ML20154N776
| ML20154N776 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/24/1988 |
| From: | Perkins K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20154N782 | List: |
| References | |
| NUDOCS 8806030075 | |
| Download: ML20154N776 (11) | |
Text
{{#Wiki_filter:-__. e f louq'o d UNITED STATES ' g, NUCLEAR REGULATORY COMMISSION o h WASHING TON, D. C. 20555 o %,...../a s UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. STN 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 License No. NPF-30 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment filed by Union Electric Company (UE, the licensee) dated July 31, 1987, supplemented February 19, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 1 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 Comission'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 applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2)~ of Facility Operating License No. NPF-30 is hereby amended to read as follows: 8806030075 e00524 PDR ADOCK 05000403 P PDR _., _.. - _ _ _.. _ _. _ _ _ _ ~. _ , - _ _.. _ _ _. - _ _ _ _. _ ~ _.
. (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.36, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION Kenneth E. Perkins, Dire or Project Directorate III-3 Division of Reactor Projects - III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: May 24, 1988 i .-. a
i i ATTACHMENT TO LICENSE AMENDMENT N0. 36 OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Corresponding overleaf pages are provided to maintain document completeness. REMOVE INSERT 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 3/4 4-36 3/4 4-36 8 3/4 4-7 8 3/4 4-7 l l . r... ---,,. -.. ~
REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CON 0! TION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit li es shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing 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, and A maximum temperature change of less than or ecual to 10'F in any c. 1-hour period during inservice hydrostatic and leak testing operations 3bove the heatup and cooldown limit curves. APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the ef fects 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 operation or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T,yg arid pressure to less than 200'F and 500 psig, respectively, within the following 30 hours. SURVEILLANCE REQUIREMENTS l 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least one.a per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2, 3.4-3, and 3.4-4. CALLAWAY - UNIT 1 3/4 4-29
MATERIAL PROPERTY BASIS CO EROLLING MATERIAL: R.V. LOWER SHELL COPPER CONTENT: 0.07 WT5 NICKEL CONTENT: 0.59 WT5 1 INITIAL RTg: 50'T RT AFTER 9 EFPY: 1/47, 116.4 F 3/47, 108 3 F I CURVES APPLICABLE FOR lEATUP RATES UP TO 100 F/HR FOR IHE i SERVICE PERIOD UP TO 9 IFPY AND CONTAINS MARGINS OF 107 AND 60 PSIG FOR P m m F INSTRLMENT ERRORS 2500 i.,,,.,,,., i. i IE.AK M I D(ITm '%J l I i r i r 1 2250 / / 1 i J l J I I I I I I I I 2000 f f CRITICAuTf DMit f 1 1 / J sASED ON 80'/HR _ [ [ { HEATUP CURVE f 1750 UNACCEPTABLE / / / / OPERATION r i r 1 i i1 1 I i I I I 6 II I I I I I / / f CRITICAUTY UM IT ' ((~ j g 1500 HEATUP RLTES UP / f k sAsED ON 1001/HR... HN URfE TO 1007/HR g / / ( s / / 8 w x r / r / i 1250 [ / ACCEPTABL'E - HEATUP 'TES UP '7 j / ERATION TO 60 /HR s f y x r r -~' g 1000 l / '~'~ / / / / h P I 2 E / / / BASED ON INSERVICE CRITICALITY LIMIT g 750 2 I / HYDROSTATIC TEST f 0 / / TD(PERATURE (261 F) 500 / FOR THE SERVICE PERIOD UP TO 9 EFPY 250 00 50 100 150 200 250 300 350 400 450 500 INDICATED TEWPERAT'URE (DES.F) REACTOR COOLANT SYSTEti HEATUP LD11TATIONS APPLICABLE FOR THE FIRST 9 EFPY. UPRATED TO 3565 FfWt. I Figure 3.4-2 CALLAWAY - UNIT 1 3/4 4-30 Amendment No. 36 [
1 \\ l MATERTAL PROPERTY BASIS CO ERG. LING MATERIAL: R.V. LOWER SIG 1 COPPER COMEE: 0.07 WIS NICKEL COM ENT: 0.L9WT5 INITIAL RTNDT: 50 7 RT AFTER 9 EF7Y: 1/4T,116.4% NDT 3/4T,108.3% CURVES APPLICABLE FOR C00LDOWN RATES UP TO 100%/HR FOK_THE SERVICE PERICD UP TO 9 FPY AND COEAINS MARGINS T 107 AND 60 PSIG FOR POSSIBLE INSTRLMEE EPRORS a 2500 i 1 i I I J 2250 i i I I / j 2000 r' I ~ 1750 f I I p 1500 UNACCEPIABLE [ OPERATION [ l 1250 ACCEPIABLE e / OPERATION f1000 / s 750 j RATES
- /
- -/, O , <, -w r ~' go m ,,ns/ 500 V C' 40 - 60 - / C 100 - 250 a 0 SO 100 150 200 250 300 350 400 450 S00 0 IN0lCATED TEMPERATURE (DES.F) REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST 9 EFPY. UPRATED TO 3565 ffdt. FIGURE 3.4-3 CALLAWAY - UNIT 1 3/4 4-31 Amendment No. 36 I -.wemwy-e--e.-.---,--,-wr,w,-ew~~m,-,---ww..-.. ---w-. = - = - - - -. ..--,,e.---
g TABLE 4.4-5 r-E REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM ~- WITHDRAWAL SCHEDULE i - R [ CAPSULE VESSEL LEAD z NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY).' -e U 58.5* 4.06 1st Refueling' i Y 241 3.70 5. ~ V 61* 3.70 9 s X 238.5 4.06 15 + b ~ W 121.5* 4.06 Standby m Q Z 301.5* 4.06 Standby i i N E-k 5 Ar
REACTOR COOLANT SYSTEM SpRVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORY a. actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; &nd Verifying the PORV isolation valve is open at least once per 72 hours c. when the PORV is being used for overpressure protection. 4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows: a. For RHR suction relief valve 87088: 1) By verifying at least once per 31 days that RHR RCS Suction Isolation Valve (RRSIV) 8701B is open with power to the valve operator removeo, and 2) By verifying at least once per 12 hours that RRSIV 87028 is open, b. For RHR suction relief valve 8708A: I 1) By verifying at least once per 31 days that RRSIV 8702A is open with power to the valve operator removed, and 2) By verifying at least once per 12 hours that RRSIV 8701A i is open, c. Testing pursuant to Specification 4.0.5. 4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours
- when the vent (s) is being used for overpressure protection.
l l
- Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
CALLAWAY - UNIT 1 3/4 4-35
2$00-1 i .I a j v-i FIGURE 3.4-4 1 l a i ? MAXIMUM ALLOWED PORY SETPOINT l " FOR THE COLD OVERPRESSURE MITIGATION SYSTEM,- T I d. .~j -1 i .+ t L i l .. q. __l._ ..i_ .. q i -q. t t 4 1 4 _g ._ u _.L_ _L_..__ k_ +l _ .y_ t j: .l } s l. ( 2000- {. -l.
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_. ~.. _ l _..,.! i i. .i . -j I i T P RTD SMX i i I (Fo) (PSIG) t-l a ..x ._.L..-.~ 1 i i- , ~ 4 70 495 120 495 __ 180 505 I i l 220 560 i ._. 260 715 I~ ~ i --~~ ~~T - ] 1500 320 1215 " y- ---j--j---{ j _g. -. 350 1670 I o p _1 ..l _ A' I ~ .i_ i i x ___ _[ a r _ _j __...L_ _. _.. __....] g ._.J_.- I .._.m.. w .__ i_.~. m 1 i i 1 _._4 ._i _._. j , _ _,i-_ CKa l i ^ _, 1000: I .,._1.-.._&-_. _ _... _... ~.. _. _. 4......a..... _...,i_ q l l _ _.__da y _ _ _j i i I I j i I i I I 1 i r-- .j...j .J - - - ~ .=.,.. _. i .y _ e. 500-i. i .i: 1. i l-j j. t i ~ i i t -1.. ..h. l l' i -j l I i
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i i - -j .e I y...- y -e.. _= u-I i i i i I i i i ..l-4 1 f i' I i i + m,_- q j I 6 1 50 100 200 300 400 MEASURED RTO TEMPERATURE,oF [ CALLAWAY - UNIT 1 3/4 4-36 Amendment t!o, 36
REACIOR COOLANT SYSTEM BASES P3 55URE/ TEMPERATURE LIMITS (Continued) 2. These limit lines shall be calculated periodically using methods provided below. 3. The secondary side of the steam generator must not be pressucized above 200 psig if the temperature of the steam generator is below 70'F. 4. The pressurizer heatup and cooldown rates shall not exceed 100 F/h and 200*F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than S83'F. 5. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI. The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1972 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code. Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end f 9 effec-tive full p n er years (EFPY) of service life. The 9 EFPY service life period is chosen such that the limiting RT at the 1/4T location in the core region NDT is greater than the RT of the limiting unirradiated material. The selection NDT of such a limiting RT assures that all components in the Reactor Coolant NDT System will be operated conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to-determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast. neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, NDT. based upon the fluence and copper content and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of aRI computed by either Regulatory Guide 1.99, Revision 2, "Ef fects of l NDT Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjust-ments for this shift in RT at the end of 9 EFPY as well as adjustments for l NDT possible errors in the pressure and temperature sensing instruments. Values of ART de e n n 5 manner may be used W 1 W M ts NDT from the material surveillance program, evaluated according to ASTM E185, are .iv a l l ab l e. Capsules will be removed in accordance with the requirements of ASIM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen withdrawal CAtLAWAY - llNIT 1 B 3/4 4-7 Amendment No. 36 )
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) schedule is shown in Table 4.4-5. The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the rear cor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated ART NDT NDT for the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A. The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickne s, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall is well as at the outside of the vessel wall. The dimensions of this postulate.1 crack, referred to in Appendix G of ASME Section III as the reference flw, amply exceed the current capabilities of inservice inspection techniques. Ti,eefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection agains' nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNOT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, l K g, for the combined thermal and pressure stresses at any time during heatup j or cooldown cannot be grea.er than the reference stress intensity factor, K IR' for the metal temperature at that time. K is obtained from the reference I gg fracture toughness curve, defined in Appendix G to the ASME Code. The K curve is given by the equation: IR K = 26.78 + 1.223 exp [0.0145(T-RTNOT + 160)) (1) IR Where: K is the reference stress intensity factor as a function cf the metal IR temperature T and the metal nil-ductility reference temperature RT
- Thus, NOT.
I C Al l AWAY - UN I T 1 8 3/4 4-8 _}}