ML20034C611

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Amend 131 to License DPR-20,revising Tech Specs Re pressure-temp Limits & low-temp Overpressure Protection
ML20034C611
Person / Time
Site: Palisades Entergy icon.png
Issue date: 04/26/1990
From: Diianni D
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20034C612 List:
References
NUDOCS 9005040209
Download: ML20034C611 (22)
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Text

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g e-b 8 W ASHlfdo TON. D. C. 20555 0 CONSUMERS' POWER' COMPANY-PALISMES PLANT: DOCKET NO. 50-255 A AMENDMENT-TO PROVISIONAL OPERATING LICENSE' ., ; 1, f, q\\ i d Amendment No.131. License No.- DPR-20 g g 1, The Nuclear Regulatory Commission (the Commission) has:found that:- 3' .s A. The application for amendment by Consumers. Power Company (the p$ licensee) dated September 12,.1989, as supplemented by letters-i dated September 22 and 25,11989, and March:2,:1990,-complies with

)

Y the standards and re as amended (the Act)quirements of.the Atomic Energy Act of 1954, 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-t Commission; ~,/ 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 tI conducted in compliance with the Commission'.s regulations;: >/ D. The issuance of this amendment'will'not be inimical to the common J, defense and security or to the health and. safety:of the public; and d-E. The issuance of-this amendment is in accordance with-10 CF.R Part 5]i k of the Commission's regulations and 'all applicable requirements h ch

A been satisfied.

i 2. Accordingly, the license is amended by. changes.to the Technical'. ] Specifications as indicated in the attachment.to this license amendment and Paragraph 3.B. of Provisional Operating License No. DPR-20'is hereby amended to read as follows: Technical Specifications, / k The Technical Specifications contained in Appendices A and B, as. revised through Amendment'No.131 ,areherebyincorporatedint$ '3 license. The licensee shall operate the facility in accordance d th the Technical Specifications. ( ,, i l P 4\\!

O.- n'r-r 2.. o. ,li i '\\. l. 3. This license amendmtmt is effective. as.of ther date of. ts issuance i l .andshallbeimplementednotlaterthan,1une.1/.f90.X a g 4 ) l f) FOR..THE NUCLEAR' REGULATORY COMMISSl/!)N - f(, y ' f, a' th x?,F ' ),a mm r I ' romish). C. Dilanni, Actij1 Dithlt i 4 DivisionofReaCorProje(ifts - III,'(" a Project Directoiate III.) 's li . s h ' ~4 IV, Y & Specill Projects l Office of Nuclear) Reactor Regulation-- t Y 1/1. L Attad;.nent: Changes to the Technical Specifications gDate of Issuance: April 26, 1990 l u c w . I, L - 4 d' y' ' 'l 0,4 ,,,yy' ,) U I k' j 'N 'r'# 4 \\ t ? i r S-s 'f i', 1,'s c ff,; 4 \\ f' + t_ )) . y-4' . i e

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5 ATTACHMENT TO LICENSE _ AMEt!Di'gT 110. m_ tr.0V1510NAL OPEFJTIJ!O LICEllSE h&. DPR-20 DOCKET NO. 50-255: l i i Revise Apperidix A Technical S pecifications by removing the pages identified below and it.sertirig the attacled pages. The revised pages are ider.tified by the capticr.cd ari,endt..ent nuriber and contain tr.arginel lir,es indicating the area ) of change. REMOVE INSERT /,..t t 3-Id 3-idf F y4 32 3-2 ;( t j-it g 3-3 3-3 l N s 3,4 3,4 t 3-5 3-5 I 3-6 <3-6 3-7 3-7 3-8 3-8 3-9 3-9 3-10 3-10 3-11 3-11 3 25a 3-25a i 3-25b 3-25b 3 s 3-25c 3-25c 3-30 3.30 3-#J 3-33 4-2 4-2 l 4-39 4-39 3 '{ d 4-41 4-41 j f h 4 f k +, . hi, j , o

A vy, x, l l ( e,, \\ 3.1 FRIMARY C00LAN1 SYSTEM (Cont'd) [ 3.1.1 Operable Components _ (Cent 'd) b h. Terced circulation (starting the first primary coolant pump) / shall not be initiated unless one of the following conditions is / mett / (1) Primary coolant cold leg temperature is > 430'T. / (2) PCS cold leg temperature is s 430*T and S/G secondary / temperature is less than PCS cold leg temperature. / (3) _ Shutdown cooling is isolated from the PCS AND PCS cold leg / temperature is > 210'T AND S/G secondary temperture is less / ,than 100'T higher than PCS temperature.- / 14)fShutdowncoolingisisolatedfromthePCSANDPCScoldleg / F teeperature is t 170'T and 5 210'T AND S/G secondary / u temperature is less than 20'T higher than PCS cold leg / l 4 temperature. / i (!) Shutdown cooling is isolated from the PCS AND PCS cold leg / temperature is k 120'T and < 170'T AND S/G secondary / temperature is less than 100'r higher than PCS cold leg / l temperature. / i 1 ( i." The PCS thell not be heated or maintained above 325'T unless a minimum of 375 kV of pressuriter heater capacity is available from both buses ID and IE. Should heater capacity from either t bus ID and it fall below 375 kW. either restore the inoperable heaters to provide at leart 375 kW of heater capacity from both '~ buses ID and IE within 7; bours or be in hot shutdown within , the next 12 hours. I r .B.u. f.a. T Vhen primary coolant boron concentration is being changed, the process must he uniform throughout the primary coolant system

e volume to prevent stratification of primary coolant at lower boron concentration which could reFult in a reactivity insertion L

Sufficient mining of the primary coolant is assured if one shutdown cooling or one prdnary coolant pump is in operation.(I} The l ahinedown ecoling pump will circulate the primary system volume in / less than 60 minutes when operated at rated capacity. By imposing a minimuu shutdown cooling pump flow rate of 2810 gpm. sufficient time is provided for the operato i l asymmetricflevconditions.y33oterminatetheborondilutionunder The pressurfter volume is relatively innettve, therefore vill tend to have a boren concentration hither than rest of the primary coolant system -during a dilution operation. Administrative procedures vi11 provide for use of pressuriter spravs to cahrain a nominal spread between the boren concentration in the pressuriter and the primarv system iluring the addition of boron.(2) 3-16 Amendttere No (7, Fy, y17, Tyr,131 OcnMO-0262-KL0t. o- .e wh=~ w~ e-

l 301 PRIMARY COOLANT SYSTDi (Contd) r (, Basis (Contd) i t. The FSAR safety analysia was performed assuming four primary coolant i l pumps were operating for accidents that occur during reactor i operation. There. fore. reactor startup above hot shutdown is not t permitted unless all four primary coolant pumps are operating. Operation with'three primary coolant pumps is permitted for a limited time to allow the restart of a stopped pump or for reactor internals vibration monitoring and testing. Requiring the plant to be in hot shutdown with the reactor tripped from the C-06 panel, opening the 42-01 and 42-02 circuit breakers. i assures an inadvertent rod bank withdrawal will not be initiated by the control room operator. Both steam generators are required to be operable whenever the temperature of the primary coolant is greater.than the design temperature of the shutdown cooling system to assure a redundant hsat removal system for the reactor.' l 2 Calculations have been performed to demonstrate that a pressure i differential of 1380 psi (3) can be withstood by a tabe uniformily thinned to 36% of its original nominal wall thickness (64% degradation), while maintainingt (1) A factor of safety of three between the actual pressure differential and the pressure differential required to i l cause bursting. (2) Stresses within the yield stress for Inconel 600 at ( operating temperature. (3) Acceptable stresses during accident conditions. Secondary side hydrostatic and leak testing requirements are consistent with ASME BPV Section XI (1971). The differential maintains stresses in the steam generator tube walls within code allowable stresses. The minimum temperature of 100'F for pressurizing the steam generator secondary side is set by-the NDTT of the aanway cover / of + 40'F. 1 The transient analyses were performed assuming a vessel flow at hot zero power (532*F) of 124.3 x 106 lb/hr minus 6% to account for flow measurement uncertainty and core flow bypass. A DNB analysis was performed in a parametric fashion to determine the core inlet temperature as a function of pressure and flow for which the minimum DNBR is equal to-1.17. This analysis includes the j following uncertainties and allowances: 2% of rated power for power 3-2 Amendment No 29. 5!. !!8,131 TSP 0889-0101-MD01-NI.04 w-r-c ,-ce - -,,~- -. -. ~., + ,r .,u..

3.1 PRIMARY COOLANT SYSTEN (Cont'd) jaL[,g(Cont'd) j measurementi 20.06 for ASI measurementi 150 pel for pressuriser 1 pressural 17'r for intet temperaturel and 31 measurement and 3% bypass for core flow. In addition, transient biases were included in l the derivatio f the following equation for limiting reactor inlet temperature I Tinlet 1543.3 +.0575(P-2060) + 0.00005(P-2060)**2 + 1.173(W-120) - ~ .0102(W-120)**2 The limits of validity of this equation are: 1800 ( Pressure <.2200 Psia ~ 100.0 m 10e i Vessel Flow $ 130 a los Lb/h l A81 as shown in FAgure 3.0 j i With measured primary coolant system flow rates > 130 M lbe/hr, l limities the maximum allowed inlet temperature to the TInlet LOO at } 130 M lbm/hr increases the margin to DNS for higher PCS flow rates. l The Aalel Shape Indes alarm channel is being used to monitor the l 1 ASI to ensure that the assumed asial power profiles used in the j [ development of the inlet temperature LCO bound measured asial power profiles. The signal representing core power (Q) is the { auctioneered higher of the neutron flus power and the Delta-T power. 1he measured ASI calculated f rom the escore detector signals and 4 adjusted for shape annealing (Y ) and the core power constitute an 3 t l! ordered pair (Q,Y ). An alare signal is activated before the I ordered pair exceed the boundaries specified in Figure 3.0. L The requirement that the steam generator temperature be j the { PCS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the secondary system to the PCS will not occur. This requirement applies only to the initiation of forced circulation (the start of the first primary coolant pump) when the PCS cold 'les temperature is < 430'F. However, analysis (Reference 6) shows / that under limited conditions when the Shutdown Cooling System / is isolated from the PCS, forced circulation may be,initisted / when the steam generator temperature is higher than the PCS cold / les temperature. / f References s [ (1) Updated 'FSAR, Section 14.3.2. (2) Updated FSAR, Section 4.3.7. i (3) Palisades 1983/1984 Steam Cenerator Evaluation and Repair j. Program Report, Section 4, April 19,1984 j. (4) ANF-87-150(NP), Volume 2. Section 15.0.7.1 (5) ANF-88-108 / t (6) Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 // 3-3 Amendment No 31, II, !!?, !!8,131 ~ TSP 0889-0101-MD01-NLO4

j l 3.1 PRIMARY COOLAWT SYSTEM (Continued) 3.1.2 Westuo and Cooldown Rates The primary coolant pressure and the system heatup and cooldown rates shall be !!aited in accordance with Figure 3-1, Figure 3-2 and as follows. { i Allowable combinations of pressure and temperature for any heatup a. or cooldown rate shall be below and to the right of the applicable / limit line as shown on Figures 3-1 and 3-2. The average heatup / f or cooldown rate in any one hour time period shall not escoed / the heatup or cooldown rate limit when one or more PCs cold leg / is less than the corresponding " Cold Les Temperature" below.

  • Cold Lea Temperature Westvo/Cooldown Rate Limit

/ 1 1 170*F 20'F/Wr // l 2. > 170'F and i 250'r 40'F/We // e 3. > 250'F and < 350'r 60*F/Wr / 4. > 350'F 100'F/Wr / Whenever the shutdown cooling isolation valves (M0V3015 and '/ ) M0V3016) are open, the primary coolant system shall not be heated / at a rate of more than 40'F/Wr. when the " Cold Leg Temperature" / k is >170'F. // l b. Allowable combinations of pressure and temperature for inservice, / l testing-during heatup are as shown in Figure 3-3. The maximum heatup and cooldown rates required by Section a. above, are / applicable. Interpolation between limit lines for other than the noted temperature change rates is permitted in 3.1.2a. / The average heatup or cooldown rates for the pressuriser shall / c. not exceed 200*F/hr in any one hour time period. Whenever the / Shutdown Cooling isolation valves (Mov3015 and MOV3016) are OPEW, / the pressuriser shall not be heated at a rate of more than /- 60'F/Hr.

  • Use shutdown cooling return temperature if the shutdown cooling system is in operation and all PCP's are off.

3-4 Amendment No. 21, d!, 55, 97, !!1, 231 h TSP 0889-0101-MD01-NLO4 i

1 3.1.2 Heatuo and Cooldown Rates (Continued) j d. Before the radiation esposure of the reactor vessel exceeds the / esposure for which the figures apply, Figures 3-1, 3-2 and 3-3 shall be updated in accordance with the following criteria and procedure

  • r 1.

Us Wuclear Regulatory Commission Regulatory Guide 1.99 Revision 2 has been used to predict the. increase in / transition temperature based on integrated fast neutron flus and surveillance test data. If measurements on the irradiated specimens show increase above this curve, a new j curve shall be constructed such that it is above and to i the left of.all applicable data points. 2. Before the and of the integrated power period for which Figures 3-1, 3-2 and 3-3 apply, the limit lines on the i figures shall be updated for a new integrated power period. The total integrated reactor thernst power from start-up to the and of the new power period shall be converted to i an equivalent integrated fast neutron esposure (E k 1 MeV). 1 Such a conversion shall be made consistent with the dosimetry evaluation of capsule W-290(12). I 3. The limit lines in Figures 3-1, 3-2 and 3-3 are based on the requirements of Reference 9, paragraphs IV.A.2 and IV.A.3. These lines reflect a preservice hydrostatic test pressure of 2400 psig and a vesssl flange material reference temperature of.60'F(8). k Basis + All components in the primary coolant system are designed to withstand the ef fects of cyclic loads due to primary system temperature and pressure changes.(l) These cyclic loads are I introduced by normal unit load transients, reactor trips and start-up and shutdown operation. During unit start-up and shutdown, the rates of temp 6rature and pressure changes are l limited. A maximum plant heatup and cooldown limit of 100'F per hour is consistent with the design numiser of cycles and i satisfies stress limits for cyclic operation.(2). The reactor vessel plate and material opposite the core has been purchased to a specified Charpy V-Wotch test result of 30 f t-lb or greater at an WDTT of + 10'F or less. The vessel weld has the highest RTWDT of Plate, weld and HAZ materials at the fluence to which the Figures 3-1, 3-2 and 3-3 apply.(10) The unirradiated RTWDT has been determined to be -56*F.Ill) An RTNDT of ~56'F is used as an-unirradiated value to which irradiation effects are added. In addition, 3-5 Amendment No. 21, E, 55, 89, 97, !!7,131 tsp 0889-0101-MD01-NLO4 b

-- - ~ _ _ _.. - - _ - _.-. - i. l s l 3.1.2 Weatup and Cooldown Rates (Continued) i l the plate has been 1001 volumetrically inspected by ultrasonic test using both longitudinal and shear wave methods. The remaining l material in the reactor vessel, and other primary coolant sys, tem components, meets the appropriate design code requirements and specific component function and has a maximum WDTI of +40*F.(5) As a result of fast neutron irradiation in this region of the core. there will be an increase in the RT with operation. The' integrated / fast neutron (E > 1 MeV) flumes of the reactor vessel are / calculated using Reference 13, utilitsing D07 III Code with the / gAILOR set of-cross-sections. / gince the neutron spectra and the flus measured at the samples and reactor vessel inside radius should.be nearly identical,' the i measured transition shift from a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flua magnitude.. The maalmum esposure of the reactor vessel will be obtained from the measured sample esposure by application of the calculated asinuthat neutron flua variation. The predicted RTWDT shift for the base metal-has been predicted based upon surveillance 6ata and i the Us WRC Regulatory Guide.(10) To compensate for any incrJase in the RT caused by irradiation, limits on the pressure

  • temperature relationship are periodically changed to stay within the stress

,~ limits during heatup and cooldown. Reference 7 provides a procedure for obtaining the allowable loadings for ferritic pressure-retaining materials in Class 1 components. This procedure is based on the principles of linear elastic fracture mechanics and involves a stress intensity factor prediction which is a lower bound of atnic, dynamic and crac{I) arrest critical values. The stress intensity factor computed is a function of RTNDT, operating temperature, and vessel well temperature gradients. Pressure-temperature limit calculational procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Reference 7. The limit lines of Figures 3-1 through 3-3 consider a 54 psi pressure allowance to account for the fact that pressure is measured in the pressuriser rather than at the vessel beltline and to account for PCP discharge pressure. In addition, / for calculational purposes, 5'F was taken as measurement error / allowance for calculation of criticality temperature. gy / i Reference 7, reactor vessel wall locations at 1/4 and 3/4 thickness are limiting. It is at these locations that the crack propagation associated with the hypothetical flaw must be arrested. At these locations, fluence attenuation and thermal gradients have been 3-6 Amendment No. 27, d!, 55, 89, 97,131 !!7, i -TSP 0889-0101-MD01-WL04 t . - - ~ - - - - - - - -

I 3.1.2 Hestuo and Cooldown Rates (Continued). l 13,31g(Cont'd) evaluated. During cooldown, the 1/4 thickness location is always j more limiting in that the RTWDT is higher than that at the 3/4 thickness location and thermal gradient stresses are tenslie there. During heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate. l t Figures 3-1 through 3-3 define stress limitations only from a fracture mechanica point of view.' Other considerations may be more restrictive with respect to pressure-temperature limits. - For normal operation, other inherent j plant characteristics may limit the heatup and cooldown rates which can be achieved. Pump parameters and pressuriser heating capacity tends to restrict both normal heatup and cooldown rates + to less than 60'F per hour.- j The revised pressure-temperature limits are applicable to reactor vessel inner wall fluences of up to 1.8 a 1088nvt.. S e application-of appropriate fluence attenuation factors (Reference '10) at the 1/4 and 3/4 thickness locations results in RT DT shifts of 241'F N i and 177'F, respectively, for the limiting weld material. he / criticality condition which defines a temperature below which

  • the core cannot be made critical (strictly based upon fracture mechanics' considerations) is 371'F. The most limiting wall location is at 1/4 thickness. The minimum criticality temperature, 371'F is the minimum permissible temperature for the inservice system hydrostatic pr::ssure test., That temperature is calculated based upon 2310 psig inservice hydrostatic test pressure.

P he restriction of averste heatup and cooldown rates to 100'F/h when all PCS cold legs are ?,350'F and the maintenance of a / pressure-temperature relationship under the heatup, cooldown and inservice test curves of Figures 3-1, 3-2 and 3-3, respectively, i ensures that the requirements of References 7, 8 and 9 are met. / Calculation.of average hourly cooldown rate after cooling to a / temperature range requiring a lower cooldown rate shall be only / from the time the lower cooldown rate is required. The core / i operational limit applies only when the reactor is critical. 3-7 r Amendment No. 21, d!, gy, 55, 39,'131 9/, !!1, i TSP 0889-0101-MD01-NLO4 i ? e ,--.--,--a, -n,- n,-n , w ..w n ,n, e,.. -., , nn

3.1.2 Heatus and Cooldown Rates (Continued) Ip_ gig (Continued) The heatup and cooldown rate restrictions are consistent with the / analyses performed for icw temperature. overpressure protection (LTOP) (References 13, 14 and 15). Below 430'F, the Power operated Relief / Valves (PORVs) provide overpressure protections at 430'F or.above, / the PCS safety valves provide overpressure protection. / The criticality temperature la det' ermined per Reference 4 and.the core operational curves adhere to the requirements of Reference 9.- The inservice test curves incorporate altosances for the thermal gradients associated with the heatup curve used to attain inservice test pressure. These curves differ from heatu respect.to margin for primary membrane stress.gurves only with Due to the shifts in RTwot, NDTI requirements associated with nonreactor vessel materials are, for all practical purposes,~no longer limiting. References (1) FSAR, Section 4.2.2. (2) ASME Boiler and Pressure Vessel Code, Section III, A-2000. (3) Battelle Columbus Laboratories Report, " Palisades Pressure Vessel Irradiation Capsule Prograat Unitradiated Mechanical Properties," August 25, 1977. (4) Battelle Columbus Laboratories Report, " Palisades Nuclear Plant Reactor Vessel Surveillance Progrant Capsule A-240," March 13, 1979, subeltted to the NBC by Consumers Power Company letter dated July 2, 1979. (5) FSAR, Section 4.2.4. (6) (Dele ted) / (7) ASME Boiler and Pressure Vessel Code, Section III, Appendix C, " Protection Against Non-Ductile Fa!!ure," 1974 Edition. -(8) US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section 5.3.2, " Pressure-Temperature Limits." (9) 10 CFR Part 50, Appendix C., '.' Fracture Toughness Requirements," May 31, 1983 as amended November 6,1986. / (10) US Nuclear Regulatory Commission, Regulatory Cuide 1.99, / Revision 2, May, 1988. / (11) Combustion Engineering Report CEN-189, December,1981. (12) " Analysis of Capsules T-330 and W-290 f rom the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program," WCAP-10637, September, 1984. (13) " Analysis of Fast Neutron Exposure of the Palisades Reactor / Pressure Vessel" by Westinghouse Electric Corporation, March 1989. / (14) Consumers Power Company Engineering Analysis EA-FC-809-13, Rev 1 // " Pressure Response Ef fect of VLTOP with Replacement PORVs." / (15) Consumers Power Company Engineering Analysis EA-A-PAL-89-98 / " Palisades Pressure and Temperature Limits." / / / 3-8 Amendment No. 27, d!, 55, 89, 97, !!7, 131 TSP 0889-0101-MD01-WLO4

FIGUTIE 3-1 ~ -PALISADES PRESSURE AND TEMPERATURE LIMITS FOR HEATUP PESS PSIG F - i.8 x 10 n/cm (No measurement uncertainty incitaleef) 3000-- I l gg l 1 l [ U i i e t t 2750 I-j l i sa 8 t --- - l --- ; i ?- i ? t j i 60 i i [ l 8 1 2500'fI l I 40 t 1 -l 2- - l-Ji .' 2 2 1 --i [0-l I k .f i i I _ L .... _.... ;........j.._ 4 -l .s-- L.---... l -g[- 2250 0 l l .I i j l ~' I I I I i I t I t

g. }

I .....J,....

j. _ _...-k.

s e ...-g-.-~.- ~l 4 3-g I i ._. 4 e ig .i w 4 i 1150 e n --- - i --- -, --- -- ?. - - -

t. -

0 --i t c ,s .t i .i 6 i .i 4,,.' l - ----- !- - -- -- I --- - -4. --- - I --- i - - L -- - -? - - --i -- - --- -} -- -, - - - - - - - + - 1500 I 1 i 1250 s --4------ } r i t-- a' I F----------

+-----8---

i i i i ? I s 2-- -....[..- $.-..h..-- .... -.....(! ~ k .l-. 1000 ... -.]. b ,r, i .,s i 1 i i i /50.e ",------,-----,1 p o------- r----- - - - - - - + - - - - .i ---g------ i ,F-- i L - -- -- - i --- ' ---- l - - - - - - '. - - ,i i. 500 a i [. i --l-r --- [ i - i -- 'a 250-. I l n- - - - - -- * - -w x -d - j- --- ' n i O: j w ' m r vre r; t re e i e e rr, er e re e v 1 e e r rl 3 r e e, e re v, e ers ' re ri ; rre r; re t r: c-* m 8L rere-ri,,- vv.. 50 M IM iM iM iM M M 'N M M M M M M M 6 TEMPERATINEE DEGEES F

FIGURE 3-2 f PALISADES PRESSURE G TEMPERATURE LIMITS FOR COOLDOWN PRESS PSIG "^" 8 ** **""at "wa t aint y tw s u.6%s) j 3000 8 8 J - 2750 1IM I i i ] 2s00 I IFil 1 ,250 I 11 - i i l 2000 l I 17s0 l 1500 I 2" w w 40 e soo i 2250 ' o 2000 k A l 75. a r/ O - Jd y w ": W U

===h p 250 6 i s n 2* O2 = o 1".!1 50 7s s00 sas sso s7s 200 aas ase 27s 30o aas ase 37s m e m TEMPERATURE DEGEES F m i E ..y ...s-y _ y +v-. -ye w; .,,.v- , g cm ,s.., ,. + -.. m m.-. m

I ~ A SADES PRESSURE AND TE PERATURE LIMITS FOR HYDRO 2 f

  • 1.8 X 10

,,fm e fl [ S [>/ / t + s i p2 j 1 r t g) e 1 pay ~ i A rj syy7 s,9 7/ -0 ,r ^ j 3 500' s y j, 7' l 7 , / [ / l . h, 1 / y ano %~ I 5 t

==

sijyvy, x

.,- ae so 7s so. 33, ,a, aso m y-s 3,, JO', TEMPERATURE DE771EES F i w e i t J 1 .~ _, y y c 9 w __.~,,.>--...,-me.,, - <-.=.-r,, .___i,.

2. a-

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c 3.1.8 OVERPRESSURE PROTECTION SYSTEMS LIMITINC 00WDITIOWS FOR OPERATION 3.1.8.1 REQUIREMEWTS / Two power operated relief valves (PORVs) with a lif t setting / below and/or to the right of the curve ;In Figure 3-4 shall be / o pe rable. 'l APPLICA8ILITY When the temperature of one or more of the primary coolant systes cold legs is less than 430'F. / .A.EIEW8 a.- With one PORY inoperable, either restore the inoperable PORY to operable status trithin 7 days or depressuriae within the next 8 hours and either vent the PCS through a 11.3 square inch vent or open both PORY valves and both PORV block valves. / b. With both PORVs inoperable, depressurise within the nest. 8 hours and either vent the PCS through a 1 1.3 square inch vent or open both PORY valves and both PORY block valves. / c. The provisions of. Specifications 3.0.3 and 3.0.4 are not applicable. .M There are three pressure transients which could cause the PCS presnure to exceed the pressure limits required by 10CFR50 Apper dix C. They aret (1) a charging / letdown labalance, (2) the start of high pressure safety injection (HPSI), and (3) initiation of forced circulation in the PCs when the steam generator temperature is higher than the PCS temperature. Analysis (R'Aference 3) shows that when three charging pumps are / operatins and letdown is isolated and a spurious HPSI occurs / i between 260*F and 430'F, the PORY rrepoints ensure that 10CFR50- // 1 Appendia C pressure limits will no; ce exceeded. Below 260*F, // overpressure protection is still provided by the PORVs but HPSI / operability. is precluded by the limitations of Technical '/ Specification 3.3.2 g. Above 430'F, the pressuriser safety / valves prevent 10CFR50 Appendix C timits f rom being exceeded. / 3-25a Amendment No. 52, 72, !!7, 131 i TSP 0889-0101-MD01-WLO4 l l s

~_-- i .i 3.1.4 OVERPRESSURE PROTECTION SYSTEMS l LIMIT!WC C0WDfTroWS FOR OPERAT!0W 3.1.4. 13, gig (continued) l Assurance that the Appendia C limits for the reactor pressure / l l vessel will not be violated while operating at low temperature / is provided by the variable setpoint of. the Low Temperature / i l Overpressure Protection (LTOP) systen. The LTOP system is / programmed and calibrated to ensure opening of the pressuriser / i l power operated relief valve (PORV) when the combination of primary / coolant system (PCS) pressure and tenparature is above or to the / j lef t of the limit displayed in Figure 3-4. That limit is developed / t from the more limiting of the heating or cooling limits for the / specific temperature of the PCs while heating or cooling at the / l maximum permissible rate for that temperature. The limit in / i Figure 3-4 includes an allowance for pressure overshoot during the / interval between the time pressuriser pressure reaches the limit, / and the time a PORV opens enough to terminate the pressure rise. / LTOP is provided by two independent channels of measurement, / s control, actuation, and valves, either one of which is capable of / providing full protection. The octual setpoint of PORY actuation / for LTOP will be lowered from the 11 silt of Figure 3-4 to allow / fer potential instrument inaccuracies, measurement error, and / instrument drift. This will ensure that at no time betwan / calibration intervals will the combination of'PCS temperature / and pressure escoed the limits of Figure 3-4 without PORY / actuation. / / When d.: ehutdown cooling system u not isolated (M0-3015 and / MO-3016 open) from the PCS, assuranc. that the shutdown cooling / system will not be pressurised above its Jesign pressure 16 / aNM h the relief valves on the shutdown cooling systes, / and the limitations of sections 3.1.1.h., 3.1.2.a & c, and / 3.3.2 3 / / The requirement for the PCS.to,be depressurised and vented by an opening ?,1.3 square inches (Reference 4) or by opening both / PORY valves and both PORY block valves when one or both PORVs are inoperable ensures that the 10CFR50 Appendia C pressure limits will not be exceeded when one of the PORVs is assumed to fait per the " single failure" criteria 10CFR50 Appendix A, Criterion 34. Since the PORV solenoid is strong enough to overcome spring pressure and valve disc weight, the PORVs may be held open by. / keeping the control switch in the open position. /. References 1. Technical Specification 3.3.2 / 2. Technical Specification 3.1.2. / 3. Consumers Power Company Engineering Analysis EA-FC-809-13, Rev 1 // 4. " Palisades Plant Overpressurization Analysis" June 1987 and / t " Palisades Plant Primary Coolant System Overpressurization / Su'usystem Description" October 1977. / / 3-25b Amendment No. !!1,131 TSP 0889-0101-MD01-WL04 t

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,.. ~. - -... -., - i 3.3 EMERCENCY CORE COOL}WC SYSTEM (Continued) 3 WPS! pump operability shall be as follows: i

1) If the reactor head is installed, both HPSI pumps shall

/ i be rendered inoperable whent / a. The PCS temperature is ( 260'F, or // b. Shutdown cooling isolation valves Mo-3015 and MO-3016 / are open. /

2) Two NPSI pumps shall be operable when the PCS temperature

/ l is > 325'F. / I

3) One NPS! pump may be made. inoperable when the reactor is

/ i subcritial provided the requirements of Section 3.3.2.c, / are met. /

4) HPSI pump testing may be performed when the PCS temperature

/ is (430'F provided the HPS! pump manual discharge valve is / j closed. / 3.3.3 Prior to returning to the Power Operation Condition af ter every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours and testing of Specification 4.3.h has not been accomplished in the previous 9 months, or prior to returning,the check valves in Table 4.3.1 to service af ter maintenance, repair or replacement, the following l conditions shall be mets a. All pressure isolation valves 18-ted in Table 4.3.1 shall be functional as a pressure isolatiot device, except as specified + in b. Valve leakage shall not exc:ed Lt.e amounts indicated. b. In the event that integrity of any pressure isolation valve specified in Table 4.3.1 cannot be demonstrated, at least two valves in each high pressure line having a non-functional valve must be in and remain in, the mode corresponding to the isolated condition.(1) b ~ j AMo' tor-operated valves shall be placed in the. closed position and power supplies deenergized. 4 3-30 Amendment No. 5!, 191, !!1, 131 TSP 0889-0101-MD01-NLd4 f w .. ~.,, -,. -. -m._,--,.%c-,, -_---._--_,--w ~.. - -, +. -.,

3.3 EMERCENCY CORE C00LINC SYSTEM jAgig(continued) de2onstrate that the maximum fuel clad temperatures that could occur over the break site spectrum are well below the setting temperature of airconium (3300'F). Malfunction of the 14 't' essure Safety injection Flow control valve could defeat tho ., Pressure injection feature of the 1 ECCSI therefore, it is sisabled in the 'open' mode (by is41 sting j the air supply) during plant operation. This action assures i that it will not block flow during Safety Injection. 1 I The inadvertent closing of any one of the Safety Injection bottle isolation valves in conjunction with a LOCA has not. i been analysed. To provide assurance that this will not occur, these valves are electrically locked open by a key switch in - i the control roos. In addition, prior to critical the valves are checked open, and then the 430 volt breakers are opened. [ i Thus, a failure of a breaker and a switch are required for any of the valves to close. Insuring both HPSI pumps are inoperable when the PCS temperature / is ( 260'F er the shutdown cooling isolation valves are open // i eliminates PCS mass additions dLa to inadvertent WPSI pump starts. / i Soth HPSI pumps starting in conjunction with a charging / letdown / imbalance may cause 10CFt50 Appendis C limits to be exceeded / when tte PCS temperature is t 260'F. When the PCS temperature // t is 1430'F, the pressuriser safety valves ensure that the PCS / pressure will not escoed 10CFR50 Appendis C. / i l The requirement to have both HPSI trains operable above 325'F / provides added assurance that the effects of a 14CA occuring l i under LMP conditions would be aitigated. If a LOCA occurs when-the primary system temperaturo is less than or equal to 325'F, / l the pressure would drop to the level where low pressure safety injection can prevent core damage. Therefore, when the PCS / temperature is 1260'F and $325'T operation of the HPSI systes // r would not cause the 10CFR$0 Appendix C limits to be exceeded / nor is HPSI system operation necessary for core cooling. / l l I \\ HPSI pump testing with the HPSI pump manual discharge valve i closed is permitted since the closed valve eliminates the possibility of pump testing being the cause of a mass addition to the PCS. i Re'erences (1) FSAR, Section 9.10.31 (2) FSAR, Section 6.1, 3-33 Amendment No. !!, ll, !#!, !!7,131 TSP 0889-0101-MD01-WLO4 h m~__.. ... - ~,. m._ ____,_.,__,,.____.._,_..._.,__,__.,_.._,_..,_,__.,,_,__...,__..._,_.m__. ,_..m,,,,__ ._,.mm_,_

b. The PCs vent (s) shall be verified to be open at least once per 12 hours when the. vent (s) is being used for overpressure protection escept when the vent pathway is provi.ded 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. I' c. When both open PotV volves are used as an alternative to / l venting the PCs, then verify both PORY valves and both PORY / block valves are open at least once per 7 days. IA11.f. Failures such as blown instrument fuses, defective indicators, and faulted aspitfiers which result in " upscale" or "downscale" l indication can be easily recognised by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases,. revealed by alors or annunciator action and a check supplements this type of built-in surveillance. i Based on esperience in operation of both conventional and nuclear plant systems when the plant is in operation, a checking l frequency of once per-shif t is deemed edequate for reactor and steam system instrumentation. Calibrations are performed to insure the presentation and acquisition of accurate information. l l 'the power range safety channels and AT power channels are are calibrated daily against a heat. balance standard to account for i errors induced by changing rod patterns and core physics parameters. l-Other channels are subject only to the "drif t" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process system instrumentation errors induced by drif t can be espected to remain within acceptable tolerances if recalibration is performed at each. refueling shutdown interval. substantial calibration shifts within a channel (essentially a i channel failure) will be revealed during routine checking and i testing procedures. Thus, minimum calibration frequencies of one per-day for the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate. The minimum testing frequency for those instrument channels connected to the reactor protective system is based on an estimated average unsafe failure rate of 1.14 x 10'S failure / hour per channel. This estimation is based on limited operating experience at conventional and nuclear plants. An " unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or attempt; to respond to a bonafide signal. 4-2 Amendment No. 15, 5!, !!7, !!8,131 , TSP 0889-0101-MD01-NLO4-i

e e b. The PCS vent (s) shall be varified to be open at least once per 12 hours when the vent (s) 16 being used for overpressure protection escept 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. c. When both open PORV velves are used as an alternative to / venting the PCs, then verify both PORY valves and both PotV / block valves are open at least once per 7 days. hill. Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in " upscale" or "downstale" indication can be easily recognised by slople observatten of the functioning of an -instrument or system.- Furthermore, such failures are, in many cases, revealed by alors or annunciator action and a check supplements this type of built-in surveillance. gased on esperience in operation of both conventional and nuclear plant systems when the plant is in operation, a checking frequency of once per-shif t is deemed adequate for reactor and staan systee instrumentation. Calibrations are performed to insure the presentation and acquisition of accurate information. We power range safety channels and AT power channels are are calibrated daily against a heat balance standard to account for I errors induced by changing rod patterns and core physics parameters. Other channels are subject only to the " drift" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process system instrumentation errors induced by drif t can be espetted to remain within acceptable tolerances if recalibration is performed at each refueling shutdown interval. Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures. Thus, minimum calibration f requencies of one-per-day for the power range safety channels, and once each refueling shutdown for the process system channels, are considered

adequate, h e minimum testing frequency for those instrument channels connected to the reactor protective systes is based on an estimated average unsafe failure rate of 1.14 a 10-5 failure / hour per channel. This estimation is based on limited operating experience at conventional and nuclear plants. An " unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or attempts to respond to a bonafide signal.

4-2 Amendment No. 15, 5!, !!1, !!8,131 , TSP 0889-0101-MD01-NLO4 ( J

4.6 SAFETY IWJECTf 0W AND CONTA!WHEWT SPRAY SYSTE.MS TESTS } Aonlicability i. Applies to the safety injection system, the containment spray systen, chemical injection system and the contalnment cooling i system tests. Ob3ective To verify that the subject systems will respond promptly and perfore their latended functions, if required. ) i Sneelfications 4.6.1 Safety injection Systes a. System tests shall be performed at each reactor refueling i interval. A test safety injection signal will be applied to initiate operation of the system. n e safety injection l and shutdown cooling systes pump actors may be de-energised i for this test. De system will be considered satisfactory if control board indication and visual observations indicate l that all components have received the safety injection l signal in the proper sequence and timing (ie, the appropriate pump breakers shall have opened and closed, and all valves i shall have completed their travel). b goth high pressure safety injection pumps, p-66A ggd p-663 shall be demonstrated-inoperable at least once per-12 hours whenever the temperature of one or more of the PCS cold tegs is < 260'F or if shutdown cooling valves M0-3015 and // f NO-3016 are open unless the reactor head is removed. / 4.6.2 contelneent Soray Systen Systes test shall be perfomed at each reactor refueling a. interval. The test shall be performed with the isolation valves in the spray supply lings at the containment blocked closed. Operation of the system is initiated by tripping the normal actuation instrumentation. b. At least every five years the spray nosales shall be verified to be open. t c. The test will be considered satisfactory if visual observations indicate all components have operated satisfactority. I f . i 4-39 Amendment No. 5!,13, 96, !!7l.131 i i l TSP 0889-0101-MD01-N1,04 j i m .<r w v,- m,- .~,.s 4-----,m

1 4.6 SAFETY TWJECTION AND COWTA!WEWT SptAY SYSTEMS TESTS (Continued) M (continued) f During reactor operation, the instrumentation which is depended { on to initiate safety injection and containment spray is r generally checked daily and the initiating circuits are tested i monthly. In addition, the active components'(pumps and valves) i are to be tested every three months to check the operatica of l the starting circuits and to verify that the pumps are in satisfactory running order. The test interval of three months is based on the judseent that more frequent testing would not significantly increase the reliability (ie, the probability that the component would operate when required), yet more { frequent test would result in increased wear over a long period i of time. Verification that the spray piping and nossige are open will be made initially by a smoke test or other suitably i sensitive method, and at least every five years thereaf ter. Since the material is all stainless steel, normally in a dry condition, and with no plugging mechanism available, the retest every five years is considered to be more than adequate. ? I Other systest that are also important to the emergency cooling function are the $1 tanks, the compocent cooling system, the service water systen and the containment air coolers. The $1 tanks are a passive safety feature.- In accordance with the specifications, the water volume and pressure in the $1 tanks are checked periodically. The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory performance. t With the reactor vessel head installed when the PCS cold leg / temperature is less than 260*F, or if the shutdown cooling // system isolation valves MO-3015 and NO-3016 are open, the start / of one HPSI pump could cause the Appendix C or the shutdown / cooling system pressure limits to be exceeded; therefore, both. / pumps are rendered inoperable. / References (1) FSAR, Section 6.1.3. (2) FSAR, Section 6.2.3. r I l t l C P 4-41 Amendment No. !!1, 131 i TSP 0889-0101-MD01-WL,04 .+w.- .,m_.--

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