g. Deleted M h. Drywell Pressure - High/ < 1.68 psig/ < 1.88 psig/
| |
| Reactor Pressure - Low > 455 psig (decreasing) a 435 psig (decreasing)
| |
| + . Primary Containment Instrument a 2.0 psi > 1.9 psi Gas to Drywell A Pressure - Low U~1 J. Manual Initiation N.A. N.A.
| |
| -a
| |
| '-a
| |
| | |
| r-4 Mr JIAiLE.3.3.21 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SETPOINTS TRIP FUNCTION ALLOWABLE TR~IP SPINTf VALUE
| |
| " 7. SECONDARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level -
| |
| Low, Low - Level 2 -38 inches* -45 inches
| |
| : b. Drywell Pressure - High : 1.68 psig s 1.88 psig c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High : 2.0 mR/h 5 2.2 mR/h
| |
| : 2. Refueling Area Unit 2 Ventilation 4%- ExhaustOuct Radiation - High : 2.0 mR/h s 2.2 mR/h
| |
| : d. Reactor Enclosure Ventilation Exhaust N Duct Radiation -High : 1.35 mR/h s 1.5 mR/h
| |
| : e. Deleted
| |
| : f. Deleted
| |
| : g. Reactor Enclosure Manual Initiation N.A. N.A.
| |
| 1r+
| |
| 0= h. Refueling Area Manual Initiation N.A. N.A.
| |
| * 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.
| |
| | |
| 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 I <1 .0###* I
| |
| : b. DELETED DELETED
| |
| : c. Main Steam Line Pressure - Low <1.0###*
| |
| : d. Main Steam Line Flow - High _<0.5###*
| |
| : 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. I
| |
| : b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High N.A.
| |
| : c. Manual Initiation N.A.
| |
| : 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. RWCS
| |
| * Flow - High N.A.##
| |
| : b. RWCS Area Temperature - High N.A.
| |
| : c. RWCS Area Ventilation 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.
| |
| JAN 0 7 1990 Amendment No. 29,89,132 LIMERICK - UNIT I 3/4 3-23
| |
| | |
| 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
| |
| & 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 A Pressure - High N.A. II
| |
| : 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 A Temperature - High N.A.
| |
| : f. RCIC Pipe Routing Area Temperature - High N.A.
| |
| : g. Manual Initiation N.A.
| |
| JAN 0 7 1999 LIMERICK - UNIT I 3/4 3-24 Amendment pb. 33-,132
| |
| | |
| TABLE 3.3.2-3 (Continued)
| |
| ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME 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.
| |
| : 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 & 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 JAN 0 71999 LIMERICK - UNIT 1 3/4 3-25 Amendment No. -,:12-,132
| |
| | |
| 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 I for the valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME 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.
| |
| OCT 18 MwxO LIMERICK - UNIT I 314 3-26 Amendment No. 6, 89, -I-2, I2n, 146
| |
| | |
| TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHE CK.(.). TESTT CALI BRAT IONL-a-). SURVEILLANCE REDUIRED
| |
| : 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
| |
| : 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
| |
| : a. Reactor Vessel Water Level##
| |
| Low - Level 3 1, 2, 3
| |
| : b. Reactor Vessel (RHR Cut-In 1, 2, 3 Permissive) Pressure - High
| |
| : c. Manual Initiation N.A. N.A. 1, 2, 3 LIMERICK - UNIT 1 3/4 3-27 Amendment No. 298, 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(._ TEST .T_ CALIBRAT I ON(.al. SURVEILLANCE REQUIREO
| |
| : 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
| |
| : a. RWCS A Flow - High 1, 2, 3
| |
| : 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 I
| |
| : 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 Area 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. --3, 69, 186
| |
| | |
| TABLE 4.3.2.1-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RmFIIlREMFNTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK (a) TEST (a) CALIBRATION(a) SURVEILLANCE REQUIRED
| |
| : 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 I 3/4 3-29 Amendment No. *-3, ý&9,186
| |
| | |
| TABLE 4.3.2.1-1 (Continued)
| |
| ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK.C.a.X TEST(.A.)- CALIBRATION.(.0). SURVEILLANCE REOUIRED
| |
| : 6. PRIMARY CONTAINMENT ISOLATION
| |
| : a. Reactor Vessel Water Level
| |
| : 1) Low, Low - Level 2 1, 2, 3
| |
| : 2) Low, Low, Low - Level 1 1, 2, 3
| |
| : b. Drywell Pressure## - 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
| |
| : 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
| |
| : j. Manual Initiation N.A. N.A. 1, 2, 3 LIMERICK - UNIT 1 3/4 3-30 Amendment No. 6, 49, 64, 4--2*,
| |
| 186
| |
| | |
| TABLE 4.3.2.1-1 (Continued)
| |
| ISOLATION ACTUATION INSTRIIMFNTATInN S11RVFIll, ANrF 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. Drywell 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 (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
| |
| **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. 2-3, 40, -53, &P, 9, -*-i*, 49, 186
| |
| | |
| 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 I 3/4 3-32 Amendment No. -- , 186
| |
| | |
| 3901000402 TABLE 3.3.3-1 I-EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRI P ( a) OPERATIONAL z TRIP FUNCTION FUNCTION(a) CONDITIONS ACTION
| |
| : 1. CORE SPRAY SYSTEM***
| |
| : a. 2/pump (b)
| |
| Reactor Vessel Water Level - Low Low Low, Level 1 2/pump'-" 1, 2, 3, 4*, 5* 30
| |
| : b. Drywell Pressure - High 1, 2, 3 30 6(b) 1, 2, 3 31
| |
| : c. Reactor Vessel Pressure - Low (Permissive) 4*, 5* 32
| |
| : d. Manual Initiation 1, 2, 3, 4*, 5* 33 UJ 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM***
| |
| 11.a Io
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1, 21 3,, 4* 5* 30 Ia
| |
| : b. Drywell Pressure - High 2 1, 2, 3 io 2 1, 2, 3 31 C. Reactor Vessel Pressure - Low (Permissive) 1, 2, 3, 4*, 5* 31
| |
| : d. Injection Valve Differential Pressure-Low 1/valve (Permissive) 33
| |
| : e. Manual Initiation I 1, 2, 3, 4*, 5*
| |
| : 3. HIGH PRESSURE COOLANT INJECTION SYSTEMO 34 Reactor Vessel Water Level - Low Low Level 2 4 1, 2, 3 1, 2, am
| |
| : a. 21 3 34
| |
| : b. Drywell Pressure - High 4(c) 3 35 C. Condensate Storage Tank Level - Low 1, 2, 2 35
| |
| : d. Suppression Pool Water Level - High 2(d) 1, 2, 3 31 1, 2, 3
| |
| : e. Reactor Vessel Water Level High, Level 8 2, 3 33
| |
| : f. Manual Initiation I/system C-_
| |
| | |
| 3901000402 TABLE 3.3.3-1 (Continued)
| |
| I--
| |
| EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER APPLICABLE TRIP OPERATIONAL
| |
| --4 TRIP FUNCTION FUNCTION(a) CONDITIONS ACTION W- 4. AUTOMATIC DEPRESSURIZATION SYSTEMIq**
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1, 2, 3 30
| |
| : b. Drywell Pressure - High 2 1, 2, 3 30 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
| |
| : g. Manual Initiation 2 1, 2, 3 33
| |
| 'A'
| |
| : h. ADS Drywell Pressure Bypass Timer 2 1, 2, 3 31 IA MINIMUM APPLICABLE TOTAL NO. CHANNELS CHANNELS OPERATIONAL OF CHANNELS(f) TO TRIP OPERABLE 'CONOIT IONS ACTION
| |
| : 5. LOSS OF POWER
| |
| : 1. 4.16 kV Emergency Bus Under-voltage (Loss of Voltage) I/bus I/bus I/bus 1,2,3,4**,5** 36
| |
| : 2. 4.16 kV Emergency Bus Under- 1/source/ 1,2,394**,5** 37 voltage (Degraded Voltage) 1/source/ 1/source/
| |
| bus bus bus A**The Minimum OPERABLE Channels Per Trip Function is per subsystem.
| |
| Ummm*
| |
| | |
| 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 conditijon 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 uB" 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.
| |
| * When the system is required to be OPERABLE per Specification 3.5.2.
| |
| I 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 less than or equal to 200 psig.
| |
| LIMERICK - UNIT 1 3/4 3-35 Amendment No. 53 DEC 17 IN91 I
| |
| | |
| 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 - 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 24 hours.
| |
| 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 I 3/4 3-36 Amendment No. 4*-, --3, 158
| |
| | |
| TABLE 3.3.3-I (Continued)
| |
| EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION STATEMENTS 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- 11XOX 127Y-11XOX and 127Z-11XOX operable 127Y-11XOX 127-11XOX and 127Z-IIXOX operable 127Z-11XOX 127-11XOX and 127Y-11XOX operable.
| |
| 127Z-11YOY 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 performance of the next required CHANNEL FUNCTIONAL TEST.
| |
| LIMERICK - UNIT I 3/4 3-36a Amendment No. 158
| |
| | |
| INTENTIONALLY LEFT BLANK I
| |
| TABLE 3.3.3-2 3901000402 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE z
| |
| __4
| |
| : 1. CORE SPRAY SYSTEM
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 > -129 inches* > -136 inches
| |
| : b. Drywell Pressure - High Z 1.68 psig < 1.88 psig C. Reactor Vessel Pressure - Low 5 455 psig,(decreasing) 5 435 psig, (decreasing)
| |
| : d. Manual Initiation H.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 Z 1.68 psig 1.88 psig C,,
| |
| C. Reactor Vessel Pressure - Low 455 psig,(decreasing) 435 psig, (decreasing)
| |
| : d. Injection Valve Differential Pressure - Low 5 74 psid, (decreasing) > 64 psid and < 84 psid (A)
| |
| : e. Manual Initiation H.A. N.A.
| |
| (A)
| |
| : 3. HIGH PRESSURE COOLANT INJECTION SYSTEM
| |
| : a. Reactor Vessel Water Level - Low Low, Level 2 > -38 inches*
| |
| Z 1.68 psig
| |
| > -45 inches Z 1.88 psig I
| |
| : b. Drywell Pressure - High C. Condensate Storage Tank Level - Low 5 167.8 inches** > 164.3 inches
| |
| : d. Suppression Pool Water Level - High < 24 feet 1.5 inches < 24 feet 3 inches Reactor Vessel Water Level - High, Level 8 Z 54 inches < 60 inches e.
| |
| : f. Manual Initiation N.A. N.A.
| |
| : 4. AUTOMATIC DEPRESSURIZATION SYSTEM
| |
| : a. Reactor Vessel Water Level - Low Low Low, Level 1 > -129 inches* > -136 inches I
| |
| Z 1.68 psig Z 1.88 psig C=, b. Drywell Pressure - High < 117 seconds
| |
| *c. ADS Timer < 105 seconds
| |
| -4J 5 145 psig,(increasing) > 125 psig, (increasing),
| |
| : d. Core Spray Pump Discharge Pressure - High > 115 psig, (increasing) z e. RHR LPCI Mode Pump Discharge Pressure-High 5 125 psig,(increasing) 0 5 12.5 inches > 11.0 inches
| |
| : f. Reactor Vessel Water Level-Low, Level 3
| |
| : g. Manual Initiation < 4A.
| |
| < 420 seconds < 450 seconds (A'
| |
| : h. ADS Drywell Pressure Bypass Timer (A)
| |
| *See Bases Figure B 3/4.3-1.
| |
| **Corresponds to 2.3 feet indicated.
| |
| I
| |
| | |
| 3901000402 TABLE 3.3.3-2 (Continued)
| |
| EMERGENCY CORE COOLINGSYTEMA-TrATION INSTRUMENTATION SETPOINTS
| |
| '-4 ALLOWABLE TRIP SETPOINT VALUE TRIP FUNCTION
| |
| : 5. LOSS OF-POWER RELAY
| |
| : a. 4. 16 kV Emergency Bus Undervoltage NA NA z 127-IIX (Loss of Voltage)
| |
| : b. 4.16 kV Emergency Bus Undervoltage RELAY (Degraded Voltage) 127-1IXOX a. 4.16 kV Basis and 2905 t 115 volts 2905 t 145 volts 102-IIXOX b. 120 V Basis 83 t 3 volts 83 f 4 volts
| |
| : c. < 1 second time < 1.5 second time aelay aelay 127Y- IIXOX** a. 4.16 kV Basis and 3640 t 91 volts 3640 t-182 volts WI 127Y- 1-11KOX b. 120 V Basis 104 t 3 volts 104 t 5.2 volts C. < 52 second time < 60 second time.
| |
| Ob delay aelay 127Z-IIXOX a. 4.16 kV Basis
| |
| 'hi and 162Y- IIXOX b.
| |
| 3910 t 11 volts 120 V Basis 3910 t 19 volts I
| |
| 111.7 t 0.3 volts 111.7 t 0.5 volts 30 C. < 10 second time < 11 second time aelay 3elay 127Z- IlXOX a. 4.16 kV Basis and 3910 t 11 volts 3910 t 19 volts 162Z- IIXOX b. 120 V Basis 111.7 t 0.3 volts 111.7 t 0.5 volts C. < 61 second time < 64 second time aelay aelay not result in
| |
| **This is an inverse time delay voltage relay. The voltages trip shown are the maximum that will in decreased times.
| |
| a trip. Some voltage conditions will result
| |
| | |
| TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds) 1.
| |
| CORE SPRAY SYSTEM
| |
| * 27#
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM : 40#
| |
| : 3. AUTOMATIC DEPRESSURIZATION SYSTEM N.A.
| |
| : 4. HIGH PRESSURE COOLANT INJECTION SYSTEM , 60#
| |
| : 5. LOSS OF POWER N.A.
| |
| # ECCS actuation instrumentation is eliminated from response time testing.
| |
| JAN 0 7 1999 LIMERICK - UNIT I 3/4 3-39 Amendment No. 4-*, 1 32
| |
| | |
| TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CH*EL. TESTCa). CALIBRATION al SURVEILLANCE REQUIRED
| |
| : 1. CORE SPRAY SYSTEM
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low Low, Level 1 1, 2, 3, 4*, 5*
| |
| : b. Drywell Pressure - High 1, 2, 3
| |
| : c. Reactor Vessel Pressure - Low 2, 3, 4*, 5*
| |
| N.A. N.A. 1,
| |
| : d. Manual Initiation 2, 3, 4*, 5*
| |
| : 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low Low, Level 1 1, 2, 3, 4*, 5*
| |
| : 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, 4*, 5*
| |
| : e. Manual Initiation N.A. N.A. 1, 2, 3, 4*, 5*
| |
| : 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 i, 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. -53, ;4, 186
| |
| | |
| 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 CHEECKLa TESTCaj CALIBRATION(a.1 SURVEILLANCE REQUIRED
| |
| : 4. AUTOMATIC DEPRESSURIZATION SYSTEM#
| |
| : a. Reactor Vessel Water Level -
| |
| Low Low Low, Level 1 1, 2, 3
| |
| : b. Drywell 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.
| |
| * When the system is required to be OPERABLE per Specification 3.5.2.
| |
| ** 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 127-11X is not field setable.
| |
| LIMERICK - UNIT I 3/4 3-41 Amendment No. 8-3,186
| |
| | |
| 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 tripsystem 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 thi-s 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 REQUIREMENTS 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, -74, 186
| |
| | |
| TABLE 3.3.4.1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER TRIP FUNCTION TRIP SY*ST* *
| |
| : 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 required surveillance provided the other channel is OPERABLE. I LIMERICK - UNIT I 3/4 3-43 Amendment No. 70 APR 2 6 1994
| |
| | |
| TABLE 3.3.4.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE TRIP FUNCTIO SETPOINT VALUE
| |
| : 1. Reactor Vessel, Water Level -
| |
| Low Low, Level 2 a -38 inches* > -45 inches
| |
| : 2. Reactor Vessel Pressure - High < 1149 psig - 1156 psig I
| |
| * See Bases Figure B3/4 3-1.
| |
| LIMERICK - UNIT I 3/4 3-44 Amendment No. 106 FEB 12 199
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 3-45 Amendment No. 4-0, 186
| |
| | |
| INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-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 30% of RATED THERMAL POWER.
| |
| : 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 channel(s) in the tripped condition within 12 hours. I
| |
| : 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 turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 12 hours.
| |
| I
| |
| : 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 takethe 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.
| |
| APR 26 1994 LIMERICK- UNIT 1 3/4 3-46 Amendment No. 70
| |
| | |
| 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. -4,186
| |
| | |
| TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION HINI IN OPERABLE CHANNELS TIP FUNCITf PER TRIP SYSTEM'
| |
| : 1. Turbine Stop Valve - Closure 2*0
| |
| : 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 301 of RATED THERMAL POWER.
| |
| APR 2 6 1994 LIMERICK --UNIT I 3/4 3-48 Amendment No. 70
| |
| | |
| 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 AU6E 1985 LIMERICK - UNIT I 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 LIMERICK - UNIT 1 3/4 3-50 AUG s 1965
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED I
| |
| LIMERICK - UNIT 1 3/4 3-51 Amendment No. 4-0,186
| |
| | |
| 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 Frequenc-y 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. 186
| |
| ;7_4,
| |
| | |
| TABLE 3.3.5-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS FUNCTIONAL UNITS PER TRIP FUNCTION* ACTION
| |
| : a. Reactgr-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 52
| |
| : d. Manual Initiation I/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.
| |
| LIMERICK - UNIT 1 3/4 3-53 Amendment No. 53 DEC 17 1991
| |
| | |
| 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.
| |
| I
| |
| : 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.
| |
| 1 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 RCIC system inoperable.
| |
| LIMERICK - UNIT 1 3/4 3-54 Amendment No. 53 DEC 17 199)
| |
| | |
| .- ');u IuVu4U2 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.
| |
| I LIMERICK - UNIT 1 3/4 3-55 Imendment no. 3 OCT 3 0 W3
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT I 3/4 3-56 Amendment No -53, 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, 4-3-1, 4-44, 186
| |
| | |
| TABLE 3.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION
| |
| : 1. ROD BLOCK MONITOR (a)
| |
| : a. Upscale
| |
| : b. Inoperative 2 1I* 60
| |
| : c. Downscale 2 1*" 60 2 60
| |
| : 2. APRM
| |
| : a. Simulated Thermal Power - Upscale 3
| |
| : b. Inoperative 61 C. Neutron Flux - Downscale 3 1, 2 61
| |
| : d. Simulated Thermal Power - Upscale 3 61 (Setdown) 3 2
| |
| : e. Recirculation Flow - Upscale 1 61 3 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(c) 3 2 61 2 5 61
| |
| : d. Downscal~e() 3 2 61 2 5 61
| |
| : 4. INTERMEDIATE RANGE MONITORS
| |
| : a. Detector not full in 6 2, 61
| |
| : b. Upscale 6 2, 61 C. Inoperative 6 2, 61
| |
| : d. Downscale(e) 6 2, 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 LIMERICV - UNIT 1 3/4 3-5P Amendment No. 4, 4-4, -4A4, 171
| |
| | |
| 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 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 IRN channels are on range 3 or higher.
| |
| (e) This function is automatically bypassed when the IRM channels are on range 1.
| |
| (f) DELETED I
| |
| z APR 1 20 LIMERICK - UNIT I 3/4 3-59 Amendment No. 4. 41, *,
| |
| 10, 141
| |
| | |
| TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1.ROD BLOCK MNITOR
| |
| : a. Upscale"'
| |
| : 1) Low Trip Setpoint (LTSP) * *
| |
| * *
| |
| : 2) Intermediate Trip Setpoint (ITSP)
| |
| : 3) High Trip Setpoint (HTSP) * *
| |
| : b. Inoperative N/A N/A
| |
| : c. Downscale (DTSP)
| |
| : d. Power Range Setpoint(b)
| |
| : 1) Low Power Setpoint (LPSP) 28.1% 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 POWER 2.APRM
| |
| : a. Simulated Thermal Power - Upscale:
| |
| - Two Recirculation Loop Operation : 0.66 W + 55.2% and ! 0.66 W + 55.7% and
| |
| : 108.0% of RATED < 108.4% of RATED THERMAL POWER THERMAL POWER
| |
| - Single Recirculation Loop Operation**** : 0.66 (W-7.6%) + 55.2% and
| |
| * 0.66 (W-7.6%) + 55.7% and I
| |
| * 108.0% of RATED
| |
| * 108.4% of RATED THERMAL POWER THERMAL POWER
| |
| : b. Inoperative N.A. N.A.
| |
| : c. Neutron Flux - Downscale Ž 3.2% of RATED THERMAL _>2.8% of RATED THERMAL POWER POWER
| |
| : d. Simulated Thermal Power - Upscale
| |
| * 12.0% of RATED 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 axial level
| |
| : 3. SOURCE RANGE MONITORS
| |
| : a. Detector not full in N.A. N.A.
| |
| : b. Upscale < 1 x 105 cps < 1.6 x 10 cps
| |
| : c. Inoperative N.A. N.A.
| |
| : d. Downscale >_ 3 cps** Ž!1.8 cps**
| |
| LIMERICK - UNIT 1 3/4 3-60 Amendment 4No. , 7-1-7 t, 454, :4.", 4-4-4, 177
| |
| | |
| TABLE 3.3.6-2 (-.,tinued)
| |
| 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 >_ 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 Ž 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 I TSP 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 1 3/4 3-60a Amendment No. 3, -39, -4, -7,
| |
| &&, 4-4-, 177
| |
| | |
| ,I..j,.I I~nri -.0O, .I 30 StA& TO V0132 O=
| |
| SIN COMNT Iun TM SZCID, TO 30751O Z0 nV=r 3.3.6-1 3/14 3-Mb Dendmnt fO. 34 DEC 13 M
| |
| | |
| TABLE 4.,,6-1 CONTROL ROA BLOCK INSTRUMFNTATION S!HRVULI ANf.E RFf)1RFMFNTS 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) I*
| |
| : b. Inoperative N.A. (c) N.A. i*
| |
| 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. NoA. 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**
| |
| : 6. DELETED
| |
| : 7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. (g) N.A. 3, 4 LIMERICK - UNIT 1 3/4 3-61 Amendment No. 4-4, -54, &6, -- ,
| |
| -444, -- 7-7,186
| |
| | |
| TABLE 4.3.6-1 (Continued)
| |
| CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS 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 applicabl.e 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.
| |
| (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. I LIMERICK - UNIT 1 3/4 3-62 Amendment No. 4-1, 66, 9, 1-44, 186
| |
| | |
| 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. 44,186
| |
| | |
| 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 10s [1Ci/cc 70 Fresh Air Supply Radiation and
| |
| * Monitor
| |
| : 2. Area Monitors
| |
| : a. Criticality Monitors
| |
| : 1) Spent Fuel 2 (a) Ž 5 mR/h and *20mR/h(b) 71 Storage Pool
| |
| : b. Control Room Directl At All Times N.A. (b) 73 Radiation Monitor
| |
| : 3. Reactor Enclosure Cooling Water Radiation Monitor 1 At All Times < 3 x Background(b) 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 or during operations with a potential for draining the reactor vessel 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. 185
| |
| | |
| 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 I LIMERICK - UNIT 1 3/4 3-66 Amendment No. -;g, 4-!-, 186
| |
| | |
| 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 or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
| |
| (a) With fuel in the spent fuel storage pool.
| |
| (b) 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.
| |
| (c) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| LIMERICK - UNIT 1 3/4 3-67 Amendment No. 185, 186
| |
| | |
| Section 3.3.7.2 (Deleted)
| |
| I THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR. TECHNICAL SPECIFICATIONS PAGES 3/4 3-69 THROUGH 3/4 3-72 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| .UC- 2 9 1994 LIMERICK - UNIT I 3/4 3-68 Amendment No. ZZ, 40, 75
| |
| | |
| Section 3.3.7.3 (DelIeted)
| |
| THE INFOR1ATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE OOCN. TECHNICAL SPECIFICATIONS PAGES 3/4 3-74 THROUGH 3/4 3-75 OF THIS SECTION HAVE BEEN INTENTIONALLY ONITTED.
| |
| LIMERICK - UNIT I 3/4 3-73 Amendment No.
| |
| JAN 02 1331 48 I
| |
| | |
| 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.
| |
| * Control is not required to be transferred to perform the CHANNEL CHECK.
| |
| LIMERICK - UNIT I 3/4 3-76 Amendment No. ;41, 4-69, 186
| |
| | |
| r- TABLE 3.3.7.4-1 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS 6-4 MINIMUM INSTRUMENTS INSTRUMENT OPERABLE
| |
| : 1. Reactor Vessel Pressure 1 3Ca 2. Reactor Vessel Water Level C"
| |
| : 3. Safety/Relief Valve Position, 3 valves
| |
| : 4. Suppression Chamber Water Level 1 5.. Suppression Chamber Water Temperature
| |
| : 6. Drywell Pressure 1 1
| |
| : 7. Drywell Temperature 1
| |
| : 8. RHR System Flow 1
| |
| : 9. RHR Service Water Pump Discharge Pressure 1
| |
| : 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
| |
| CL 14. Condensate Storage Tank Level
| |
| (-D0 1
| |
| : 15. RHR Heat Exchanger Bypass Valve (HV51-1FO48A) Position Indication (0 - 100%)
| |
| 1
| |
| : 16. RCIC Turbine Tripped Indication 1
| |
| : 17. RCIC Turbine Bearing Oil Pressure Low Indication
| |
| : 18. RCIC LP Bearing Oil Temperature High Indication
| |
| : 19. RHR Heat Exchanger Discharge Line High Radiation Indication
| |
| | |
| 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-49-1F076 Control-Steam Line warmup bypass valve HV-49-1F060 Control-RCIC turb exhaust to suppression pool isolation HV-50-112 Control-Turb trip throttle valve HV-50-1F045 Control-Turbine steam supply-valve HV-49-1F008 Control-Turbine steam line outboard isolation valve HV-49-1F007 Control-Turbine steam line inboard isolation valve HV-49-1F031 Control-RCIC pump suction from suppression pool HV-49-1F029 Control-RCIC pump suction from suppression pool HV-49-1F010 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-50-1F046 Control-RCIC turbine cooling water valve HV-49-IF012 Control-RCIC pump disch valve HV-49-1F013 Control-RCIC pump disch valve 10P220 Control-Vacuum tank condensate puMp 1OP219 Control-Barometric condenser vacuum pump HV-49-1F002 Control-Barometric condenser vacuum pump disch 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-IFOSO ContrOl-Vacuue breaker Outboard isolation valve HV-49-1FO&4 Control-Vacuum breaker inboard isolation valve FIC-49-JRO01 Controller-RCIC discharge flow control E51-S45 RCIC Turbine Trip Bypass NUCLEAR BOILE -"
| |
| HSS-41-191 Control-Transfer switch PSV-41-1FO13A Control-Main stem line safety/relief valve PSV-41-1FO13C Control-Main steam line safety/relief valve PSV-41-1FO13N Control-Main steam line safety/relief valve RHR SYSTEM HSS-51-192 Control-Transfer switch HSS-51-193 Control-Transfer switch HSS-51-194 Control-Transfer switch HSS-51-195 Control-Transfer switch HSS-51-196 Control-Transfer switch MSS-51-197 Control-Transfer switch SS-51-198 Control-Transfer switch HV-51-1FO09 Control-RNA pump shutaown cooling suction inbcaard isolation HV-S1-1FOO8 Control-RNR shutdown cooling suction outboard isolation HV-S-1F06A Control-lA RHR loop shutdown cooling suction HV-Si- IFOO68 Contrql-2B ARNR loop -shutduwcoolIng suction HV-S1-1FOO4A Control-lA RHR pump suction 1AP202 Control-lA RNR pum LIMERICK - UNIT I 3/4 3-79
| |
| | |
| TABLE 3.3.7.4-1 (Continued)
| |
| RHR SYSTEM (Continued)
| |
| HV-43-1F023A Control-Recirculation pump A suction valve HSS-43-]91 Control-Transfer switch HV-51-1FOO7A Control-IA RHR pump minimum flow bypass valve HV-51-1FO48A Control-lA heat exchanger shell side bypass HV-51-1FO15A Control-IA shutdown cooling injection valve HV-51-1FO16A Control-Reactor containment spray HV-51-1FO17A Control-lA RHR loop injection valve HV-51-IFO24A Control-lA RHR loop test return HV-51-1F027A Control-Suppression pool sparger isolation HV-51-1FO47A Control-lA Heat exchanger shell side inlet HV-;51-1FOO3A Control-IA Heat exchanger shell side outlet HV-51-1F049 Control-RHR Discharge to radwaste outboard isolation HV-51-125A Control-lA/IC test return line to suppression pool RHR SERVICE WATER SYSTEM HSS-12-015A-2 Control-Spray pond/cooling tower select HSS-12-015C-2 Control-Spray pond/cooling tower select HSS-12-016A-2 Control-Spray/bypass select HSS-12-016C-2 Control-Spray/bypass select AUG 23 1994 LIMERICK - UNIT 1 3/4 3-80 Amendment No. i, 05, 74
| |
| | |
| 3901000402 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-51-1FO14A 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 outlet EMERGENCY SERVICE WATER SYSTEM I
| |
| 0AP548 Control-A emergency servi ce water pump HV-11-O11A Control-A emergency servi ce 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-Dll Safeguard SWGR feeder bkr.
| |
| 152-I1609/CSR 101-012 Safeguard SWGR feeder bkr.
| |
| 152-11709/CSR 101-013 Safeguard SWGR feeder bkr.
| |
| 152-11502/CSR 201-011 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 0114 Safeguard LC XFMR breaker LIMERICK - UNIT 1 3/4 3-81 Amendment No. 33 OCT 3 0 L38
| |
| | |
| 3901000402 Table 3.3.7.4-1 (Continued)
| |
| STANDBY AC POWER SUPPLY (Continued) 152-11605/CSR D124 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 hus~ B1i LIMERICK - UNIT 1 3/4 3-82
| |
| | |
| 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.*.-. 1 ACCIDENT MONITORING INSTRUMENTATION MINIMUM APPLICABLE REQUIRED NUMBER CHANNELS OPERATIONAL INSTRUMENT OF CHANNELS OPERABLE 1,2IIONS 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 1 2 1,2 80
| |
| : 4. Suppression Chamber Water Temperature 8, 6 locations 6, 1,2 80 I/location
| |
| : 5. Suppression Chamber Air Temperature 1 1 1,2 80 I
| |
| : 6. Drywell Pressure 2 80 1,2
| |
| : 7. Drywell Air Temperature 1 1,2 80
| |
| : 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-&-, 4-7-, 179
| |
| | |
| Table 3.3.7.5-1 (Continued)
| |
| ACCIDENT MONITORING INSTRUMENTATION TABLE NOTATIONS
| |
| *Three noble gas detectors with overlapping ranges (10-7 to 10I, 10-4 to 102, 10-' to I01 RCi/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 No. 4-64, 173
| |
| | |
| TABLE 4,.,-5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECKL(-a4 CALIBRATIONLa)-
| |
| : 1. Reactor Vessel Pressure
| |
| : 2. Reactor Vessel Water Level
| |
| : 3. Suppression Chamber Water Level
| |
| : 4. Suppression Chamber Water Temperature
| |
| : 5. Suppression Chamber Air Temperature
| |
| : 6. Primary Containment Pressure
| |
| : 7. Drywell Air Temperature
| |
| : 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, i-7-3, 4-7-9, 186
| |
| | |
| 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 REQUIREMENTS 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, J74, 9ý86
| |
| | |
| INSTRUMENTATION Section 3/4.3.7.7 THE INFORMATION FROM THIS TECHNICAL SPECIFICATION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM)
| |
| LIMERICK - UNIT I 3/4 3-89 Amendment No. 11, 117 imL I1I9In
| |
| | |
| 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 t.he chlorine isolation mode of operation.
| |
| b., With both chlorine detection subsystems inoperable, within I 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. 4-4,4,,-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 REQUIREMENTS 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. l-4, ;4, 4,186
| |
| | |
| 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.
| |
| DEC 2 0 995 LIMERICK - UNIT 1 3/4 3-92 Amendment No. 69,104
| |
| | |
| INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 3-97 Amendment No. 44, -*4, 153
| |
| | |
| Section 3.3.7.11 (Deleted)
| |
| 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.
| |
| LIMERICK - UNIT I 3/4 3-98 Amendment No. ),48 JAN 02 IS3M
| |
| | |
| INSTRUMENTATION OFFGAS GAC MONITIORNG INSTRUMENTATION LIMITING CONDITIONI FOR OPERATION 3.3.7.12 The offgas monitoring instrumentation channels shown in Table 3.3.7.12-1 shall be OPERABLEwith their alarm/trip setpoints set to ensure that the limits of Specifications 3.11.2.5 and 3.11.2.6:respectively, are not exceeded.
| |
| APPLICABILITY: As shown in Table 3.3.7.12-1 ACTION:
| |
| : a. With an offgas monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, declare the channel inoperable, and take the ACTION shown in Table 3.3.7.12-1.
| |
| : b. With less than the minimum number of offgas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain why this inoperability was not corrected in a timely manner in the next Annual Radioactive Effluent Release Report.
| |
| : c. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.7.12 Each offgas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.7.12-1.
| |
| LIMERICK - UNIT 1 3/4 3-103 Amendment No. 4-9, 4-3-1, 4-6., 186
| |
| | |
| t TABLE 3.3.7.12-1 I
| |
| *-4 OFFGAS MONITORING INSTRUHENTATION
| |
| '-4 S
| |
| MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION
| |
| '-a 1. MAIN CONDENSER OFFOAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
| |
| : a. Hydrogen Monitor 1 110 2.
| |
| 3.
| |
| 4.
| |
| (Deleted)
| |
| (Deleted)
| |
| MAIN CONDENSER OFFOAS PRE-TREATHENT I
| |
| 4*
| |
| RADIOACTIVITY MONITOR 1 115
| |
| : a. Noble Gas Activity Monitor S. (Deleted)
| |
| I a
| |
| I t.
| |
| w 0
| |
| 0
| |
| | |
| TABLE 3.3.7.12-1 (Continued)
| |
| TABLE NOTATIONS (Deleted) F--
| |
| During operation of the main condenser steam jet air ejector and. offgas treatment system.
| |
| . (Deleted)
| |
| I ACTION STATEMENTS ACTION 110 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser offgas treatment system may continue for up to 30 days provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours.
| |
| ACTION 111-114 (Deleted)
| |
| I ACTION 115 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, releases to the environment may continue for up to 72 hours'provided that the North Stack Effluent Noble Gas Activity Monitor is OPERABLE; otherwise, be in at least HOT SHUTDOWN within 12 hours.
| |
| LIMERICK - UNIT I 3/4 3-105 Amendment No. 48 JAN 02 IS I
| |
| | |
| INTENTIONALLY LEFT BLANK 1
| |
| LIMERICK - UNIT I 3/4 3-106 Amendment No.48 JAN 02 13
| |
| | |
| TABLE 4.3.7,12-1 OFFGAS MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK(59. CHECKC5.1 CALIBRATION(51 TEST . IS REQUIRED
| |
| : 1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
| |
| : a. Hydrogen Monitor N.A. (3) **
| |
| : 2. (Deleted)
| |
| : 3. (Deleted)
| |
| : 4. MAIN CONDENSER OFFGAS PRE-TREATMENT RADIOACTIVITY MONITOR (STEAM JET AIR EJECTOR)
| |
| **
| |
| : a. Noble gas activity monitor (2) (1)
| |
| : 5. (Deleted)
| |
| LIMERICK - UNIT 1 3/4 3-107 Amendment No. 48, 186
| |
| | |
| TABLE 4.3.7.12-1 (Continued)
| |
| TABLE NOTATIONS
| |
| * (Deleted)
| |
| ** During operation of the main condenser steam jet air ejector and offgas treatment system.
| |
| (Deleted)
| |
| (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
| |
| : 1. Instrument indicates measured levels above the alarm/trip setpoint.
| |
| : 2. Circuit failure.
| |
| : 3. Instrument indicates a downscale failure.
| |
| : 4. Instrument controls not set in operate mode.
| |
| (2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST), previously National Bureau of Standards, or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. 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.
| |
| (3) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
| |
| : 1. 0.0 volume percent hydrogen, balance nitrogen, and
| |
| : 2. 4 volume percent hydrogen, balance nitrogen (4) (Deleted)
| |
| (5) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.
| |
| LIMERICK - UNIT 1 3/4 3-108 Amendment No. 48, 186
| |
| | |
| INTENTIONALLY LEFT BLANK LIMERICK - UNIT 1 3/4 3-109 Amendment No.48 JAN 0 2 1991 I
| |
| | |
| Section 3/4.3.8 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
| |
| TECHNICAL SPECIFICATIONS PAGE 3/4 3-111 HAS BEEN INTENTIONALLY OMITTED.
| |
| SEP 21 1995 LIMERICK - UNIT 1 Amendment No. 100 3/4 3-110
| |
| | |
| PAGE INTENTIONALLY LEFT BLANK 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 shallbe 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. , 4-7,186
| |
| | |
| 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*
| |
| * With Thermal Power greater than or equal to 25% of Rated Thermal Power.
| |
| LIMERICK - UNIT 1 3/4 3-113 Amendment No. 91 JUl 0 7 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 1/4 3-n4 RAlE 8 9
| |
| | |
| INFORMATION ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 3-115 Amendment No. -7., 9-4, 186
| |
| | |
| PAGE INTENTIONALLY LEFT BLANK 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.
| |
| APPLICABILITY: 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
| |
| * 76.2% 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 I 3/4 4-1 Amendment No. =40, 46, 4i06, 177
| |
| | |
| 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 speci fi cati ons.
| |
| : 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 4-1la Amendment No. 49, 66, 4-44, 449, 177
| |
| | |
| REACTOR COOLANT SYSTEM q11PVF II ANCF RF oIIRFMFNTS 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
| |
| * 76.2% 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.ý 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. 3,30,4,-7-6,7-1,06, 4-42, 4-7-7, 186
| |
| | |
| INTENTIONALLY LEFT BLANK CONTENTS OF THIS PAGE HAVE BEEN DELETED LIMERICK - UNIT I 3/4 4-3 Amendment No. , 4-96, 177
| |
| | |
| REACTOR COOlANT SYSTEM JET 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.
| |
| SURVEILLANCE 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%.
| |
| *To be determined from the startup test program data.
| |
| LIMERICK - UNIT I 3/4 4-4 Amendment No. 4G, 186
| |
| | |
| REACTOR COOLANT SYSTEM SIIRVF I ANCF RFOIUIRFMENTS (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:
| |
| I. 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.
| |
| *To be determined from the startup test program data.
| |
| LIMERICK - UNIT 1 3/4 4-4a Amendment No. -39, 186
| |
| | |
| 3895241160 THIS PAGE IR"EmflOIhLT LEFT BLMIK
| |
| | |
| 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.
| |
| APPLICABILITY: 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 REQUIREMENTS 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 (corrcc~tcd)
| |
| | |
| 3892373290 REACTOR COOLANT SYSTEM IDLE RECIRCULATIOW LOOP STARTUP LIMITZNG 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 145*F, 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 504F, 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 SOF 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.
| |
| An aSM5 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%
| |
| APPLICABILITY: 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 previously set 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. -6,47-4,406-,J,-l37,4-5-7,44, 186
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.3 REACTOP COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONLJFION FOR OPERATION 3.4.3.1 The following reactor coclant leakage detection systems shall be OPERABLE:
| |
| : a. The primary containment atmosphere gaseous radioactivity monitoring system,
| |
| : b. The drywell floor drain SUmD flow 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 floor drain sump flow monitoring system INOPERABLE, restore the drywell floor drain sump flow 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 Sections 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 I 3/4 4-8 Amendment ;4, 4-40, 169
| |
| | |
| REACTOR COOLANT SYSTEM ACTIONS (Continued)
| |
| F. With any other two or more leak detection systems inoperable other than ACTION E above OR with required Actions and associated Completion Time of ACTIONS A, B, C, D or E not met, be in HOT SHUTDOWN within 12 hours AND in COLD SHUTDOWN within the next 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 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. 1440, 186
| |
| | |
| INTENTIONALLY LEFT BLANK REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 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. I gpm leakage at a reactor coolant system pressure of 950 +/-10 psig from any reactor coolant system pressure isolation valve.**
| |
| : 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 andlor 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 3/4 4-9 Amendment No. 28, 49, 4-74, 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,
| |
| : c. Monitoring the drywell unit coolers condensate flow rate in accordance with the Surveillance Frequency Control Program,
| |
| : 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 Surveillance Frequency Control Program (not a means of quantifying leakage).
| |
| 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 I 3/4 4-10 Amendment No. 3, 4-9, ;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 I 3/4 4-11 Amendment No. 4-4, 182
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.4 (DELETED)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM). TECHNICAL SPECIFICATIONS PAGES 3/4 4-13 AND 3/4 4-14 HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT 1 3/4 4-12 Amendment No. 4-6., 174
| |
| | |
| 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 1-131.
| |
| : 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 1-131 but less than or equal to 4 microcuries per gram, DOSE EQUIVALENT 1-131 for more than 48 hours during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT 1-131, 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 of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT 1-131, 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 I 3/4 4-15 Amendment No. 2-9, 4-9, 174
| |
| | |
| 3893224150 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to
| |
| .1 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 19 U
| |
| | |
| TABLE 4. ,-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM OPERATIONAL CONDITIONS TYPE OF MEASUREMENT SAMPLE AND ANALYSIS IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSTS IS REOIJIRFD
| |
| : 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 I 3/4 4-17 Amendment No. 3-3, 44-4,186
| |
| | |
| 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.
| |
| SURVFILLANCF RFOUIREMFNTS 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 asoapplicable, in accordance with the Surveillance Frequency Control Program.
| |
| LIMERICK - UNIT I 3/4 4-18 Amendment No. -, 446, --4-5,186
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (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 flange and 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 0 F, in accordance with the Surveillance Frequency Control Program.
| |
| : 2.
| |
| * 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 I 3/4 4- 19 Amendment No. 2-9, -16, 1-2-6, , 186
| |
| | |
| 140(
| |
| 130 0
| |
| ) I I I A22A TTW lll II B C 12010 110(
| |
| 1000 a.900 0
| |
| 800 0 0 uJ A, B, C - CORE BELTLINE AFTER O 700 ASSUMED 69"F SHIFT FROM AN INITIAL PLATE RTNDT OF 20°F lu: 600 z A - SYSTEM HYDROTEST WITH FUEL INTHE VESSEL I- B - NON-NUCLEAR HEATUP/COOLDOWN LIMIT C - NUCLEAR (CORE CRITICAL)
| |
| LIMIT 300 B/ CURVES A, B, C ARE VALID UP 200 TO 32 EFPY OF OPERATION CURVE A2 IS VALID UP.TO 100 80F 22 EFPY OF OPERATION 0
| |
| 0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 375 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)
| |
| MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE FIGURE 3.4.6.1-1 LIMERICK - UNIT 1 3/4 4-20 Amendment No. -6, 4-46, 44-5,1515, 163
| |
| | |
| INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 4-21 Amendment No. 4--6, 4--5-, 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 1 3/4 4-22 Amendment No. 446,186
| |
| | |
| RFArT(1R COnl ANT SYýTFM 3/4 4 7 MAIN ¶TFAM IINF TSOIATION VAIVFS 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.
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| ACTION:
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| With one or more MSIVs inoperable:
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| : a. Maintain at least one MSIV OPERABLE in each affected main steam line that I is open and within 8 hours, either:
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| : 1. Restore the inoperable valve(s) to OPERABLE status, or
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| : 2. Isolate the affected main steam line by use of a deactivated MSIV in I the closed position.
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| : b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD I SHUTDOWN within the following 24 hours.
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| 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.
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| LIMERICK - UNIT 1 3/4 4-23 Amendment No. 169
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| REACTOR COOLANT SYSTEM 389 11002560 3/4.4.8 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.8 The structural integrity of ASME Code Class 1, 2. and 3 components shall be maintained in accordance with Specification 4.4.8.
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| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.
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| ACTION:
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| : a. With the structural integrity of any ASME Code Class 1-component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the reactor coolant system temperature more than 50F above the minimum temperature required by NDT considerations.
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| : b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the reactor coolant system temperature above 2000F.
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| C. With the structural integrity of any ASNE Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) from service.
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| I SURVEILLANCE REQUIREMENTS 4.4.8 No requirements other than Specification 4.0.5..
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| LIMERICK - UNIT 1 3/4 4-24 Amendment No. 11 KVIg lU~o
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| 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. * ** ***
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| Each independent RHR shutdown cooling subsystem shall consist of at least:
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| : a. One OPERABLE RHR pump, and
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| : b. One OPERABLE RHR heat exchanger, not common to the two (2) independent subsystems.
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| APPLICABILITY: OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint.
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| ACTION:
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| : 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.
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| 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.****
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| : 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.
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| SURVEILLANCE REOUIREMENTS 4.4.9.1 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.
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| *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.
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| **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.
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| ***The independent RHR shutdown cooling subsystem may be removed from operation during hydrostatic testing.
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| ****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.
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| LIMERICK - UNIT 1 3/4 4-25 Amendment No. 9-7, 4-14, 186
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| 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. * ** ***
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| APPLICABILITY: OPERATIONAL CONDITION 4.
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| ACTION: #
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| : a. With one (1) or two (2) RHR shutdown cooling subsystems inoperable:
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| : 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.
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| : b. With no RHR shutdown cooling subsystems in operation and no recirculation pump in operation:
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| : 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
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| : 2. Once per hour monitor reactor coolant temperature and pressure.
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| SURVEILLANCE REQUIREMENTS 4.4.9.2 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.
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| Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up two (2) hours per eight (8) hour period.
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| *** One (1) RHR shutdown cooling subsystem may be inoperable for up to two (2) hours for the performance of Surveillances.
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| S** The shutdown cooling subsystem may be removed from operation during hydrostatic testing.
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| # Separate Action entry is allowed for each shutdown cooling subsystem.
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| LIMERICK - UNIT I 3/4 4-26 Amendment No. W, 4-1-9, 186
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| 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:
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| : a. The core spray system (CSS) consisting of two subsystems with each subsystem comprised of:
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| : 1. Two OPERABLE CSS pumps, and
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| : 2. An OPERABLE flow patch capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.
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| : b. The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of four subsystems with each subsystem comprised of:
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| : 1. One OPERABLE LPCI pump, and
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| : 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
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| : c. The high pressure coolant injection (HPCI) system consisting of:
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| : 1. One OPERABLE HPCI pump, and
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| : 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
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| : d. The automatic depressurization system (ADS) with at least five OPERABLE ADS valves.
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| APPLICABILITY: OPERATIONAL CONDITION 1, 2* ** #, and 3* ** ##.
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| *The HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.
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| **The ADS is not required to be OPERABLE when the reactor steam dome pressure is less that or equal to 100 psig.
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| #See Special Test Exception 3.10.6.
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| ##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.
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| LIMERICK - UNIT 1 3/4 5-1 Amendment No. a, 6,131 IdJV 1 6 199
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| EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
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| ACTION:
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| : a. For the core spray'system:
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| : 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.
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| : 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.
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| : 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..
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| : 5. W'th 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.
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| : 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.*
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| *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.
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| LIMERICK - UNIT 1 3/4 5-2 Amendment No. 86,94,131
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| -NOV 1 6 1998
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| EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
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| ACTION: (Continued)
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| : c. For the HPCI system:
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| : 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.
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| : 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.
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| : 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
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| * 100 psig within the next 24 hours.
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| : 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
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| * 100 psig within the next 24 hours.
| |
| : e. With a CSS and/or LPCI header AP instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours or determine the ECCS header AP locally at least once per 12 hours; otherwise, declare the associated CSS and/or LPCI, as applicable, inoperable.
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| : f. In the event an ECCS system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
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| LIMERICK UNIT 1 3/4 5-3 Amendment No. 33, -94, 169
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| | |
| FMFRGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:
| |
| : a. In accordance with the Surveillance Frequency Control Program:
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| : 1. For the CSS, the LPCI system, and the HPCI system:
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| a) Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.
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| 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.
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| : 2. For the LPCI system, verifying that both LPCI system subsystem cross-tie valves (HV-51-182 A, B) are closed with power removed from the valve operators.
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| : 3. For the HPCI system, verifying that the HPCI pump flow controller is in the correct position.
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| : 4. For the CSS and LPCI system, performance of a CHANNEL FUNCTIONAL TEST of the injection header AP instrumentation.
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| : b. Verifying that, when tested pursuant to Specification 4.0.5:
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| : 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 Ž 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 2 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 ressure of 1040 psig plus head and line losses when steam is eing 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.
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| * 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.
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| LIMERICK - UNIT 1 3/4 5-4 Amendment No. 2-9, ;,4, *46, 186
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| | |
| FMFRGFNCY CORF COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
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| : 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 2200 psig plus head and line losses, when steam is being supplied to the turbine at 200 + 15, 0 psig.**
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| 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.
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| : 3. Performing a CHANNEL CALIBRATION of the CSS, LPCI, and HPCI system discharge line "keep filled" alarm instrumentation.
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| : 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.
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| : 5. Performing a CHANNEL CALIBRATION of the LPCI header AP instru-mentation and verifying the setpoint to be S the allowable value of 3.0 psid.
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| : 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.
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| LIMERICK - UNIT 1 3/4 5-5 Amendment No. 4-9,-34, ;4--, 42-O, 4-2-,186
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| | |
| EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:
| |
| : a. Core spray system (CSS) subsystems with a 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 subsystems with a 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 .**
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| APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5*.
| |
| ACTION:
| |
| : a. With one of the above required subsystems inoperable, restore at least two subsystems to OPERABLE status within 4 hours or suspend all operations with a potential for draining the reactor vessel.
| |
| : b. With both of the above required subsystems inoperable, suspend CORE ALTERATIONS and all operations with a potential for draining the reactor vessel. Restore at least one subsystem to OPERABLE status within 4 hours or establish SECONDARY CONTAINMENT INTEGRITY within the next 8 hours.
| |
| *The ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.
| |
| **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.
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| LIMERICK - UNIT 1 3/4 5-6 Amendment No. 95 A 0719IM
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| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1.*
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| 4.5.2.2 The core spray system shall be determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the condensate storage tank required volume when the condensate storage tank is required to be OPERABLE per Specification 3.5.2a.2.b).
| |
| *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.
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| LIMERICK - UNIT I 3/4 5-7 Amendment No. -95, 186
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| | |
| -EMIERGENCY CORE COOLING SYSTEMS 3/4.5.3 SUPPRESSXON CHANUR LI*MNG CONDMTON FOR OPERATION 3.5.3 The 5uppressiOn chamber shall be OPERASLE:
| |
| : a. In OPERATIONAL CODIZTIOS 2.. 2. end)3 with a contained water volume of at least IU,220 fts, equivalent to a level of 2200.
| |
| : b. 1n OPERATIMNL COSSION 4 end P with a contained watex volume of at least 18,125 fts. eqialnt to a le'vel of 26100, except that the sUppMUsimi chamber level NO be leis than the liit or ma be drained provided th t-
| |
| : 1. No wperaions are Porto.me that have a potential for draining the reactor vessel, 2.* The reactor mode witch is locked In the Sluitdewn or Nefmml postion,
| |
| : 3. The condensate storage tu* contain at least 135,000 available 1gallons otf Mter equivalent to a level of 2 feet., ad
| |
| : 4. The core sprW rsti. is OPAgLE per Speciticgtmn L5.2 with enOPERABLE flow path caale of taking auction from the condonate storage tank and tiwn*sfering the water through the*
| |
| spW sParger to the reactor vessel.
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| APPLI*CAILTIY: OPERATION*L m Tm N .2. 3. 4. and 5'.
| |
| AMTON:
| |
| : a. In OPERATIOAL CO TIN 1, 2 or 3 with the suppression chmer water level less then the hbove limit, restore the water level ts within the limit within 1-hur or be In at least HOT SHUTOD within the next 32 hours end in COLD SlIJlD04 within the following 24 hours.
| |
| : b. In OPERATIONAL COITION 4 or-5 with the suppression chm r water level ess than the a1bov limit or drained and the above required csnditions not satisfied, suspend COE ALTERATIONS and all operations that have a potential for draining the reactor vessel and lock the reactor mode witch in the Shutdon position. Establish SECOINA CWWTAWIIIT **N within 2 hour.
| |
| 'The sUpression chaier Is not required to be OPERABLE provided tbat the reactor vessel head Is removed, the cavity Is flooded or being elooded fro the sUppression pool, the spent fuel pool pates an movd when the caviV is flooded, and the water level is intained within the limits of Specifications 3.9.8 and 3.9.9.
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| 36em LIMMICK - UNrT 2 1J/4 5-8
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| | |
| 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. I
| |
| : b. 16'0" in accordance with the Surveillance Frequency Control Program. I 4.5.3.2 With the suppression chamber level less than the above limit or drained in OPERATIONAL CONDITION 4 or 5*, in accordance with the Surveillance Frequency Control Program:
| |
| : a. Verify the required conditions of Specification 3.5.3b. to be satisfied, or
| |
| : b. Verify footnote conditions
| |
| * to be satisfied.
| |
| *The suppression chamber is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.
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| LIMERICK - UNIT 1 3/4 5-9 Amendment No. 186
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| | |
| INTENTIONALLY LEFT BLANK 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
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| | |
| 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 I 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. ' 18 LIMERICK - UNIT I 3/4 6-2 Amendment No. +47, 448, 146
| |
| | |
| 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 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 3/4 6-3 Amendment No. &-7, 4-*;, 44-8, 446,185
| |
| | |
| THIS PAGE IS INTENTIONALLY LEFT BLANK JAN 2 4 1997 LIMERICK -. UNIT 1 3/4 6-4 Amendment No. 2, U, 71, 93, 118
| |
| | |
| 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.
| |
| I
| |
| : 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 I 3/4 6-5 Amendment No. , 4-1-8, 169
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 The primary containment air lock shall be demonstrated OPERABLE:
| |
| : a. By verifying the sea] 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 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. 44-8, 186
| |
| | |
| CONTAINMENT SYSTEMS MSIV LEAKAGE ALTERNATE DRAIN PATHWAY I LIMITING CONDITION FOR OPERATION 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.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.4 The MSIV Leakage Alternate Drain Pathway shall be demonstrated OPERABLE:
| |
| : a. In accordance with 4.0.5, by cycling each motor operated valve, required to be repositioned, through at least one complete cycle of full travel.
| |
| LIMERICK - UNIT I 3/4 6-7 Amendment No. 71, 107 FEB 1 2 W96
| |
| | |
| 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. I 4.6.1.5.2 Reports Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days. This report shall include a description of the condition of the liner and concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.
| |
| JAN 2 4 1997 LIMERICK - UNIT 1 3/4 6-8 Amendment No. 118
| |
| | |
| 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.
| |
| 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 REQUIREMENTS 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:
| |
| 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 6-1i0 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.
| |
| APPLICABILITY: 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 REQUIREMENTS 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 I 3/4 6-11 Amendment No. 4---5, 44A, 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 ft 3 , equivalent to a level between 22' 0" and 24' 3", and a
| |
| : 2. Maximum average temperature of 95°F except that the maximum average temperature my be permitted to increase to:
| |
| a) 105*F during testing which adds heat to the suppression chamber.
| |
| b) 110°F with THERMAL POWER less than or equal to 11 of RATED THERMAL POWER.
| |
| c) 120*F 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/1K design value of 0.0500 ftW.
| |
| : c. At-least eight suppression pool water. temperature instruentation 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 95*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 1050F 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 950F 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) 95OF for more than 24 hours and THERMAL POWER greater than 21 of RATED THERMAL POWER, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
| |
| b) 1100F, place the reactor mode switch in the Shutdown position and operate at least one residual heat r val loop in the suppression pool cooling mods.
| |
| *Includes the volume inside the pedestal.
| |
| LIMERICK - UNIT 1 3/4 6-12 Amendment No. 29 JUL 2 2
| |
| | |
| CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
| |
| ACTION: (Continued)
| |
| : 3. With the suppression chamber average water temperature greater than 120'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'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'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 950 F by verifying suppression chamber average water temperature less than or equal to 120 0 F.
| |
| LIMERICK - UNIT 1 3/4 6-13 Amendment No. 186
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (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//kI 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/NIi 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. 68,7-9,1,7-08,44-8, 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.
| |
| *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 I 3/4 6-15 Amendment No. 86, 1-3-1, 186
| |
| | |
| 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 REOUIREMENTS 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.
| |
| *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 6-16 Amendment No. 7, -5g, 86, 4-34, 186
| |
| | |
| CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each primary containment isolation valves and the instrumentation line excess flow check valves 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. 2-9, 4-46, 4-16, 169
| |
| | |
| 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. -,219,4-3,*-7,-46,44-9, 186
| |
| | |
| TABLE 3.6.3-] (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.
| |
| OCT 18 moo LIMERICK - UNIT I 3/4 6-19 Amendment No. -, -3,89, 4-07, 146
| |
| | |
| PAGE INTENTIONALLY LEFT BLANK CONTAINMENT SYSTEMS 3/4.6.4 VACUUM RELIEF SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS LIMITING CONDITION FOR OPERATION 3.6.4.2 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 SHUTDOM 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.
| |
| LIMERICK - UNIT 1 3/4 6-44 Amendment NO. 46 OCT 02IN
| |
| | |
| INTENTIONALLY LEFT BLANK CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS 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 Ž0.120".
| |
| LIMERICK - UNIT I 314 6-45 Amendment No. -74, 186
| |
| | |
| 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.
| |
| 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.
| |
| 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.
| |
| : 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'.
| |
| LIMERICK - UNIT 1 3/4 6-46 Amendment No. 8, -74, 446, 4-2-9, 186
| |
| | |
| 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, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.
| |
| ACTION:
| |
| Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. 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.
| |
| : 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.
| |
| LIMERICK - UNIT 1 3/4 6-47 Amendment No. 2-9, 7,
| |
| .- 4-*4, 186
| |
| | |
| CONTAINMENT SYSTEMS REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES IMITING 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 REQUIREMENTS 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, 3, 74, 4-4-5,186
| |
| | |
| .4 THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
| |
| .7 DEC 2 0 1995 LIMERICK - UNIT I 3/4 6-49 Amendment No. 23,105
| |
| | |
| 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, or during operations with a potential for draining the reactor vessel, with the vessel head removed and fuel in the vessel.
| |
| 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, and operations with a potential for draining the reactor vessel. 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 I 3/4 6-50 Amendment No. 6,4,-1,4-00; ,-1--,5, 186
| |
| | |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
| |
| DEC 2 0 1995 LIMERICK - UNIT I 3/4 6-51 Amendment No. 3,105
| |
| | |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
| |
| DEC 2 0 195 LIMERICK - UNIT 1 3/4 6-51a Amendment No. 21,105
| |
| | |
| 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 ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
| |
| ACTION:
| |
| : a. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:
| |
| : 1. 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.
| |
| : 2. When (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel, suspend handling of IRRADIATED FUEL in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.
| |
| : b. With both standby gas treatment subsystems inoperable, if in progress, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS or operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3. are not applicable.
| |
| SURVEILLANCE REOUIREMENTS 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 I 3/4 6-52 Amendment No. 2-, 44, 4-9, 186
| |
| | |
| INTENTIONALLY LEFT BLANK 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'C (86°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. , -3, -7, 4-2-,, -144, 186
| |
| | |
| C0MTA1ImENTSVSTEMS_
| |
| SURVEILLANCE R[EU1REMENTS (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 a 15°F when tested in accordance with ANSI $SI4-19S0.
| |
| : 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 NS1-1980 while operating the system at a flow rate of 57"4 cfm
| |
| * 10%-
| |
| f_ After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the tnplace penetration and leakage testing acceptance criteria of less than 0-05%
| |
| in accordance with ANSI NSID-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a. flow rate of 5764 cfm
| |
| * 1C%.
| |
| : g. After any major system alteration:
| |
| : 1. Verify that when the SGTS fan is running the subsystem flo-rate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cft 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 g16 seconds with the reactor enclosure recirculation system in operation and the adjacent reactor enclosure and refueling area zones are in their isolation modes.
| |
| LIMERICK - UNIT 1 3/4 6-54 mendment No. 4, 10#, 122 FEB 1 8 199?G
| |
| | |
| 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 REOUIREMENTS 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. ;4, 444, 186
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (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 I 3/4 6-56 Amendment No. 4-, 4-,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. 2-8, -74, 173
| |
| | |
| CONTAINMENT SYSTEMS DRYWELL HYDROGEN MIXING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.2 Four independent drywell unit cooler hydrogen mixing subsystems (IAV212, 1BV212, 1GV212, 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 REQUIREMENTS 4.6.6.'2 Each drywell unit cooler hydrogen mixing subsystem shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:
| |
| : a. Starting the system from the control room, and
| |
| : b. Verifying that the system operates for at least 15 minutes.
| |
| LIMERICK - UNIT I 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.
| |
| SUIRVEILLANCE 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.
| |
| *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
| |
| | |
| INTENTIONALLY LEFT BLANK 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING'CONDITION FOR OPERATION 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 I 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.
| |
| APPLICABILITY: 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 or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. j
| |
| : 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 required by the ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. I LIMERICK - UNIT 1 3/4 7-1 Amendment No. So, %6,131 NOV 1 6 1998
| |
| | |
| PLANT SYSTEMS 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.
| |
| SURVFILLANCF REOUIREMENTS 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, 4-0, 4-69, 186
| |
| | |
| PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 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 I and common safety-related equipment, shall be OPERABLE:
| |
| : a. In OPERATIONAL CONDITIONS 1, 2, and 3, two loops.
| |
| : b. In OPERATIONAL CONDITIONS 4, 5, and *, one loop.
| |
| APPLICABILITY: 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 inoperable**, 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 be declared inoperable and the actions of 3.8.1.1 taken.
| |
| I LIMERICK - UNIT 1 3/4 7-3 Amendment No. ff, 40, 06, 131 NOV 1 6 1998
| |
| | |
| 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 HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
| |
| : b. In OPERATIONAL CONDITION 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 cooling remains available for the diesel 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.
| |
| LIMERICK - UNIT 1 3/4 7-4 Amendment No. 2-, 4-, ;-4, 186
| |
| | |
| 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 0 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 32 0 F.
| |
| : 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 I 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.
| |
| APPLICABILITY: All OPERATIONAL CONDITIONS and when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel.
| |
| ACTION:
| |
| : a. In OPERATIONAL CONDITION 1, 2, or 3 with one control room emergency fresh air supply 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. In OPERATIONAL CONDITION 4, 5, or when RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining the reactor vessel:
| |
| : 1. With one control room emergency fresh air supply subsystems inoperable, 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, suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REOUIREMENTS 4.7.2 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%.
| |
| LIMERICK - UNIT I 3/4 7-6 Amendment No. 40, ;4, 4*-, 186
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
| |
| : 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 0 F) and a relative humidity of 70%.
| |
| : 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 and the control room is maintained at a positive pressure of at least 1/8 inch water gauge relative to the turbine enclosure and auxiliary equipment room and outside atmosphere during subsystem operation with an outdoor air flow rate less than or equal to 525 cfm:
| |
| a) Outside air intake high radiation, and b) Manual initiation from control room.
| |
| LIMERICK - UNIT 1 3/4 7-7 Amendment No. 4, 44, -4,444, 186
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued).
| |
| : 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%.
| |
| : g. 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.
| |
| ADS 8 W5 LIMERICK - UNIT I 3/4 7-8
| |
| | |
| 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. In the event the RCIC system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.
| |
| : c. Specification 3.0.4.b is not applicable to RCIC.
| |
| SURVEILLANCE REOUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:
| |
| : a. In accordance with the Surveillance Frequency Control Program by:
| |
| : 1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is 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 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 pressure to less than 150 psig within the following 72 hours.
| |
| LIMERICK - UNIT 1 3/4 7-9 Amendment No. .2-, 4*96, 4-64,186
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : c. In accordance with the Surveillance Frequency Control Program by:
| |
| : 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, ;,4,186
| |
| | |
| PLANT SYSTEMS 3/4.7.4 SNUBBERS LIMITING CONDITION 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 Specification 4.7.4g on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.
| |
| SURVEILLANCE REOUIREMENTS 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
| |
| : a. Inspection Types As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.
| |
| : b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 4.7.4-1. The visual inspection interval for each type of snubber shall be determined based upon the criteria provided in Table 4.7.4-1 and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before amendment no. 51.
| |
| LIMERICK - UNIT 1 3/4 7-11 Amendment No. 51 I
| |
| | |
| TABLE 4.7.4-1 SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTABLE SNUBBERS Population Column A Column B Column C or Category Extend Interval Repeat Interval Reduce Interval (Notes 1 and 2) (Notes 3 and 6) (Notes 4 and 6) (Notes 5 and 6) 1 0 0 1 80 0 0 2 100 0 1 4 150 0 3 8 200 2 5 13 300 5 12 25 400 8 18 36 50i'? 12 24 48 750 20 40 78 1000 or greater 29 56 109 Note 1: The next visual inspection interval for a snubber population or category size shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during that interval. Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible.
| |
| These catergories may be examined separately or jointly. However, the licensee must make and document that decision before any inspection and shall use that decision as the basis upon which to determine the next inspection interval for that category.
| |
| Note 2: Interpolation between population or category sizes and the number of unacceptable snubbers is permissible. Use next lower integer for the value of the limit for Columns A, B, or C if that integer includes a fractional value of unacceptable snubbers as determined by interpolation.
| |
| Note 3: If the number of unacceptable snubbers is equal to or less than the number in Column A, the next inspection interval may be twice the previous interval but not greater than 48 months.
| |
| Note 4: If the number of unacceptable snubbers is equal to or less than the number in Column B but greater than the number in Column A, the next inspection interval shall be the same as the previous interval.
| |
| LIMERICK - UNIT 1 3/4 7-11a Amendment No. 51
| |
| | |
| TABLE 4.7.4-1 (continued)
| |
| SNUBBER VISUAL INSPECTION INTERVAL Note 5: If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced proportionally by interpolation, that is, the previous interval shall be reduced by a factor that is one-third of the ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Columns B and C.
| |
| Note 6: The provisions of Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months.
| |
| LIMERICK - UNIT 1 3/4 7-11b Amendment No. 51
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : c. Visual Inspection Acceptance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage are secure.
| |
| Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, providing that: (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type on that system that may be generically susceptible; and/or (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifications 4.7.4f. For those snubbers common to more than one system, the OPERABILITY of such snubbers shall be considered in assessing the surveillance schedule for each of the related systems. A review and evaluation shall be performed and documented to justify continued operation with an unacceptable snubber. If continued operation cannot be justified, the snubber shall be declared inoperable and the ACTION requirements shall be met.
| |
| : d. Transient Event Inspection An inspection shall be performed of all snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients, as determined from a review of operational data or a visual inspection of the systems, within 72 hours for accessible systems and 6 months for inaccessible systems following this determination. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following: (1) manually induced snubber movement; or (2) evaluation of in-place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.
| |
| LIMERICK - UNIT 1 3/4 7- 12 Amendment No. 51
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : e. Functional Tests At least once per 24 months a representative sample of each type ofsnubber shall be tested using the following sample plans. The sample plan(s) shall be selected for each type prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan(s) selected for each type prior to the test period or the sample plan(s) used in the prior test period shall be implemented:
| |
| : 1) At least 13.3% of the total population of a snubber type shall be functionally tested. For each snubber of that type that does not meet the functional test acceptance criteria of Specification 4.7.4f., an additional sample of at least 1/2 the size of the initial sample shall be tested until the total number tested is equal to the initial sample multiplied by the factor, 1+C/2, where C is the total number of unacceptable snubbers or until all the snubbers of that type have been tested; or
| |
| : 2) A representative sample of 37 snubbers of a snubber type shall be functionally tested in accordance with Figure 4.7.4-1. "C" is the total number of snubbers of that type found not meeting the acceptance requirements of Specification 4.7.4f. The cumulative number of snubbers of the type tested is denoted by "N". If at any time the point plotted falls in the "Accept" region, testing of snubbers of thattype may be terminated. When the point plotted lies in the "Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the "Accept" region, or all the snubbers of that type have been tested.
| |
| LIMERICK - UNIT 1 3/4 7-13 Amendment No. 2U, U,81 NOV 0 2 !994
| |
| | |
| PLANrT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| The representative sample selected for the function test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure as far as practical that they are representative of the various configurations, operating environments, range of size, and capacity of snubbers of that type. Snubber's placed in the same locations as snubbers which failed in the previous functional test period shall be retested at the time of the next functional test period but shall not be included in the sample plan, and failure of this functional test shall not be the sole cause for increasing the sample size under the sample plan. Testing equipment failure during functional testing may-invalidate the day's testing and allow that day's testing to resume anew at a later time provided all snubber's tested with the failed equipment during the day of equipment failure are retested.
| |
| If during the functional testing, additional testing is required due to failure of snubber's, the unacceptable snubber's may be catergorized into failure mode group(s). A failure mode group shall include all unacceptable snubbers that have a given failure mode and all other snubber's subject to the same failure mode. Once a failure mode group has been established, It can be separated for continued testing &part from the general population of snubber's. However, all unacceptable snubber's in the failure mode group shall be counted as one unacceptable snubber for additional testing in the general population. Testing In the failure mode group shall be based on the number' of unacceptable snubber's and shall continue in accordance with the sample plan selected for the type or' until all snubber's in the failure mode group have been tested. Any additional unacceptable snubber's found in the failure mode group shall be counted for continued testing only for that test failure mode group. In the event that a snubber(s) becomes included in more than one test failure mode group, it shall be counted in each failure mode group and shall be subject to the corrective action of each test failure mode group.
| |
| : f. Functional Test Acceptance Criteria The snubber functional test shall verify that:
| |
| : 1) Activation (restraining action) is achieved within the specified range in both tension and compression;
| |
| : 2) Snubber bleed, or release rate where required, is present in both tension and compression, within the specified range (hydraulic snubber's only);
| |
| : 3) For mechanical snubber's, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel; and
| |
| : 4) For snubber's specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement.
| |
| LIMERICK - UNIT 1 3/4 7-14 Amrm No. 54 MAY II I"S?
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.
| |
| : g. Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
| |
| For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service.
| |
| : h. Functional Testing of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test result shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 1? months before being installed in the unit.
| |
| I. Snubber Service Life Replacement Program The service life of all snubbers shall be monitored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for various seals, springs, and other critical parts shall be extended or shortened based on monitored test results and failure history. Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber Is required to be OPERABLE. The parts replacements shall be documented and the documentation shall be retained in accordance with Specification 6.10.3.
| |
| LIMERICK - UNIT 1 3/4 7-15 Amendment No. 54 I
| |
| MAY 11 197
| |
| | |
| 10 9
| |
| 8 7
| |
| 6 05 4
| |
| 3 2
| |
| 1 0
| |
| 0 10 20 30 40 50 60 70 80 90 100 N
| |
| Figure 4.7.4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST LIMERICK - UNIT 1 3/4 7-16 Amendment No. 54 I MAY 111992
| |
| | |
| 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 1 3/4 7-17 Amendment No. -14, 186
| |
| | |
| PLANT SYSTEMS 3089100256.0 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 fis$ton detectors transferred without a certificatt 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 Reeorts - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination.
| |
| ...... 09 8Ulm LIMERICK - WIN I 3/4 7-18
| |
| | |
| 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 (TRM) FIRE PROTECTION SECTION. TECHNICAL SPECIFICATIONS PAGES 3/4 7-19 THROUGH 3/4 7-32 HAVE BEEN INTENTIONALLY OMITTED.
| |
| DEC 2 0 195 LIMERICK - UNIT I 3/4 7-19 Amendment No. 41, 104
| |
| | |
| INTENTIONALLY LEFT BLANK PLANT SYSTEMS 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 REOUIRFMENTS 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. -, 7-4,186
| |
| | |
| PAM fl1rDTSM\tY TLEFT BLAN 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 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 1E distribution system, and
| |
| : b. Four separate and independent diesel generators, each with:
| |
| : 1. A separate day tank containing a minimum of 200 gallons of fuel,
| |
| : 2. A separate fuel storage system containing a minimum of 33,500 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 SurveillanceRequirement 4.8.1.1.la. within 24 hours and at least once per 7 days thereafter. If the diesel generator became inoperable for any reason other than preplanned preventative maintenance or testing, demonstrate the OPERABILITY of the remaining operable diesel generators by performing Surveillance Requirement 4.8.1.1.2a.4 for one diesel generator at a time, within 24 hours* and at least once per 7 days thereafter.
| |
| 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 4.8.1.1.1a. within I hour and at least once per 8 hours thereafter. Demonstrate the OPERABILITY of the remaining diesel generators by performing Surveil-lance Requirement 4.8.1.1.2a.4 for one diesel generator at a time, within 8 hours. 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.
| |
| *Contrary to the provisions of Specification 3.0.2, this test is required to be completed regardless of when the inoperable diesel generator is restored to OPERABILITY for failures that are potentially generic to the remaining Unit 1 diesel generators. and for which appropriate alternative testing cannot be designed.
| |
| LIMERICK - UNIT 1 3/4 8-1 Amendment No. 32, 40 1W 30 1"1
| |
| | |
| -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 1 hour and at least once per 8 hours thereafter; and perform Surveillance Requirement 4.8.1.1.2a.4 for the remaining diesel generator, within I hour. Restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in wt 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 I hour andat least once per 8 hours thereafter.
| |
| -If the diisel generator became inoperable for any reason other than preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining diesel generators by performing Surveillance Requirement 4.8.1.1.2a.4 for one diesel generator at a time, within 8 hours*. Restore at least two offsite circuits to OPERABLE status within 72 hours from the 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.
| |
| *Contrary to-'the provisions of Specification 3.0.2, this test is required to be completed regardless of when the inoperable diesel generator is restored to OPERABILITY for failures that are potentially generic to the remaining Unit 1 diesel generators.and for which appropriate alternative testing cannot be
| |
| :;Le*dned.
| |
| LIMERICK - UNIT 1 3/4 8-1a Amendment No. 32. 40 OW 30 19H
| |
| | |
| ELECTRICAL POWER SYSTEM L1IMITING CONDITION FOR OPERATION (Continued)
| |
| 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, coponents, and devices is OPERABLE and Its associated diesel generator is OPERABLE. Otherwise, restore either the inoperable diesel generator or the inoperable system subsystm 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 LoCI system subsystems, trains, comonents, and devices are OPERABLE and its associated diesel 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 systais covered in Specifications 3.7.1.1. and 3.7.1.2.
| |
| Amendment No. 31. 40
| |
| .LIMERICK - UNIT 1 3/4 8-2 I MAY3 0 WO
| |
| | |
| ELECTRICAL POWER SYSIEMS 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; and if any of the diesel generators have not been successfully tested within the proceding 24 hours, demonstrate OPERABILITY of these diesel generators by periforming Surveillance Requirement 4.8.1.1.2.a.4 for one diesel at a time, within 24 hours. 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, demonstrate the OPERABILITY of all of the above required 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 diesel generators are already operating; 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 4.8.1.1.1a. within 1 hour and at least once per 8 hours thereafter; 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. Restore at least one of the above required inoperable A.C. sources to OPERABLE status within 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 I 3/4 8-2a Amendment No. 40, 169
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVFILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class iE 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 frequency specified in Table 4.8.1.1.2-1 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. 42, ;4, 44--, 186
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (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'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 verifying 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 while maintaining voltage at 4285 +/- 420 volts and frequency at 60 +/- 1.2 hz and after steady state conditions are reached, voltage is maintained at 4280 +/- 120 volts.
| |
| LIMERICK - UNIT 1 3/4 8-4 Amendment No. 4, ;-4, 4-4, 4-24, 4-36, 4-5-, 186
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (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. -, Z-1,4-4-,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. 4, -74, ;4, 40-3,
| |
| -36, 186
| |
| | |
| ELECTRICAL POWER SYSTEMS rIIH \IFTI I 6NrF PplHITPFMFlJTq (r~nnin"PH)
| |
| : 13. Verifying that the following diesel generator lockout 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 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 I 3/4 8-7 Amendment No. -, -4,186
| |
| | |
| ELECTRICAL POWER SYSTEMS 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** in less than or equal to 200 seconds, and operate for at least 60 minutes. This test, if it is performed so it 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 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 30 days. Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.
| |
| *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.
| |
| ***If diesel generator started manually from the control room, 10 seconds after the automatic prelube period.
| |
| LIMERICK - UNIT 1 3/4 8-7a Amendment No. -32, 404, 186
| |
| | |
| TABLE 4.8.1.1.2-1 DIFSFI G.FNFRATOR TFST SCHFDIJI F NUMBER OF FAILURES IN LAST 20 VALID DEMANDS* TEST FREQUENCY less than or equal In accordance with the to 1 or, Surveillance Frequency Control Program greater than or In accordance with the equal to 2** Surveillance Frequency Control Program
| |
| *Criteria for determining the number of failures and number of valid demands shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis.
| |
| **The associated test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 demands have been reduced to one.
| |
| LIMERICK - UNIT I 3/4 8-8 Amendment No. -32, 186
| |
| | |
| 3892373290 PAM YaEMCIZ.:Y VEFT EMAN
| |
| | |
| ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimu, the following A.C. electrical power sources shall be OPERABLE:
| |
| : a. One circuit between the offslte transmission network and the onsite Class 1E distribution system,. and
| |
| : b. Two diesel generators each with:
| |
| : 1. A day fuel tank containing a minimum of 200 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. electri al power sources OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when Vn OPERATIONAL CONDITION S with the water level less than 22 feet above
| |
| ,the reactor pressure vessel flange, imeediately 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 EUIREMENTS 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 4.8.1.1.2, and 4.8.1.1.3.
| |
| I NWhn handling irradiated fuel in the secondary contauinent.,.
| |
| LIMERICK - UNIT 1 3/4 8-9 Aendmnt No. 32 SEP 28
| |
| .I
| |
| | |
| ELECTRICAL POWER SYSTEMS
| |
| .3/4.8.2 D.C. SOURCES D.C. SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical power sources shall be OPERABLE:
| |
| : a. Divisi.on 1, Consisting of:
| |
| : 1. 125-Volt Battery 1A1 (1AID101).
| |
| : 2. 125-Volt Battery 1A2 (1A2D101).
| |
| : 3. 125-Volt Battery Charger 1BCA1 (1AID103).
| |
| : 4. 125-Volt Battery Charger 1BCA2 (1A2D103).
| |
| : b. Division 2, Consisting of:
| |
| : 1. 125-Volt Battery IB1 (1BID101).
| |
| : 2. 125-Volt Battery 1B2 (1B2D1O1).
| |
| : 3. 125-Volt Battery Charger 1BCBI (1B1D103).
| |
| : 4. 125-Volt Battery Charger 1BCB2 (IB2DI03).
| |
| : c. Division 3, Consisting of:
| |
| : 1. 125-Volt Battery 1C (1CD101).
| |
| : 2. 125-Volt Battery Charger 1BCC (ICD103).
| |
| : d. Division 4, Consisting of:
| |
| : 1. 125-Volt Battery ID (1DD101).
| |
| : 2. 125-Volt Battery Charger IBCD (1DD103).
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one or two battery chargers on one division inoperable:
| |
| : 1. Restore battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours,
| |
| : 2. Verify associated Division 1 or 2 float current < 2 amps, or Division 3 or 4 float current < I 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 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) and restore battery float current to within limits within 18 hours.
| |
| LIMERICK - UNIT I 3/4 8-10 Amendment No. 164
| |
| | |
| ELECTRICAL POWER SYSTEMS LIMITING 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.
| |
| (*) 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.
| |
| LIMERICK - UNIT I 3/4 8-10a Amendment No. 164 I
| |
| | |
| THIS PAGE INTENTIONALLY LEFT BLANK 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 < 2 amps, and Division 3 and 4 battery float current is < 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 I 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 I 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)
| |
| IBCAI 300 1BCA2 300 1BCB1 300 1BCB2 300 1BCC 75 IBCD 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_, 49, 444186 norrccte hby letter- d,-t'-d Ju-lne 19, 1995
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILIANCE REQUIREMENTS (Continued)
| |
| : e. In accordance with the Surveillance Frequency Control Program by verifying 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 battery has reached 85% of expected life with battery capacity 100% of manufacturer's rating.
| |
| LIMERICK - UNIT 1 3/4 8-12 Amendment No. ;4, 92, 4-64,186 Corr..ted by letter dated June 19, 195
| |
| | |
| TABLE 4.8.2.1-1 (DELETED) I THE INFORMATION ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT 1 3/4 8-13 Amendment No. 4-2--, 164
| |
| | |
| FIFCTRICAl POWER SYSTEMS D.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two of the following four divisions of the D.C.
| |
| electrical power sources system shall be OPERABLE with:
| |
| : a. Division 1, Consisting of:
| |
| : 1. 125-Volt Battery 1A1 (1AID101).
| |
| : 2. 125-Volt Battery 1A2 (1A2D101).
| |
| : 3. 125-Volt Battery Charger 1BCA1 (IA1D103).
| |
| : 4. 125-Volt Battery Charger 1BCA2 (1A2D103).
| |
| : b. Division 2, Consisting of:
| |
| : 1. 125-Volt Battery 1B1 (1BiD101).
| |
| : 2. 125-Volt Battery 1B2 (1B2D1O1).
| |
| : 3. 125-Volt Battery Charger 1BCB1 (1B1D103).
| |
| : 4. 125-Volt Battery Charger 1BCB2 (1B2D103).
| |
| : c. Division 3, Consisting of:
| |
| : 1. 125-Volt Battery IC (1CD101).
| |
| : 2. 125-Volt Battery Charger 1BCC (1CD103).
| |
| : d. Division 4, Consisting of:
| |
| : 1. 125-Volt Battery 1D (1DD101).
| |
| : 2. 125-Volt Battery Charger 1BCD (1DD103).
| |
| APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and
| |
| * ACTION:
| |
| : a. With one or two required battery chargers on one required division inoperable:
| |
| : 1. Restorebattery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours,
| |
| : 2. Verify 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.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. 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/or b.,
| |
| Suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
| |
| : d. The provisions of Specification 3.0.3 are not applicable.
| |
| SURVEILLANCE REOUIREMENTS 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 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.
| |
| LIMERICK - UNIT I 3/4 8-14a Amendment No. 164 1
| |
| | |
| THIS PAGE INTENTIONALLY LEFT BLANK 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: DlI (1OAl15) b) 480-VAC Load Center: D114 (10B201)
| |
| C) 480-VAC Motor Control Centers: D114-R-C1 (10B219)
| |
| D114-R-C (10B213)
| |
| D114-R-G (10B211)
| |
| D114-R-GI (10B215)
| |
| D114-D-G (10B515) d) 120-VAC Distribution Panels: 10Y101 10Y206
| |
| : 2. Unit 1 Division 2, Consisting of: I a) 4160-VAC Bus: D12 (1OA116) b) 480-VAC Load Center: D124 (10B202) c) 480-VAC Motor Control Centers: D124-R-CI (10B220)
| |
| D124-R-C (10B214)
| |
| D124-R-G (10B212)
| |
| D124-R-G1 (IOB216)
| |
| D124-D-G (10B516) d) 120-VAC Distribution Panels: lOY102 10Y207
| |
| : 3. Unit 1 Division 3, Consisting of:
| |
| a) 4160-VAC Bus: D13 (10A117) b) 480-VAC Load Center: D134 (IOB203)
| |
| C) 480-VAC Motor Control Centers: D134-R-H1 (IOB221)
| |
| D134-R-H (10B217)
| |
| D134-R-E (10B223)
| |
| D134-C-B (00B131)
| |
| D134-D-G (10B517) d) 120-VAC Distribution Panels: IOY103 10Y163
| |
| : 4. Unit 1 Division 4, Consisting of:
| |
| a) 4160-VAC Bus: D14 (10A118) b) 480-VAC Load Center: D144 (108204)
| |
| Amendment No. 24, 131 LIMERICK - UNIT I 3/4 8-15 NOV 1 6 1998
| |
| | |
| ELECTRICAL POWER SYSTEMS ITMTTTNf~ CONIJITTON FOR OPERATION (Continued)
| |
| LIMITING CONOTTION FOR OPERATION (Continued) c) 480-VAC Motor Control Centers: D144-R-G (10B222)
| |
| D144-R-H (108218)
| |
| D144-R-E (10B224)
| |
| D144-C-B (OOB132)
| |
| D144-D-G (10B518) d) 120-VAC Distribution Panels: 10Y104 lOY164
| |
| : 5. Unit 2 and Common Division 1, Consisting of:
| |
| a) 4160-VAC bus: 021 (20A115) b) 480-VAC load center: 0214 (20B201) c) 480-VAC motor control centers: D114-S- L (008519) 0214-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: 022 (20A116) b) 480-VAC load center: 0224 (20B202) c) 480-VAC motor control centers: 0124-S-L (00B520) 0224-D-G (20B516) d) 20-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: 0234 (20B203) c) 480-VAC motor control centers: 0234-S-L (008521)
| |
| D234-D-G (20B517) d) 120-VAC distribution panels: 03Y501 20Y103 20Y163
| |
| : 8. Unit 2 and Common Division 4, Consisting of:
| |
| a) 4160-VAC bus: 024 (20A118) b) 480-VAC load center: DZ24 (20B204) c) 480-VAC motor control centers: 0244-S-L (008522)
| |
| D244-D-G (20B518) d) .20-VAC distribution panels: 04Y501 20Y104 20Y164 Amendment No. 24 " 2 2 18 LIMERICK - UNIT 1 3/4 8-16
| |
| | |
| ELECTRICAL POWER SYSTEMS LIkITING CONDITION FOR OPERATION (Continued)
| |
| : b. D.C. Power Distribution Panels
| |
| : 1. Unit I Division I, Consisting of:
| |
| a) 250-V DC Fuse Box: IFA (IAD105) b) 250-V DC Motor Control Center: IDA (10D201)
| |
| C) 125-V DC Distribution Panels: IPPAI (1ADI2)
| |
| ]PPA2 (lAD501)
| |
| IPPA3 (1AD162)
| |
| : 2. Unit I Division 2, Consisting of:
| |
| a) 250-V DC Fuse Box: 1FB (IBDI05) b) 250-V DC Motor Control Centers: IDB-i (10D202) 1DB-2 (IOD203) c) 125-V DC Distribution Panels: 1PPB1 (1BD]02)
| |
| IPPB2 (1BD501) 1PPB3 (1BD]62)
| |
| : 3. Unit I Division 3, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FC (ICD1O5) b) 125-V DC Distribution Panels: 1PPC1 (]CD102)
| |
| PPC2 (ICD501)
| |
| PPC3 (1CD162)
| |
| : 4. Unit I Division 4, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FD (1DD105) b) 125-V DC Distribution Panels: IPPD1 (IDD1O2) 1PPD2 (1DD501) 1PPD3 (IDD162)
| |
| S. 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 (2BDI02) 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: 2PPC1 (2CD102)
| |
| : 8. Unit 2 and Common Division 4, Consisting of: 2PPC2 (20501) a) 125-V DC Fuse Box: 2FD (2DD105) b) 125-V DC Distribution Panels: 2PPD1 (2DD102) 2PPD2 (2DD501)
| |
| LIMERICK - UNIT 1 3/4 8-16a Amendment No. Z4, 139 MAP 142000
| |
| | |
| 3892300520 THIS PAE INTD=TON=LY =EF BLAB
| |
| | |
| 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 REQUIREMENTS 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 I 3/4 8-17 Amendment No. 2-4, 186
| |
| | |
| 3892300520 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: D11 (10A115) b) 480-VAC Load Center: 0114 (108201) c) 480-VAC Motor Control Centers: 0114-R-Cl (108219) 0114-R-C (108213) 0114-R-G (108211)
| |
| I DU14-R-G1 (105215)
| |
| D114-D-G (108515) d) 120-VAC Distribution Panels: 1OYO1o 10Y206
| |
| : 2. Unit 1 Division 2, Consisting of:
| |
| a) 4160-VAC Bus: 012 (10A116) b) 480-VAC Load Center: 0124 (108202) c) 480-VAC Motor Control Centers: D124-R-C1 (108220)
| |
| D124-R-C (108214)
| |
| D124-R-G D124-R-G1 (108212)
| |
| (108216)
| |
| I D124-D-G (108516) d) 120-VAC Distribution Panels: 10Y102 10Y207
| |
| : 3. Unit I Division 3, Consisting of:
| |
| a) 4160-VAC Bus: 013 (10Afl7)
| |
| I b) 480-VAC Load Center: 0134 (105203) c) 480-VAC Motor Control Centers: 0134-R-Hl (108221)
| |
| D134-R-N (10B217)
| |
| D134-R-E (108223) 0134-C-B (005131) 0134-D-G (105517) d) 120-VAC Distribution Panels: 10Y103 10Y163 I
| |
| : 4. Unit I Division 4, Consisting of:
| |
| a) 4160-VAC Bus: 014 (1OA1lS)
| |
| II b) 480-VAC Load Center: 0144 (108204)
| |
| LIMERICK - UNIT 1 3/4 8-18 Amendment No. 24 I
| |
| juN 2 2 10
| |
| | |
| 3892300520 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) 0144-C-B (008132)
| |
| D144-D-G (10B518) I d) 120-VAC Distribution Panels: 10YI04 10Y164 I
| |
| : 5. Unit 2 and Common Division 1, Consisting of:
| |
| a) 4160-VAC Bus: 021 (20A115) b) 480-VAC Load Center: D214 (208201) c) 480-VAC Motor Control Centers: 0114-S-L (008519)
| |
| D214-R-C (20B213) 0214-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: 0124-S-L (008520)
| |
| D224-D-G (208516)
| |
| : d. 120-VAC Distribution Panels: 02Y501 20Y102 20Y207
| |
| : 7. Unit 2 and Common Divison 3, Consisting of:
| |
| : a. 4160-VAC Bus: 023 (20A117)
| |
| : b. 480-VAC Load Center: D234 (208203)
| |
| C. 480-VAC Motor Control Centers: D234-S-L (006521) 0234-D-G (208517)
| |
| : d. 120-VAC Distribution Panels: 03Y501 20Y103 20Y163
| |
| : 8. Unit 2 and Common Divison 4, Consisting of:
| |
| : a. 4160-VAC Bus: 024 (20A118)
| |
| : b. 480-VAC Load Center: 0224 (208204)
| |
| C. 480-VAC Motor Control Centers: D244-S-L (00B522)
| |
| D244-D-G (208518)
| |
| : d. 120-VAC Distribution Panels: 04Y501 20Y104 20Y164 LIMERICK - UNIT 1 3/4 8-18a JUN 2 2 1 Amendment No. 24 I
| |
| | |
| 3892300520 THIS PAGE INTENTIONALLY LEFT BLANK
| |
| | |
| ELECTRICAL POWER SYSTEMS I TM1T1MI~ rANflYTIAN Ffl~ APFRATTAN (rnntinu~d~
| |
| TL.Amu1I nT
| |
| -OtT" F, "(Continued)
| |
| : b. D.C. Power Distribution:
| |
| : 1. Unit I Division 1, Consisting of:
| |
| a) 250-V DC Fuse Box: 1FA (1AD1O5) b) 250-V DC Motor Control Center: 1DA (10D201) c) 125-V DC Distribution Panels: IPPA1 (1AD102) 1PPA2 (IAD5DO)
| |
| IPPA3 (IAD162)
| |
| : 2. Unit 1 Division 2, Consisting of:
| |
| a) 250-V DC Fuse Box: IFB (1BD105) b) 250-V DC Motor Control Centers: 1DB-i (10D202) 1DB-2 (10D203) c) 125-V DC Distribution Panels: IPPBI (IBDI]2)
| |
| IPPB2 (IBD501)
| |
| IPPB3 (1BDI62)
| |
| : 3. Unit I Division 3, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FC ()CD105) b) 125-V DC Distribution Panels: 1PPC1 (ICDIO2) j PPC2 (1CD501) 1PPC3 (1CO162)
| |
| : 4. Unit 1 Division 4, Consisting of:
| |
| a) 125-V DC Fuse Box: 1FD (1DD1O5) b) 125-V DC Distribution Panels: 1PPDI (IDDIO2) 1PPD2 (IDD501) 1PPD3 (1DD162)
| |
| : 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 (2ADSO1)
| |
| : 6. Unit 2 and Common Division 2, Consisting of:
| |
| a) 250-V DC Fuse Box: 2FB (2BDIOS) 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: 2PPC] (2CD102) 2PPC2 (2CD501)
| |
| : 8. Unit 2 and Common Division 4, Consisting of:
| |
| a) 125-V DC Fuse Box: 2FD (20D105) I b) 125-V DC Distribution Panels: 2PPDI (2DO102) 2PPD2 (2DD501)
| |
| .AERICK - UNIT I 3/4 8-19 Amendment No. 10, 139 MAR 14 200
| |
| | |
| 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, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
| |
| : b. With less than two divisions of the above required Unit 1 D.C. dis-tribution systems energized, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
| |
| : 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.
| |
| SIRVFII I ANC.F RFnII RFMFNTS 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 I 3/4 8-20 Amendment No. 2-4, 186
| |
| | |
| 3ection 3/4.8.4.1 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS BEEN RELOCATED TO THE UFSAR. TECHNICAL SPECIFICATIONS PAGES 3/4 8-21 THROUGH 3/4 8-26 OF THE SECTION HAVE BEEN INTENTIONALLY OMITTED.
| |
| LIMERICK - UNIT I 3/4 8-21 Amendment No. -2, 44, 74, 93 JUL 0 7 1995
| |
| | |
| PAGE INTENTIONALLY LEFT BLANK ELECTRICAL POWER SYSTEMS MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION LIMITING CONDITION FOR OPERATION 3.8.4.2 The thermal overload protection of all Class 1E motor operated valves shall be either:
| |
| : a. Continously bypassed for all valves with maintained position control switches; or,
| |
| : b. Bypassed only under accident conditions for all valves with spring-return-to-normal control switches.
| |
| APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE.
| |
| ACTION:
| |
| With the thermal overload protection for one or more of the above required valves not bypassed continuously or only under accident conditions, as applicable, restore the thermal overload bypass within 8 hours or declare the affected valve(s) inoperable and apply the appropriate ACTION statement(s) for the affected system(s).
| |
| SURVEILLANCE REOUIREMENTS 4.8.4.2.1 The thermal overload protection for the above required valves which are continuously bypassed and temporarily placed in force only when the valve motor is undergoing periodic or maintenance testing shall be verified to be bypassed following periodic or maintenance testing during which the thermal overload protection was temporarily placed in force.
| |
| 4.8.4.2.2 In accordance with the Surveillance Frequency Control Program, a CHANNEL FUNCTIONAL TEST of all those valves which are bypassed only under accident conditions (valves with spring- return-to-normal control switches) shall be performedto verify that the thermal overload protection will be bypassed under accident conditions.
| |
| LIMERICK - UNIT I 3/4 8-27 Amendment No. ;74_186
| |
| | |
| 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, undervoltage, 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 : 127.6 VAC,
| |
| : 2. Undervoltage Ž 110.7 VAC,
| |
| : 3. Underfrequency Ž 57.05 Hz.
| |
| LIMERICK - UNIT 1 3/4 8-28 Amendment No. -74, 4-,-4, 186
| |
| | |
| 3.4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH LIMITING CONDITION FOR OPERATION 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. 4-1-4, 149
| |
| | |
| 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 I 3/4 9-2 Amendment No. 444,186
| |
| | |
| 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 REOUIREMENTS 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, 49, 4-5.g, 186
| |
| | |
| REFUELING OPERATIONS 3/4 9.3 CONTROL ROD POSITION LIMITING CONDITION FOR OPERATION 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.
| |
| *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. -49,186
| |
| | |
| 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.
| |
| APPLICABIL*:' 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 subcrlticallty prior to movement of Irradiated fuel in the reactor pressure vessel.
| |
| LIMERICK - UNIT 1 3/4 9-6 Rut 8 il
| |
| | |
| 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 REQUIREMENTS 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.*
| |
| *Except movement of control rods with their normal drive system.
| |
| LIMERICK - UNIT 1 3/4 9-7 Amendment No. 4-O9, 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 REOUIREMENTS 4.9.6.1 The refueling platform main hoist used for handling of fuel assemblies 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 normal water level.
| |
| LIMERICK - UNIT 1 3/4 9-8 Amendment No. 43 1
| |
| | |
| REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 4.9.6.2 The refueling platform frame-mounted auxiliary hoist used for handling of contfol 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 t 50 pounds.
| |
| : b. Demonstrating operation of the uptravel mechanical stop on the frame 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. I 4.9.6.3 The refueling platform monorail mounted auxiliary hoist used for handling of control rods within the reactor pressure vessel shall be 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 t 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 I
| |
| level.
| |
| I LIMERICK - UNIT 1 3/4 9-9 Amendment No. 43 AM 16 a I
| |
| | |
| REFUELING OPERATIONS 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 inthe 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 crane operation.
| |
| LIMERICK - UNIT 1 3/4 9-10 Amendment No.186
| |
| | |
| REFUELING OPERATIONS 3/4.9.8 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION 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. 49, 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.
| |
| SURVFILLANCE 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 disarmedor 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 SURVEILLANCE REQUIREMENTS 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 No186
| |
| | |
| 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 SURVEILLANCF REQUIREMENTS 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 REOUIREMENTS 4.9.11.1 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.
| |
| The required RHR shutdown cooling subsystem may be removed from operation for up to two (2) hours per eight (8) hour period.
| |
| LIMERICK - UNIT I 3/4 9-17 Amendment No. 4-7, 44-1,186
| |
| | |
| 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 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.
| |
| 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 3/4 9-18 Amendment No. 44, 186
| |
| | |
| 3/4.10 SPECIAL TEST EXCEPTIONS 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'F, immediately place the reactor mode switch in the Shutdown position.
| |
| SURVEILLANCE REQUIREMENTS 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
| |
| | |
| 3893087830 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 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 when THERMAL POWER is less than or equal to i0% of RATED THERMAL POWER.
| |
| ACTION:
| |
| With the requirements of the above specifications not satisfied, verify that the RWM is OPERABLE per Specification 3.1.4.1.
| |
| SURVEI LLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed by the RWM are bypassed, verify;
| |
| : a. That movement of control rods is blocked or limited to the approved control rod withdrawal sequence during scram and friction tests. l
| |
| : 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 NU 2 2 1!"
| |
| | |
| 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 REOUIREMENTS 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
| |
| | |
| SPFCIAI TFST FXCFPTIONS 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 REQUIREMENTS 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 I 3/4 10-4 Amendment No. 186
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/410.5 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.
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| APPLICABILITY: OPERATIONAL CONDITION 1.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, be in at least STARTUP within 6 hours.
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| SURVEILLANCE REQUIREMENTS 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 I 3/4 10-5 Amendment No. 186
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| | |
| 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'F.
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| APPLICABILITY: OPERATIONAL CONDITION 2, during training startups.
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| ACTION:
| |
| With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position.
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| SURVEILLANCE REQUIREMENTS 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 training startups.
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| LIMERICK - UNIT 1 3/4 10-6 Amendment NoJ86
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| SPECIAL TEST EXCEPTIONS 3/4.10.7 RESERVED-CURRENTLY NOT USED LIMERICK - UNIT I 3/4 10-7 Amendment No. 133 FEB 1.1 1999
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| | |
| 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 212°F, 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:
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| : 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;
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| : c. 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY;
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| : 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.
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| APPLICABILITY: OPERATIONAL CONDITION 4, with average reactor coolant temperature greater than 200°F and less than or equal to 2127.
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| ACTION:
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| With the requirements of the above Specifications not satisfied:
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| : 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 200°F or less within 24 hours.
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| SURVEILLANCE REQUIREMENTS 4.10.8 Verify applicable OPERATIONAL CONDITION 3 surveillances for the Specifications listed in 3.10.8 are met.
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| LIMERICK - UNIT 1 3/4 10-8 Amendment No. 133 I FEB I 1 1999
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| | |
| Section 3/4 11.1.1 through 3/4 11.2.4 (Deleted)
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| 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.
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| LIMERICK - UNIT 1 3/4 11-1 Amendment No. ?,48 JAN 02 1991 I
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| THIS PAGE INTEN~TIONALLY LEFT BLANK 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.
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| : b. The provisions of Specification 3.0.3 are not applicable.
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| 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.
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| LIMERICK - UNIT 1 3/4 11-7 Amendment No. 44_,;-a, 186
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| | |
| Section 3/4 22.2.1 through Section 3/4 11.2.4 (Deleted) .I THE INFORMiTION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 ni-9 THROUGH 3/4 11-14 OF THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED.
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| LIMERICK - UNIT 1 3/4 U-8 Amendment No. g,48 JAN 02 1531 I
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| RADIOACTIVE EFFLUENTS 38391002S60 EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of hydrogen in the main condenser offgas treatment system shall be limited to less than or equal to 4% by volume.
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| APPLICABILITY: Whenever the main condenser air ejector system is ir; operation.
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| ACTION:
| |
| : a. With the concentration of hydrogen in the main condenser offgas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours.
| |
| : b. The provisions of Specification 3.0.3 are not applicable. I SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen in'the main condenser offgas treatment system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser offgas treatment system with the hydrogen monitors required OPERABLE by Table 3.3.7.12-1 of Specification 3.3.7.12.
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| LIMERICK - UNIT 1 3/4 U-iS Amendment No. ]1 NovIOU1 I
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| | |
| RADIOACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION 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*.
| |
| ACTION:
| |
| 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.
| |
| SURVFILLANCF REOUIREMENTS 4.11.2.6.1 The rate of the sum of the activities of noble gases at the recombiner after-condenser discharge shall be continously monitored in accor-dance with Specification 3.3.7.12.
| |
| 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. I
| |
| : b. Within 4 hours following an increase, as indicated by the Main Condenser Off-Gas Pretreatment Radioactivity Monitor, 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 coolant.
| |
| : c. The provisions of Specification 4.0.4 are not applicable.
| |
| *When the main condenser air ejector is in operation.
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| LIMERICK - UNIT I 3/4 11-16 Amendment No. 186
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| | |
| Section 3/4 11-2.7 (Deleted)
| |
| THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM.
| |
| Amendment No. )e-48 LIMERICK - UNIT 1 3/4 11-17 JAN 02 1991 I
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| | |
| Section 3/4 11.3 and Section 3/4 11.4 (Deleted)
| |
| I THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE PCP OR 0OCK. TECHNICAL SPECIFICATIONS PAGES 3/4 11-19 THROUGH 3/4 11-20 OF THESE SECTIONS HAVE BEEN INTENTIONALLY OMITTED.
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| LIMERICK - UNIT 1- 3/4 11-18 Amendment No. ,]48 I JAK 02 1991
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| | |
| 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.
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| LIMERICK - UNIT 1 3/4 12-1 Amendment No. 48 JAN 02 1991 I
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| | |
| THIS PAGE INTENTIONALLY LEFT BLANK BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS AUG 8 1985
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| | |
| INTENTIONALLY LEFT BLANK NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Sections 3.0 and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.
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| AUG 8 1985
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| | |
| 3890020620 THIS PAGE INTENTIONALLY LEFT BLANK
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| | |
| 3/4.0 APPLICABILITY BASES 3891002560 Specifications 3.0.1 through 3.0.4 establish the general requirements applicable to Limiting Conditions for Operation. These requirements are based on the requirements for Limiting Conditions for Operation stated in the Code of Federal Regulations, 10 CFR 50.36(c)(2):
| |
| "Limiting Conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation oY the facility. When a limiting condition for operation of a nuclear'reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification until the condition can be met."
| |
| Specification 3.0.1 establishes the Applicability statement within-each individual specification as the requirement for when (i.e., in which OPERATIONAL CONDITIONS or other specified conditions) conformance to the Limiting Conditions for Operation is required for safe operation of the facility. The ACTION requirements establish those remedial measures that must be taken within specified time limits when the requirements of a Limiting Condition for Operation are not met. It is not intended that the shutdown ACTION requirement be used as an operational convenience which permits (routine) voluntary removal of a system(s) or component(s) from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.
| |
| There are two basic types of ACTION requirements. The first specifies the remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements. In this case, conformance to the ACTION requirements provides an acceptable level of safety for unlimited continued operation as long as the ACTION requirements continue to be met. The second type of ACTION requirement specifies a time limit in which conformance to the conditions of the Limiting Condition for Operation must be met. This time limit is the allowable outage time to restore an inoperable system or component to OPERABLE status or for restoring parameters within specified limits. If these actions are not completed within the allowable outage time limits, a shutdown is required to place the facility in an OPERATIONAL CONDITION or other specified condition in which the specifi-cation no longer applies.
| |
| The specified time limits of the ACTION requirements are applicable from the point of time it is identified that a Limiting Condition for Operation is not met. The time limits of the ACTION requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational problems. Individual specifications may include a specified time limit for the completion of a Surveillance Requirement when equipment is removed from service. In this case, the allowable outage time limits of the ACTION requirements are applicable when this limit expires if the surveillance has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered an OPERATIONAL CONDITION in which a new specification becomes applicable. In this case, the time limits of the ACTION requirements would apply from the point in time that the new specification becomes applicable if the requirements of the Limiting Condition for Operation are not met.
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| LIMERICK - UNIT 1 B 3/4 0-1 Amendment No. I1 HOV 7 19W
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| | |
| 3891002560 3/4.0 APPLICABILITY BASES Specification 3.0.2 establishes that noncompliance with a specification exists when the requirements of the Limiting Condition for Operation are not met and the associated ACTION requirements have not been implemented within the specified time interval. The purpose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time inter-val constitutes compliance with a specification and (2) completion of. the remedial measures of the ACTION requirements is not required when compliance with a Limiting Condition of Operation is restored within the time interval specified in the associated ACTION requirements.
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| Specification 3.0.3 establishes the shutdown ACTION requirements that must be ipplemented when a&limiting Conditionfor Operation iszmot:met and~the con-dition is not specifically addressed by the associated ACTION requireme-ts.
| |
| The purpose of this specification is to delineate the time limits for placing the unit in a safe shutdown CONDITION when plant operation cannot be maintained within the limits for safe operation defined by the Limiting Conditions for Operation and its ACTION requirements. It is not intended to be used as an operational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to prepare for an orderly shutdown before initiating a change in plant operation. This time permits the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower CONDITIONS of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the primary coolant system and the potential for a plant upset that could challenge safety systems under conditions for which this specification applies.
| |
| If remedial measures permitting limited continued operation of the facility under the provisions of the ACTION requirements are completed, the shutdown may be terminated. The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition for Operation. Therefore, the shutdown may be terminated if the ACTION require-ments have been met or time limits of the ACTION requirements have not expired, thus providing an allowance for the completion of the required actions.
| |
| The time limits of Specification 3.0.3 allow 37 hours for the plant to be in COLD SHUTDOWN when a shutdown is required during POWER operation. If the plant is in a lower CONDITION of operation when a shutdown is required, the time limit for reaching the next lower CONDITION of operation applies. However, if a lower CONDITION of operation is reached in less time than allowed, the total allowable time to reach COLD SHUTDOWN, or other OPERATIONAL CONDITION, is not reduced. For example, if STARTUP is reached in 2 hours, the time allowed to reach HOT SHUTDOWN is the next 11 hours because the total time to reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours. Therefore, if remedial measures are completed that would permit a return to POWER operation, a penalty is not incurred by having to reach a lower CONDITION of operation in less than the total time allowed.
| |
| LIMERICK - UNIT 1 B 3/4 0-2 Amendment No. 11 NOV 7 1sa&
| |
| | |
| 3/4_n APPLICABILITY BASES The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACTION requirements for one specification results in entry into an OPERATIONAL CONDITION or condition of operation for another specification in which the requirements of the Limiting Condition for Operation are not met. If the new specification becomes appli-cable in less time than specified, the difference may be added to the allowable outage time limits of the second specification. However, the allowable outage time limits of ACTION requirements for a higher CONDITION of operation may not be used to extend-the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lower CONDITION of-operation.
| |
| The shutdown requirements of Specification 3.0.3 do not apply in CONDITIONS 4 and 5, because the ACTION requirements of individual specifications define the remedial measures to be taken.
| |
| Specification 3.0.4 establishes limitations on changes in OPERATIONAL CONDITIONS or other specified conditions in the Applicability when a Limiting Condition for Operation is not met. It allows placing the unit in an OPERATIONAL CONDITION or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the Limiting Condition for Operation would not be met, in accordance with Specification 3.0.4.a, Specification 3.0.4.b, or Specification 3.0.4.c.
| |
| Specification 3.0.4.a allows entry into an OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not met 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. Compliance with ACTION requirements that permit continued operation of the unit for an unlimited period of time in an OPERATIONAL CONDITION or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the OPERATIONAL CONDITION change. Therefore, in such cases, entry into an OPERATIONAL CONDITION or other specified condition in the Applicability may be made in accordance with the provisions of the ACTION requirements.
| |
| Specification 3.0.4.b allows entry into an OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not met 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.
| |
| The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires that risk impacts of maintenance activities be assessed and managed. The risk assessment, for the purposes of Specification 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These documents address general guidance for conduct of the risk LIMERICK - UNIT 1 B 3/4 0-3 Amendment No. 41-, -621, 169
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| | |
| 3/4.0 APPLICABILITY BASES assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed OPERATIONAL CONDITION change is acceptable.
| |
| Consideration should also be given to the probability of completing restoration such that the requirements of the Limiting Condition for Operation would be met prior to the expiration of the ACTION requirement's specified time interval that would require exiting the Applicability.
| |
| Specification 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.
| |
| The results of the risk assessment shall be considered in determining the acceptability of entering the OPERATIONAL CONDITION or other specified condition in the Applicability, and any corresponding risk management actions. The Specification 3.0.4.b risk assessments do not have to be documented.
| |
| The Technical Specifications allow continued operation with equipment unavailable in OPERATIONAL CONDITION 1 for the duration of the specified time interval. Since this is allowable, and since in general the risk impact in that particular OPERATIONAL CONDITION bounds the risk of transitioning into and through the applicable OPERATIONAL CONDITIONS or other specified conditions in the Applicability of the Limiting Condition for Operation, the use of the Specification 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above. However, there is a small subset of systems and components that have been determined to be more important to risk and use of the Specification 3.0.4.b allowance is prohibited. The Limitihg Condition for Operations governing these system and components contain Notes prohibiting the use of Specification 3.0.4.b by stating that Specification 3.0.4.b is not applicable.
| |
| Specification 3.0.4.c allows entry into a OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not met based on a Note in the Specification which states Specification 3.0.4.c is applicable. These specific allowances permit entry into OPERATIONAL CONDITIONS or other specified conditions in the Applicability when the associated ACTION requirements to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTION requirements or to a specific ACTION requirement of a Specification. The risk assessments performed to justify the use of Specification 3.0.4.b usually only consider systems and components. For this reason, Specification 3.0.4.c is typically applied to Specifications which describe values and parameters (e.g., Reactor Coolant Specific Activity), and may be applied to other Specifications based on NRC plant-specific approval.
| |
| The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated OPERATIONAL CONDITION or other specified condition in the Applicability.
| |
| The provisions of Specification 3.0.4 shall not prevent changes in OPERATIONAL CONDITIONS or other specified conditions in the Applicability that are required to comply with ACTION requirements. In addition, the provisions of Specification 3.0.4 shall not prevent changes in OPERATIONAL CONDITIONS or other specified LIMERICK - UNIT I B 3/4 0-3a Amendment No. .1-1, 4-62*, 169 I
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| | |
| 3/4.0 APPLICABILITY BASES conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in OPERATIONAL CONDITION or other specified condition in the Applicability associated with transitioning from OPERATIONAL CONDITION 1 to OPERATIONAL CONDITION 2, OPERATIONAL CONDITION 2 to OPERATIONAL CONDITION 3, and OPERATIONAL CONDITION 3 to OPERATIONAL CONDITION 4.
| |
| Upon entry into an OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not met, Specification 3.0.1 and Specification 3.0.2 require entry into the applicable Conditions and ACTION requirements until the Condition is resolved, until the Limiting Condition for Operation is met, or until the unit is not within the Applicability of the Technical Specification.
| |
| Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by Specification 4.0.1. Therefore, utilizing Specification 3.0.4 is not a violation of Specification 4.0.1 or Specification 4.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected Limiting Condition for Operation.
| |
| Specification 4.0.1 through 4.0.5 establish the general requirements applicable to Surveillance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations 10 CFR 50.36(c)(3):
| |
| "Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met."
| |
| Specification 4.0.1 establishes the requirement that SRs must be met during the OPERATIONAL CONDITIONS or other specified conditions in the Applicability for which the requirements of the Limiting Condition for Operation apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Surveillance time interval and allowed extension, in accordance with Specification 4.0.2, constitutes a failure to meet the Limiting Condition for Operation.
| |
| Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:
| |
| : a. The systems or components are known to be inoperable, although still meeting the SRs; or
| |
| : b. The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.
| |
| Surveillances do not have to be performed when the unit is in an OPERATIONAL CONDITION or other specified condition for which the requirements of the associated Limiting Condition for Operation are not applicable, unless otherwise specified. The SRs associated with a Special Test Exception Limiting Condition for Operation are only applicable when the Special Test Exception Limiting Condition for Operation is used as an allowable exception to the requirements of a Specification.
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| LIMERICK - UNIT 1 B 3/4 0-3b Amendment No. 44, 449, 169 I
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| | |
| 3/4.0 APPLICABILITY BASES Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given OPERATIONAL CONDITION or other specified condition.
| |
| Surveillances, including Surveillances invoked by ACTION requirements, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with Specification 4.0.2, prior to returning equipment to OPERABLE status.
| |
| Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with Specification 4.0.2. Post maintenance testing may not be possible in the current OPERATIONAL CONDITION or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to an OPERATIONAL CONDITION or other specified condition where other necessary post maintenance tests can be completed.
| |
| Some examples of this process are:
| |
| : a. Control Rod Drive maintenance during refueling that requires scram testing at > 950 psi. However, if other appropriate testing is satisfactorily completed and the scram time testing of Specification 4.1.3.2 is satisfied, the control rod can be considered OPERABLE.
| |
| This allows startup to proceed to reach 950 psi to perform other necessary testing.
| |
| : b. High pressure coolant injection (HPCI) maintenance during shutdown that requires system functional tests at a specified pressure.
| |
| Provided other appropriate testing is satisfactorily completed, startup can proceed with HPCI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.
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| LIMERICK - UNIT I B 3/4 0-3c Amendment No. 4-_,449, 169 I
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| | |
| THIS PAGE INTENTIONALLY LEFT BLANK 3/4.0 APPLICABILITY BASES Specification 4.0.2 establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveil-lance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with an 24-month surveillance interval. It is not intended that this provision be used repeatedly as a convenience to extend the surveillance intervals beyond that specified for surveillances that are not performed during refueling outages.
| |
| Likewise, it is not the intent that REFUELING INTERVAL surveillances be performed during power operationunless it is consistent with safe plant operation. The limitation of Specification 4.0.2 is based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.
| |
| Specification 4.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Surveillance time interval and allowed extension. A delay period of up to 24 hours or up to the limit of the specified Surveillance time interval, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with Specification 4.0.2, and not at the time that the specified Surveillance time interval and allowed extension was not met.
| |
| This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with ACTION requirements or other remedial measures that might preclude completion of the Surveillance.
| |
| The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements. When a Surveillance with a Surveillance time interval based not on time interyals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering OPERATIONAL CONDITION 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to have not been performed when specified, Specification 4.0.3 allows for the full delay period of up to the specified Surveillance time interval to perform the Surveillance. However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.
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| Specification 4.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of OPERATIONAL CONDITION changes imposed by ACTION requirements.
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| Failure to comply with specified Surveillance time intervals and allowed extensions for SRs is expected to be an infrequent occurrence. Use of the delay period established by Specification 4.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.
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| LIMERICK - UNIT I B 3/4 0-4 Amendment No. *4, 38, -74, 162
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| | |
| 3/4.0 APPLICABILITY BASES While up to 24 hours or the limit of the specified Surveillance time interval is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determina-tion of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance. This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, 'Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants.' This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk managementaction thresholds, and risk management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the Corrective Action Program.
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| If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the ACTION requirements for the applicable Limiting Condition for Operation begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period or the variable is outside the specified limits, then the equipment is inoperable and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.
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| Completion of the Surveillance within the delay period allowed by this Specification, or within the allowed times specified in the ACTION requirements, restores compliance with Specification 4.0.1.
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| Specification 4.0.4 establishes the requirement that all applicable SRs must be met before entry into an OPERATIONAL CONDITION or other specified condition in the Applicability.
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| This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into OPERATIONAL CONDITIONS or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an assoc ated OPERATIONAL CONDITION or other specified condition in the Applicability.
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| A provision is included to allow entry into an OPERATIONAL CONDITION or other specified condition in the Applicability when a Limiting Condition for Operation is not met due to a Surveillance not being met in accordance with Specification 3.0.4.
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| However, in certain circumstances, failing to meet an SR will not result in Specification 4.0.4 restricting an OPERATIONAL CONDITION change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated LIMERICK - UNIT I B 3/4 0-5 Amendment No. 44, 4-9, 42-5, , 169
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| 3/4.0 APPLICABILITY BASES SR(s) are not required to be performed, per Specification 4.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, Specification 4.0.4 does riot apply to the associated SR(s) since
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| *the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) withi the specified Surveillance time interval does not result in a Specification 4.0.4 restriction to changing OPERATIONAL CONDITIONS or other specified conditions of the Applicability. However,, since the Limiting Condition for Operation is not met in this instance, Specification 3.0.4 will govern any restrictions that may (or may not) apply to OPERATIONAL CONDITION or other specified condition changes. Specification 4.0.4 does not restrict changing OPERATIONAL CONDITIONS or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Surveillance time interval, provided the requirement to declare the Limiting Condition for Operation not met has been delayed in accordance with Specification 4.0.3.
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| The provisions of Specification 4.0.4 shall not prevent entry.into OPERATIONAL CONDITIONS or other specified conditions in the Applicability that are required to comply with ACTION requirements. In addition, the provisions of Specification 4.0.4 shall not prevent changes in OPERATIONAL CONDITIONS or other-specified conditions in the Applicability that result from any unitshutdown. In this context, a unit shutdown is defined as a change in OPERATIONAL CONDITION or other specified condition in the Applicability associated with transitioning from OPERATIONAL CONDITION 1 to OPERATIONAL CONDITION 2, OPERATIONAL CONDITION 2 to OPERATIONAL CONDITION 3, and OPERATIONAL CONDITION 3 to OPERATIONAL CONDITION 4.
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| Specification 4.0.5 establishes the requirement that inservice inspection of.ASME' Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Additionally, the Inservice Inspection Program conforms to the NRC staff oositions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austinetic Stainless Steel Piping," as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990, or in accordance with alternate measures approved by the NRC staff.
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| This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section X1 of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarifica-tion is provided to ensure consistency in surveillance intervals throughout the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.
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| Under the terms of this specification, the more restrictive requirements of the Technical Specifications take Drecedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. The requirements of Specification 4.0.4 to perform surveillance activities before entry into an OPERATIONAL CONDITION or other specified condition takes precedence over the ASME Boiler and Pressure Vessel Code provision that allows pumps and valves to be tested up to one week after return to normal operation. The Technical Specification definition of OPERABLE does not allow a grace period before a component, which is not capable of performing its specified function, is declared inoperable and takes prece-dence over the ASME Boiler and Pressure Vessel Code provision that allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable.
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| LIMERICK - UNIT I B 3/4 0-6 Amendment No. 44, 4-9, 171 1-6-9,
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| 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
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| Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.38%a k/k or R + 0.28%A k/k, as appropriate.
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| The 0.38%a k/k includes uncertainties and calculation biases. The value of R in units of Z A k/k is the difference between the calculated value of minimum shutdown margin during the operating cycle and the calculated shutdown margin at the time of the shutdown margin test at the beginning of cycle. The value of R must be positive or zero and must be determined for each fuel loading cycle.
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| Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.
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| The highest worth rod may be determined analytically or by test.' The SHUTDOWN MARGIN is demonstrated by (an insequence) control rod withdrawal at the beginning of life fuel cycle conditions, and. if necessary, at any.future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure. Observation of subcriticality in this condition assures subcriticality with the most reactive control rod fully withdrawn.
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| This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.
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| 3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns.
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| Since the comparisons are easily done, frequent checks are not an imposition on normal operations. A 1% change* is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated. A change as large as IX would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.
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| LiMtKI6& -
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| .........
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| ........
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| URIUI
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| ,
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| 0 J/
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| * 1_1 AU6 8 95
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| REACTIVITY CONTROL SYSTEMS
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| * BASES 3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
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| Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control 'rods.
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| Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.
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| The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.
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| The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the fuel cladding safety limit during the limiting power transient analyzed in Section 15.2 of the FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifi-cations, provided the required protection and MCPR remains greater than the fuel cladding safety limit. The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.
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| Scram time testing at zero psig reactor coolant pressure is adequate to ensure that the control rod will perform its intended scram function during startup of the plant until scram time testing at 950 psig reactor coolant pressure is performed prior to exceeding 40% rated core thermal power.
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| The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.
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| The OPERABILITY of all SDV vent and drain valves ensures that the SDV vent and drain valves will close during a scram to contain reactor water discharged to the SDV piping. The SDV has one common drain line and one common vent line. Since the vent and drain lines are provided with two valves in series, the single failure of one valve in LIMERICK - UNIT I B 3/4 1-2 Amendment No. , 9, 168
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| REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) the open position will not impair the isolation function of the system. Additionally, the valves are required to open on scram reset to ensure that a path is available for the SDV piping to drain freely at other times.
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| When one SDV vent or drain valve is inoperable in one or more lines, the valves must be restored to OPERABLE status within 7 days. The allowable outage time is reasonable, given the level of redundancy in the lines and the low probability of a scram occurring while the valve(s) are inoperable. The SDV is still isolable since the redundant valve in the affected line is OPERABLE. During these periods, the single failure criterion may not be preserved, and a higher risk exists to allow reactor water out of the primary system during a scram.
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| If both valves in a line are inoperable, the line must be isolated to contain the reactor coolant during a scram. When a line is isolated, the potential for an inadvertent scram due to high SDV level is increased. ACTION "e" is modified by a note ("****") that allows periodic draining and venting of the SDV when a line is isolated. During these periods, the line may be unisolated under administrative control. This allows any accumulated water in the line to be drained, to preclude a reactor scram on SDV high level. This is acceptable since the administrative controls ensure the valve can be closed quickly, by a dedicated operator, if a scram occurs with the valve open. The 8 hour allowable outage time to isolate the line is based on the low probability of a scram occurring while the line is not isolated and the unlikelihood of significant CRD seal leakage.
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| Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. The drive water pressure normal operating range is specified in system operating procedures which provide ranges for system alignment and control rod motion (exercising). Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor. A control rod is considered trippable if it is capable of fully inserting as a result of a scram signal.
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| LIMERICK - UNIT 1 B 3/4 1-2a Amendment No. 30, 9-, 4168, 178
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| THIS PAGE INTENTIONALLY LEFT BLANK REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)
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| Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving' criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demon-stration.
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| In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
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| The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
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| The required surveillances are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
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| 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm. Thus requiring the RWM to be OPERABLE when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate control.
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| The RWM provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
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| The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3. Additional pertinent analysis is also contained in Amendment 17 to the Reference 4 topical report.
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| The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density over the range of power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods. RBM OPERA-BILITY is required when the limiting condition described in Specification 3.1.4.3 exists.
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| LIMERICK - UNIT 1 B 3/4 1-3 Amendment No. 4-4, 66186
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| REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that-the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. To allow for potential leakage and improper mixing, this concentration is increased by 25%. The required concentration is achieved by having available a minimum quantity of 3,160 gallons of sodium pentaborate solution containng a minimum of 3,754 lbs of sodium pentaborate having the requisite Boron-t0 atom % enrichment of 29% as determined from Reference 5.
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| This quantity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the portion which cannot be injected.
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| The above quantities calculated at 29% Boron-lO enrichment have been demonstrated by analysis to provide a Boron-lO weight equivalent of 185 lbs in the sodium pentaborate solution. Maintaining this Boron-lO weight in the net tank contents ensures a sufficient quantity of boron to bring the reactor to a cold, Xenon-free shutdown.
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| The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution.
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| With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
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| The SLCS system consists of three separate and independent pumps and
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| ',1 osive valves. Two of the separate and independent pumps and explosivve valves are required to meet the minimum requirements of this technical specification and, where applicable, satisfy the single failure criterion.
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| The SLCS must have an equivalent control capacity of 86 gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants). As part of the ARTS/MELLL program the ATWS analysis was updated to reflect the new rod line. As a result of this it was determined that the Boron 10 enrichment was required to be increased to 29% to prevent exceeding a suppression pool temperature of 1901F. This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.
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| The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.
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| LIMERICK - UNIT 1 B 3/4 1-4 Amendment No. 2-9, 6,185
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| REACTIVITY CONTROL SYSTEMS BASES STANDBY LIQUID CONTROL SYSTEM (Continued)
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| Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume assures that the solution is available for use.
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| Replacement of the explosive charges in the valves will assure that these valves will not fail because of deterioration of the charges.
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| The Standby Liquid Control System also has a post-DBA LOCA safety function to buffer Suppression Pool pH in order to maintain bulk pH above 7.0. The buffering of Suppression Pool pH is necessary to prevent iodine re-evolution to satisfy the methodology for Alternative Source Term. Manual initiation is used, and the minimum amount of total boron required for Suppression Pool pH buffering is 240 lbs. Given that at least 185 lbs of Boron-lD is maintained in the tank, the total boron in the tank will be greater than 240 lbs for the range of enrichments from 29% to 62%.
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| ACTION Statement (a) applies only to OPERATIONAL CONDITIONS 1 and 2 because a single pump can satisfy both the reactor control function and the post-DBA LOCA function to control Suppression Pool pH since boron injection is not required until 13 hours post-LOCA. ACTION Statement (b) applies to OPERATIONAL CONDITIONS 1, 2 and 3 to address the post-LOCA safety function of the SLC system.
| |
| : 1. C. J. Paone, R. C. Stirn and J. A. Woolley, "Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NEDO-10527, March 1972.
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| : 2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NEDO-10527, July 1972.
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| : 3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, "Exposed Cores."
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| Supplement 2 to NEDO-10527, January 1973.
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| : 4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel".
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| : 5. "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," NEDC-32193P, Revision 2, October 1993.
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| LIMERICK - UNIT 1 B 3/4 1-5 Amendment No. 24, 66, 5, 186
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| PAGE INTENTIONALLY LEFT BLANK 3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the fuel design analysis limits specified-in NEDE-24011-P-A (Reference 2) will not be exceeded.
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| Mechanical Design Analysis: NRC approved methods (specified in Refer-ence 2) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 2.
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| No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is usec as the basis for the fuel design analysis MAPLHGR limit.
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| LOCA Analysis: A LOCA analysis is performed in accordance with 10 CFR 50 Appendix K to demonstrate that the permissible planar power (MAPLHGR) limits comply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant, using the evaluation model described in Reference 9.
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| The MAPLHGR limit as showm in the CORE OPERATING LIMITS REPORT is the most limiting composite of the fuel mechanical design anaylsis MAPLHGR and the ECCS MAPLHGR limit.
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| Only the most limiting MAPLHGR values are shown in the CORE OPERATING LIMITS REPORT for multiple lattice fuel. Compliance with the specific lattice MAPLHGR operating limits, which are available in Reference 3, is ensured by use of the process computer.
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| As a result of no longer utilizing an APRM trip setdown requirement, additional constraints are placed on the MAPLHGR limits to assure adherence to the fuel-mechanical design bases. These constraints are introduced through the MAPFAC(P) and MAPFAC(F) factors as defined in the COLR.
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| B 3/4 2-1 Amendment No. 7, 30, 37, 66 LIMERICK - UNIT I FEB 10 1994
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| POWER DISTRIBUTION LIMITS BASES 3/4.2.2 (DELETED)
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| INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED LIMERICK - UNIT I B 3/4 2-2 Amendment No. 7, 30, 66 FEB 10 1994
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| LEFT IMTViT@Ay sLMw LWDh1CK - 11M1 I 1 3/4 %.3
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| ~Am a st Ms. I AN6 1 4 1147
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| POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.
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| To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
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| The evaluation of a given transient begins with the system initial para-meters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate transients are discussed in Reference 2.
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| The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR(F),. and MCPR(P), respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Ref. 6). Flow dependent MCPR limits (MCPR(F)) are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 7) to analyze slow flow runout transients.
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| Power dependent MCPR limits (MCPR(P)) are determined mainly by the one dimensional transient code (Ref. 8). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPR(P), operating limits are provided for operating between 25% RTP and 30% RTP.
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| The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPR(F), and MCPR(P) limits.
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| JUN 0 3 1999 LIMERICK - UNIT 1 B 3/4 2-4 Amendment No. -7,4, 30, P, ECR LG. 99-01138
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| POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
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| At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void-eontent will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The require-ment for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.
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| 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.
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| | |
| ==Reference:==
| |
| : 1. Deleted.
| |
| : 2. "General Electric Standard Application for Reactor Fuel,"
| |
| NEDE-24011-P-A (latest approved revision).
| |
| : 3. "Basis of MAPLHGR Technical Specifications for Limerick Unit I,"
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| NEDO-31401, February 1987 (as amended).
| |
| : 4. Deleted
| |
| : 5. Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit I Cycle 1, NEDC-31323, October 1986 including Errata and Addenda Sheet No. I dated November 6, 1986.
| |
| : 6. NEDC-32193P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units I and 2," Revision 2, October 1993.
| |
| : 7. NEDO-30130-A, "Steady State Nuclear Methods," May 1985.
| |
| : 8. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, "October 1978.
| |
| : 9. NEDC-32170P, "Limerick Generating Station Units I and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," June 1993.
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| ITMEIII TT 1 0 1*A I-- Amendment No. 7, 19, p7, 66 I
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| LI. rI.1.* ,NI -- I ,, ,iL FEB 10 1994
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| PAGE INTENTIONALLY LEFT BLANK 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
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| : a. Preserve the integrity of the fuel cladding.
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| : b. Preserve the integrity of the reactor coolant system.
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| : c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and
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| : d. Prevent inadvertent criticality.
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| This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
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| The reactor protection system is made up of two independent trip systems.
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| There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The APRM system is divided into four APRM channels and four 2-Out-Of-4 Voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed.
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| The system meets the intent of IEEE-279 for nuclear power plant protection systems. Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined in accordance with NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System" and NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function." The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.
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| The APRM Functions include five Functions accomplished by the four APRM channels (Functions 2.a, 2.b, 2.c, 2.d, and 2.f) and one accomplished by the four 2-Out-Of-4 Voter channels (Function 2.e). Two of the five Functions accomplished by the APRM channels are based on neutron flux only (Functions 2.a and 2.c), one Function is based on neutron flux and recirculation drive flow (Function 2.b) and one is based on equipment status (Function 2.d). The fifth Function accomplished by the APRM channels is the Oscillation Power Range Monitor (OPRM) Upscale trip Function 2.f, which is based on detecting oscillatory characteristics in the neutron flux. The OPRM Upscale Function is also dependent on average neutron flux (Simulated Thermal Power) and recirculation drive flow, which are used to automatically enable the output trip.
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| The Two-Out-Of-Four Logic Module includes 2-Out-Of-4 Voter hardware and the APRM Interface hardware. The 2-Out-Of-4 Voter Function 2.e votes APRM Functions 2.a, 2.b, 2.c, and 2.d independently of Function 2.f. This voting is accomplished by the 2-Out-Of-4 Voter hardware in the Two-Out-Of-Four Logic Module. The voter includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs). The analysis in Reference 2 took credit for this redundancy in the justification of the 12-hour allowed out-of-service time for LIMERICK - UNIT 1 B 3/4 3-1 Amendment No. .- 3, 94, .1-3.2, 4-44, 4-74, 186
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| 3/4.3 INSTRUMENTATION
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| _11-3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)
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| Action b, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable. The voter Function 2.e does not need to be declared inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module.
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| Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. To provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel. In addition, no more than 9 LPRMs may be bypassed between APRM calibrations (weekly gain adjustments). For the OPRM Upscale Function 2.f, LPRMs are assigned to "cells" of 3 or 4 detectors. A minimum of 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in that channel.
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| References 4, 5 and 6 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.
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| An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in any cell, with period confirmations and ielative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithms detect growing oscillatory changes in the neutron flux for one or more cells in that channel.
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| The OPRM Upscale Function is required to be OPERABLE when the plant is at 2 25% RATED THERMAL POWER. The 25% RATED THERMAL POWER level is selected to provide margin in the unlikely event that a reactor power increase transient occurring while the plant is operating below 30% RATED THERMAL POWER causes a power increase to or beyond the 30%'RATED THERMAL POWER OPRM Upscale trip auto-enable point without operator action. This OPERABILITY requirement assures that the OPRM Upscale trip automatic-enable function will be OPERABLE when required.
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| Actions a, b and c define the Action(s) required when RPS channels are discovered to be inoperable. For those Actions, separate entry condition is allowed for each inoperable RPS channel. Separate entry means that the allowable time clock(s) for Actions a, b or c start upon discovery of inoperability for that specific channel. Restoration of an inoperable RPS channel satisfies only the action statements for that particular channel. Action statement(s) for remaining inoperable channel(s) must be met according to their original entry time.
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| Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to be acceptable (NEDC-30851P-A and NEDC-32410P-A) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided that the associated Function's (identified as a "Functional Unit" in Table 3.3.1-1) inoperable channel is in one trip system and the Function still maintains RPS trip capability.
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| LIMERICK - UNIT 1 B 3/4 3-1a Amendment No. 43, 89, 4,32-, 4-44-, 177
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| 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)
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| The requirements of Action a are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal.
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| For the typical Function with one-out-of-two taken twice logic, including the IRM Functions and APRM Function 2.e (trip capability associated with APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f are discussed below), this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip).
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| For Function 5 (Main Steam Isolation Valve--Closure), this would require both trip systems to 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 in trip (or the associated trip system in trip).
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| For Function 9 (Turbine Stop Valve-Closure.), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip).
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| The completion time to satisfy the requirements of Action a is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
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| With trip capability maintained, i.e., Action a satisfied, Actions b and c as applicable must still be satisfied. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Action b requires that the channel or the associated trip system must be placed in the tripped condition.
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| Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
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| As noted, placing the trip system in trip is not applicable to satisfy Action b for APRM functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of one required APRM channel affects both trip systems. For that condition, the Action b requirements can only be satisfied by placing the inoperable APRM channel in trip. Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and the requirement to satisfy Action a.
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| The requirements of Action c must be satisfied when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, normally the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system (see additional bases discussion above related to loss of trip capability and the requirements of Action a, and special cases for Functions 2.a, 2.b, 2.c, 2.d, 2.f, 5 and 9).
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| LIMERICK - UNIT I B 3/4 3-1b Amendment No. -4A4, 177
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| 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)
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| The requirements of Action c limit the time the RPS scram logic, for any Function, would not accommodate single failure in both tip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reduced reliability of this logic arrangement was not evaluated in NEDC-30851P-A for the 12 hour Completion Time. Within the 6 hour allowance, the associated Function must have all required channels OPERABLE or in trip (or any combination) in one trip system.
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| Completing the actions required by Action c restores RPS to a reliability level equivalent to that evaluated in NEDC-30851P-A, which justified a 12 hour allowable out of service time as allowed by Action b. To satisfy the requirements of Action c, the trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e.,
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| what OPERATIONAL CONDITION the plant is in). If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.
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| The 6 hour allowable out of service time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.
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| As noted, Action c is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f.
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| Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-Out-Of-4 voter and other non-APRM channels for which Action c applies. For an inoperable APRM channel, the requirements of Action b can only be satisfied by tripping the inoperable APRM channel. Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions that will restore capability to accommodate a single APRM channel failure.
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| If it is not desired to place the channel (or trip system) in trip to satisfy the requirements of Action a, Action b or Action c (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Action d requires that the Action defined by Table 3.3.1-1 for the applicable Function be initiated immediately upon expiration of the allowable out of service time.
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| Table 3.3.1-1, Function 2.f, references Action 10, which defines the action required if OPRM Upscale trip capability is not maintained. Action lOb is required to address identified equipment failures. Action lOa is to address common mode vendor/industry identified issues that render all four OPRM channels inoperable at once. For this condition, References 2 and 3 justified use of alternate methods to detect and suppress oscillations for a limited period of time, up to 120 days. The alternate methods are procedurally established consistent with the guidelines identified in Reference 7 requiring manual operator action to scram the plant if certain predefined events occur. The 12-hour allowed completion time to implement the alternate methods is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. The 120-day period during which use of alternate methods is allowed is intended to be an outside LIMERICK - UNIT I B 3/4 3-1c Amendment No. l-A4, 177
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| 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) limit to allow for the case where design changes or extensive analysis might be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. The evaluation of the use of alternate methods concluded, based on engineering judgment, that the likelihood of an instability event that could not be adequately handled by the alternate methods during the 120-day period was negligibly small. Plant startup may continue while operating within the allowed completion time of Action lOa. The primary purpose of this is to allow an orderly completion, without undue impact on plant operation, of design and verification activities in the event of a required design change to the OPRM Upscale function. This exception is not intended as an alternative to restoring inoperable equipment to OPERABLE status in a timely manner.
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| Action 10a is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to be accomplished within the completion times allowed for LCO 3.3.1 Action a or Action b, as applicable.
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| Action lOb applies when routine equipment OPERABILITY cannot be restored within the allowed completion times of LCO 3.3.1 Actions a or b, or if a common mode OPRM deficiency cannot be corrected and OPERABILITY of the OPRM Upscale Function restored within the 120-day allowed completion time of Action 10a.
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| The OPRM Upscale trip output shall be automatically enabled (not-bypassed) when APRM Simulated Thermal Power is 2 30% and recirculation drive flow is < 60%
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| as indicated by APRM measured recirculation drive flow. NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated core flow. This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations may occur. As noted in Table 4.3.1.1-1, Note c, CHANNEL CALIBRATION for the OPRM Upscale trip Function 2.f includes confirming that the auto-enable (not-bypassed) setpoints are correct. Other surveillances ensure that the APRM Simulated Thermal Power properly correlates with THERMAL POWER (Table 4.3.1.1-1, Note d) and that recirculation drive flow properly correlates with core flow (Table 4.3.1.1-1, Note g).
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| If any OPRM Upscale trip auto-enable setpoint is exceeded and the OPRM Upscale trip is not enabled, i.e., the OPRM Upscale trip is bypassed when APRM Simulated Thermal Power is e 30% and recirculation drive flow is < 60%, then the affected channel is considered inoperable for the OPRM Upscale Function.
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| Alternatively, the OPRM Upscale trip auto-enable setpoint(s) may be adjusted to place the channel in the enabled condition*(not-bypassed). If the OPRM Upscale trip is placed in the enabled condition, the surveillance requirement is met and the channel is considered OPERABLE.
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| As noted in Table 4.3.1.1-1, Note g, CHANNEL CALIBRATION for the APRM Simulated Thermal Power - Upscale Function 2.b and the OPRM Upscale Function 2.f, includes the recirculation drive flow input function. The APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function both require a valid drive flow signal. The APRM Simulated Thermal Power - Upscale Function uses drive flow to vary the trip setpoint. The OPRM Upscale Function uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS. A CHANNEL CALIBRATION of the APRM recirculation drive flow input function requires both calibrating the drive flow transmitters and establishing a valid drive flow /
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| core flow relationship. The drive flow / core flow relationship is established once per refuel cycle, while operating within 10% of rated core flow and within LIMERICK - UNIT I B 3/4 3-1d Amendment No. 177 I
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| 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) 10% of RATED THERMAL POWER. Plant operational experience has shown that this flow correlation methodology is consistent with the guidance and intent in Reference 8.. Changes throughout the cycle in the drive flow / core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power - Upscale Function and the OPRM Upscale Function.
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| As noted in Table 3.3.1-2, Note "*", the redundant outputs from the 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels, so N = 8 to determine the interval between tests for application of Specification 4.3.1.3 (REACTOR PROTECTION SYSTEM RESPONSE TIME). The note further requires that testing of OPRM and APRM outputs shall be alternated.
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| Each test of an OPRM or APRM output tests each of the redundant outputs from the 2-Out-Of-4 Voter channel for that function, and each of the corresponding relays in the RPS. Consequently, each of the RPS relays is tested every fourth cycle. This testing frequency is twice the frequency justified by References 2 and 3.
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| Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NEDO-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated May 15, 1991).
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| The measurement of response time at the frequencies specified in the Surveillance Frequency Control Program provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. No credit was taken for those channels with response times indicated as not applicable except for the APRM Simulated Thermal Power - Upscale and Neutron Flux - Upscale trip functions and the OPRM Upscale trip function (Table 3.3.1-2, Items 2.b, 2.c, and 2.f). Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.
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| Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NEDO-32291-A. Response time testing for the remaining channel components is required as noted. For the digital electronic portions of the APRM functions, performance characteristics that determine response time are checked by a combination of automatic self-test, calibration activities, and response time tests of the 2-Out-Of-4 Voter (Table 3.3.1-2, Item 2.e).
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| LIMERICK - UNIT 1 B 3/4' 3- le Amendment No. 444, 4-7-7, 186
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| (INTENTIONALLY LEFT BLANK)
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| INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
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| Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined: in accordance with NEDC-30851P, Supplement 2, "Technical Specification Improvement Analysis for BWR Instrumentation Common to RPS and ECCS Instrumentation" as approved by the NRC and documented in the NRC Safety Evaluation Report (SER)
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| (letter to D.N. Grace from C.E. Rossi dated January 6, 1989) and NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," as approved by the NRC and documented in the NRC SER (letter to S.D. Floyd from C.E. Rossi dated June 18, 1990).
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| Automatic closure of the MSIVs upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NEDO-31400A. The NRC, approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated May 15, 1991).
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| Some of the tripsettings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
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| Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C.
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| power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time of 13 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrentwith the 10-second diesel startup and the 3 second load center loading delay. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay. It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.
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| Response time testing for sensors are not required based on the analysis in NEDO 32291-A. Response time testing of the remaining channel components is required as noted in Table 3.3.2-3.
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| Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. Primary containment isolation valves that are actuated by the isolation signals specified in Technical Specification Table 3.3.2-1 are identified in Technical Requirements Manual Table 3.6.3-1.
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| 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
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| LIMERICK - UNIT 1 B 3/4 3-2 Amendment No.-,-3,1,4,9,4-446,186
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| INSTRUMENTATION BASES 3/4.3.3 EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION (Continued)
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| Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, "Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D. N. Grace from A. C.
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| Thadani dated December 9, 1988 (Part 1) and letter to D. N. Grace from C. E.
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| Rossi dated December 9, 1988 (Part 2)).
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| Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability of adequate power for energizing various components such as pump motors, motor operated valves, and the associated control components. If the loss of power instrumentation detects that voltage levels are too low, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources. The loss of power relays in each channel have sufficient overlapping detection characteristics and functionality to permit operation subject to the conditions in Action Statement 37. Bases 3/4.8.1, 3/4.8.2, and 3/4.8.3 provide discussion regarding parametric bounds for determining operability of the offsite sources.
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| Those Bases assume that the loss of power relays are operable. With an inoperable 127Z-11XOX relay, the grid voltage is monitored to 230kV (for the 101 Safeguard Bus Source) or 525kV (for the 201 Safeguard Bus Source) to increase the margin for the operation of the 127Z-11XOX relay.
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| Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable-on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
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| 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NEDO-10349, dated March 1971, NEDO-24222, dated December 1979, and Section 15.8 of the FSAR.
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| The end-of-cycle recirculation pump trip (EOC-RPT) system is a supplement to the reactor trip. During turbine trip and generator load rejection events, the EOC-RPT will reduce the likelihood of reactor vessel level decreasing to level
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| : 2. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
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| A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.
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| LIMERICK - UNIT 1 B 3/4 3-3 Amendment No. , 6-9, 9, 449,186
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| INSTRUMENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued)
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| Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.
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| The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,
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| 175 ms. Included in this time are: the response time of the sensor, the time allotted for breaker arc suppression, and the response time of the system logic.
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| LIMERICK - UNIT 1 B 3/4 3-3a Amendment No. 158 1
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| INSTRUMENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued)
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| Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined in accordance with GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D. Binz, IV, from C.E. Rossi dated July 21, 1992).
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| Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
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| 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. This instrumentation does not provide actuation of any of the emergency core cooling equipment.
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| Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been specified in accordance with recommendations made by GE in their letter to the BWR Owner's Group dated August 7, 1989,
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| ==SUBJECT:==
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| "Clarification of Technical Specification changes given in ECCS Actuation Instrumentation Analysis."
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| Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
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| 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section.3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.
| |
| Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," as approved by the NRC and documented in the SER (letter to D. N. Grace from C. E. Rossi dated September 22, 1988).
| |
| Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.
| |
| LIMERICK - UNIT 1 B 3/4 3-4 Amendment No. 49, -53, -Q-0,186
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| | |
| INTENTIONALLY LEFT BLANK INSTRUMENTATION BASES 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.
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| The surveillance interval for the Main Control Room Normal Fresh Air Supply Radiation Monitor is determined in accordance with the Surveillance Frequency Control Program.
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| 3/4.3.7.2 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.
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| 3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.
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| 3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50, Appendix A.
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| 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"
| |
| December 1975 and NUREG-0737, "Clarification of TMI Action Plan Requirements,"
| |
| November 1980.
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| LIMERICK - UNIT 1 B 3/4 3-5 Amendment No. 48, - , -ýO, ;-&,
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| 4-5-, ECR 02 00470, 4-7-3, 186
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| | |
| -INSTRUMENTAT ION BASES 3/4.3.7 MONITORING INSTRUMENTATION (Continued) 3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown.
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| At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the intermediate range monitors are on scale, adequate information is avdilable without the SRMs and they can be retracted.
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| LIMERICK - UNIT 1 B 3/4 3-5a Amendment No. 48, 53, .70, 7-5, 151
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| | |
| INSTRUMENTATION BASES 3/4.3.7.7 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.
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| 3/4.3.7.8 CHLORINE AND TOXIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine isolation mode of operation to provide the required protection. Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation to provide the required protection. The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95, "Pro-tection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975.
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| There are three toxic gas detection subsystems. The high toxic chemical concentration alarm in the Main Control 'Room annunciates when two of the three subsystems detect a high toxic gas concentration. An Operate/Inop keylock switch is provided for each subsystem which allows an individual subsystem to be placed in the tripped condition. Placing the keylock switch in the INOP position initiates one of the two inputs required to initiate the alarm in the Main Control Room.
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| Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined in accordance with GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D.
| |
| Binz, IV, from C.E. Rossi dated July 21, 1992).
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| 3/4.3.7.9 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.
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| LIMERICK - UNIT 1 B 3/4 3-6 Amendment No. 49,40,]7-9,84,4*4,4*-ý,186
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| | |
| (INTENTIONALLY LEFT BLANK)
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| INSTRUMENTATION BASES 3/4.3.7.10 (Deleted) 3/4.3.7.11 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.
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| 3/4.3.7.12 OFFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.
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| 3/4.3.8 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.
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| 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand.
| |
| | |
| ==REFERENCES:==
| |
| : 1. NEDC-30851P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.
| |
| : 2. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.
| |
| : 3. NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," November 1997.
| |
| : 4. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
| |
| : 5. NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
| |
| : 6. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
| |
| : 7. Letter, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action," June 6, 1994.
| |
| : 8. GE Service Information Letter No. 516, "Core Flow Measurement - GE BWR/3, 4, 5 and 6 Plants," July 26, 1990.
| |
| : 9. GE Letter NSA 00-433, Alan Chung (GE) to Sujit Chakraborty (GE),
| |
| "Minimum Number of Operable OPRM Cells for Option III Stability at Limerick 1 & 2," May 02, 2001.
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| LIMERICK - UNIT 1 B 3/4 3-7 Amendment No. 3, 48, ý7-0, 4-9Q, --04, -53, 177
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| RECTRVESL1 ATRLmE 4U, 13 LIMEIK-UI / -
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| LIMERICK - UNIT I 8 3/4 3-6
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| | |
| BASES FIGURE B 3/4.3-2 APRM CONFIGURATION APR 12 2W0 LIMERICK - UNIT 1 3 3/4 3-9 Amendment No. 141
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| | |
| (INTENTIONALLY LEFT BLANK)
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| :j
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| .
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| 3/4.4.REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively.
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| Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive internals vibration.
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| The surveillance on differential temperatures below 30% RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.
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| An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.
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| Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recir-culation loop mode.
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| In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50°F of each other prior to startup of an idle loop. The loop temperature must also be within 50°F of the reactor pressure vessel coolant temperature to prevent thermal shock tothe recirculation pump and recirculation nozzles. Sudden equalization of a temperature difference > 145°F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.
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| LIMERICK - UNIT 1 B 3/4 4-1 Amendment No. _29, 64, 177
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| REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety/
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| relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.
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| Demonstration of the safety/relief valve lift settings will occur only during shutdown. The safety/relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations at the frequency specified in the Surveillance Frequency Control Program.
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| Cerreeted by Ltr. Dgated 3/10/0O LIMERICK - UNIT 1 B 3/4 4-2 Amendment No. -3, 4-3-'-, 4-17-T,186
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| REACTOR COOLANT SYSTEM BASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS BACKGROUND UFSAR Safety Design Basis (Ref. 1), requires means for detecting and, to the extent practical, identifying the location of the source of Reactor Coolant System (RCS)
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| PRESSURE BOUNDARY LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.
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| Limits on leakage from the reactor coolant pressure boundary (RCPB) are required so that appropriate action can be taken before the integrity of the RCPB is impaired (Ref.
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| 2). Leakage detection systems for the RCS are provided to alert the operators when leakage rates above normal background levels are detected and also to supply quantitative measurement of leakage rates.
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| Systems for separating the leakage of an identified source from an unidentified source are necessary to provide prompt and quantitative information to the operators to permit them to take immediate corrective action. Leakage from the RCPB inside the drywel is detected by at least one of four (4) independently monitored variables which include drywell drain sump level changes over time yielding drain flow rates, and drywell gaseous radioactivity, drywell unit cooler condensate flow rate and drywell pressure/temperature levels. The primary means of quantifying leakage in the drywell are the drywell floor drain sump flow monitoring system for UNIDENTIFIED LEAKAGE and the drywell equipment drain tank flow monitoring system for IDENTIFIED LEAKAGE.
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| IDENTIFIED leakage is not germane to this Tech Spec and the associated drywell
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| !quipment drain tank flow monitoring system is not included.
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| The drywell floor drain sump flow monitoring system monitors UNIDENTIFIED LEAKAGE collected in the floor drain sump. UNIDENTIFIED LEAKAGE consists of leakage from RCPB components inside the drywell which are not normally subject to leakage and otherwise routed to the drywell equipment drain sump. The primary containment floor drain sump has transmitters that supply level indication to the main control room via the plant monitoring system. The floor drain sump level transmitters are associated with High/Low level switches that open/close the sump tank drain valves automatically. The level instrument processing unit calculates an average leak rate (gpm) for a given measurement period which resets whenever the sump drain valve closes. The level processing unit provides an alarm to the main control room each time the average leak rate changes by a predetermined value since the last time the alarm was reset. For the drywell floor drain sump flow monitoring system, the setpoint basis is a I gpm change in UNIDENTIFIED LEAKAGE.
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| In addition to the drywell floor drain sump flow monitoring system described above, the discharge of each sump is monitored by an independent flow element. The measured flow rate from the flow element is integrated and recorded. A main control room alarm is also provided to indicate an excessive sump discharge rate measured via the flow element. This system, referred to as the "drywell floor drain flow totalizer', is not credited for drywell floor drain sump flow monitoring system operability.
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| APR Z 0 20C0 IMERICK - UNIT I B 3/4 4-3 Amendment-40 7 49; 140
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| REACTOR COOLANT SYSTEM BASES BACKGROUND (Continued)
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| The primary containment atmospheric gaseous radioactivity monitoring system continuously monitors the primary containment atmosphere for gaseous radioactivity levels. A sudden increase of radioactivity, which may be attributed to RCPB steam or reactor water leakage, is annunciated in the main control room. The primary containment atmospheric gaseous radioactivity monitoring system is not capable of quantifying leakage rates, but is sensitive enough to detect increased leakage rates of I gpm within I hour. Larger changes in leakage rates are detected in proportionally shorter times (Ref. 4).
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| Condensate from the eight drywell air coolers is routed to the drywell floor drain sump and is monitored by a series of flow transmitters that provide indication and alarms in the main control room. The outputs from the flow transmitters are added together by summing units to provide a total continuous condensate drain flow rate. The high flow alarm setpoint is based on condensate drain flow rate in excess of I gpm over the currently identified preset leak rate. The drywell air cooler condensate flow rate monitoring system serves as an added indicator, but not quantifier, of RCS UNIDENTIFIED LEAKAGE (Ref. 5).
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| The drywell temperature and pressure monitoring systems provide an indirect method for detecting RCPB leakage. A temperature and/or pressure rise in the drywell above normal levels may be indicative of a reactor coolant or steam leakage (Ref. 6).
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| APPLICABLE SAFETY ANALYSES A threat of significant compromise to the RCPB exists if the barrier contains a crack that is large enough to propagate rapidly. Leakage rate limits are set low enough to detect the leakage emitted from a single crack in the RCPB (Refs. 7 and 8). Each of the leakage detection systems inside the drywell is designed with the capability of detecting leakage less than the established leakage rate limits and providing appropriate alarms of excess leakage in the control room.
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| A control room alarm allows the operators to evaluate the significance of the indicated leakage and, if necessary, shut down the reactor for further investigation and corrective action. The allowed leakage rates are well below the rates predicted for critical crack sizes (Ref. 8). Therefore, these actions provide adequate response before a significant break in the RCPB can occur.
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| RCS leakage detection instrumentation satisfies Criterion 1 of the NRC Policy Statement.
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| LIMITING CONDITION FOR OPERATION (LCO)
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| The drywell floor drain sump flow monitoring system is required to quantify the UNIDENTIFIED LEAKAGE from the RCS. The other monitoring systems provide early alarms to the operators so closer examination of other detection systems will be made to determine the extent of any corrective action that may be required. With any leakage detection system inoperable, monitoring for leakage in the RCPB is degraded.
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| APR Z 0 ZUO0 LIMERICK - UNIT I B 3/4 4-3a Amendment 1140
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| REACTOR COOLANT SYSTEM BASES APPLICABILITY In OPERATIONAL CONDITIONS 1, 2, and 3, leakage detection systems are required to be OPERABLE to support LCO 3.4.3.2. This applicability is consistent with that for LCO 3.4.3.2.
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| A. ACTION A. With the primary containment atmosphere gaseous monitoring system inoperable, grab samples of the primary containment atmosphere must be taken and analyzed to provide periodic leakage information. [Provided a sample is obtained and analyzed once every 12 hours, the plant may be operated for up to 30 days to allow restoration of the radioactivity monitoring system. The plant may continue operation since other forms of drywell leakage detection are available.]
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| The 12 hour interval provides periodic information that is adequate to detect leakage. The 30 day Completion Time for Restoration recognizes other forms of leakage detection are available.
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| B. With the drywell floor drain sump flow monitoring system inoperable, no other form of sampling can provide the equivalent information to quantify leakage at the required 1 gpm/hour sensitivity. However, the primary containment atmospheric gaseous monitor [and the primary containment air cooler condensate flow rate monitor] will provide indication of changes in leakage.
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| With the drywell floor drain sump flow monitoring system inoperable, drywell condensate flow rate monitoring frequency increased from 12 to every 8 hours, and UNIDENTIFIED LEAKAGE and total leakage being determined every 8 hours (Ref. SR 4.4.3.2.1.b) operation may continue for 30 days. T6 the extent practical, the surveillance frequency change associated with the drywell condensate flow rate monitoring system, makes up for the loss of the drywell floor drain monitoring system which had a normal surveillance requirement to monitor leakage every 8 hours.
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| Also note that in this instance, the drywell floor drain tank flow totalizer.will be used to comply with SR 4.4.3.2.1.b. The 30 day Completion Time of the required ACTION is acceptable, based on operating experience, considering the multiple forms of leakage detection that are still available.
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| LIMERICK - UNIT 1 B 3/4 4-3b Amendment 440, 169
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| REACTOR COOLANT SYSTEM BASES ACTIONS (Continued)
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| C. With the required primary containment air cooler condensate flow rate monitoring system inoperable, SR 4.4.3.1.a must be performed every 8 hours to provide periodic information of activity in the primary containment of more frequent interval than the routine frequency of every 12 hours. The 8 hour interval provides periodic information that is adequate to detect leakage and recognizes that other forms of leakage detection are available. The required ACTION has been clarified to state that the additional surveillance requirement is not applicable if the required primary containment atmosphere gaseous radioactivity monitoring system is also inoperable. Consistent with SR 4.0.3, surveillances are not required to be performed on inoperable equipment. In this case, ACTION Statement A. and E. requirements apply.
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| D. With the primary containment pressure and temperature monitoring system inoperable, operation may continue for up to 30 days given the system's indirect capability to detect RCS leakage. However, other more limiting Tech Spec requirements associated with the primary containment pressure/temperature monitoring system will still apply.
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| E. With both the primary containment atmosphere gaseous radioactivity monitor and the primary containment air cooler condensate flow rate monitor inoperable, the only means of detecting leakage is the drywell floor drain sump monitor and the drywell pressure/temperature instrumentation. This condition does not provide the required diverse means of leakage detection. The required ACTION is to restore either of the inoperable monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a degraded configuration for a lengthy time period. While in the primary containment atmosphere gases radioactivity monitor is INOPERABLE, Primary containment atmospheric grab samples will be taken and analyzed every 12 hours since ACTION Statement A. requirements also apply.
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| F. If any required ACTION of Conditions A, B, C, D or E cannot be met within the associated Completion Time, the plant must be brought to an OPERATIONAL CONDITION in which the LCO does not apply. To achieve this status, the plant must be brought to at least HOT SHUTDOWN within 12 hours and COLD SHUTDOWN within the next 24 hours.
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| The allowed Completion Times are reasonable, based on operating experience, to perform the ACTIONS in an orderly manner and without challenging plant systems.
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| SURVEILLANCE REQUIREMENTS SR 4.4.3.1.a This SR is for the performance of a CHANNEL CHECK of the required primary containment atmospheric monitoring system. The check gives reasonable confidence that the channel is operating properly.
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| LIMERICK - UNIT B 3/4 4-3c Amendment 44-, 4-64,186
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| REACTOR COOLANT SYSTEM BASES SURVEILLANCE REQUIREMENTS (Continued)
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| SR 4.4.3.1.b This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their function in the desired manner. The test also verifies the alarm setpoint and relative accuracy of the instrument string.
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| SR 4.4.3.1.c The SR is for the performance of a CHANNEL CALIBRATION of required leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.
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| SR 4.4.3.1.d This SR provides a routine check of primary containment pressure and temperature for indirect evidence of RCS leakage.
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| REFERENCES
| |
| : 1. LGS UFSAR, Section 5.2.5.1.
| |
| : 2. Regulatory Guide 1.45, May 1973.
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| : 3. LGS UFSAR, Section 5.2.5.2.1.3.
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| : 4. LGS UFSAR, Section 5.2.5.2.1.5.
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| : 5. LGS UFSAR, Section 5.2.5.2.1.4.
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| : 6. LGS UFSAR, Section 5.2.5.2.1.1(2).
| |
| : 7. GEAP-5620, April 1968
| |
| : 8. NUREG-75/067, October 1975.
| |
| : 9. LGS UFSAR, Section 5.2.5.6.
| |
| 3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action. The limit of 2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period and the monitoring of drywell floor drain sump and drywell equipment drain tank flow rate at least once every eight (8) hours conforms with NRC staff positions specified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," as revised by NRC Safety Evaluation dated March 6, 1990. The ACTION requirement for the 2 gpm increase in UNIDENTIFIED LEAKAGE limit ensures that such leakage is identified or a plant shutdown is initiated to allow further investigation and corrective action. Once identified, reactor operation may continue dependent upon the impact on total leakage LIMERICK - UNIT 1 B 3/4 4-3d Amendment No. 4-4-9,186
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| | |
| REACTOR COOLANT SYSTEM BAU -
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| 3/4.4.3.2 OPERATIONAL LEAKAGE (Continued)
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| The function of Reactor Coolant System Pressure Isolation Valves (PIVs) is to separate the high pressure Reactor Coolant System from an attached low pressure system.
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| The ACTION requirements for pressure isolation valves are used in conjunction with the system specifications for which PIVs are listed in the Technical Requirements Manual and with primary containment isolation valve requirements to ensure that plant operation is appropriately limited.
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| The Surveillance Requirements for the RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is not included in any other allowable operational leakage specified in Section 3.4.3.2.
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| 3/4.4.4 (Deleted) INFORMATION FROM THIS SECTION RELOCATED TO THE TRM LIMERICK - UNIT 1 B 3/4 4-3e Amendment No. 44G, 44-, 4-74, 182
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| REACTOR COOLANT SYSTEM BASES 314.4.4 (Deleted) INFORMATION FROM THIS SECTION RELOCATED TO THE TRM 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour thyroid and whole body doses resulting from a main steam line failur'e outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
| |
| The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT 1-131, but less than or equal to 4 microcuries per gram DOSE EQUIVALENT 1-131, accommodates possible iodine spiking phenomenon which may occur following changes in the THERMAL POWER. This action is modified by a Note that permits the use of the provisions of Specification 3.0.4.c. This allowance permits entry into the applicable OPERATIONAL CONDITION (S) while relying on the ACTION requirements. Operation with specific activity levels exceeding 0.2 microcurie per gram DOSE EQUIVALENT 1-131 but less than or equal to 4 microcuries per gram DOSE EQUIVALENT 1-131 must be restricted since these activity levels increase the 2-hour thyroid dose at the SITE BOUNDARY following a postulated steam line rupture.
| |
| Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant Will be detected in sufficient time to take corrective action.
| |
| 3/4.4.6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown; the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
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| LIMERICK - UNIT I B 3/4 4-4 Amendment No. 2-0, 4-0, 4-9, 4-6-9, 174
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| INTENTIONALLY LEFT BLANK REACFOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)
| |
| The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section XI, Appendix G. The curves are based on the RTNr and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Chapter 5, Para-graph 5.3.1.5, "Fracture Toughness."
| |
| The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in Table B 3/4.4.6-1. Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTNDI. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." The pressure/temperature limit curves, Figure 3.4.6.1-1, include a shift in RTNDT for conditions at 32 EFPY. The A, B and C limit curves are predicted to be bounding for all areas of the RPV until 32 EFPY. In addition, an intermediate A curve has been provided for 22 EFPY.
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| The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C, and A, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.
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| LIMERICK - UNIT I B 3/4 4-5 LAmendment No. -36, 4--6, 4--4-5, 167
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| RFACTnR COOLANT SYSTFM BASES _____ __________________ _
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| 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.
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| Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.
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| 3/4.4 8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity'of these components will be maintained at an acceptable level throughout the life of the plant.
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| Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.
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| The inservice inspection program for ASME Code Class 1,2, and 3 components will be performed in accordance with Section XI of the ASME. Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a. Additionally, the Inservice Inspection Program conforms to the NRC staff positions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,"
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| as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990, or in accordance with alternate measures approved by the NRC staff.
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| 3/4.4.9 RESIDUAL HEAT REMOVAL The RHR system is required to remove decay heat and sensible heat in order to maintain the temperature of the reactor coolant. RHR shutdown cooling is comprised of four (4) subsystems which make two (2) loops. Each loop consists of two (2) motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Two (2) redundant, manually controlled shutdown cooling subsystems of the RHR System can provide the required decay heat removal capability. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection pathway. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.
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| An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.
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| In HOT SHUTDOWN condition, the requirement to maintain OPERABLE two (2) independent RHR shutdown cooling subsystems means that each subsystem considered OPERABLE must be associated with a different heat exhanger loop. i.e., the "A" RHR heat exchanger with the "A" RHR pump or the "C" RHR pump, and the "B" RHR heat exchanger with the "B" RHR pump or the "D" RHR pump are two (2) independent RHR shutdown cooling subsystems. Only one (1) of the two (2) RHR pumps associated with each RHR heat exchanger loop is LIMERICK - UNIT I B 3/4 4-6 Amendment No. 4-9, -7, 4-1-9, 4-2-5, 171
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| 3/4.4.9 RESIDUAL HEAT REMOVAL (Cont'd) required to be OPERABLE for that independent subsystem to be OPERABLE. During COLD SHUTDOWN and REFUELING conditions, however, the subsystems not only have a common suction source, but are allowed to have a common heat exchanger and common discharge piping. To meet the LCO of two (2) OPERABLE subsystems, both pumps in one (1) loop or one (1) pump in each of the two (2) loops must be OPERABLE. Since the piping and heat exchangers are passive components, that are-assumed not to fail, they are allowed to be common to both subsystems. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one (1) subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
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| Alternate decay heat removal methods are available to operators. These alternate methods of decay heat removal can be verified available either by calculation (which includes a review of component and system availability to verify that an alternate decay heat removal method is available) or by demonstration, and that a method of coolant mixing be operational. Decay heat removal capability by ambient losses can be considered in evaluating alternate decay heat removal capability.
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| LIMERICK - UNIT 1 B 3/4 4-6a Amendment No. 119 FEB 3 1997
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| INTENTIONALLY LEFFT BLANK BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS*
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| HEAT/SLAB MIN.UPPER BELTLINE WELD SEAM I.D. OR STARTING SHELF COMPONENT OR MAT'L TYPE HEAT/LOT CU M~ Ni M RTNm (OF ART **(OF) (LFT-LBS) ART (OF)
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| Plate SA-533 Gr. BCL. 1 C 7677-1 .11 .5 +20 +35 NA +89 Weld AB (Field Weld) 640892/ .09 1.0 -60 +58 NA +54 J424B27AE NOTES:
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| * Based on 110% of original rated power.
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| ** These values are given only for the benefit of calculating the end-of-life (EOL/32 EFPY) RTwr*
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| NON-BELTLINE MT'L TYPE OR HEAT/SLAB OR HIGHEST STARTING COMPONENT WELD SEAM I.D. HEAT/LOT RTmr (0F)
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| Shell Ring SA 533, Gr. B, CL. 1 C7711-1 +20 Bottom Head Dome C7973-1 +12 Bottom Head Torus C7973-1 +12 Top Head Dome A6834-1 +10 B1993-1 +10 Top Head Torus.
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| 123B195-289 +10 Top Head Flange SA-508, CL. 2 2V1924-302 -20 Vessel Flange Q2Q22W-412 -20 Feedwater Nozzle All -12 Weld Non-Beltline Q2Q25W -6 LPCI Nozzle*** SA-508, CL. 2 All Meet requirements of 45 ft-lbs Closure Studs SA-540, Gr. B-24 and 25 mils Lat. Exp. at +10°F Note: *** The design of the LPCI nozzles results in their experiencing an EOL fluence in excess of 2 0 10" N/Cm which predicts an EOL (32 EFPY) RTn of +41 F.
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| Amendment No. -3, 106 B 3/4 4-7 LIMERICK - UNIT 1 145 SEP15
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| '-IA Mi a
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| ,~4.~
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| 1.2 0
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| co 1.0 C>
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| 0.8 uJ 04 E 0.6 C- 0.4 a
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| 0.2 0.0 Service Life (Years*)
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| BASES FIGURE B 3/4.4.6-1 FAST NEUTRON FLUENCE (E> I MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE*
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| At 90g of Rated Thermal Power and 90%availability )
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| LIMERICK-UNIT I B 3/4 4-8 Amendment No. 3,106 FEB 1 2 1996
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| 3/4.5 E-_*RGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPEPATING and SHUTDOWN The core spray system (CSS), together with the LPCI mode of the RHR system, is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-enoed reactor recirculation line break, and for smaller breaks following depressurization by the ADS.
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| The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.
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| The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
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| The low pressure coolant injection (LPCI) mode of the RHR system is provided :o assure that the core is adequately cooled following a loss-of-coolant accident. Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.
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| The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional. test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
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| The -high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HCPI system continues to operate until reactor vessel pressure is below the pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling.
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| The capacity of the system is selected to provide the required core cooling.
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| The HPCI pump is designed to deliver greater than or equal to 5600 gpm at reactor pressures between 1182 and 200 psig and is capable of delivering at least 5000 gpm between 1182 and 1205 psig. In the system's normal alignment, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for condensate storage tank water.
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| HAYN2 00 5
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| LIMERICK - UNIT 1 B 3/4 5-1 Amendment No. .4?6, 127 ECR 00-00177
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| EMERGENCY CORE COOLING SYSTEM BASES ECCS - OPERATING and SHUTDOWN (Continued)
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| With the HPCI system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the CS and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCI out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems and the RCIC system. The HPCI system, and one LPCI subsystem, and/or one CSS subsystem out-of-service period of 8 hours ensures that sufficient ECCS, comprised of a minimum of one CSS subsystem, three LPCI subsystems, and all of the ADS will be available to 1) provide for safe shutdown of the facility, and 2) mitigate and control accident conditions within the facility. A Note prohibits the application of Specification 3.0.4.b to an inoperable HPCI subsystem. There is an increased risk associated with entering an OPERATIONAL CONDITION or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions of Specification 3.0.4.b, which allow entry into an OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
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| The surveillance requirements provide adequate assurance that the HPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.
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| Upon failure of the HPCI system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automa-tically causes selected safety/relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200'F. ADS is conserva-tively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig.
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| This pressure is substantially below that for which the low pressure core cool-ing systems can provide adequate core cooling for events requiring ADS.
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| ADS automatically controls five selected safety-relief valves. The safety analysis assumes all five are operable. The allowed out-of-service time for one valve for up to fourteen days is determined in a similar manner to other ECCS sub-system out-of-service time allowances.
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| Verification that ADS accumulator gas supply header pressure is t90 psig ensures adequate gas pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator at least two valve actuations can occur with the drywell at 70% of design pressure. The ECCS safety analysis assumes only one actuation' to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of Ž90 psig is provided by the PCIG supply.
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| LIMERICK - UNIT 1 B 3/4 5-2 Amendment No. 8/10/94 4t-9,.4,4-521,4-619, 186
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| FMERGENCY CORE COOLING SYSTEM BASES ECCS - OPERATING and SHUTDOWN (Continued) 3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCI, CS and LPCI systems in the event of a LOCA. This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression chamber in OPERATIONAL CONDITION 1, 2, or 3 is also required by Specification 3.6.2.1.
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| Repair work might require making the suppression chamber inoperable. This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.
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| In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 200 0 F. Since pressure suppression is not required below 212°F, the minimum water volume is based on NPSH, recirculation volume and vortex prevention plus a safety margin for conservatism.
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| LIMERICK - UNIT 1 B 3/4 5-2a Amendment No. 152 1
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| INTENTIONALLY LEFT BLANK 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 314.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate-rials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
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| 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value calculated in the safety analyses for the design basis LOCA maximum peak containment pressure of s 44 psig, Pa. As an added conservatism, the measured overall integrated leakage rate (Type A Test) is further limited to less than or equal to 0.75 6a during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
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| Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.
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| The surveillance testing for measuring leakage rates is consistent with the Primary Containment Leakage Rate Testing Program.
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| 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCK The limitations on closure and leak rate for the primary containment air lock are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the Primary Containment Leakage Rate Testing Program. Only one closed door in the air lock is required to maintain the integrity of the containment.
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| 3/4.6.1.4 MSIV LEAKAGE'ALTERNATE DRAIN PATHWAY Calculated doses resulting from the maximum leakage allowances for the main steamline isolation valves in the postulated LOCA situations will not exceed the criteria of 10 CFR Part 100 guidelines, provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been continuously maintained. The requirement for the MSIV Leakage Alternate Drain Pathway serves to reduce the offsite dose.
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| JAN 2 4 1997 LIMERICK - UNIT I B 3/4 6-1 Amendment No. 33,-4O6,-OQ, 118
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| CONTAINMENT SYSTEMS BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment will withstand the maximum calculated pressure in the event of a LOCA. A visual inspection in accordance with the Primary Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.
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| 3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the calculated containment peak pressure does not exceed the design pressure of 55 psig during LOCA conditions or that the external pressure differ-ential does not exceed the design maximum external pressure differential of 5.0 psid. The limit of - 1.0 to + 2.0 psig for initial containment pressure will limit the total pressure to
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| * 44 psig which is less than the design pressure and is consistent with the safety analysis.
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| 3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the con-tainment peak air temperature does not exceed the design temperature of 340°F during steam line break conditions and is consistent with the safety analysis.
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| 3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for inerting, deinerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open. Limiting the use of the drywell and suppression chamber purge system to specific criteria is imposed to protect the integrity of the SGTS filters. Analysis indicates that should a LOCA occur while this pathway is being utilized, the associated pressure surge through the (18 or 24") purge lines will adversely affect the integrity of SGTS. This condition is not imposed on the 1 and 2 inch valves used for pressure control since a surge through these lines does not threaten the operability of SGTS.
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| Surveillance requirement 4.6.1.8 ensures that the primary containment purge valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable.
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| The SR is modified by a Note stating that primary containment purge valves are only required to be closed in OPERATIONAL CONDITIONS 1, 2 and 3. The SR is also modified by a Note stating that the SR is not required to be met when the purge valves are open for the stated reasons. The Note states that these valves may be opened for inerting, deinerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open. The 18 or 24 inch purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time.
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| LIMERICK - UNIT 1 B 3/4 6-2 Amendment No. 2-9,446,14-5,4-98,43-0, 186
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| CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 55 psig during primary system blowdown from full operating pressure.
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| The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system. The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from rated conditions. Since all of the gases in the drywell are purged into the suppression chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
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| Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design bisis accident is below the design pressure. Maximum water volume of 134,600 ft results in a downcomer submergence of 12'30 and the minimum volume of 122,120 ft 3 results in a submergence approximately 2'30 less. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170°F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature above 170°F.
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| Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.
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| Under full power operating conditions, blowdown through safety/relief valves assuming an initial suppression chamber water temperature of 950 F results in a bulk water temperature of approximately 140°F immediately following blowdown which is below the 190°F bulk temperature limit used for complete condensation via T-quencher devices. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.
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| LIMERICK - UNIT I B 3/4 6-3 Amendment No. Ul,%7,00,106 FEB 1 2199
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| -1/4.6.2 DEPRESSURTZATION SYSTEMS (Cont.)
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| One of the surveillance requirements for the suppression pool cooling (SPC) mode of the RHR system is to demonstrate that each RHR pump develops a flow rate
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| :10,000 gpm while operating in the SPC mode with flow through the heat exchanger and its associated closed bypass valve, ensuring that pump performance has not degraded during the cycle and that the flow path is operable. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component operability, trend performance and de~tect incipient failures by indicating abnormal performance. The RHR heat exchanger bypass valve is used for adjusting flow through the heat exchanger, and is not designed to be a tight shut-off valve. With the bypass valve closed, a portion of the tctal flow still travels through the bypass, which can affect overall heat transfer. However, no heat transfer performance requirement of the heat exchanger is intended by the current Technical Specification surveillance requirement. This is confirmed by the lack of any flow requirement for the RHRSW system in Technical Specifications Section 3/4.7.1.
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| Verifying an RHR flowrate through the heat exchanger does not demonstrate heat removal capability in the absence of a requirement for RHRSW flow. LGS does perform heat transfer testing of the RHR heat exchangers as part of its response to Generic Letter 89-13, which verified the commitment to meet the requirements of GDC 46.
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| Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 2001F during any period of relief valve operation for T-quencher devices.
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| Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.
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| Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
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| In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open. As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppres-sion pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety/relief valve to assure mixing and uniformity of energy insertion to the pool.
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| During a LOCA, potential leak paths between the drywell and suppression chate-airspace could result in excessive containment pressures, since the steam flow into the airspace would bypass the heat sink capabilities of the chamber. Potential sour-..;
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| of bypass leakage are the suppression chamber-to-drywell vacuum breakers (VBs),
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| penetrations in the diaphragm floor, and cracks in the diaphragm floor and/or liner plate and downcomers located in the suppression chamber airspace. The containment pressure response to the postulated bypass leakage can be mitigated by manually actuating the suppression chamber sprays. An analysis was performed for a design bypass leakage brea of A/fk equal to 0.0500 ft' to verify that the operator has sufficient time to initiate the sprays prior to exceeding the containment design pressure of 55 psig. The limit o0 1°o0of the design value of 0.0500 ft, ensures that the design basis for the steam bypass anallsis is met.
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| LIMERICK - UNIT 1 B 3/4 6-3a Amendment No. 57, 68 IS!"
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| FEB 17
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| CONTAINMENT SYSTEMS BASES DEPRESSURIZATION SYSTEMS (Continued)
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| The drywell-to-suppression chamber bypass test at a differential pressure of at least 4.0 psi verifies the overall bypass leakage area for simulated LOCA conditions is less than the specified limit. For those outages where the drywell-to-suppression chamber bypass leakage test in not conducted, the VB leakage test verifies that the VB leakage area is less than the bypass limit, with a 76% margin to the bypass limit to accommodate the remaining potential leakage area through the passive structural components. Previous drywell-to-suppression chamber bypass test data indicates that the bypass leakage through the passive structural components will be much less than the 76% margin. The VB leakage limit, combined with the negligible passive structural leakage area, ensures that the drywell-to-suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled.
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| 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A of 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environ-ment will be consistent with the assumptions used in the analysesvfor a LOCA.
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| The scram discharge volume vent and drain valves serve a dual function, one of which is primary containment isolation. Since the other safety functions of the scram discharge volume vent and drain valves would not be available if the normal PCIV actions were taken, actions are provided to direct the user to the scram discharge volume vent and drain operability requirements contained in Specification 3.1.3.1.
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| However, since the scram discharge volume vent and drain valves are PCIVs, the Surveillance Requirements of Specification 4.6.3 still apply to these valves.
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| The opening of a containment isolation valve that was locked or sealed closed to satisfy Technical Specification 3.6.3 Action statements, may be reopened on an intermittent basis under administrative controls. These controls consist of stationing a dedicated individual at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.
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| Primary containment isolation valves governed by this Technical Specification are identified in Table 3.6.3-1 of the TRM.
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| This Surveillance Requirement requires a demonstration that a representative sample of reactor instrument line excess flow check valves (EFCVs) is OPERABLE by verifying that the valve actuates to the isolation position on a simulated instrument line break signal. The representative sample consists of an approximately equal number of EFCVs, such that each EFCV is tested in accordance with the Surveillance Frequency Control Program. In addition, the EFCVs in the sample are representative of the various plant configurations, models, sizes, and operating environments. This ensures that any potentially common problem with a specific type or application of EFCV is detected at the earliest possible time. This Surveillance Requirement provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during a postulated instrument line break event. Furthermore, any EFCV failures will be evaluated to determine if additional testing in the test interval is warranted to ensure overall reliability is maintained. Operating experience has demonstrated that these components are highly reliable and that failures to isolate are very infrequent. Therefore, testing of a representative sample was concluded to be acceptable from a reliability standpoint. For some EFCVs, this Surveillance can be performed with the reactor at power.
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| LIMERICK - UNIT 1 B 3/4 6-4 Amendment No. 46,68,-14-4,-1,4-6.8,186
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| *BASES 3/4.6.4 VACUUM RELIEF Vacuumn relief valves are provided to equalize the pressure between the suppression chamber and drywell. This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.
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| The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident. Two pairs of valves are required to protect containment structural integrity. There are four pairs of valves (three to provide minimum redundancy) so that operation may continue for up to 72 hours with no more than two pairs of vacuum breakers inoperable in the closed position.
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| Each vacuum breaker valve's position indication system is of great enough sensitivity to ensure that the maximum steam bypass leakage coefficient of A
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| 4k=0.05 ft' for the vacuum relief system (assuming one valve fully open) will not be exceeded.
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| LIMERICK - UNIT I B 3/4 6-4a 1 UNI L~h~ICK No. 148
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| -4aAmendme B3/4
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| CONTAINMENT SYSTEMS BASES 3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Enclosure and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.
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| Establishing and maintaining a vacuum in the reactor enclosure secondary containment with the standby gas treatment system in accordance with the Surveillance Frequency Control Program, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.
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| The OPERABILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA. The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY and Control Room radiation doses associated with containment leakage. The operation of these systems and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analysis. Provisions have been made to continuously purge the filter plenums with instrument air when the filters are not in use to prevent buildup of moisture on the adsorbers and the HEPA filters.
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| As a result of the Alternative Source Term (AST) project, secondary containment integrity of the refueling area is not required during certain conditions when handling irradiated fuel, during CORE ALTERATIONS, or during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel. The control room dose analysis for the fuel handling accident (FHA) is based on unfiltered releases from the South Stack and therefore, does not require the standby gas treatment system to be aligned to the refueling area whenever irradiated fuel is being handled during CORE ALTERATIONS or operations are being conducted that have the potential to draion the reactor vessel. The OPERABILITY of the standby gas treatment system however, ensures that release pathways do not result in a control room dose higher than determined in the refueling accident dose analysis. This is accomplished by assuring that the Standby Gas Treatment System is OPERABLE if releases from the refueling area are made through the North Stack following a FHA.
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| Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 930 seconds, these surveillance require-ments specify a draw down time of 916 seconds. This 14 second difference is due to the diesel generator starting and sequence loading delays which is not part of this surveillance requirement.
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| The reactor enclosure secondary containment draw down time analyses assumes a starting point of 0.25 inch of vacuum water gauge and worst case SGTS dirty filter flow rate of 2800 cfm. The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting the adjacent reactor enclosure and refueling area to the SGTS to split the exhaust flow between the three zones and verifying a minimum flow rate of 2800 cfm from the test zone. This simulates the worst case flow alignment and verifies ade-Suate flow is available to drawdown the test zone within the required time.
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| he Technical Specification Surveillance Requirement 4.6.5.3.b.3 is intended to be a multi -zone air balance verification without isolating any test zone.
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| The SGTS fans are sized for three zones and therefore, when aligned to a single zone or two zones, will have excess capacity to more quickly drawdown the affected zones. There is no maximum flow limit to individual zones or pairs of zones and the air balance and drawdown time are verified when all three zones are connected to the SGTS.
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| The three zone air balance verification and drawdown test will be done after any major system alteration, which is any modification which will have an effect on the SGTS flowrate such that the ability of the SGTS to drawdown the reactor enclosure to greater than or equal to 0.25 inch of vacuum water gage in less than or equal to 916 seconds could be affected.
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| LIMERICK - UNIT 1 B 3/4 6-5 Amendment No. 6,4-,74,4-0J6,J2-9,-18-5,186
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| CONTAINMENT SYSTEMS BASES 3/4.6.5 SECONDARY CONTAINMENT (Continued)
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| The field tests for bypass leakage across the SGTS charcoal adsorber and HEPA filter banks are performed at a flow rate of 5764 +/- 10% cfm. The laboratory analysis performed on the SGTS carbon samples will be tested at a velocity of 66 fpm based on the system residence time.
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| The SGTS filter train pressure drop is a function of air flow rate and filter conditions. Surveillance testing is performed using either the SGTS or drywell purge fans to provide operating convenience.
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| Each reactor enclosure secondary containment zone and refueling area secondary containment zone is tested independently to verify the design leak tightness. A design leak tightness of 2500 cfm or less for each reactor enclosure and 764 cfm or less for the refueling area at a 0.25 inch of vacuum water gage will ensure that containment integrity is maintained at an acceptable level if all zones are connected to the SGTS at the same time.
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| The Reactor Enclosure Secondary Containment Automatic Isolation Valves and Refueling Area Secondary Containment Automatic Isolation Valves can be found in the UFSAR.
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| The post-LOCA offsite dose analysis assumes a reactor enclosure secondary containment post-draw down leakage rate of 2500 cfm and certain post-accident X/Q values. While the post-accident X/O values represent a statistical inter-pretation of historical meteorological data, the highest ground level wind speed which can be associated with these values is 7 mph (Pasquill-Gifford stability Class G for a ground level release). Therefore, the surveillance requirement assures that the reactor enclosure secondary containment is verified under meteorological conditions consistentwith the assumptions utilized in the design basis analysis. Reactor Enclosure Secondary Containment leakage tests that are successfully performed at wind speeds in excess of 7 mph would also satisfy the leak rate surveillance requirements, since it shows compliance with more conservative test conditions.
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| 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL The primary containment atmospheric mixing system is provided to ensure adequate mixing of the containment atmosphere to prevent localized accumulations of hydrogen and oxygen from exceeding the lower flammability limit during post-LOCA conditions.
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| All nuclear reactors must be designed to withstand events that generate hydrogen either due to the zirconium metal water reaction in the core or due to radiolysis. The primary method to control hydrogen is to inert the primary containment. With the primary containment inert, that is, oxygen concentration
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| <4.0 volume percent (v/o), a combustible mixture cannot be present in the primary containment for any hydrogen concentration. The capability to inert the primary containment and maintain oxygen <4.0 v/o works together with Drywell Hydrogen Mixing System to provide redundant and diverse methods to mitigate events that produce hydrogen.
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| LIMERICK - UNIT 1 B 3/4 6-6 Amendment No. 8, 4-05, 4-2-9,
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| .PR-nn00122, 173
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| 3901038720 3/4.7 PLANT SYSTEMS BASES .
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| -3/4.7.1 SERVICE WATER SYSTEMS - COMMON SYSTEMS The OPERABILITY of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and -accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.
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| The RHR and ESW systems are common to Units 1 and 2 and consist of two independent subsystems each with two pumps. One pump per subsystem (loop) is powered from a Unit 1 safeguard bus and the other pump is powered from a Unit 2 safeguard bus. In order to ensure adequate onsite power sources to the systems during a loss of offsite power event, the inoperability of these supplies are restricted in system ACTION statements.
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| RHRSW is a manually operated system used for core and containment heat removal. Each of two RHRSW subsystems has one heat exchanger per unit. Each RHRSW pump provides adequate cooling for one RHR heat exchanger. By limiting operation with less than three OPERABLE RHRSW pumps with OPERABLE Diesel Generators, each unit is ensured adequate heat removal capability for the -t design scenario of LOCA/LOOP on one unit and simultaneous safe shutdown of the other unit.
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| Each ESW pump provides adequate flow to the cooling loads in its associated loop. With only two divisions of power required for LOCA mitigation of one unit and one division of power required for safe shutdown of the other unit, one ESW pump provides sufficient capacity to fulfill design requirements. ESW pumps are automatically started upon start of the associated Diesel Generators. Therefore, the allowable out of service times for OPERABLE ESW pumps and their associated Diesel Generators is limited to ensure adequate cooling during a loss of offsite power event.
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| LIMERICK - UNIT I B 3/4 7-1 AmendmentNo-.27 I
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| PLANT SYSTEMS BASES 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM The OPERABILITY of the control room emergency fresh air supply system ensures that the control room will remain habitable for operations personnel during and following all design basis accident conditions. Constant purge of the system at I cfm is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less Total Effective Dose Equivalent. This limitation is consistent with the requirements of 10 CFR Part 50.67, Accident Source Term.
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| Since the Control Room Emergency Fresh Air Supply System is not credited for filtration in OPERATIONAL CONDITIONS 4 and 5, applicability to 4 and 5 is only required to support the Chlorine and Toxic Gas design basis isolation requirements.
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| 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the emergency core cooling system equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure ex-ceeds 150 psig. This pressure is substantially below that for which low pressure core cooling systems can provide adequate core cooling.
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| The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.
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| With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCI system and justifies the specified 14 day out-of-service period. A Note prohibits the application of Specification 3.0.4.b to an inoperable RCIC system. There is an increased risk associated with entering an OPERATIONAL CONDITION or other specified condition in the Applicability with an inoperable RCIC subsystem and the provisions of Specification 3.0.4.b, which allow entry into an OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
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| The surveillance requirements provide adequate assurance that RCIC will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.
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| LIMERICK - UNIT I B 3/4 7-1a Amendment No. 2-7, 4-0, 4-6-9, 185
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| PLANT SYSTEMS BASES 3/4.7.4 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety related system.
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| Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip, lO-kip, and 100-kip capacity manufactured by Company WA" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.
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| A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall be determined and approved by the Plant Operations Review Committee. The determination shall be based upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addition or deletion of any snubber shall be made In accordance with Section 50.59 of 10 CFR Part 50.
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| The visual inspection schedule is based on the number of unacceptable snubbers found during the previous inspection in proportion to the sizes of the various snubber populations or categories. The visual inspection interval is based on a fuel cycle of up to 24 months and may be as long as two fuel cycles or 48 months depending on the number of unacceptable snubbers found during the previous visual inspection. The visual inspection schedule provides confidence that a constant level of snubber protection is being maintained and generally allows performance of visual inspections and corrective actions during plant outages. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results required a shorter inspection interval will override the previous schedule.
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| A snubber is considered inoperable if it fails the acceptance criteria of the visual inspection.
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| LIMERICK - UNIT 1 B 3/4 7-2 Amendient No. 17, 40, 51 JUN 2 5 19l
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| PLANT SYSTEMS BASES SNUBBERS (Continued)
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| To provide assurance of snubber functional reliability one of two functional testing methods is used with the stated acceptance criteria:
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| : 1. Functionally test 13.3% sample of a type of'snubber with an additional 1/2 sample tested for each functional testing failure, or
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| : 2. Functionally test 37 snubbers and determine sample acceptance using Figure 4.7.4-1.
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| Functional Testing sample plans are based on ASME/ANSI OMc-1990 Addenda to ASME/ANSI OM-1987, Part 4.
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| Figure 4.7.4-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in Quality Control and Industrial Statistics" by Acheson J. Duncan.
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| Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the completion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.
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| The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (i.e., newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.
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| 3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e,, sealed sources within radiation monitoring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
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| LIMERICK - UNIT I B 3/4 7-3 Amendment No. 4, 81 NOV 0)2 1994
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| PLANT SYSTEMS BASES 3/4 7.6 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.
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| 3/4.7.7 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.
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| DEC 2 0 IM9 LIMERICK - UNIT 1 B 3/4 7-4 Amendment No. 104
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| PLANT SYSTEMS BASES 3/4 7.8 MAIN TURBINE BYPASS SYSTEM The required OPERABILITY of. the main turbine bypass system is consistent with the assumptions of the feedwater controller failure analysis in the cycle specific transient analysis.
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| The main turbine bypass system is required to be OPERABLE to limit peak pressure in the main steam lines and-to maintain reactor pressure within acceptable limits during events that cause rapid pressurization such that the Safety Limit MCPR is not exceeded. With the main turbine bypass system inoperable, continued operation is based on the cycle specific transient analysis which has been performed for the feedwater controller failure, maximum demand with bypass failure.
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| LIMERICK - UNIT 1 B 3/4 7-5 Amendment No. 52 OCT 2 4 1991
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| 3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2. and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.
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| An offsite power source consists of all breakers, transformers, switches, interrupting devices, cabling, and controls required to transmit power from the offsite transmission network to the onsite Class 1E emergency bus or buses. The determination of the OPERABILITY of an offsite source of power can be made using three factors, that when taken together, describe the design basis calculation requirements for voltage regulation. The combination of these factors, described below, ensures that the offsite source(s), which provide power to the plant emergency buses, will be fully capable of supporting the equipment required to achieve and maintain safe shutdown during postulated accidents and transients.
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| An offsite source of electrical power is considered OPERABLE if it is within the bounds of analyzed conditions. The most limiting analysis provides the following bounds:
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| : 1. The Startup Transformer (#10 and/or #20) Load Tap Changer(s) (LTC) are in service and in automatic operation,
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| : 2. offsite source grid voltages are maintained above 218.5 kV on the 230 kV system and 498 kV on the 525 kV networks,
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| : 3. electrical buses and breaker alignments (13 kV and 4 kV) are maintained within the bounds of approved plant procedures.
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| Based on specific design analysis, variations to any of these parameters can be determined, usually at the sacrifice of another parameter, based on plant conditions. Specifics regarding these variations must be controlled by plant procedures or by operability determinations, backed by specific design calculations.
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| The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are con-sistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least two of the onsite A.C. and the corresponding D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. or D.C. source. At least two onsite A.C. and their corresponding D.C. power sources and distribution systems providing power for at least two ECCS divisions (1 Core Spray loop, 1 LPCI pump and 1 RHR pump in suppression pool cooling) are required for design basis accident mitigation as discussed in UFSAR Table 6.3-3. Under Modes 1, 2 and 3, an offsite circuit is considered to be inoperable if it is not capable of supplying at least three Unit 1 4 kV emergency buses. If both offsite sources are capable of supplying LIMERICK - UNIT I B 3/4 8-1 Amendment No. -321, 4G, 164 ECR g0 00937, ECR 99 00682
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| 3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS only three Unit 1 4 kV emergency buses, then each of the four Unit 1 4 kV emergency buses must be supplied from at least one operable offsite source.
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| Onsite A.C. operability requirements for common systems such as RHRSW and ESW are addressed in the appropriate system specification action statements.
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| The A.C. source allowable out-of-service times are based on Regulatory Guide 1.93, "Availability of Electrical Power Sources," December 1974. When one or more diesel generators are inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components, and devices, that depend on the remaining OPERABLE diesel generators as a source of emergency power, are also OPERABLE. The LPCI mode of the RHR system is considered a four train system, of which only two trains are required. The verification for LPCI is not required until two diesel generators are inoperable.
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| This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period when one or more of the diesel generators is inoperable. The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.
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| A Note prohibits the application of Specification 3.0.4.b to an inoperable diesel generator. There is an increased risk associated with entering an OPERATIONAL CONDITION or other specified condition in the Applicability with an inoperable diesel generator subsystem and the provisions of Specification 3.0.4.b, which allow entry into an OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
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| With one division with one or two battery chargers inoperable (e.g., the voltage limit of 4.8.2.1.a.2 is not maintained), the ACTIONS provide a tiered response that focuses on returning the battery to the fully charged state and restoring a fully qualified charger to OPERABLE status in a reasonable time period.
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| Action a.1 requires that the battery terminal voltage be restored to greater than or equal to the minimum established float voltage within 2 hours. This time provides for returning the inoperable charger to OPERABLE status or providing an alternate means of restoring battery terminal voltage to greater than or equal to the minimum established float voltage. Restoring the battery terminal voltage to greater than or equal to the minimum established float voltage provides good assurance that, within 18 hours, the battery will be restored to its fully charged condition (Action a.2) from any discharge that might have occurred due to the charger inoperability.
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| A discharged battery having terminal voltage of at least the minimum established float voltage indicates that the battery is on the exponential charging current portion (the second part) of its recharge cycle. The time to return a battery to its fully charged state under this condition is simply a function of the amount of the previous discharge and the recharge characteristic of the battery.
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| Thus there is good assurance of fully recharging the battery within 18 hours, avoiding a premature shutdown with its own attendant risk.
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| If established. battery terminal float voltage cannot be restored to greater than or equal to the minimum established float voltage within 2 hours, and the charger is not operating in the current-limiting mode, a faulty charger is indicated. A faulty charger that is incapable of maintaining established battery terminal float voltage does not provide assurance that it can revert to and operate properly in the current limit mode that is necessary during the recovery period following a battery discharge event that the DC system is designed for.
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| LIMERICK - UNIT I B 3/4 8-1a ECR 00 00937, ECRD 9900682, 4-64 Amendment No. 169
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| 3/4.8 ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
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| If the charger is operating in the current limit mode after 2 hours that is an indication that the battery is partially discharged and its capacity margins will be reduced. The time to return the battery to its fully charged condition in this case is a function of the battery charger capacity, the amount of loads on the associated DC system, the amount of the previous discharge, and the recharge characteristic of the battery. The charge time can be extensive, and there is not adequate assurance that it can be recharged within 18 hours (Action a.2).
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| Action a.2 requires that the battery float current be verified for Divisions 1 and 2 as < 2 amps, and for Divisions 3 and 4 as < 1 amp. This indicates that, if the battery had been discharged as the result of the inoperable battery charger, it has now been fully recharged. If at the expiration of the initial 18 hour period the battery float current is not within limits this indicates there may be additional battery problems.
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| Action a.3 limits the restoration time for the inoperable battery charger to 7 days. This action is applicable if an alternate means of restoring battery terminal voltage to greater than or equal to the minimum established float voltage has been used (e.g., balance of plant non-Class IE battery charger). The 7 days reflects a reasonable time to effect restoration of the qualified battery charger to OPERABLE status.
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| With one or more cells in one or more batteries in one division < 2.07 V, the battery cell is degraded. Per Action b.l, within 2 hours, verification of the required battery charger OPERABILITY is made by monitoring the battery terminal voltage (4.8.2.1.a.2) and of the overall battery state of charge by monitoring the battery float charge current (4.8.2.1.a.1). This assures that there is still sufficient battery capacity to perform the intended function. Therefore, with one or more cells in one or more batteries < 2.07 V, continued operation is permitted for a limited period up to 24 hours.
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| Division 1 or 2 with float current > 2 amps, or Division 3 or 4 with float current
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| > 1 amp, indicates that a partial discharge of the battery capacity has occurred. This may be due to a temporary loss of a battery charger or possibly due to one or more battery cells in a low voltage condition reflecting some loss of capacity. Per Action b.2, within 2 hours verification of the required battery charger OPERABILITY is made by monitoring the battery terminal voltage.
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| Since Actions b.l and b.2 only specify "perform," a failure of 4.8.2.1.a.1 or 4.8.2.1.a.2 acceptance criteria does not result in this Action not being met. However, if one of the Surveillance Requirements is failed the appropriate Action(s), depending on the cause of the failures, is also entered-If the Action b.2 condition is due to one or more cells in a low voltage condition but still greater than 2.07 V and float voltage is found to be satisfactory, this is not indication of a substantially discharged battery and 18 hours is a reasonable time prior to declaring the battery inoperable.
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| With one or more batteries in one division with one or more cells electrolyte level above the top of the plates, but below the minimum established design limits (i.e.,
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| greater than minimum level indication mark), the battery still retains sufficient capacity to perform the intended function. Per Action b.3, within 31 days the minimum established design limits for electrolyte level must be re-established.
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| LIMERICK - UNIT I B 3/4 8-lb ECR 00 00937, ECR 99-00682, 164
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| 3/4.8 ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
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| With electrolyte level below the top of the plates there is a potential for dryout and plate degradation. Action b.3 addresses this potential (as well as provisions in Specification 6.8.4.h, "Battery Monitoring and Maintenance Program"). Within 8 hours level is required to he restored to above the top of the plates. The Action requirement to verify that there is no leakage by visual inspection and the Specification 6.8.4.h item to initiate action to equalize and test in accordance with manufacturer's recommendation are taken from Annex D of IEEE Standard 450-1995. They are performed following the restoration of the electrolyte level to above the top of the plates. Based on the results of the manufacturer's recommended testing the battery may have to be declared inoperable and the affected cell(s) replaced.
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| Per Action b.4, with one or more batteries in one division with pilot cell temperature less than the minimum established design limits, 12 hours is allowed to restore the temperature to within limits, A low electrolyte temperature limits the current and power available. Since the battery is sized with margin, while battery capacity is degraded, sufficient capacity exists to perform the intended function and the affected battery is not required to be considered inoperable solely as a result of the pilot cell temperature not met.
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| Per Action b.5, with one or more batteries in more than one division with battery parameters not within limits there is not sufficient assurance that battery capacity has not been affected to the degree that the batteries can still perform their required function, given that multiple divisions are involved. With multiple divisions involved, this potential could result in a total loss of function on multiple systems that rely upon the batteries. The longer restoration times specified for battery parameters on one division not within limits are therefore not appropriate, and the parameters must be restored to within limits on all but one division within 2 hours.
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| When any battery parameter is outside the allowances of Actions b.l, b.2, b.3, b.4, or b.5, sufficient capacity to supply the maximum expected load requirement is not ensured and a 2 hour restoration time is appropriate. Additionally, discovering one or more batteries in one division with one or more battery cells float voltage less than 2.07 V and float current greater than limits indicates that the battery capacity may not be sufficient to perform the intended functions. The battery must therefore be restored within 2 hours.
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| The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that (1) the facility can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. Under Modes 4, 5 and *, an offsite source is considered operable if it is capable of supplying all 4 kV emergency buses necessary for operating in that Mode.
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| The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, Regulatory Guide 1.137 "Fuel-Oil Systems for Standby Diesel Generators,"
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| Revision 1, October 1979 and Regulatory Guide 1.108, LIMERICK - UNIT 1 B 3/4 8-1c L3EC, fECR 99006&2, 164
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| ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
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| "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977 except for paragraphs C.2.a(3), C.2.c(1), C.2.c(2), C.2.d(3) and C.2.d(4), and the periodic testing will be performed in accordance with the Surveillance Frequency Control Program.
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| The exceptions to Regulatory Guide 1.108 allow for gradual loading of diesel generators during testing and decreased surveillance test frequencies (in response to Generic Letter 84-15). The single largest post-accident load on each diesel generator is the RHR pump.
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| The Surveillance Requirement for removal of accumulated water from the fuel oil storage tanks is for preventive maintenance. The presence of water does not necessarily represent failure of. the Surveillance Requirement, provided the accumulated water is removed during performance of the Surveillance.
| |
| Accumulated water in the fuel oil storage tanks constitutes a collection of water at a level that can be consistently and reliably measured. The minimum level at which accumulated water can be consistently and reliably measured in the fuel oil storage tank sump is 0.25 inches. Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and
| |
| -cause fouling, but all must have a water environment in order to survive.
| |
| Removal of accumulated water from the fuel storage tanks once every (31) days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria.
| |
| Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequencies are established by Regulatory Guide 1.137.
| |
| The surveillance requirements for demonstrating the OPERABILITY of the units batteries are in accordance with the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications."
| |
| Verifying battery float current while on float charge (4.8.2.1.a.1) is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state. The float current requirements are based on the float current indicative of a charged battery.
| |
| Use of float current to determine the state of charge of the battery is consistent with IEEE Standard 450-1995.
| |
| This Surveillance Requirement (4.8.2.1.a.1) states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of 4.8.2.1.a.2. When this float voltage is not maintained, the Actions of 3.8.2.1 Action a., provides the necessary and appropriate verifications of the battery condition. Furthermore, the float current limits are established based on the float voltage range and is not directly applicable when this voltage is not maintained.
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| LIMERICK - UNIT 1 B 3/4 8-2 Amendment No. 4-0, .-
| |
| 7, 2, Q, 164,186 correction ltr. 6/19/95 ECR 97 01067
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| | |
| T2IS PAGE 3lfl2fOHZM&Y LEFT EXRAM AU9 21 M ECR 00-00937
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| | |
| ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
| |
| Verifying, per 4.8.2.1.a.2, battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady state loads of the associated DC subsystem. On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the minimum float voltage established by the battery manufacturer (2.20 Vpc, average, or 132 V at the battery terminals). This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 20 years).
| |
| Surveillance Requirements 4.8.2.1.b.1 and 4.8.2.1.c require verification that the cell float voltages are equal to or greater than 2.07 V.
| |
| The limit specified in 4.8.2.1.b.2 for electrolyte level ensures that the plates suffer no physical damage and maintains adequate electron transfer capability.
| |
| Surveillance Requirement 4.8.2.1.b.3 verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 60 degrees Fahrenheit). Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements. Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity.
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| Surveillance Requirement 4.8.2.1.d.1 verifies the design capacity of the battery chargers. According to Regulatory Guide 1.32, the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied.
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| Surveillance Requirement 4.8.2.1.d.1 requires that each battery charger be capable of supplying the amps listed for the specified charger at the minimum established float voltage for 4 hours. The ampere requirements are based on the output rating of the chargers. The voltage requirements are based on the charger voltage level after a response to a loss of AC power. This time period is sufficient for the charger temperature to have stabilized and to have been maintained for at least 2 hours.
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| A battery service test, per 4.8.2.1.d.2, is a special test of the battery's capability, as found, to satisfy .the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements as specified in the UFSAR.
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| LIMERICK - UNIT 1 B 3/4 8-2a Amendment No. 4-23, 4-64,186 ECR 97 01067
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| FIFCTRICAI POWER SYSTEMS BASES A.C. SOURCES. D.C. SOURCES. and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
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| A battery performance discharge test (4.8.2.1.e and f) is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage. Degradation (as used in 4.8.2.1.f) is indicated when the battery capacity drops more than 10% from its capacity on the previous performance test, or is below 90% of the manufacturer's rating.
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| Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying 4.8.2.1.e and 4.8.2.1.f; however, only the modified performance discharge test may be used to satisfy the battery service test requirements of 4.8.2.1.d.2.
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| LIMERICK - UNIT I B 3/4 8-2b Amendment No. 164 I
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| ELECTRICAL POWER SYSTEMS BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Bypassing motor operated valves thermal overload protection ensures that the thermal overload protection will not prevent safety related valves from performing their function. For motor operated va1ves with spring return-to-center control switches, the thermal overload is bypassed by the automatic control signals associated with the Class 1E valves. For Class IE motor operated valves with maintained contact control switches, the thermal overloads do not interrupt the valve motor power circuit, but they alarm on an overload condition in the control room. The Surveillance Requirements for demonstrating the bypassing of the thermal overload protection continuously are met by functionally testing the automatic operation of the motor operated valve and ensuring that the motor thermal overload protection design does not change and is in accordance with Regulatory Guide 1.106 "Thermal Overload Protection for Electric Motors on Motor Operated Valves", Revision 1, March 1977.
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| The RPS Electric Power Monitoring System is provided to isolate the RPS bus from the RPS/UPS inverter or an alternate power supply in the event of overvoltage, undervoltage, or underfrequency. This system protects the loads connected to the RPS bus from unacceptable voltage and frequency conditions. The essential equipment powered from the RPS buses includes the RPS logic, scram solenoids, and valve isolation logic.
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| The Allowable Values are derived from equipment design limits, corrected for calibration and instrument errors. The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection and include allowances for instrumentation uncertainties, calibration tolerances, and instrument drift.
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| The Allowable Values for the instrument settings are based on the RPS roviding power within the design ratings of the associated RPS components (e.g.,
| |
| PS logic, scram solenoids). The most limiting voltage requirement and associated line losses determine the settings of the electric power monitoring instrument channels.
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| JUN 0 9 1999 LIMERICK - UNIT I B 3/4 8-3 A78-&Mt No. 39, 93,
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| ! 141 = 134
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| | |
| THIS PAGE INTENTIONALLY LEFT BLANK 3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 REACTOR MODE SWITCH Locking the OPERABLE reactor mode switch in the Shutdown or Refuel position, as specified, ensures that the restrictions on control rod withdrawal and refueling platform movement during the refueling operations are properly activated. These conditions reinforce the refueling procedures and reduce the probability of inadvertent criticality, damage to reactor internals or fuel assemblies, and exposure of personnel to excessive radioactivity.
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| 3/4.9.2 INSTRUMENTATION The OPERABILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. The minimum count rate is not required when sixteen or fewer fuel assemblies are in the core. During a typical core reloading, two, three or four irradiated fuel assemblies will be loaded adjacent to each SRM1 to produce greater than the minimum required count rate. Loading sequences are selected to provide for a continuous multiplying medium to be established between thie required oper-able SRMs and the location of the core alteration. This enhances the ability of the SRMs to respond to the loading of each fuel assembly. During a core un-loading, the last fuel to be removed is that fuel adjacent to the SRMs.
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| 3/4.9.3 CONTROL ROD POSITION The requirement that all control rods be inserted during other CORE ALTERATIONS ensures that fuel will not be loaded into a cell without a control rod.
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| 3/4.9.4 DECAY TIME The minimum requirement for reactor subcriticality prior to fuel movement ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assump-tions used in the accident analyses.
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| 3/4.9.5 COIMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel.
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| MAYC 13N7 LIMERICK - UNIT 1 B 3/4 9-1 Amerceet 14o. 4
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| | |
| REFUELING OPERATIONS BASES 3.4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each hoist has sufficient load capacity for handling fuel assemblies and control rods, (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations, and (4) inadvertent criticality will not occur due to fuel being loaded into a unrodded cell.
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| 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a fuel assembly and associated lifting device over other fuel assemblies in the storage pool ensures that in the event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.
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| 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis.
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| 3/4.9.10 CONTROL ROD REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.
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| 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed by the RHR system to maintain adequate reactor coolant temperature.
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| RHR shutdown cooling is comprised of four (4) subsystems which make two (2) loops. Each loop consists of two (2) motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Two (2) redundant, manually controlled shutdown cooling subsystems of the RHR system provide decay heat removal. Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via the associated recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water System.
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| An OPERABLE RHR shutdown cooling subsystem consists of one (1) OPERABLE RHR pumn.
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| one (1) heat exchanger, and the associated piping and valves. The requirement for LIMERICK - UNIT 1 B 3/4 9-2 Amendment No. 88T-97, 11!
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| FEB 3 1997
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| | |
| 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION (Cont'd) having one (1) RHR shutdown cooling subsystem OPERABLE ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140'F, and 2) sufficient coolant circulation would be available through the reactor core to assure accurate temperature indication.
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| The requirement to have two (2) RHR shutdown cooling subsystems OPERABLE when there is less than 22 feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 22 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR subsystem, adequate time is provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core.
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| To meet the LCO of the two (2) subsystems OPERABLE when there is less than 22 feet of water above the reactor vessel flange, both pumps in one (1) loop or one (1) pump in each of the two (2) loops must be OPERABLE. The two (2) subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping. Additionally, each shutdown cooling subsystem can provide the required decay heat removal capability; however, ensuring operability of the other shutdown cooling subsystem provides redundancy.
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| The required cooling capacity of an alternate method of decay heat removal should be ensured by verifying its capability to maintain or reduce reactor coolant temperature either by calculation (which includes a review of component and system availability to verify that an alternate decay heat removal method is available) or by demonstration. Decay heat removal capability by ambient losses can be considered in evaluating alternate decay heat removal capability.
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| With the required decay heat removal subsystem(s) inoperable and the required alternate method(s) of decay heat removal not available in accordance with Action "a",
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| additional actions are required to minimize any potential fission product release to the environment. This includes ensuring Refueling Floor Secondary Containment is OPERABLE; one (1) Standby Gas Treatment subsystem is OPERABLE; and Secondary Containment isolation capability (i.e., one (1) Secondary Containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components.
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| If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
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| If no RHR subsystem is in operation, an alternate method of coolant circulation is required to be established within one (1) hour. The Completion Time is modified such that one (1) hour is applicable separately for each occurrence involving a loss of coolant circulation.
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| LIMERICK - UNIT 1 B 3/4 9-2a Amendment No. 44--9, ECR 01-00386
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| INTENTIONALLY LEFT BLANK 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3.4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS.
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| 3/4.10.2 ROD WORTH MINIMIZER In order to perform the tests required in the technical specifications it is necessary to bypass the sequence restraints on control rod movement. The additional surveillance requirements ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.
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| 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified in this LCO.
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| 3/4.10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.
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| 3/4.10.5 OXYGEN CONCENTRATION Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the initial startup and testing phase of operation. Without this access the startup and test program could be restricted and delayed.
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| 3/4.10.6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.
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| 3/4.10.7 RESERVED - CURRENTLY NOT USED LIMERICK - UNIT I B 3/4 10-1 Amendment No. 77, w', 133 FEB 1 1 1999
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| 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING This special test exception permits certain reactor coolant pressure tests to be performed in OPERATIONAL CONDITION 4 when the metallurgical characteristics of the reactor pressure vessel (RPV) or plant temperature control capabilities during these tests require the pressure testing at temperatures greater than 200'F and less than or equal to 212'F (normally corresponding to OPERATIONAL CONDITION 3). The additionally imposed OPERATIONAL CONDITION 3 requirements for SECONDARY CONTAINMENT INTEGRITY provide conservatism in response to an operational event.
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| Invoking the requirement for Refueling Area Secondary Containment Integrity along with the requirement for Reactor Enclosure Secondary Containment Integrity applies the requirements for Reactor Enclosure Secondary Containment Integrity to an extended area encompassing Zones I and 3. Operations with the Potential for Draining the Vessel, Core alterations, and fuel handling are prohibited in this secondary containment configuration. Drawdown and inleakage testing performed for the combined zone system alignment shall be considered adequate to demonstrate integrity of the combined zones.
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| Inservice hydrostatic testing and inservice leak pressure tests required by Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code are performed prior to the reactor going critical after a refueling outage. The minimum temperatures (at the required pressures) allowed for these tests are determined from the RPV pressure and temperature (P/T) limits required by LCO 3.4.6, Reactor Coolant System Pressure/Temperature Limits. These limits are conservatively based on the fracture toughness of the reactor vessel, taking into account anticipated vessel neutron fluence. With increased reactor fluence over time, the minimum allowable vessel temperature increases at a given pressure.
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| LIMERICK - UNIT 1 B 3/4 10-2 Amendment No. 4-3-3 ECR 99 00864, 167
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| 3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1.1 and 3/4. 11. 1.2 (Deleted)
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| THE INFORMATION FROM THESE SECTIONS HAS BEEN RELOCATED TO THE OCM.
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| LIMERICK - UNIT 1 B 3/4 11-1 Amendment No..2*,48 JAN 02 3l- I
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| RADIOACTIVE EFFLUENTS BASES 3/4.11.1.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.
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| 3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
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| Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
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| 3/4.11.2.1 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.
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| LIMERICK - UNIT 1 B 3/4 11-2 Amendment No. X, 48 JAN 02 1391
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| RADIOACTIVE EFFLUENTS BASES 3/4.11.2.2 through 3/4.11.2.4 (Deleted)
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| THE INFORMATION FROM THESE SECTIONS HAS BEEN RELOCATED TO THE ODCM.
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| B 3/4 11-3 Amendment No. 48 LIMERICK - UNIT 1 JAN 02 1991 I
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| RADIOACTIVE EFFLUENTS BASES 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the main condenser offgas treat-ment system is maintained below the flammability limits of hydrogen and oxygen.
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| Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
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| LIMERICK - UNIT 1 B 3/4 11-4 Amendment No.48 JAN 02 1991
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| RADIOACTIVE EFFLUENTS BASES 3/4.11.2.6 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to. an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.
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| 3/4 11.2.7, 3/4 11.3, and 3/4 11.4 (Deleted) - INFORMATION FROM THESE SECTIONS RELOCATED TO THE ODCM OR PCP.
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| LIMERICK - UNIT"1 B 3/4 11-5 Amendment No.48 JAN 02 1991
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| PA(ZE TNTENTTONLLY =EF BIANK 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES
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| .Section 3/4.12 (Deleted)
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| THE INFORMATION FROM THIS SECTION HAS BEEN RELOCATED TO THE ODCM. BASES PAGE B 3/4 12-2 HAS BEEN INTENTIONALLY OMITTED.
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| LIMERICK- UNIT 1 B 3/4 12-1 Amendment No.48 JAN 02 1991
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| PAGE TNTENTTONPILY L BLANK SECTION 5.0 DESIGN FEATURES AVG6 81
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| INTENTIONALLY LEFT BLANK 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.
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| LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.
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| MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 Information regarding 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 5.1.3-1a and 5.1.3-1b.
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| 5.1.4 (Deleted) 5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel lined reinforced concrete structure consisting of a drywell and suppression chamber. The drywell 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 air 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.
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| DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:
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| : a. Maximum internal pressure 55 psig.
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| : b. Maximum internal temperature: drywell 3400 F.
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| suppression pool 220 0 F.
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| : c. Maximum external to internal differential pressure 5 psid.
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| : d. Maximum floor differential pressure: 30 psid, downward.
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| 20 psid, upward.
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| LIMERICK - UNIT 1 5-1 Amendment No. 3-3,48 JAN 02 1951
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| 3901000402
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| *- 1? ' PRICK .ua t m .. _
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| LIEI fi LOKtie1 Y--" N SIO EXCL ARE FIGURE 5. 1. 1-1 EXCLUSION AREA AUG 8 1;8 LIMERICK - UNIT 1 5-2
| |
| | |
| FIGURE 5.1.2-1 LOW POPULATION ZONE LIMERICK - UNIT I 5-3
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| FIGURE 5.1.3-la MAP DEFINING UNRESTRICTED AREAS AND SITE'BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS AUS S I9*5 LIMERICK - UNIT 1 5-4
| |
| | |
| OMML RILIAU PONTS
| |
| *AlSAVC COOAOG.ACs.
| |
| / ....
| |
| 1 flJf*J~ s1 e.
| |
| e**,.ooU,* viol POINTS OF/
| |
| GASEOUS --...(--Z -foc.1-.In o Naco.
| |
| RELEASE **CI... , SOUTH STACK) 50l fl .22..CV0.V mO*jRA
| |
| *a Dim ,
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| FIGURE 5.1.3-lb MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS LIMERICK - UNIT 1 5-5 AVE&rgr
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| THE FIGURE ON THIS PAGE HAS BEEN RELOCATED TO THE ODCM.
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| LIMERICK - UNIT I 5-6 Amendment No.48 JAK 02 111
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| | |
| DESIGN FEATURES SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of three distinct isolatable zones.
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| Zones I and II are the Unit I and Unit 2 reactor enclosures respectively.
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| 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.
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| The common refueling area (Zone III) completely encloses and provides secondary containment for the 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.
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| 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall consist of not moretthan 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).
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| CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform-shaped control rod assemblies.
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| 5.4 REACTOR COOLANT SYSTEM 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 to. 19 APR 2 4 1989
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| | |
| D-ESIGN FEATURES nrTENI DOMMIID AND TEMPERATURE (Continued)
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| DLJSIN F ATURES
| |
| : 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 575°F.
| |
| 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 saturatipn temperature of 552*F. I 5.5 FUEL STORAGE CRITIALITY 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.
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| 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.
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| 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 I 5-8 Amendment No. 72,8Z,106 FEB 12 1996
| |
| | |
| K m TABLE 5.6.1-1 7,
| |
| COMPONENT CYCLIC OR TRANSIENT LIMITS C
| |
| z'-4 CYCLIC OR
| |
| -4 DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT
| |
| '-a Reactor 120 heatup and cooldown cycles 70°F to 560°F to 70°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 I0
| |
| | |
| PA('E TN=I¶PONAIJY LEP7 BUZNK SECTION 6.0 ADMINISTRATIVE CONTROLS
| |
| | |
| INTENTIONALLY LEFT BLANK 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall 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.
| |
| AUG 0 8 1995 LIMERICK - UNIT 1 6-1 Amendment No. 10, 35, 96
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| | |
| 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. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions (e.g., licensed Senior Operators, licensed Operators, health physi-cists, auxiliary operators, and key maintenance peisonnel).
| |
| Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40-hour week while the unit is operating.
| |
| However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shut-down for refueling, major maintenance, or major unit modifications, on a temporary basis the following guidelines shall be followed:
| |
| : 1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time.
| |
| : 2. An individual should not be permitted to work more than 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period, all excluding shift turnover time.
| |
| : 3. A break of at least 8 hours should be allowed between work periods, including shift turnover time.
| |
| : 4. Except during extended shutdown periods, the use of overtime should be considered on an Individual basis and not for the entire staff on a shift.
| |
| *The Health Physics Technician position may be less than the minimum requirements for a period of time not to exceed 2 hours, in order to accommodate unexpected absence, provided immediate action is taken to fill the required position.
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| LIMERICK - UNIT 1 6-2 Amendment No. 20, 5, 00, 104, 139 MAR 14 2000
| |
| | |
| ADMIINISTRATIVE CONTRL UNIT STAFF (Continued)
| |
| Any deviation from the above guidelines shall be authorized by the Plant Manager or personnel designated in administrative procedures or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager, or the appropriate designated personnel to assure that excessive hours have not been assigned.
| |
| Routine deviation from the above guidelines is not authorized; and
| |
| : 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.
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| LIMERICK - UNIT 1 6-3 Amendment No. 0, 3-, Q-3, 8-0, -6, 162
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| | |
| I This page is intentionally left blank.
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| LIMERICK - UNIT 1 6-4 DEC 19 7e98
| |
| | |
| TA&A.E 6.Z.2-i MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMMON CONTROL ROOM WITH UNIT 2 IN CONDITION 4 OR 5 OR DEFUELED 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 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.
| |
| I
| |
| ** One of the two required individuals may fill the same position on Unit 2.
| |
| '**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 I 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 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.
| |
| AUG 0 8 1995 LIMERICK - UNIT 1 6-5 Amendment No. 28,3234,96
| |
| | |
| ADMI!NISTRATIVE CONTROLS 6.2.3 DELETED. The information from this section is located in the UFSAR.
| |
| 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 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 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1-1978 for comparable positions, except for the Manager - Radiation Protection who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and the licensed operators who shall comply with the requirements of 10CFR55.
| |
| I LIMERICK - UNIT I Amendment No.YO, ý7, ýý, 04, 96 6-6 Corrected by Letter dated October 17, 1995
| |
| | |
| ADMINISTRATIVE CONTROLS 6.4 DELETED 6.5 DELETED THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE OATR LIMERICK - UNIT 1 6-7 Amendment No. -10, 33, S, 96, 4-19, 4-0, 4*24, 176 6,0P[eU L
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| | |
| ADMINISTRATIVE CONTROLS THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT 1 6-8 Amendment No. -I-0, 4-7, 4-04, 4-34, 4--54, 176
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| | |
| ADMINISTRATIVE CONTROLS THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT I 6-9 Amendment No. 0, 3-5, -6, 1-4, 176
| |
| | |
| AnwMNTSTPATI\IF cflNTRfli THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT I 6-10 Amendment No. -%, 176
| |
| | |
| THIS PAGE IS INTENTIONALLY LEFT BLANK.
| |
| AUG 0 8 1995 LIMERICK - UNIT I 6-]1 Amendment No. 10,39,96
| |
| | |
| ADMINISTRATIVE CONTROLS THE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATR LIMERICK - UNIT I 6-12 Amendment No. 4-,5,2 4-, 96, 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.
| |
| 6.7 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.
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| LIMERICK - UNIT I 6-12a Amendment No. 4-7, -, 41-4, 176
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| | |
| INTENTIONALLY LEFT BLANK 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 coveri ing 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. Ouality 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 I 6-13 Amendment No. 710, 4-7, -94, 4-3-9, 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 I 6-14 Amendment No. 29, 166
| |
| | |
| ADMINISTRATIVE CONTROLS PROCEDURES 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 operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
| |
| : 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,
| |
| : 2) Limitations on the concentrations of radioactive material released in liquid effluents to-UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2,
| |
| : 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 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 unitto UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,
| |
| : 5) Determination of cumulative and projected 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,
| |
| : 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) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1, LIMERICK - UNIT 1 6-14a Amendment No.48 JAN 02 1951
| |
| | |
| 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-lives greater 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 Program", 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 provided to monitor the radiation and radionuclides in the environs 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 of environmental exposure pathways. The program shall (1) be contained in the ODCM, (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 monitoring program 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 Xmendment No.48 JAN 02 1991
| |
| | |
| ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
| |
| : g. Primary Containment Leakage Rate Testina ProQram A program shall 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.
| |
| The peak calculated containment internal pressure for the design basis loss of coolant accident, P,, 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 P,.
| |
| : 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 Rate Testing 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. 44-18, 162
| |
| | |
| ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
| |
| : i. Battery Monitoring and Maintenance Program This Program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries For Stationary Applications," of the following:
| |
| : a. Actions to restore battery cells with 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 NEI 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.
| |
| LIMERICK - UNIT I 6-14d Amendment No. 4-44, 186
| |
| | |
| 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 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level,-(3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.
| |
| 6.9.1.2 The startup report shall address each of the tests identified in Sub-.
| |
| section 14.2.12 of the Final Safety Analysis Report and shall include a descrip-tion of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
| |
| 6.9.1.3 Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commence-ment of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.
| |
| 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 I 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 I 6-15 Amendment No. 4-3-5, 175
| |
| | |
| ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued)
| |
| : b. (Deleted)
| |
| : c. Any other unit unique reports required on an annual basis.
| |
| : d. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Cleanup system flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
| |
| 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 I 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. 2-9, 4A-,48, 1-3-5, 4-9, 175
| |
| | |
| 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.
| |
| report shall include a summary of the quantities of radioactive The I 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.
| |
| *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.
| |
| JUN 2 2 1999
| |
| .I . UT 1en6-t1No. P, &9, ,135 LIMERICK - UNIT 1 6-17
| |
| | |
| I INTENTIONALLY LEFT BLANK LIMERICK - UNIT I 6-18 Amendment No.,,,r48 JAN 02 1991
| |
| | |
| ADMINISTRATIVE CONTROLS 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 MAPFAC(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,
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| : e. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,
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| : f. The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6,
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| : g. The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,
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| : h. The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1.
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| 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:
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| : a. NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Latest approved revision),*
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| : b. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August. 1996.
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| 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.
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| 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.
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| 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.
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| * For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98) to support License Amendment No. (127).
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| LIMERICK - UNIT 1 6-18a Amendment No. 4-21-, 4-42, 177
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| INTENTIONALLY LEFT BLANK ADMINISTRATIVE CONTROLS 6.10 DELETED THE INFORMATION FROM SECTION 6.10 HAS BEEN RELOCATED TO THE QATR.
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| LIMERICK - UNIT I 6-19 Amendment No. 4-, 176
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| 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.
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| 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
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| .paragraph 20.1601(a) and (b) of 10 CFR Part 20:
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| 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):
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| : 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.
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| LIMERICK - UNIT I 6-20 Amendment No. 4-8, 176 41-3-5,
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| ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)
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| : 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 -
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| rates) is included on the ladiation surveys associated with the RWP or equivalent.
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| 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.
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| : d. Each individual or group entering such an area shall be provided with or accompanied by one or more of the following:
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| : 1. A radiation monitoring device that continuously displays radiation dose rates in the area ("radiation monitoring and indicating device"), OR
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| : 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
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| : 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
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| : 4. A direct reading dosimeter AND:
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| a) A health physics qualified individual (i.e.,
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| 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),
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| responsible for controlling personnel radiation exposure in the area.
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| : 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.
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| LIMERICK - UNIT 1 6-20a JU 2 219 ATencirent Nob. 131;
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| ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) 6.12.2 In addition to the requirements of Section 6.12.1, High 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:
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| : 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:
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| : 1. All such door and gate keys shall be maintained under the administrative control of radiation protection supervision.
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| : 2. Doors and gates shall remain locked or guarded except during periods of personnel or equipment entry or exit.
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| : 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.
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| : c. Each individual entering such an area shall be provided with or accompanied by one or more of the following:
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| : 1. A dose rate survey meter and a radiation monitoring device with the capability to display accumulated dose and an integrating alarm setpoint, OR
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| : 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
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| : 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
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| : 4. A direct reading dosimeter AND:
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| JUN 2 2 1999 LIMERICK - UNIT 1 6-21 ATendTent Nb. 48, 135
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| ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) a) A health physics qualified individual (i.e.,
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| 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),
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| responsible for controlling personnel radiation exposure in the area, and with the means to communicate with the individuals in the area.
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| 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 Changes to the PCP:
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| : a. Shall be documented with the following information:
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| : 1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and LIMERICK - UNIT I 6-21a Amendment No. 4-3-5, 176
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| | |
| 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.
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| : b. Shall become effective upon review and acceptance by the PORC and approval of the Plant Manager.
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| 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 Changes to the ODCM:
| |
| : a. Shall be documented with the following information:
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| : 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 20.106, 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 PORC 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.
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| 6.15 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.
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| LIMERICK - UNIT I 6-22 Amendment No. 3-5, 48, 7-3, 176
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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 I Amendment No. 4-34-, 44-, 180,
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| INTENTIONALLY LEFT BLANK LIMERICK GENERATING STATION UNIT I ENVIRONMENTAL PROTECTION PLAN (NONRADIO LOGICAL)
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| TABLE OF CONTENTS Section Paqe 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 .................... 5-1 5.4 Plant Reporting Requirements ................................ 5-2 Amendment No. 180
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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 stuoy 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)
| |
| (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 Ij
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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
| |
| | |
| 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.
| |
| 3.2 Reporting Related to the NPDES Permit and State Certification Changes to, or renewals of, the NPDES Permits or the State certification shall be reported to the NRC within 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
| |
| | |
| 4.0 Environmental Conditions 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.
| |
| 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.
| |
| 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
| |
| | |
| 4.2.2 Terrestrial Monitoring No terrestrial monitoring is.required.
| |
| 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
| |
| | |
| 4.2.4 Noise Monitoring All initial Environmental Noise Assessments have been completed.
| |
| The information in Subsections 4.2.4.1, 4.2.4.2.
| |
| 4.2.4.3, and 4.2.4.4 has been deleted.
| |
| 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
| |
| | |
| INTENTIONALLY LEFT BLANK INTENTIONALLY LEFT BLANK 5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the EPP.
| |
| 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.
| |
| 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.
| |
| 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.
| |
| All other records relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
| |
| 5-1 Amendment No. 180
| |
| | |
| 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.
| |
| 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.
| |
| 5.4 Plant Reporting Requirements 5.4.1 Deleted 5-2 Amendment No.180
| |
| | |
| 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
| |
| | |
| 4T".Ion 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.
| |
| 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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| PAGE NTERTION-LLY LEFT BIANK APPENDIX C ADDITIONAL CONDITIONS OPERATING LICENSE NO. NPF-39 DELETED Amendment No. 428,4-34,4-47,184
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| INTENTIONALLY LEFT BLANK}}
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