ML20086C859

Amend 154 to License DPR-51,revising RCS TS Pressure/Temp Operating Limits for First 15 Effective Full Power Yrs

ML20086C859
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/14/1991
From:	Larkins J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20086C863	List: ML20086C859 ML20086C866
References
NUDOCS 9111250127
Download: ML20086C859 (10)

Text

{{#Wiki_filter:y.e /p. oe.u p, UNITED STATES 2 NUCLEAR REGULATORY COMMISSION 3m n i ~. WASHINGTON, D C.10665 o %,'...+,/ ENTERGY OPERATIONS INC. DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO.1 AMENDtiENT TO FACILITY OPERATING LICENSE Ateret.it No.154 License No. U R 51 1. The huclear Regulatory Cornission (the Conn.ission) h s found that: A. The application for anendtient by Entergy Operations, Inc. (the licensee)datedSeptember 20, 1990, as supplemented by letters dated February 28, and August 14, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as nended (the Act), and the Concission's rules and regulations set furth in 10 CFR Chapter I; B. The f acility will operate in conf rmity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (1) that the activitits authorized by this c.rendirent can be conducted without eridangering the health and St.f ety of the public, ar.d (ii) tnat such activities will be conducted iri complianct with the Conmission's regulations; D. The issuance of this license amenduent will rot be inimical tc the cot.vaun defense and security or to the health and safety of the public; and E. The issuance of this amendnent is in accorcance with 10 CFR part 51 of the Cormission's regulations and all applicable requirements have been satisfied. 9111250127 '711114 PDR ADOCK 0U000313 p PDH ( l

V '7 i 2. Accordingly, the license is amended by changes to the Technical Specificatior.:. as indicated in the attachment to this license arendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-51 is hereby amended to read as follows: 2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.154, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. The license amendment is effective 30 days from the date of issuance. FOR THE NUCLEAR REGULATORY COMMIS$10N l f,s John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects 111 IV, and V Office of Nuclear Reactor Regulation Attathment: Changes to the Technical Specifications Date ci Issuance: November 14. 1991

9 l ATTACitMENT TO LICENSE AMENDMENT NO.154 FACILITY OPERATING L,1 CENSE NO. DPR-51 DOCKET NO. 50-313 Revise the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment nuntier and contain vertical lines indicating the area of change. REQ VE PAGES INSERT PAGES 1 18 IBa 18a 19 19 20 20 20a 20a 20b tob 23c 20c I ..c,,,---r---,-w...

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3.1.2 Pressurization, }Lestuo. end Cooldown Limitations i $pecificaliQD 3.1.2.1 Hydro Tests Tor thermal steady state system hydro tests, the system may be pressurized to the limits set forth in Specification 2.2 when there are fuel assemblies in the core, under the provisions of 3.1.2.3, and to ASME Code limits when no fuel assemblies are present provided the reactor coolant system limits are to the right of and below the limit line in Figure 3.1.2-1. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. 3.1.2.2 Leak Tests Leak tests required by Specification 4.3 shall be conducted under the provision of 3.1.2.3. The provisions of Specification 3.0.3 are not applicabic. 3.1.2.3 The reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited ir accordance with Figure 3.1.2-2 and Figure 3.1.2-3, and are as follows: Hestup: AllowaHe combinations of pressure and temperature shall be to the right of and below the limit line in Tigure 3.1.2-2. The heatup rates shall not exceed those shown in Figure 3.1.2-2. Cooldown: Allowable combinations of pressure and temperature for a specific cooldowr. shall be to the right of and below the limit line in Figure 3.1.2-3. Cooldown rates shall not exceed those shown in Figure 3.1.2-3. 3.1.2.4 The secendary side of the steam generator shall not be pressurired above 200 psig if the terperature of the steen generator shell is below 100T. 3.1.2.5 .he pressurizer bestup and cooldown rates shall not exceed 100r/hr. The spray shall not be used if the temperature dif ference between the pressuriter and the spray fluid is greater then 430T. 3.1.2.6 Vith the limits of Specifications 3.1.2.3 or 3.1.2.4 or 3.1.2.5 exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evalustion to determine the ef fects of the out-of-linit condition on the fracture toughness properties of the Reactor Coolant Syster; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HDT STANDPY vitM n the next f hours and reduce the FCS Tavg to less than 2M T, while reintaining RCS temperat ure and pressure below the curve. w it bin the fnllowirg 3C hours. At endr e ' \\'. 21, 2 f. 11, 154 1F i

~. (? J.1.2.7 Prior to reaching fifteen effective full powrr years of operation, Tigures 3.1.2-1, 3.1.2*2 and 3.1.2-3 shall be updated for the next service period in accordance: with 10CpR50, Appendix G, Section V.B. The service period shall be of sufficient duration to permit the scheduled evaluation of a portion of the surveillance data scheduled in accordance with the latest revision of Topical Report BAW-1543(5). The highest predicted adjusted reference temperature of all the beltline region materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.8. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. 3.1.2.6 The updated proposed technical specifications referred to in 3.1.2.7 shall be submitted for NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CpR part 50, Appendix G. Section V.C. 3.1.2.9 With the exception of ASME Section XI testing and when the core flood tank is.depressurized, during a plant cooldown the core flood tank discharge valves shall be closed and the circuit breakers for the motor operators opened before depressurizing the reactor coolant system below 600 psig. 3.1.2.10 With the exception of ASME Section XI testing, fill and vent of the reactor coolant system, and to allow maintenance of the valves, when the reactor coolant terperature is less than 300'p the four High l Pressure Injection motor operated valves shall be closed with their opening control circuits for the motor operators disabled. 3.1.2.11 The plant shall not be operated in a water solid condition when the RCS pressure boundary is intact except as allowed by Emergency Operating procedures and during System Hydrotect. l l l Amendment No. 1/, f7, 75, ygg, ygg, 1Ba ARKANSAS - UNIT 1 154

- ~. - -. -. _ ) ASIS All reactor coolant system components are designed to withstand the ef fects of cyclic loads due to system temperature and pressure changes.(8) These cyclic loads are introduced by unit load transienta, reactor trips, and unit heatup and cooldown operations. The number of thermal end loading cycles used for design purposes are shown in Table 4-8 of the FSAR. The maximum unit heatup and cooldown rates satisfy stress limits for cyclic operation.(8) The 200 psig l i pressure limit for the secondary side of the steam generator at a temperature less. than 100F satisfies stress levels for temperatures below the DTT.(') The major comporp+ s of the reactor coolant pressure boundary have been analyzed in cecordance with Appendix G to 10CTR50. Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in BAW-2106('). The limiting weld material is being l frradiated as part of the B&W Owners Group Integrated Reactor Vessel Material Surveillance Program and the identification and locations of the capsules containing the limiting weld snaterial is discussed in the latest revision to B&W report, BAW-1543. (') The chemical compos ion of the limiting weld material is reported in the B&W Report, BAW-1511P.(') The effect of neutron irradiction on the R1 of the limiting weld material is reported in the B&W Report, BAW 2075(h l Figures 3.1.2-1, 3.1.2-2, and 3.1.2-3 present the pressure-temperature limit curves for hydrostatic test, normal bentup, and normal cocidown respectively. The livnit curves are applicable through the fifteenth effective full power year of operation. The pressure limit is adjusted fer the pressure dif ferential l between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations. The pressure-temperature limit lines shown on Figure 3.1.2-2 for reactor criticality and on Tigure 3.1.2-1 for hydrostatic testing have been provided to assure coep}iance with the minirnum temperature requirements of Appendix G to 10CFK50 for reactor criticality and for inservice hydrostatic testing. The actual shift in Ri of the beltline region material will be established periodicallyduringop'r1tionbyremovingandevaluating,inaccordancewith g e Appendix H to 10CTK50, reactor vessel material irradiation surveillante specimens which are installed near the inside wall of this or a similar reactor vessel in the core region. The spray temperature difference restriction based on a stress analysis of the sprar line nozzle is irnposed to mofntain the thermal stresses et the pr< ssurizer spray line nozzle below th design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell. I I l Amendment No. 22, 2f,17, f f, 154 19 l _y .,_m ,m .,c., ~. ____.m.,

n~ 4 4 The heatup and cooldown rates stated in this specification are intended as the maximum changes in temperature in one direction in a one hour period. The actual temperature linear ramp rate may exceed the stated limits for a time period provided that the maximum total temperature difference does not exceed the limit and that a temperature hold is observed to prevent the total temperature difference from exceeding the limit for the one hour period. Specification 3.1.2.9 is to ensure that the core flood tanks are not the source for pressurizing the reactot coolant system when in cold shutdown, Specification 3.1.2.10 is to ensure that high pressure injection is not the source of pressurizing the reacior coolant system when in cold shutdown. Specification 3.1.2.11 is to ensure that the reactor coolant system is not operated in a manner which would allow overpressurization due to a temperature transient. RETEEENCES (1) FSAR, Section 4.1.2.4 (2) ASME Boiler and Pressure Code, Section III, N-415 (3) TSAR, Section 4.3.11.5 (4) PAW-2106 (5) BAV-1543, latest revision l (6) BAW-1511P (7) BAV-2075, Kevision 1 Amendment No. 2, 22, 2f, ft, 71, 154 20 ARKANSAS - UNIT 1

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