ML20203L249
| ML20203L249 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 02/27/1998 |
| From: | Bajwa S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203L252 | List: |
| References | |
| NUDOCS 9803050416 | |
| Download: ML20203L249 (10) | |
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{{#Wiki_filter:. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ _ - Cs:vq po t UNITED STATES s j NUCLEAR REGULATORY COMMISSION "o WASHINGTON, D.C. 306M 0001 %...../ POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50 3?) JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 242 Ucense No. OPR 59 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Power Authority of the State o' New York (the licensee) dated December 14,1995, as supplemented September 26,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the applicaticn, the provisions of the Act, and the rules and regulatior,s of the Oommission; C. There is reasonable assurance (1) that the activities authorized by this amendment 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 amendment 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 amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applit.able requirements have been satisfied. 2. Acccrdingly, the license is emended by changet to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR 59 is hereby amended to read as follows: 9803050416 980227 PDR ADOCK 05000333 p PDR
2-(2) Technical Specificatio.nl The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 242, are hereby incorporated in the license. Tbe licensee shall operate the facility in e a,cordance with the Technical Specifications. 3. This license amendment is effective as of the date of its issuunce to be implemented within 90 days. FOR THE NUCLEAR REGULATORY COMMISSION S. Singh Bajwa, Director Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical y Specifications Date of Issuance: February 27, 1998 ________.___________________-_____.________________m
ATTACHMENT TO LICENSE AMENDMENT NO. 242 FACILITY OPERATING LICENSE NO. DPR-59 QQQKET NO. 50-331 Revise Appendix A as follows: Remove Paaes Insert Paoes 132 132 -133 133 177 177 178 178 185 185 186 186 197 197 k K-------
JAFNPP 4.5 BASES The testing interval for the Core and Containment Cooling With components or subsystems out-of-service, overall core and Systems is based on a quantitative reliability analysis, industry containment cooling reliability is maintained by verifying the practice, judgement, and practicality. The Emergency Core operability of the remaining cooling equipment. Consistent with Cooling Systems have not been designed to be fully testable the definition of operable in Sectior. 4.0.C, demonstrate means ( during operation. For example, the core spray final admission conduct a test to show; verify means that the associated vstves do not open until reactor pressure has fallen to 450 psig; surveillance activities have been satisfacterify performed within the l thus, during operation even if high drywell pressure were specified time interval. simulated, the final valves would not open. In the case of the The RCIC ficw rate is described in the UFSAR. The flow rates to HFCI, automatic initietion during power operation would result in be delivered to the rasctor core for HPCI, the LPCI mode of RHR, pumping cold water into the reactor vessel which is not and CS are based on the SAFER /GESTR LOCA analysis. The flow desirable. rates for the LPCI mode of RHR and CS are modified by a 10 i percent reduction from the SAFER /GESTR LOCA analysis. The The systems will be automatically actuated once per 24 months. reductions are based on a sensitivity analysis (General Electnc in the case of the Core Spray System, condensate storage tank water will be pumped to the vessel to verify the operability of MDE-93-0786) performed for the parameters used in the the core spray heade. On a monthly basis, correct alignment SAFER /GESTR analysis. shall be verified for manual, power operated, or automatic valves in ECCS and RCIC system flow pa.hs to provide assurance that The CS surveillance requirement includes an allowance for system proper flow paths will exist for system operation. For the HPCI leakage in addition to the flow rate required to be delivered to the and RCIC Systems, this reouirement also includes the steam reactor core. The leak rate from the core spray piping inside *he flowpath for the turbines and the flow controller position. This reactor but outside the core shroud is assumed in the UFSAN d surveihance requirement does not apply to valves that cannot be includes a known loss of less than 20 gpm from the 1/4 mch inadvertently misaligned such as check valves, or to valves that diameter vent hole in the core spray T-box connection in each of are locked, sealed, or otherwise secured in position. A valve the loops, and in the B loop, a potential additional loss of less than that receives an initiation signal is allowed to be in a non-40 gpm from a clamshell repair whose structural weld covers only accident position provided the valve willautomatically reposition 5/6 of the circumference of the pipe. Both of these identified in the proper stroke time upon receipt of the initiation signal. sources of leakage occur in the space between the reactor vessel The monthly frequency of this requirement is based upon wall and the core shroud. Therefore flow lost through these leak engineering judgement r 1 is supported by procedural controls sources does not contribute to core cooling. goveming valve operatic; 1 hat ensure correct valve positions. This frequency is further supported by the Inservice Testing Program, which demonstrates system pump and power operated valve operability. This combination ef automatic actuation tests, periodic pump and valve testing, and monthly flow path verification is adequate to demonstrate operability of these systems. Amendment No. 'i, ' 4 8, Mi, 233 242 132 ~ a
JAF9fPP / i a.5 BASES (cent'd The aunmiNonce requirements *o onoure that the dW piping of the core spray, LPCI mode of the RHR, HPCl, and RCIC Systems are fined provides for a visual observation that water flows frorn a high point vent. This ensures that the line is in a fun condition. Instrumenteelon has been provided in the Core Spray Syshun and LPCI System to monitor the presence of water ir - the diecharge piping. This instrumentation is functionety tested montMy to ensure that Juring ths hterval l betweem the montMy checks the etstas of the discharge piping is moedtered on a continuous bes:4. i Normally the low pressure EN subsystems required by l Speci'ication 3.5.F4 are den..h operable tiy the ourveillence testsin SMhtions 4.li.A.1 and 4.5.A.3. Section i 4.E.F cpecMies periorAc esvoillence te for the low pressure ECCS Nbsystems which are applicable whos, the reactor is in the cold condition. These tests in conjunction with the l requirements on fiBod docharge piping CM 3.5.G), and the requirements 'on ECCS actustkee ins:trumsntation j (Specification 3.2.2), assure adequete ECCS capetdlity in the cohl condition. The wotor level h the e-; =e::':-r. pool, er the l Cond-mete Storage Tanks (CST, when the suppreseien pool is j inoperab6e, is checked once occh shift to ensure that sufficioni water is reveliable for core cooling. [ f t t i v F AmenderEnt No. *', ?SS 242 -{ 133
JAFNPP 4.7 { cont'd) 5.7 (cont'd) - e 4. Pressure Suppression Chamber Reactor Building Vacuum
- 4. Pressure Suppression Chamber-Reactor Buildig Vacuum Breakers Breakers The pressure suppression chamber-reactor building Except as specified in 3.7.A.4.b belovi, two a.
a. Pressure Suppression Chamber Reactor Building vacuum breakers shall be checked for proper Vacuum Breakers shall be operable at all times operation in accordance with the inservice Testing when the primary containmer.t integrity is required. Program. The setpoint of the differential pressure instrumentation which actuates the pressure b. Instrumentation associated with pressure suppiession chamber reactor bui! ding vacuum suppression chamber-reactor building vacuum breakers shall be 50.5 e.u below reactor building breakers shall be functionally tested once per 92 days. pressure. b. From and after t'ie date that one of the pressure suppression chamber reactnr building vacuum breakers is made or found to be inoperable for any reason. reactor operation is permissible only during the succeeding 7 days, unless such vacuum Amendment No. 130,131,139 242 177
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l JAFNPP 4.7 (c mt'd) 3.7 (cont'd) breaker is soor**r made opecable, provided that the repair proceduta does not vidate primary containment integrity. 5. l Pressure Suppressinn Chamber - Drywell Vacuum 5. Piessura Supprestsion Chem a - DryweN Vacuum Breakers Breakers a. Each dryweN suppression cher9bor vacuum brooker When Mmary contaiomont integrity is quired, all a. drvweN suppression chamber vacuum breakers shall shen be exercised through an opening - closing cycle be operable and positioned in the fully closed monthly. position except during testing and as specified h 3.7.A.5.b below. b. One drywell suppression chamber vacuuri breaker b. When it is riotermirnd that one vacuum breaker in may ba non-fully closed so long as it is determined inopereble io-fully closing when operebility is to be not more than 1
- open as indicated by the required, the operable breskers shaN be exercised immediately, and every 15 days thereafter untn the position lights.
inoperable valve has been returned to normal service. c. Each vacuum breaker valve shell be viscelly l l c. One drywell suporession chamber vacuum breaker may be determined to be inoperable for opening. inspected to insure proper mantenance and l operation in accordance with the Inservice Testing Program. d. Deleted d. A leak test of the drywell to suppression chamber ucture shall be conducted once per 24 months; me acceptable let k rate is s0.25 in. water / min, over a 10 min period, with t' dr /well at 1 lesid. Amendment No. 131,192,232 242 178 \\ >
JAFNPP 3.7 (cent'd) 4.7 (cont'd) c. Secondary containment capsbility to maintain a 1/4 in. t of water vacuum under calm wind conditions with a i filter train flow rate of not morts than 6,000 cim, shall be demonstrated at cach refueling outage prior to refueling. G. Primarv Containment isolation Valves D. Primary Containment I?olation Valves 1 .Whenever primary cm::aintnent integrity is required per 1. The pnmary containment isolation valves surveillance shall
- 3.7.A.2, containment esolation valves and allinstrument be performed as follows:
line exces: flow check valves shall be operable, except as specified in 3.7.D.2. The containment vent and purge item Frecuerx;v vetves shall be limited to opening angles less than or equal to that specified below: a. The operabie isolation in accordance with vaives that are power the Inservice Valve Number Maxirnum Openina Anale merated and Testir.g Program 27AOV-111 40* automatically initiated 27AOV-112 40 shall be tested for 27AOV-113 40' simulated automatic 27AOV-114 50' initiation and for 27AOV-115 50* closure time. l 27AOV-116 50* i 27AOV-117 50' b. Instrument line er. cess in accordance with 27AOV-118 50' flow check valves shall the Inservice be tested for proper Testing Program operation. c. All normally open power-In accordanca with operated Isolation valves t'se Inservice (except for the main Testing Program steam isolation valves) shall be fully closed and reopened. 1 Amendment No. 15', '72, 195, 232, 233 242 185
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JAFNPP i - 4.7 BASES (cont'd) I The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability. i The primary contenment is penetrated by several small diameter instrument lines connected 9 the reactor coolant system. Each instrument line contains a 0.25 in, restricting orifice inside the primary containment and an excess flow check valve outside the primary containtrant. I A list of containment isolation valves, iceluding a brief description of each valve is included in Section 7.3 of the updated FSAR. t i i i t i i Amendment No. 151,*?3, 203, 232 242 197 l ..}}