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Proposed Tech Specs Permitting Use of Variable PORV Setpoint for Low Temp Overpressure Protection to Increase Operating Margin During Plant Heatup & Cooldown
	ML18054A979
Person / Time
Site:	Palisades
Issue date:	09/12/1989
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18054A976	List: ML18054A974; ML18054A979;
References
	NUDOCS 8909190014
	Download: ML18054A979 (54)
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I ;I *ATTACHMENT 1 Consumers Power Company Palisades Plant

Docket 50-255 PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES o=:9C>9190014 i=-t1R. ADOCK F' TSP0889-0101-MD01-NL04 September 12, 1989 19 Pages v ' 3 .1 3 .1.1 PRIMARY COOLANT SYSTEM (Cont'd) Operable Components (Cont'd) h. During initial primary coolantpump starts (initiation of f orc.ed ci rcuiat ion) at PCS cold leg temperatures 430° F, the secondary system temperature in both steam generators must be < the PCS cold leg temperature unless the Shutdown Cooling System is isolated from the PCS and one of the following conditions is met: I I I I I I 1. The -steam generator temperature shall not exceed the I PCS cold leg temperature by more than the limit below. I Cold Leg Temperature Limit I 1. ) 120°F and < 160°F 100°F I 2. ) 160°F and < 210°F -10°F I --3. ) 210°F and < 350°F 100°F I 2. Under transient conditions with only one of the steam I generator's temperatures higher than the PCS temperature, I and the. PCS temperature

> 350°F: . I I A. To start a PCP.in the. cold steam generator loop, I in the. hot steam generator loop shall be * < 100°F; or / l. B. To start a PCP in _the hot steam generator loop, I shall be < 100°F and the PCS pressure shall be I at least 100 psi lesi than the pressure limits of I Figure 3-4. I i. The PCS shall not be heated or maintained above. 325°F unless a m1n1mum of 375 kW of pressurizer heater capacity is available from both buses lD and lE. Should heater capacity from eit.her bus lD and lE fall below 375* kW, either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses lD and lE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . o.r be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Basis When primary coolant boron concentration is being changed, the process must be uniform throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion.

Sufficient mixing of the primary coolant is assured if one shutdown cooling or one primary coolant pump is i!l operation.Cl) , The shutdown cooling pump will circulate the primary system volume in less than 60 minutes when operated at rated. capacity.

By imposing a minimum shutdown cooling pump flow rate of 2810 gpm, sufficient time is provided for the operatof to terminate the boron dilution under asymmetric flow conditions.

SJ The pressurizer volume is relatively inactive, therefore will. tend to have a boron concentration higher than rest of the primary coolant system during a dilution operation.

3-ld Amendment No ,1, SS, tt1, tt8, TSP0889-0101-MD01-NL04

\:.,J " 3. 1 PRIMARY COOLANT SYSTEM (Contd) Basis (Contd) . '.

_procedures will _pro"l[:f,de f9r us_e of pressurizer sprays_ to maintain a nominal spread between the boron concentration in* . (2). the pressurizer and the primary system during the addition of_boron.

The FSAR safety analysis was performed assuming four primary coolant pumps were operating for accidents that occur during reactor operation.

Therefore, reactor startup above hot shutdown is not. permitted unless all four primary coolant pumps are 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, 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 system to assure a redundant heat removal system for the reactor. Calculations have been performed to demonstrate that a pressure dffferential of 1380 psi(3) can be withstood.by a tube uniformily thinned to 36% of its original nominal wall (64% degradation), while maintaining:

(1) A factor of safety of three between the actual pressure differential and the pressure differential required to cause bursting.

{2) Stresses within the yield stress for Inconel 600 at operating temperature

. (3) Acceptable stresses during accident 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 manway cover I of + 40°F * . ' ' . . The transient_

analyses were performed assuming a vessel .flow at hot -** ; zero power (532°F) of 124.3 x 10 6 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 * ; following uncertainties and allowances:

2% of rated power for power Ame.ndment No t", '/it, 1l8, TSP0889-0101-MD01-NL04

'3.1 PRIMARY COOLANT SYSTEM (Cont'd) Basis (Cont'd) measurement;

+/-0.06 for ASI measurement;

+/-50 psi for pressuri-zer p_ressure;

+/-7°F for inlet temperature; and 3% measurement and 3% bypass for core flow. In addition, transient biases were included in the derivatiOQ Qf the following equation for limiting reactor in.let temperature:

l 4 J Tinlet 543.3 + .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 lOQ.0-x 106 < Ve;sel Flow 130 x 106 Lb/h ASI as shown in Figure 3.0 With measured primary coolant system flow rates > 130 H lbm/hr, limiting the maximum allowed inlet temperature to the Tinlet LCO at 130 H lbm/hr increases the margin to DNB for higher PCS flow rates. The Axial Shape Index alarm channel is being use*d to monitor the AS! to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles.

The signal representing core power (Q} is the auctioneered higher of the neutron flux power and the Delta-T power. The measured AS! calculated from the excore detector signals and adjusted for shape annealing (Yr} and the core power constitute an ordered pair (Q,Yr}. AD alarm signal is activated before the ordered pair exceed the boundaries specified in Figure 3.0. The requirement that the steam generator temperature be the PCS temperature when circulation is initiated in the PCS ensures that ari 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 leg temperature is < 430°F. However' analysis (Reference 6)* shows. I *that under limited conditions when the Shutdown Cooling System I is isolated from the PCS, forced circulation may be initiated I when the steam generator temperature is higher than the PCS cold I leg temperature.

I References (1) Updated FSAR, Section 14.3.2. (2) Updated Section 4.3.7. (3) Palisades 1983/1984 Steam Generator Evaluation and Repair Program Report, Section 4, April 19, 1984 (4)

Volume 2, Section 15.0.7.1 (5) . ANF-88-108 (6) Consumers Power Company Engineering Analysis EA-A-NL-89-14 3-3 Amendment No U, U, U1, HS, TSP0889-0101-HD01-NL04 I I 3.1 PRIMARY COOLANT SYSTEM (Continued) 3.1.2, Heatup and Cooldown Rates The primary coolant the system

.!!.Pd cooldown rates shall be limited in accordance with Figure 3-1, Figure 3-2 and as follows. a. Allowable combinations of pressure and temperature for any heatup 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 or cooldown rate in any one hour time period shall not exceed the heatup or cooldoWI1 rate limit when one or more PCS cold leg is less than the corresponding "Cold Leg Temperature" below. 1 2. 3. 4. *Cold Leg Temperature

< 160°F > 160°F and < 250°F > 250°F and < 350°F > 350°F ,*_ Heatup/Cooldown Rate Limit 20°F/Hr 40°F/Hr 60°F/Hr 100°F/Hr Whenever the shutdown cooling isolation valves (MOV3015 and MOV3016) are open, the primary coolant system shall not be heated at a rate of more than 40°F/Hr. when the "Cold Leg Temperature" is >160°F. b. Allowable combinations of and temperature for inservice 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 pepnitted in 3.I.2a. c. The average heatup or cooldown rates for the pressurizer shall not exceed 200°F/hr in any one hour time period. Whenever the Shutdown Cooling isolation valves (MOV3015 and MOV3016) are OPEN, ** the pressurizer 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. l7, $$, 97, tt7, TSP0889-0101-MD01-NL04 I I I I I I I I I I I I I I I I I I I I ** .. *

** 3.1.2 Heatup and Cooldown Rates (Continued)

d. Before the radiation exposure of the reactor vessel exceeds the exposure for which the figures apply, Figures 3-1, 3-2 and 3-3 be

f,n with tbtj! fQlLowing .crit: .er:f,a and procedure:

1. US Nuclear Regulatory Commission Regulatory Guide 1.99 Revision 2 has been used to predict the increase in transition temperature based on integrated fast neutron flux and surveillance test data. If measurements on the irradiated specimens show increase above this curve, a new curve shall be constructed such that it is above and to the left of all applicable data points. 2. Before the end of the integrated power period for which Figures 3-1, 3-2 and 3-3 apply, the limit lines on the figures shall be updated for a new integrated.power period. The total integrated reactor thermal power from start-up to.the end of the.new power period shall be converted to an equivalent integrated neutron exposure (E 1 MeV). Such a conversion shall be made consistent with the dosimetry evaluation of capsule W-290(1 2). 3. The limH 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 reflec_t a preservice hydrostatic test pressure of 2400 psig and a vessel flange material reference temperature of.6 0oF(8).

Basis All components in the primary coolant system are designed to withstand the. effects of cyclic loads due to primary system temp,erature and pressure changes*. (l) .These cyclic loads are *introduced by normal :unit load transients, reactor trips and start-up and shutdown operation.

During unit start-up and shutdown, the rates of temperature and pressure changes are limited. A maximumplant heatup and cooldown limit of 100°F per hour is consistent with the design number of cycles and limits for. cyclic operation.

(2) _The reactor vesselplate and material opposite the core has.been purchased to a specified Charpy V-Notch test result of 30 ft-lb or greater at an NDTT of + 10°F. or less *. The vessel weld has -the-::*

highest RTNDT of plate, weld and HAZ materials (10) which the Figures 3-1, 3-2 and 3-3 apply.* . . 0 *(11) . . has been determined to be -56 F. An RTNDT at the fluence to The unirradiated RTNDT of -56°F is used as an unirradiated value to which irradiation effects are added. In addition, 3-5 Amendment No. t7, 1/.1, jJ, n, 1t7, TSP0889-0101-MD01-NL04 . I I 3 .1.2 Heatup and Cooldown Rates (Continued) the plate has been 100% volumetrically inspected by ultrasonic test using both longitudinal and shear wave methods. The remaining material in the reactor vessel, and other primary coolant system components, meets the appropriate design code requirements (5) and specific component function and has a maximum NDTT of +40°F. As a result of neutron irradiation in this region of the core, there will be an increase in the RT with operation.

The integrated I fast neutron (E > 1 MeV) fluxes of the reactor vessel are

I calculated using Reference 13, utilitzing DOT III Code with the I. SAILOR.set of cross-sections.

I Since the neutron spectra and. the flux measured at. the samples and reactor vessel inside radius should be nearly the 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 flux magnitude.

The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron flux variation.

The predicted RTNDT shift for the base metal has been predicted based upon surveillance data and (10) . the US NRC Regulatory Guide. To compensate for any increase in the RT caused by irradiation, limits on the pressure-temperature relationship are periodically 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 mecha,nics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and cracfl) arrest critical values. The stress intensity factor computed *is a furiction of RTNDT' operating temperature, and vessel wall 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 pressurizer 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.

By Reference

],.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. l7, $$, 97, H7, I /. I TSP0889-0101-MD01...:NLQ4

.*---._ .... *.,_ * ..

1. '*' 3 .1. 2 Heatup and Cooldown Rates (Continued)

Basis (Cont'd) evaluated.

During cooldown, the 1/4 thickness location is always more limiting in that the RTNDT is higher than that at the 3/4 thickness location and thermal gradient stresses are tensile there. During heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate. Figures 3-1 through 3-3 define stress limitations only from a fracture mechanics point of view. Other considerations may be more restrictive with respect to pressure-temperature limits. For normal operation, other inherent plant characteristics may limit the heatup and cooldown rates which can be achieved.

Pump parameters and pressurizer heating capacity tends to restrict.both normal heatup and cooldown rates to less than 60°F per hour.

The revised pressure-temperature limits are applicable to reactor vessel inner wall fluences of up.to 1,8 x l0 19 nvt. The application . of appropriate fluence attenuation factors (Reference

10) at the

1/4 and 3/4 thickness locations results in RTNDT shifts of 241°F and 177°F, respectively, for the limiting weld material.

The I 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, 37 l.°F is the minimum permissible temperature for the inservice system hydrostatic pressure test. That temperature is calculated based upon 2310 psig inservice hydrostatic test .. pressure.

The restriction of average heatup and cooldown rates to 100°F/h when all PCS cold legs are> 350°F.and the maintenance of.a I pressure-temperature relationship under the heatup, cooldown and inservice test curves of Figures 3-1, 3-2 and 3-3, respectively, ensures that the requirements of References 7, 8 .and 9 are met. I Calculation of average hourly cooldown rate after cooling to a I temperature range requiring a lower cooldown rate shall be only I from the time the lower cooldown rate is* required.

The core I operational limit applies only when the reactor is crit:f.cal.

3-7 TSP0889-0101-MD01-NL04 Amendment No. 17, $$, 117, 3.1.2 Heatup and Cooldown Rates (Continued)

'Basis (Continued)

The heatup and cooldown rate restrictions are consistent with* the I analyses performed for low temperature overpressure protection (LTOP) (References 13, 14 and 15). Below 430°F, the Power Operated Relief / Valves (PORVs) provide overpressure protection; at 430°F or above, / the PCS safety valves provide overpressure protection.

I The criticality temperature is determined per Reference 8 and the _ core operational curves adhere to the requirements of Reference

9. The inservice test curves incorporate allowances for the thermal gradients associated with the heatup curve used to attain inservice test pressure.

These curves differ from only with respect to margin for primary membrane stress. Due to the shifts in RTNDT' NDTT requirements associated with nonreactor vessel materials are, for all practical purposes,-

no longer limiting.

References (1) (2) (3) (4) FSAR, Section 4.2.2. ASME Boiler and Pressure Vessel Code,Section III, A-2000. -Battelle Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties," August 25, 1977. Battelle Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Vessel Surveillance Program: Capsule A-240," March 13, submitted to the NRC by Consumers Power Company letter dated July 2, 1979. _ (5) FSAR, Section 4.2.4. (6) (Deleted)

(7) ASME Boiler and Pressure Vessel Code,Section III, Appendix -"Protection Agaipst Non-Ductile Failure," 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 G, "Fracture Toughness Requirements," May 31, 1983 as amended November 6, 1986. (10) US Nuclear Regulatory_Comm.ission, Regulatory Guide 1.99, Revision 2, May, 1988. (11) Combustion Engineering Report CEN-189, December, 1981. (12) -"Analysis of Capsules T-330 and W-290 from the Consumers Power (13) (14) (15) -Company Palisades Reactor Vessel Radiation Surveillance Program," WCAP-10637, September, 1984. "Analysis of Fast Neutron Exposure of the Palisa_des Reactor Pressure Vessel" by Westinghouse Electric Corporation, Ma_rch 1989. Consumers Power Corqpany E11gin_eering

__ Anaiysis EA-FC-809-13 "Pressure Response Effect of VLTOP wi.th Replacement PORVs." Consumers Power Company Engineering Analysis EA-A-PAL-89-98 "Palisades Pressure and Temperature Limits." 3-8 Amendment No. n' f.t' u' SSf' Sf7, 117, TSP0889-0101-MD01-NL04 I I I I I /. I I I I I /

FIGURE 3-1 PALISADES PRESSURE AND TEMPERATURE LIMITS FOR HEATUP .' ', ! j I l ; .* 1 . I i : i I ' ' -* ... .< -,--*--+ ! "N. 250 1-4."' '$;:;; l4. 0 "' '<::A. '<::A. 50 "' 75 100 125 -150 175 200 225 250 275 .*.*. SOO S25 S50 375 400 425 450 TEMPERATURE DEGREES F FIGURE.3-2

PALISADES PRESSURE & TEMPERATURE LIMITS FOR COOLDOWN Cl> ::s Cl> ::s c+ !;;::: 0 PRESS PSIG 3000 w I 2750 2500 2250 2000 1750 1500 ::; 1250 1000 750 500 250 '"-l '"-l . 0 'M "$:;; 1'-l N '"-lv y . y '\9;l 'l:S. lY 2 F' = l.f1 X lo n/cm (No Measurement Uncertainty

Included)

!/) i '0 . I I . . : n 60 40 20. 80 -*-100 80 _t--t:::::a c. \ V"" --::;:::::::

It --20 0 . . . . . . . . .. . . ' .. ' ' ' ' .... .. . ' .. . ' . . . . . . . .... . . . . . . I I I * . .. 50 75 100 125. 150 175 200 225 250 275. 300 325 350 375 400 425 -450 TEMPERATURE DEGREES F

. §! <D :::1 <D :::1 . c+ 0 ':&.No "1"4."$:;;: ... ... '1::1:1. ... "$l . '0:$. ... FIGURE PALISADES PRESSURE AND TEMPERATURE LIMITS FOR HYDRO PRESS PSIG 3000 2750 2500 2250 2000 1750 1500 w* I 1250 ..... ..... 1000 750 50Q 250 0 '\;' 19 2 f = 1.8 X 10 n/cm ' //; l/1' . !/!! lj . 0 '* 20 /J '!//;' a w ' 40 v )) 6( W. ' 30 @ *. 4 / 100 7 v ? ... . v / . ::.....---

__,/ i.-. .JI" . ---... -V" . -. ----,,..--I If .... I I I I I I I I I I I I I I I I I I I I I I I I .... * * ' 1 .. ' I I.' 1.1 I I . . . . . . *I I 0 1 50 75 100 125 150 115. 200 225 250 275 300 325 350 375 400 425 450 TEMPERATURE DEGREES F ...

3.1.8 OVERPRESSURE PROTECTION SYSTEMS CONDITIONS FOR OPERATION REQUIREMENTS Two power operated relief valves (PORVs) with a lift setting below and/or to the right of the curve iri Figure 3-4 shall be operable.

APPLICABILITY:

When the temperature of one or more of the primary coolant system cold legs is less than 430°F. ACTION: a. With one PORV inoperable, either restore the inoperable PORV to operable status within 7 days or depressurize within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and either vent the PCS through a > 1.3 square I I I I I inch vent or open both PORV valves and both PORV block valves. I b. With both PORVs inoperable, depressurize within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and either vent the PCS through a > 1.3 square inch vent or open both PORV valves and both PORV block valves. /* c. The provisions of Specifications 3.0.3 and 3.0.4 are not

applicable.

Basis There are three pressure transients which could cause the PCS pressure tq exceed the pressure limits required by 10C.FR50 Appendix G. They are: (1) a charging/letdown imbalance, (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 (Reference

3) shows that when three charging pumps are I . operating and letdown is isolated and a spurious HPSI occurs I between 230°F and 430°F, the PORV setpoints ensure that 10CFR50 I Appendix G pressure limits will not be exceeded.

Below 230°F, I overpressure protection is still provided by the PORVs but HPSI I operability is precluded by the limitations of Technical

/ Specification 3.3.2 g. Above 430°F, the pressurizer safety I valves prevent 10CFR50 Appendix G linlits from being exceeded.

I 3-25a *Amendment No. $l, 7l, 117, TSP0889-0101-MD01-NL04 3 .1. 8 OVERPRESSURE_

PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION

3. f. 8. Basis (continued)

Assurance that the Appendix G limits for the reactor pressure I vessel will not be violated while operating at low temperature I is provided by the variable setpoint of the Low Temperature I Overpressure Protection (LTOP) system. The LTOP system is I programined and calibrated to ensure opening of the pressurizer I power operated relief valve (PORV) when the combination of primary I coolant system (PCS) pressure and temperature is above or to the I left of the limit displayed in Figure 3-4. That limit is developed I _from the more limiting of the heating or cooling limits for the I specific temperature of the PCS while heating or cooling at the I maximum permissible rate for that temperature.

The limit in I Figure 3-4 includes an allowance for pressure overshoot during the . I interval between the time pressurizer pressure reaches the limit, I and the time a PORV opens enough to terminate the pressure rise. I LTOP !s provided by two independent channels of measurement, / control, actuation, and valves, either one of which is capable of I providing full-protection.

The actual setpoint of PORV actuation I for LTOP will be lowered from the limit of Figure 3-4 to allow ./ for potenti_al instrument inaccuracies, measurement error, and I instrument drift. This will ensure that at no time between -I calibration intervals will the combination of PCS temperature I and pressure exceed the limits of Figure 3-4 without-PORV I actuation.

I I -When the shutdown cooling system is not isolated (M0-3015 and I M0-3016 open) from the PCS, assurance that the shutdown cooling I .system will not be pressurized above its design pressure is I afforded by the relief valves on the shutdown cooling system, I and_the limitations of sections 3.1.1.h., 3.1.2.a & c, and I 3.3.2.g. I I -. The requirement for the PCS to be depressurized and vented by an opening > 1.3 square inches (Reference

4) or by opening both I PORV valves and both PORV block valves when one or both PORVs are inoperable ensures that the 10CFR50 Appendix G pressure limits . will not be exceeded when one of the PORVs is assumed to fail 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 I keeping the control switch in the open position.

I -References

1. Technical Specification 3.3.2 2.* Technical Specification 3.1.2. 3. Consumers Power Company Engineering Analysis EA-FC-809-13

4. "Palisades Plant Overpressu:dzation Analysis" June 1987 and "Palisades Plant Primary Coolant System Overpressurization Subsystem Description" October 1977. 3-25b Amendment No. U7, I I I I I I TSP0889-0101-MD01-NL04

., **-* A. I w C'O . I . .

o CD .. . A. N .. a. LTOP T.S. LIMITS 2.soo .... * .......*.....

... * ... Figure *. a ... 4******.*:***************** . I ! I . . . *, 2,400 ..........

'. ..........

..........

- .........

.* .........

- .........

- ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. < 0 0 0 I t '

2,000 **********'.***********:**_*********:***********:***********:*

... *******:***

- - - - -****

- - - - . . . . . . . . . . . . . . I . : : : : : :

I : : : : : . : : .

I : 1,600 ..........................

.................................

., ..................

* * * * *

1 *. . . . . . . . i . : : : : : : :*. I . : ' * * * *

I * . . . . . . .. . . . . . . . \ . 1,200 ************************.*********'****************************************

- - - - '***.**[*******

800. *. . . ' . I ' I I I ... ' ....... ; ..........

..........

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . ' . . .. ___,.,,, .......................

.* .................... . . . . . . . . . . . . . . . . . . . . . .

50 100 150 200 . 250 300 . 350 400 *. PCS Degrees F * . . ., . -. . .. . . . .. ...... ' . ' .. * . . . '* -. -. ... . ' .. . . ; . . . . . -

-' 3 .3 EMERGENCY CORE COOLING SYSTEM (Continued)

g. HPSI pump operability shall be as follows: -------------_-------------1-)---If-the -reactor-head-is--iristaHed-, -both-HPS-I-pwnps--shaH-

/----

be rendered inoperable when: I a. The PCS temperature is < 230°F, or I b. Shutdown cooling isolation valves H0-3015 and M0-3016 I are open.

I 2) Two HPSI pumps shall be operable when the PCS temperature I is > 325°F. / 3) One HPSI pump may be made inoperable when the reactor is I subcritial provided the requirements of Section 3.3.2.c I are met. I 4) HPSI pump testing may be performed when the PCS temperature I is <430°F provided the HPSI pump manual discharge valve is I closed. I 3.3.3 Prior to returning to the Power Operation Condition aftei every time the plarit has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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*after maintenance, repair or replacement, the-following conditions shall be met: a. All pressure isolation valves listed in Table 4.3.1 shall be functional as a pressure isolation device, except as specified in b. Valve leakage shall not exceed the amounts indicated.

b. In the event that integrity of any pressure isolation valve specified in Table 4.3.1 cannot be dempnstrated, ai:: 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.Cl)

!Motor-operated valves shall be placed in the closed position and power supplies 3-30 Amendment No. j1, 101, 111, TSP0889-0101-MD01-NL04

-,_

". 3.3 EMERGENCY CORE COOLING SYSTEM Basis (continued)

c--demonstrate

.. that .. the .maximum.fueL.clad

.. temperatures-that-COUld

_________

. __ c ____ ----occur over the break .size spectrum are well below the melting

temperature of zirconium (3300°F)

Malfunction of the Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injection feature of the ECCS; therefore, it is disabled in the 'open' mode (by isolating the air supply) during plant operation.

This action assures *that it will not block flow during Safety Injection.

The inadvertent closing of any one of the Safety Injection bottle isolation valves in conjunction with a LOCA has not been analyzed.

To provide assurance that this will not occur, these valves are electrically locked open by a key switch in the control room. In addition, prior to critical the valves are checked open, and then the 480 volt breakers are opened. Thus, a failure of a breaker and a switch are required for any of the valvei to close. Insuring both HPSI pumps are inoperable when the PCS temperature I is < 230°F or the shutdown cooling isolation valves are open I eliminates PCS mass additions due to inadvertent HPSI pump starts. . I Both HPSI pumps starting in conjunction with .a charging/letdown I imbalance may cause 10CFR50 Appendix G limits to be exceeded I when the PCS temperature is < 230°F *. When the PCS temperature

/. is*> 430°F, the pressurizer

safety valves ensure that the PCS / pressure will not exceed 10CFR50 Appendix G. I

The requirement to have both HPSI trairis operable above 325°F *; provides added assurance that the effects of a LOCA occuring under LTOP conditions would be mitigated.

If a LOCA occurs when the primary system is less than or equai to 325°F, I the pressure would drop to the level where low pressure safety injection can prevent core damage. Therefore, when the PCS / temperature is >230°F and <325°F operation of the HPSI system I would not cause-the 10CFR50 Appendix G limits to be exceeded / nor is HPSI system operation necessary for core cooliqg. I HPSI pump testing with the HPSI pump manual discharge valve closed is permi.tted since. the closed valve eliminates the possibility of pump testing being the cause of a mass addition to the PCS. References.

(1) FSAR, Section 9.10.3; (2) FSAR, Section 6.1, TSP0889-0101-MD01-NL04 3-33 Amendment No. 11, jt, 101,

117, I I

b. The PCS vent(s) shall be verified to'be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the vent(s) is being used for protection except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open* posit-16rf, *i:h-en veri*fy-cne*se*vatveT<:>penat*

lea-st *onc::*e--*

per 31 days. c. When both open PORV valves are used as an alternative to venting the PCS, then verify both PORV valves and both PORV block valves are open at _least once per 7 days. Basis Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciatdr action and a check supplements this type of built-in surveiliance.

Based on experience in operation of both conventional and* nuclear plant systems when the plant is in operation, a checking frequency of once-per-shift is deemed adequate for reactor and steam system instrumentation.

Calibrations are performed to insure the presentation and acquisition of accurate information.

The power range channels and 6T power channels are are calibrated daily against a heat-balance standard to account for 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 drift be expected to remain within acceptable tolerances if recalibration 1S 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 frequencies of one-per--day for the power range safety channels, and once each refueling shutdown for the process system channels, are considered 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 failure" is defined as one which negates channel operability and which, to its nature, is revealed only when the channel is tested or attempts*to respond to a bonafide signal. 4-2 Amendment No

U , U , U 1 , t t8 , TSP0889-0101-MD01-NL04 I I *.

4 .6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS Applicability

--Applies--to-the-sa-£et--y-i-n-:i ee--t-i-on-s-ys-t-em,--t-he--c on ta-inment-spr ay -------

system, chemical injection systf'.m and the containment cooling 4.6.1 4.6.2 system tests. Objective To verify that the subject systeIJ1s will respond promptly and perform their intended functions, if required.

Specifications Safety Injection System a. System tests shall be performed at*each reactor refueling interval.

A test safety injection signal will

-to initiate operation of the system. -The safety injection and shutdown cooling system pump motors may be. de-energized for this test. The system will be considered satisfactory if control board indication and visual observations indicate that all components have received the safety injection signal the proper sequence and timing (ie, the pump breakers shall have opened and closed_, and all valves shall-have completed their travel}. b Both high pressure safety injection-pumps, P-66A and P-668 shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the temperature of one or more of the PCS cold legs is < 230°F or if shutdown cooling valves M0-3015 and M0-3016 are open unless the reactor head is removed. Containment Spray System a. Systein test shall be performed at each reactor refueling interval.

The test shall be* performed with the isolation valves in the spray supply lines at the containment blocked closed. Operation of the system is initiated by trippirrn . the normal actuation instrumentation.

b. At least every five years the spray nozzles shall be verified to be open. -c. The test will be considered satisfactory if observations indicate all components have operated satisfactorily.

4-39 Amendment No. $!, 13, tl1, TSP0889-0101-MD01-NL04

./ I

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS (Continued)

Basis (continued)

-** *------Dur.ing-r.eactor..-oper.at-ion-,--the--instt"Umentat-ion-which-is-depended---

*-------on to initiate safety injection and containment spray is generally checked daily and the initiating circuits are tested monthly. In addition, the active components (pumps and valves) are to be tested every three months to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test .interval of three months is based ori the judgment 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 of time. Verification that the spray piping and nozzles are open will be made initially by a smoke test or other suitably sensitive method, and at least every five years thereafter.

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.

Other systems that are also important to the emergency cooling function are the SI tanks, the component cooling system, the service water system and the containment air coolers. The SI tanks are a passive safety feature. In accordance with the specifications, the water volume and. pressure in the SI tanks are checked periodically.

The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory performarice.

With the reactor vessel head installed when the PCS cold leg temperature is less than 230°F' or if the shutdown cooling system isolation.valves M0-3015 and M0-3016 are open, the start .of one HPSI pump could cause the Appendix G 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. TSP0889-0101-MD01-NL04 4-41

Amendment No. 117, I I I I I I *,

ATTACHMENT II Consumers Power Company Palisades Plant Docket 50-255

EXISTING TECHNICAL SPEGIFICATION PAGES MARKED UP WITH PROPOSED CHANGES September 12, 19.89 31. Pages TSP0889-0101..:.MD01-NL04

.,,, .. ;

,, . 3. l ' ------------

3.1.1 *

PRIMARY COOLANT SYSTEM --------------ApplicabilltY--

-

.-*--Applies to the operable-status of the primary coolant system. Objective To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.

Specifications Operable Components

a. At least one prilnary coolant pump or one shutdown cooling pump with a flow rate greater than or equal to 2810 gpm shall be in operation whenever a change is beins made in the boron concentration' of the primary coolant and tha plant is operating in cold shutdown or above, except during an emergency loss of coolant flow situation.

Under these circumstances, the boron concentration may be.increased with* no primary coolant pumps or shutdown cooling pumps running. b. Four primary coolant pumps shall be in operation whene.ver*

the reactor is operated above hot shutdown, with the following Before removing a pump from service, thermal power shall be reduced as specified in Table and appropriate corrective*

action implemente".t.

With one pump out* of service, return the . pump to service rd.thin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (t eturn to

.. ,.p operation).

or be in hot (or below) '-lth the reacto1 . ,:;ripped (from the C-06 panel, opening 42-01 and 42-02 circuit breakers) within the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Start-up hot.shutdown) with less than four pumps is not permitted and operation with less than three pumps is not permittedo

c. The measured four coolant pumps reactor . . 6 . vessel flow shall be 124.3 x 10 lb/hr or greater, when corrected to 532°1. d. Both. steam generators shall be capabl& of. performing their heat transfer function whenever the average temperature of the primary coolant is above

e.

primary system pressure differentials shall not exceed-* the following:

(1) Deleted 3-lb *Amendment No JI, SJ, lt8, 119 December 12, 1988 TSP1088-0181-NL04 . . I

,

3.1 PRIMARY COOLANT SYSTEM (Continued) . ' 3.1. l ' . Operable Components (Continued)

*--

-___ .:_____ --*------------'--*--*

--* -------*----------

* --------__ : * . . (2) Hydrostatic tests shall be conducted in accordance with : . * . *

applicable paragraphs of Seed.on XI ASME Boiler & * * *. -*

Pressure Vessel Code (1974). Such tests shall be

conducted with sufficient pressure on the secondary side. *of the steam generators to restrict primary to secondary pressure differential to a maximum. of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plus SO psi where Po is nominal operating pressure.

(3) Primary side leak tests shall ba conducted at normal -operating pressure.

The temperature shall be consistent with applicable fracture toughness cri.taria for ferritic materials and shall ba selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi. (4) Maximum secondary hydrostatic teat pressure shall not exceed 1250 psia. A minimum temperature of 100°F is required *. Only ten cycles are permitted.

(5) Maximum. secondary leak test pressure shall not exceed *

1000 psia. A minimum temperature of 100°F is required.

(6) In performing the tests identified in.3.l.l.e(4) and** 3.l.l.e(5), above, the secondary pressure *shall not exceed the primary pressure by more than 350-psi. f. Nominal primary system operation pressure shall not exceed *j ., .. 2100 psia.

g. The.;*:eactor

inlet tO' .. -perature (indicated) si;all not e'l"ceed the'. value given by the following equation at steady state power *operation: . *

I . Ti l t S 543.3 + .0575(P-2060)

+ O.OOOOS(P-2060)**2

+ l.173(W"."120)

-I . n e .0102(W-120)**2 . . . * *

I* Where: Tinlet .

reactor inlet temperature in F 0 *

P

nominal operating pressure psia ** W

total recirculating mass flow in 10 6 lb/b

corrected to the operating temperature .conditions

. Whan the ASI exceeds the limits specified in Figure 3.0, within initiate corrective actions to restore the accep.table region. Restore the ASI to acceptable values within one hour or be at less than 70% of rated power within the follOwtng two hours.

If the measured primary coolant system flow rate is greater. than 130 M lbm/hr, the maximum inlet temperature shall be less.than or equal to the Tinlet LCO at 130 M lbm/hr. TSP1088-0181-NL04.

3-lc Amendment No JI, JI, U, U1, 118 November 15, 1988 I I I I I I I '() ' ..i.

*** .. . ,

3.1 SYSTEX (Cone C1J

3. l. l -------*--------------h. During init pri y coolant ?ump starts (Le., initiation of forced cir ion), secondary system temperature in che _,,,-* .. steam genera hall be < the PCS cold leg temperature

unless the S col eg temperature 450°F. *

i. .. The PCS shall not be heated or maintained above 325°F unless a minimum of 375 kW of pressurizer heater capacity is available froui both buses lD and lE. Should heater capacity from either bus 10 and lE fall below 375 kW, either restore the inoperable heaters to provide at least 37S kW of heater.capacity from both .buses lD *and lE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot shutdown with.in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Balis When primary coolant boron concentration is being changed, the process must be unif oim throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron *concentration which could result in a reactivity insertion.

.*. Sufficient mixing of the primary coolant is assured if one shutdown:**

or. one primary pump is in operation. ( 1) The shutdown cooling pump will circulate the primary system volume in. less than 60 minutes when operated at rated capacity.

By imposing a I minimum shutdown cooling pump flov rate of 2810 gpm, sufficient time I is provided for the oparatof do terminate the boron dilution under . I asynimetric flow conditions.

The pres*urizar volume is I inactive, therefore will tend to have a boron concentratioi..

higher r *. st of the prims.-.. i coolant systea .iuring a dilutiot..

!peration.

procadures vill provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in th* presnrizar and th* prtmar; syst.m *during the addition of boron. <2> Th* FSAlt° saf aty analy*is vaa padonad assuming four primUy coolant

I pumps were operating for accidents that occur during reactor . I *operation.

Therefore, reactar staHup above hot shutdown is not* I permitted

-unl*** all four primary coolant pump* are operating..

I . Op a ration with three primary coolant pump a is permit tad for . /

a limited time* to the restart of a stopped pWllf or for I reactor internals vibration monitoring and casting. I -Requiring th* plant to be in hot shutdown with th*-reactor.tripped I from th* C-06 panel9 opening the 42-01 and 42-02 circuit break*rs,_

I assures an inadvertent rod bank withdrawal will not be initiated I by.the control room operator.

Both steam generators are required*

I .to be operable whenever th* taaparatura of th* primary coolant is I greater than the design t91111eratur*

of th* .shUtdovn cooling system I to a*sura a redundant heat removal sy*t .. for th* reactor. I 3-ld Ali*ndmant No f7, 81, 111. 118 Novaaber 1,, 1988 .,

... *

3.1 PRIMARY COOLANT SYSTEM (Cont'd) 3 1 1 Operable Components (Cont'd) . .

_______ . __________ __ Q_"1rJQg_

_p_rilJlai:.y.

coQ.!_ant

__

___ Hf!i

__ Q_{ __________

forced circulation) at PCS cold leg temperatures 430°F, I the secondary system temperature in both steam generators I must be < the PCS cold leg temperature unless the Shutdown I iystem is isolated from the PCS and one of the I following conditions is met: I 1. The steam generator temperature shall PCS cold leg temperature by more than Cold Leg Temeerature

1. > 120°F and < 160°F -2. > 160°F and < 210°F --3. > 210°F and < 350°F not exceed the the 6T limit below. 6T Limit 100°F 10°F 100°F I I I I I l Under transient conditions with only one of the steam.. I generator.'

s temperatures higher than the PCS temperature, I and*the PCS temperature>

350°F: I I A. To start a PCP in the cold generator loop, I 6T in. the hot steam generator loop shall be shall be < 100°F; or . J /. B. To start a PCP in the hot steam generator loop, I* 6T shall be < 100°Fand the PCS pressure shall be I at least 100 psi less than the pressure lim"its of I* Figure 3-4. * . I . i. The PCS shalt" not be heated or maintained above 325°F unless a . minimum of 375 kW of pressurizer heater capacit}'

is available from both buses lD and lE. Should heater capacity from either bus lD and lE fall .below 375 kW, either *restore the* inoperable heaters to provide at least 375 kW of heater capacity from both buses lD. and lE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . When primary coolant boron concentration is being changed, the process.must be uriiform throughout the primary coolant system volume to prevent stratification of primary coolant at l_ower boron concentration which co*uld result in a reactivity insertion.

Sufficient mixing of the primary coolant .is assured if.one shutdown cooling _or: one pdm.a_ry co_olant pump in

The shutdown cooling pump will circulate the system volume in less than 60 minutes when operated at rated capacity.

By imposing a minimum shutdown cooling pump flow rate.of 2810 gpm, sufficient time is provided for the operatof to terminate the boron dilution under

asynmetric flow conditions.

5 J The pressurizer volume is relatively.

inactive, there*fore will tend to have a boron concentration higher than rest of the primary coolant system during a dilution opera.tion.

3-ld Amendment No 8i, ll1, ll8, TSP0889-0101-NL04

,

. ,

3.1 PRIMARY COOLANT SYSTEM (Contd) Basis (Contd) -----------------------Aami n ts t rati ve -pr-o-c-edu -p-rovi-de__:

£-err -u-se--o-f--p-r-e-s-s-i:rrize r.-sprays

_,:_ to maintain a nominal :: pread between the boron concentration in. the pressurizer and the primary system during the addition of boron.<2> The FSAR safety analysis was performed assuming four primary coolant pumps were operating for accidents that occur during reactor operation.

Therefore, reactor startup above hot shutdown is not .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 wi_th the reactor tripped from the C-06 panel, opening the 42-01 and 42-02 circuit breakers, assures an inadvertent rod bank withdrawal will.not be initiated by the control room operaior.

Both steam generators ate 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 heat removal system for the reactor. Calculatfons have been performed to demonstrate*

that a pressure differential of 1380 psi (3) can be withstood by a tube uniformiiy thinned to 36% of its original nominal wall thickness (64% degradation)., while maintaining: . . (1) A factor of safety of three between the actual pressure differential and the differential required to 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 ASHE BPV Section *xr (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 NOTT of the.,.J*&J cover of.+ 40° F. , . ,...,,...,.., The transient were performed assuming a vessel flow at hot . zero (532°F). of 124.3 x 10 lb/hr minus 6% to account for flow measurement uncertainty and core flow bypass. A DNB analysis was in a parametric fashion.to-determine the core iniet temperature as a function of pressure and flow for which the minimum DNBR is* equal to 1.17. This analysis includes the following uncertainties and 2% of rated power for power 3-2.* . . . Amendment No 10, U, US, TSP0889-0101-NL04 " ** ..

* * " ---*** *. '. 1 GGGt;.A..'l'l'-

________ _ -------*--* -----+ !!!..!..! (Cont'd) The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles.

The signal representing core power (Q) is the auctioneered higher of the*neutron flux power and the Delta-! power. The measured ASI calculated from the excora 4etector signals and adjusted for shape annealing CY 1) and the core power constitute an ordered pair (Q,Y 1). An alarm signal is activated before the ordered pair exceed boundaries specified Figure 3.0. The requirement that the steam generator temperature be < the PCS temperature when forced circulation is initiated in the PCS

ensures that* an energy addition caused by heat transferred from *the secondary sys tea to the PCS will not occur. This requirnaent applies only to the initiation of forced circulation (the start of the first primary coolant pump)' when the PCS cold le1 temperature is < 1 'I H thzit '15g*r the pcs 1du, v11l*1e* Al3D *References (1) Updated FSAll, (2) Updated FSAll, Sect (3) 1983/198 . Prorsrw.*

Report, Sa (4) ANl-87-150(NP), Vo , ,ta> . (BclccacO (J

' . (h)CPto tH,,11o1M1;,1;

on 4.3.7. . Stema Generat.or Evaluation tior-April l'9, 1984

2, Section lS.0.7.1 and Rep .. Jr ,, I I I I I I I I I I I I I I I "\ '"-/1t31r

&A-19-NJ.-19-11°"'__,,, ,4,,,_1y.,,:r , ) s how'59 rk. h"" ,:r, d c.nt el, 7;,;,.r wh. oM... "t ,:s-,:r. 1 .. r, c1 I J-n 1.r. Pe.S;i ""'d " c, 4'f'i*, Jc.. ;,,,; 7i*7id'

-Ii, .s-Ji..,,,, . e.-...,.ro.,

1$-)1ek, °"'°"" -lh C"ld /,' -r ...,,,,,P"'.

71;M.. ' . 3-3 Allandmant No JI, SI. 117, 118 November 15. 1988

'* I * *, 3.1 PRixARY COOLANT SYSTEM (

_ f '

3 .*l. 2 Heatup and Cool down Rates ---------.. -

---* * *** ** * --** . and as follows.*

114J S*l.

0 ,_ t-**'d*..,,,,.

I t a. Allowable combinations o pressure and tempera C.oo/J 0 ,.,,.,, rate shall be below an.d to the right of the plicable limit line as shown on F1gureS3-l The average heatup rate in any one hour time period shall not exceed* the heatup rate limit when one or s more PCS cold leg is less than the corresponding "Cold Leg j Temperature" below. *Cold Leg Temperature Reatup Rate Limit .. /L.fl 0 -i = ::a°:<wv-MP°F " " . 3. > .aM6r and < ti9'P 3$0*/I . 60°1/Br : .! UQ'P °F

100'1/Br *'* . . i '\ *£ ' j* ; J' ->.. i \.) :o *! ._{ &,* .. . Allowable combinations of pressure and temperature oldovn rate shall be below and to th* righc of applicable.

l

lines as shown on. Figura Th* av

cooldovn race '; , shall t ixceed th* Cooldow Rate.Lim an* one or more PCS colcl legs is.

than. th* correspond old L*I Temperature" ba1o¥i Cooldo¥a Rate Limit 100°1/Br 6o*r111r 40°1/Br 20°1/Br Allowable combinationa.of prassur* and tmparacure for warvica testtna durinl haacup are aa *hon ill Figure 3-3. Th* mazimua* heatup and cooldOVD rat** required by Section* a and b, above, are applicable.

Incarpolacion batvaan limit lin** for other than th* nocad tmqteratur*

change racaa i* parmitcad ill 3.1.Za. b or c. t. J.oo . . . . The averaae ha&CUp'l"aCaa for th* pras8urizer shall DOC ezca*d ..Wr/hr ill any one hour time tlJIP'-1 .. WI ie 1*** tb*p 618'Pa

th* shutdovwa coolinl 1yatam ia .3-4 Amandmanc No. Z7, It, JJ, 97 117

November 14, 1988 * . ..,, I */ I I I I I I I I I I I 1 I l I I I I I I I I I ./ I I I I I I I I

TSP 118.7-0218-Nl.04 ivJ.,,..,. -rAt sJ.,.,,J,,..,

,,.;..1.;;w, .--. l'lltJ!H:>("fD 1/1111" AcL Tk.

rA.11 ,,._..,,<<-

J,. T"'" .r._ >*14 .,/ M1*"1L f/IJ'"' .,o_* r /11>,. . . . * . . .*. ., .. ' ! ' .,

3.1.2 Heatup and Cooldown Rates (Continued)

-_________

'_ ___ J.1_.

radiation exposure of the reactor vessel .exceeds the . *

exposure--for-whi-ch-the--figuna-apply_, __

__

a11d_ 3-3 _ __:'""\ __ _ . shall be updated in accordance with the following criteria and ____ ----7 procedure:

. . *.

l. US Nuclear Regulatory Commission Regulatory Guide 1. 99'has : . 2 *. been used to predict the increase in transition temperature based on integrated fast neutron flux and surveillance test data. If measurements on the irradiated specimens show *increase above this curve, a new curve shall be constructed such that it is above and to the left of all applicable data points. Before the end of the integrated power period for which . Figures 3-1, 3-2 and 3-3 apply, the limit lines ou the figures shall be updated for a new integrated power period. Th* total integrated reactor therm.al.

power from start-up to th* end of the new power period shall be converted to an. equivalent integrated fast neutron exposure (! 1 MaV). Such a conversion shall. be made consistent with the dosimetry evaluation of capaula W-290(.12)

. . . . -. . . 3.. Th* limit lines in Figures 3-1, 3-2 and are on th* requirements of Reference 9, Paragraphs_IV.A.2 and IV._A.3. These lines reflect a presarvice hydrostatic test pressure of-2400 psig and.a vessel flange material reference temperature

of 60°1<8>. ** Basis .. Ali C011lponants in tile pr:imary caol*t system are d'eipad to withstand

th* ef f ecta of cyclic loads due to primary system temperature and (1) . . . pressure changaa. Th*** cyclic load* are introduced by normal unit load transianta, reactor tripa and start-uit and 11-utdovn During unit start-up and shutdown, the rates of temperature and pressure change* are limited. A maximum plant heatup and cooldovn limit of 100°1 per hour is consistent with th* design number of cycle* andsatisfiea.str***

limits for cyclic operation.

<2> Th* reactor vessel plate and material opposite th* core has bean purch&l*d.

to a specified Cbarpy V-Notch test result of 30 ft-lb or greater at an NDTr of *+ 10°1 or lesa .-The vessel weld ba.* _th* highest RTNDT of* plate. w*ld and HAZ materials at the fluenca to which the Figures 3-1, 3-2 and 3-3 apply.ClO)

Th* unirradiatad lTNDT -(11) .has been determined to b* -S6°F. An lTNDT*of -s6*r is used as an unirradiated value to which irradiation ef facts are added. In addi.tion, TSP1187-0218-NL04 3-5 Amendment No. %7, It, JJ, 89, 97, 117 November 14, 1988

.*, 3 .1.2 the plate has been 0% volumetrically inspected by ultrasonic test

______ .:_ ________

nd shear wave methods. The remaining

.* . . . material in the re-actor--ve

-el, an-d-*ottre-r primary-*caolant-1y*1t*em-

components, meeti the appropr

design code requirements and specific component function and a maximum NDTT of +4o*r.(5)

As a result of fast neutron irradiation in re1ion of the core, there will be an incl'ease in the iT with opention.

ho tiela: iqsas up'd St petii U Shi icU9rod hll RUI IA (8 I 1 tie() flW:Ui of &h* r*1°ror se111L asa in 8 asrian 1 ) 306 ai 1he PB/z8 ... 1'20 ia \aa1*'us*

13; 8ee1i1a II; 11 gsae Since the neutron spectra and the flu meaaured *at the samples and reactor vesael inaide radiua should be nearly idllAtical, the meaaured tranaition shift from a sample can be applied to tha adjacent section of tha reactor ve11al for later stages in plant life equivalent to the difference in calculated fluz maaaitude.

'Iba maai1m1m uposura of the reactor vusel will be obtained from tha measured sample uposura by application of tha calculated azimuthal neutron flua variation.

Tha predicted RT)fur shift for tha baH metal has been predicted baied upon aurvaillanca data and US NIC Regulatory Cuide.(10)

To compensate for any increase . in tha RT caused by irradiation, limits on the pre11ure-temperature relationship are periodically changed to. stay vi thin the streu limi ti during heat up and

-* -Ref ereace 1 providai a procadura for obtaining the allowable loadings for ferritic materials in Clas* 1 compoaenta.

This procedure is baaed on tha principles of linear elaat1c fracture mach,nics and involves.

a 1tre11 inteaaity factor predi.* .. r.ip*'.\ which is c -*or. .. bound of 1: '*',ir* ** dynamic ant. 1.:ro.r.k 7) arralt C\. .. itical value1 *.

1tre11 inten,\iy factor c011puted( . i1 a function of RTHur, oparatin1 temperature, and VCIHl vaU temperature gradianu.

Pressure-temperature liait calci&lational procaduras for the reactor coolant ptaaaure boundary are def inad in ief ereace 8 baied upon Rafarence

7. The limit lines of Pigura* 3-1 through 3-l consider a 54 psi allowance to account for the fact that pra11ure is maasurad in the pre11uriaer rather than at the ve11al . beltline.

In addition, for calculatioaal purpo1e1, 5*p ... JO p*i ......_ taken aa maasurlllltnt erroi-allowance.

for. tllilperatura,...._

pc a*sfss1i zl;b By Raferenca 7, rea or ve11el wall locations at 1/4 and 3/4 thickneH ara lim' ing. It is at th11e _ locations that the. cr*c;k propagation a11 1atad with the hypothetical flaw muat ba arra1ted.

A these locations, flueaca . attenuation and.thermal gradients ha bean Amendment No. 27, II, 8t, t7, 117 November 14, 1988 TSP1187-0218-NL04 0 ,,J (} ,,J. 7i,;. { c,,./-r, i." /, 7Y --** ,4 ". ' *'

e

3.1.2 Heatup and Cooldown Rates (Continued) !!.!!:! (Cont'd) . . *. * . . . * § . * -----------

-t ':)._. 3/4 thickness location and thermal gradient stresses are tensile there. . During heatup. either the. 1I4 thickness or 3 I 4 thickness -location may be limiting depending upon heatup rate. a t! , ' Figures 3-1 through 3-3 define stress limitations only from a fracture point of view. * *

Other considerations may be more restrictive with respect to pressure-temperature limits. For normal operation, other* inherent *plant characteristics may limit the heatup and cooldown rates which can be achieved.

Pump parameters and pressurizer heating capacity tends to restrict both normal heatup and cooldown rates to less than 60°P per hour. * * -4 1; f' " *,t , t 4 The revised pressure-temperature limits are applicable to reactor .* i

vessel inner wall fluences of up to 1.8 x l019nvt. The

-of appropriate fluence attenuation factors (Reference

10) at tha I * . 0 . 1/4 and 3/4 locations results in. RTNDT* shifts of 241°F *i .. 171 respectively, for the limiting weld material.

The y .. 1. ** *.

criticality condition which defines a temperature below which I T the ce>re cannot be made critical (a_ trictly based upon fracture I '-* *

mechanics' considerations) is 371°F. The most 11.Jiaiting wall *'rl .

location is at 1/4 thickness.

The minimum criticality

* * .{ temperature, 371°P is the minimum permissible temperature for " the inservice

- ystem pressure test.* That temperature

- tl "f-is calculated . -ased upon 2310 }-Sig h} Jrostatic test

R ' i pressure.

sJO i The restriction of average hea .. cooldown rates to 100°F/h . i .a.{ I when all PCS cold legs are 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, f ensures that the requirements of References 6, 7, 8 and 9 are met.:.:..J Tho core operational limit applies only when the. reactor is critical. , , , *The heatup and cooldown rate* restrictions

pp 11 ceh'* *Ni1a *h* temo*rnmr*

a' an* a* **** tf '"' RCS cold la" 1a l111a $k a SSC l are consistent-with the analyses perfoTJne4 low temperature

* .,. 5::ssure protection (LTOP) (References 13, 14, 15, 16 & 17). *or -. P or above, the PCS safety valves. provide oveii>ressure

protection for heatup or cooldown rates < 100°F/hr *. TSP118?-0218-NL04 I I I I I I

....._ ./' J.l.2 -Heatup and Cooldown Rates (Cor ........ I . -

--.-_

_**-r-*-_

-

.. ,

o. core operational curves adhere to the requirements of Reference

9. ' The inservice test curves incorporate allowances for the thermal l gradients associated with the heatup curve used to attain inservice

;( _'t test pressure.

These curves differ from heatup curves only with l t\' ' respect to margin for primary me!llbrane stress.<7) Due to the shifts :'-' in RTNDT' NDTT requirements associated with nonreactor vessel *

materials are. for all practical purposes.

no longer limiting.

i Ref ecencu ' Q .! *f (l) FSAR. -section 4.2.2. " \ * ' (2) ASM! Boiler and Pressure Vessel Code,Section III, A-2000. A ' *

f * (3) Battelle Columbus Laboratories Report, "Palisades Pressure I *" 4' !I Vessel Irradiation Capsule Program: Unirradiated Mechanical -, "> *'al . _Properties," August 2.5. 1977. \' * *'-QC *l "t (4) Battelle Columbus Laboratories Report. "Palisades Nuclear Plant ' ' Q I Reactor Vessel Surveillance Program: Capsule A-240," March 13. Q..

I t

1979, submitted to the NRC by Consumers Power Company letter * '"'-. t I: * ( dated July 2, 1979. r t { J* (.5) FSAll, Section 4.2.4. -* ) *t l -... E "" (6) Uli Weashn Beaz*Jetoqr cnmmhe1sa, Rtau_laeuz Gdide ( . I *i a I'-i "Ffhtll .. Ri&idual EliEidtS OU Ptidlttacl Raclie*6**

a._.,. Cc 9 "1 J ..! I

llHUH quul lf81Hhh;" a*l:J j U?lo

.,, (7) ASM! Boiler and Presaure Vessel Coda,Section III, Appendix G, U ' " "Protection Against Non-Ductile Failure," 1974 Edition. * ! .* J .. f i* (8) US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section S.3.2, "Pressure-Temperature LiJllits." ... 't i(\-(9) *10 CFR Part SO, Appendix G, "Fracture Toughness Requirements," r

J °" May 31, 19830.0 ...,..,J._J Nw....M 4, /916, .

J' \\ ( 10) US Nuclear Regulatory Commission, Regulatory-Guide 1. 99, 81el1 CL ( Revision 2, 'f*Uu '9 86 lllll*f,lfll'

_

I ' (11) Co11lbu*ti011 Engineering Report C!H-189, December, 1981.--C (12) "Analyli*

of Capsules T-330 and W-290 from the Consumers Power ' I > Company Pali*ad**

Reactor Vessel Radiation Surveillance

1) ( 'i Program," WCAP-10637, September, 1984. * * . 1[ '$) J 1 I' PU II 101 "C1h11llliH sf PCI i'HHuse IaeneH is1* 'Uiag I'= ua 8'* (3 .........

'***) Ideue ehe P8M'& 8paa;P Bn*ht* 4. 1 *i (14) ""' H8l19 "8alcalattca cf laqatzad eea* eapach, H Vat=tdn th* !Cl ld111r '****** &;" lwaazy U, U88. " -i! ""' _H_ I) .. PM titer 888128 *** I -** !819 HUI ClplEif ... IC !xr&CU. .--tj .. -. . L';OI c, ** , ....... ..... u. 1911 I " -Hi) M a;QI 110111 . "C11tuht1n o' Tim* fg= o,uatot co. Act *""""-M8 AH *** luhhlt" hauHy 38; Ull'o ** ('17) !111: 8991 88fl!f 8/8 81 "Paliaaclaa Plaat PtlmatJ ejotdt Sy&ta* pu .. ,,_, Tnqa*ret11r*

I 1.1 .. PH 'Ht**MI Q If Iha t:** Bctlat 'ad pgurzga guul Cocf a" h*'"** 3-8 TSP1187-0218=NL04 Amendment No. Z7, It, JJ, IJ, f7, 117 November 14_, 1988_ I /' I I I l I I I I I I i .. l

> li 0 < ID ID ::1 !l G' a ID .. '1 :;l rt ...... "'" 0 . ...... '° 61 N. CX>'9. ""-l en--.. '9.-.-:: ""-l t-4. .

I*-' \A .*I* . w . I 'O PALISADES PLANT PRESSURE AND V£MPERATURE LIMITS FOR HEATUP For fluence to 1.1x10* nvt . . , ....... 5-1 100 f /Hi:

80 f /HI u liO f /Hr

.....

£

40 f /HI .___...__..._

__

50 . 15. 100 . 125 *150 I 1:>> 200 . 225

250 J15 JOO ----------

... -. ---...--rTURE DCGllEES f I ¥ ' *

_,._:...._.

_,..,.. * '

* ,.,. ' * * ** >' '" * * * '* ' * . .* ** * "'",. ...

......... , .. *.* , *. :...,., l.t * :. ' .... . "" .. *' .*

FIGURE 3-1 . I. . I . . PALISADES PRESSURE AND TEMPERATURE LIMITS FOR H8ATlJP 1':l.. .. .. .. l::A. ":A. .. . ! 250 0 50 . 75 :100 125 . 150 175 200 225. 250 .,.;275 300 . 325 350 375 TEMPERATURE DEGREES F ... _;. ... . .;..; ..... _ ...... _. :._ I I . I t I 400 1425 450 I I . , I I r

= 0 < ID ID '1 I-' f

'" ii' ID ::s . G' rt :;:= 0 . p R E S I 't s .........

0 u R I E I I II ,, I I . --11:* *--1 .* . . I ---**--Jl-*-.. I : I 50 7S . IOI> 125 150 175 1.W 225 250

27:>> lOO l2S JI) 40o> TEMPERATURE*OEGREES r . . ' I' I . !

...

FIGURE 3-2 *PALISADES PRESSURE & TEMPERATURE LIMITS FOR COOLOOWN /_ . . . (9 2 . . ' F' = l.H X 10 n/cm (No Measurement Uncertainty Included)

I I . I I

llJ 'J v ' 0 ' : '/ I -I I I I *_._ I I .. flj I I I . '. I ' . I I {\ i 60 40 20 ' ; 80 I --I v. : I ; I 10(] I 80 I I _J !.---:'. i c.I

I ---ac I : 20 i I 0 I i *

J I ' * ... .... I* I I *' I I I I I ' ' ' '* ' ' ' .

TT*T T ** I I I I . . . . **** . ... . ' ' I *** I I 1 1* I 50' 75 100 125 150 175 200 225 250 275 . 300 I 375 400 425 . ! 450 325.* 350 TEMPERATURE DEGREES F ': ,:.

g; fl' < CD . ! i:s Ii .. ,. ... g .... 0 .

PALISADES PLANT PRESSURE AND TEMPERATURE LIMITS fOR HYDROSTATIC TESTING

for nu.nc:e TO 1.a X 10 11 nvl /

fleur* 5-J . 1S

100 JOO TEMPERATURE DEGREES F .._ __ ---""'!/ -* ' I I I . I I '" * * * * ' * -* ** **. ** * ...... , .. ..; * <-4 """ * * * * .. 1.*. ...:.. ' ' . * . * -.... ,..;.::,., .. , .........

-, ..... . . i'1 '-"

I I . I. I FIGURE I

PALISADES PRESSURE AND TEMPERATURE LIMITS FOR HYDRO .. I .* . PRESS PSIG . 3000 . . *2750 . . .. 2500 2250 . ' . 2000 . . 1750 . . 1500 . . w. I 1250 I-' I-' . . . §" 1000 . . . CD ::s . §' 750 CD . ::s c+ !<;:: 500 . 0 . . . 250 i 1-4. "$<::: 0 1-4.1-4. . .. .. * ' !

l::R. l::R. 50 .. 'l5Q ... *!. .

19 2 f = 1.8 X *10 n/cm j* ! //J V/1 II I I ' !//! I/ 0 ! 1//;' I I 20 . ' ///; w I I 40 lj/j , I i If' I .. I 6C I @ i :10 I I / ; 100 I _.......::::;;

v v ' I I I /" / v I .,,, I I ---. I. ---------..,.-/; I I I ---I -..,. I -----I I I I I I I I I I I I I I I I I I I I I I I I t I I I 1. I .... O I O I IO I 0 I I I I . ' I *-.. I ' . I 75 100 . 125 150 175 . 200 225 250 275 300 TEMPERATURE DEGREES F 325 350 375 400 425 450 i I I I * *

. __

____

l _ .:, LIMITING CONDITIONS FOR OPERATION

_____ --------------f 'C ** '

3.1.8.1 REQUIREMENTS .a . e I c

a. b. c. !!ban the ** ., ** aswso cf 92 2 a= mo=e ei ll:u ,., *** ,. liiu 1 ant s)'&t 2 m cal <I h8c ts.

388°F; ct wheneoec the shaede"" au 1 1n& ieel:al!hR

aloes meiJ 3815 and HST/ 3916) are epH;. &uo po"er aparar* 1 l ia1hec ehall 91 epeR; e* each papv p 1'ar values iipd both POiU' e leek "91"U atlellu 9 a ep ea. ui.: ** *** UnpHehn el s::c n sH* ai. eha pt!m&tJ spcem ii ' tJfJRp.1 Two power operated relief valves (PORVs) a lift

.. """.-..-..,...
.. ...i...i...,.'911....,.,.,.,_._.,...
... .. When the temperature of one or more of the system cold legs is less than 4Jo*r. With one PORV inoperable.
either restore the inoperable PORV to operable status within 7 days or depressurize and within . the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either vent the PCS through a > 1. 3 .square _inch vent or open both PORV,.....
valves and both PORV block* valves.* With both PORVs inoperable.
depresaurizo and within tho next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either vent tho PCS through a > 1.3 square inch vent or open both PORV 9611* valves and both PORV block valves.
The prOVi*ion*
of Specifications 3.0.3 and 3.0.4 are not applicable.*
Bui*-There are threo:presaure transients which co11ld cause tho PCS pres8Ur* to exceed the pressure limits required by lOCPR.50 Appendix G. They are: (1) a charging/letdown imbalance.
(2) tho start of a high pressure safety injection (BPSI) pump, and (3) initiation of forced circulation in. tho PCS when tho steam temperature is higher than tho PCS temperature.
3-2.5a Amaadmoat No. J2, 7%, 117
Novabor 14, 198&-TSP1187-0218-ML04
.... ----T*; I :. I I I I I I I I I I I I I I I
I I I l I I I I I I I I -J I l I ./ I I I I I I I I ../ I I I . ' ei
...... 3. l *. 8 , OVfRP!{ESSt:RE-SYSTE: LIMITING CONDITIONS FOR OPERATION 3.l.8. !!!!.! (continued) . _, . . Analysis (ReferenceurJ:;
4 & it shows chat when three charging pumps are operating and letdoWT! is isolated and a spurious HPSI occurs, the PORV setpoints ensure that lOCFRSO Appendix G pressure limits will not be exceeded.
Above 430°F, the pressurizer safety valves prevent lOCFR Appendix G limits from being excHded.*r 1 ehaegtu5/
Iordm,m tmbelanc* (l1fe*****
i)* Ib* ***nt1em1at tftit sc1am gaaetatot ae ' cbe RCi tiapencu**
w&:a111 iH11* einttlattoa ts taiftat&d ti: itae PCS ll"l'H clla1 .. ean11 addittua caasad by triae ttausfaud boa th* 1111***., 11ee* eu tha res *111 nut uccaz a 'Phi* u 1ali1 tu .eha "''-*iea cf fotca4 etzealaetua (Ille ua1 . gf the f1ret P*'*'"' eeelseal '".,) ""la* eB& oz liULW of aha Pe9 SPld lei ump1ser 11 r 11 < 45g*r I I I -./-o--I I I I I I I I I I I I I I I . I :;:;!;1:: !:: * ;J:MS.>1-r ene"HI Mel ehu shceduwa C0611Hi IJ&t& **ill aec b* pree111g1zed
" / ... */T>. a'ae"e. , ... .,, .. ,, .. ,H .. *-ur**** _
The requirement for the PCS to be depreaauriaed and vented by an I opening> 1.3 square inch** (Reference 3), or by openina both I POIV pilot valves and botb POIV block valves when one or both I POIVs are inoperable eneur11 that the Appendiz G preseure I *1 limits will not be exceeded When one of the POIV1 ia aaeumed to I fail per th* .... single lOCFl.50 _Appeudiz A, , I
_. 1
34. S1#1.., +. Pt#ll" tr* .* #ti ,:,,-.;r,_.,.o,_....,,._,. -ro i 8/11-1111-PH*.*".._
__ .,,,..,,,n, ,.,.,.,..r, .,-APOltl/.r;w..., ... J,,'" O/f&llt'n s *** , ** a.a.a,,,,)
.... , *** I ***6ectad opc1aeu1 wbaa d!i POI'* I *re
.. allewa* *a &eeliea Julul lulu '*1ly16* d****i)iug I thte 11****'9*
nfat1ae111
!ii cha late *** hutn 1 J, * *
I I> 'I /f/u,p1# +M.. e.,,,,,,..,-,,,., -/I.e. .,. 0 * *' flJN', Ti * . . Refereucea
* . i?c.J.,.,,,i.,/ .S-1Nc:J1c17i;.,,.,..
3.a. 'Z. * . 1.
BA PJ6 89 tel Ca1tU1aetoa at Pea Pnuuazu !wt**** h** Mdfpl n+e;* (3 lll**li**
,,..,,)
etsu P81¥u *****" xazepb*r +, l8IJ1 2. *Tecbaical Specification 3.1.2.
+!'-6 $ "Palia&d**
Plat Analysia." June 1977 ancl Palisade*
Plue Primary Coolant Sy1cea Ov1rpreHuri1ation Subay1c* De1cripci.on, *.* October. 1977.
u=1u .rroa H8U9 * "ea:tetll1:1ba a' Beqad* .. J&ar eapae:tcr t= "d*la'9 Ille JQ& Balow *ppaucUs 8 &a: .. H;P lcamj th Hiia " '" JMO, H8H8 "Pdt&adi&
tlW IOKV ildititi eapcc611*
TAan raa le9JH*l*i'*
388'P ct 8t1acaz." i***.,. 20 1saL .S. t!PC
,44*1,,.11".r 3-2S1t Ameuduac 117 Movember 14.: 1988 l I I I I I I ., I I I I /. .,
.. L t ---------_'._j-.-1-:.. __ QVERPRESSURE . PROTECTION SYSTEMS . ------------------------------------

LIMITING CONDITIONS FOR OPERATION 3.1.8. Basis (continued)

Assurance that the Appendix G limits for the reactor pressure vessel will not be violated while operating at low temperature is provided by the variable setpoint of the Low Temperature Overpressure Protection (LTOP) system. The LTOP system is programm.e_d and calib-rated to ensure opening of the pressurizer power operated relief valve (PORV) when the combination.of primary coolant system (PCS) pressure and temperature is above-or to the left of the limi_t displayed in Figure 3-4. That limit is developed from the more limiting of the heating or cooling limits for the specific temperature of the PCS while heating or cooling at the maximum permissible rate for that temperature.
The limit in Figure 3-4 includes an allowance for pressure overshoot during the interval between the time pressurizer pressure reaches the limit, and the time a PORV opens enough to terminate the pressure rise. LTOP is by two independent channels of_ measurement, -*-control, actuation, and either one of which is capable of : . providing full protection.
The actual setpoint of PORV actuation
-::* for LTOP will be lowered from the limit of Figure 3-4 to allow_ for potential instrunient inaccuracies, measurement error,.and instrument drift. This will ensure that at no time between -calibration intervals will the combination pf PCS temperature -and. pressure exceed the limits of Figure 3-4 PORV actuation.
When the shutdown cooling system is not isolated (M0-3015 and_* M0-3016 open) from the PCS, assurance that th!'l shutdown cooling sy.stem will _not be pressurized above its design pressure is ;
afforded by the relief valves on the shutdown cooling system, . and the limitations of sections 3.1.1 h., 3.1.2 a & c, and
3.3 2 g. The requirement for the PCS to be depressurized and vented by an I I l I I l I I / 1. I I / I I I I I I I I I I I I I I I I I r opening 1.3 square inches (Reference
4) or by opening both *I PORV valves and both PORV block valves when one or both PORVs are inoperable ensures that the lOCFRSO Appendix G pressure limits will not. be exceeded when one of the PORVs is assumed to* fail per the_ "single* failure" criteria lOCFRSO Appendix A, Criterion
34. -since the PORV solenoid is_
enough to overcome _spring pressure and valve disc weight, -the PORVs -may* be held open -by_ I keeping the control switch in the open position.
I References
1. Technical Specification 3.3.2 I 2. Technical Specification 3.1.2. I 3. Consumers Power Company Engineering Analysis EA-FC-809-13
4. "Palisades Plant Overpressurization Analysis"*June 1987 and I "Palisades Plant Primary Coolant System Overpressurization I Subsystem Description" October 1977. I I 3-25b Amendment No. tt1, *., .
. w I I\) .. \Jl p.* "' *-** D. . I (I) (I) CD .. a. N .. a. , LTOP T.S. LIMITS . . 2.soo . * .........
...... Figu.re a. .. 4 .... ** ....... * ... * ....................
..... * .. . . . . * . * * .*
I
2,400 ... : .. * ...... : ....... : ... : ..........
.............
...........................................
! ..... * ' . * . * . . * . . I . . . . . . . . . . . . . . . . . ' . .. . . . . . . . . . . . . . . . . . . * .. ' . . . . . . . . . . . . ............................
' ..........................................................
-. . . . ' . . . . . . . 2,000 . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' * ' * * '
II I * * * * ' '
I * * . * * . , .
I . .............. " ....................................
....... * .... * .... : .. *. .. . . . . . . . . .
..... *I* ...... . * * * * . * * * * *
1
1,600 ' 0 0 I 0
0 0 o . . . . . . . . . . *. . . . . . . . I . . . . . . . I . " : * ' * . . * ' ' I ' . . . . . . . . . .. . ' .... .... * *.... '* .. * ... i .... * ..... ' .... * .. * . . . . . . . .
.. . . . . . . . . . .
.. . . . . . . . . . . . . . . . . . . . . . . . . . . .
.. * .. ' .... 1,200 . . . . . . **********:***********:***********:-.**********:***********:**********:**********":***)********:* . . . . . . . * * * * * . * ! 800 . . . ' . . . . . . . . . . . . . . : 400 . . . . . " . . . . -L*..;...;..* . . ;..;..;* .....................
---"'-.;..,;..---..-.-:
................................................ . *. .* . i : *. I * * * ! * * . .
i . ' . : : *i : ... 50 100 150 **200 250 . 300 350 400 .. *1** . . *
PCS. Degrees F ... * . ' I. . . I .. . . . ,... .." ... .. ,.. ,, ... _,_ -. ...... I**** ..... . , .... . . ; ',.I* ,* ' . I . ..
-*-------3 . EMERGENCY CORE COOLING. SYSTEM
......
i / Applicab1lity

Appli** to tha operating statua of tha emergency core cooling system. Obhcciva.

To assure operability of equipment required to remove decay heat from the core in either emargancy or normal shutdown situations.
Specif icat1ona Safety Injection and Shutdown Coolin& Syscama 3.3.l Th* reactor shall not be made cr.itical, excapc for lov-temparatura physics taata, wll.asa all of the following.
concliciou are mac: a. Tha SillW tank contaiu DOC la** tb&D 2SO,OOO gallou of water with. a boron concancratioa.
of ac laaac 1720 ppm buc not more th&ll . 2000 ppm at a temperature*
not leaa th&ll 40.F. *. b. All four Safety Injection caa.ka are operable ad praa11urizacl c*o ac leaac
200 psig with a tank liquid level of ac leaac 186 inchea'. (SS.5%) and a maximum level of 198 illchea (59%) with a boron . concentration of at laaac 1720 ppm buc not more than 2000 ppm. l c. Oue iow-prHsure Safety IAjection pump 1a operable on each bua .* d. Ona high-pressure Safety Injection pump ia operable oa. each bua. a. Boch haac exchanger*
ancl both component cooling heac
igar* are oparaf.1,a.
f. Piping &nd valvaa shall ba operable co provide cwoflov from the SillW tank.co tile primary cooling syacea *. g. All valvaa, piping ad iilcarlocu aaaociacacl with th* above componanca and requiracl co f unctiOD during accid8nc conditio'na are opar*bla.
* *
h. The t.0¥-Pruaura Safecy Injection Plov Control Valve CV..-3006 shall be opaDacl aD4 disabled (by isolating cha.atr*supply) to prevent apurioua cloaure. . . i.* tha Safety Injection bottle isolation valvaa shall be opened _with cha _electric power supply co *th* valve lllOtor * .. disconnected.
j. the Safety Injection miniflow valvea CV-3027 and 3056 shall ba opened with and 3056 positions to maintain them open. 3-29 Amendment No. 7i. lOl February 10, 198 7

J.3 E?-!ERGENCY CORE COOLING SYSTEM (Contd) 3.3.2 During power operation.

the requirements of 3.3.l may be modified to allow one of th* following-conditions to be true at* any one ti.ma. If the system is not to meet the requirements of 3.3.l
within th* time period specified below, the reactor shall be placed in a hot shutdowu condition 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the* requirements of 3.3.1 are not mac within an.additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in a cold shutdown condition within 24 houra. a. One safaty injection*cank may be inoperable for a period of no more than one-hour.
b. c. Ona low-pressure safaty injection pump may be inoperable provided the pump is raatored to operable statu. within 24 houra. The other low-pr***ur*
safety injaction pump shall be testad to damoDStrata operability prior to initiating rapair of *
the inoperable pump. One high-pr***ura.
safaty injaction pump may be inoperable providad the pump is testorad to operable statu. within * .... 24 houra. The other higb-pras*ura safety injection pump shall _be teatad to damoutrate oparability prior to initiadng.rapair*
of th* inoperable pump. d. One ahutclovn haat ezchaDgar and Olla C011lpO'llallCC0011Dg watar heat*ezchangar may be iuoperabla for a period of no more th&ll 24 houra. * *i" e. AAy valvec. btarlocks or piping dii;actly aaaGciatad with oat\ of the above component*
and raquirad to function during accidanc conditiona shall be daemad to be part of that componant and shall meet the same raquiraaent*
listed for that component.
f. AAy valve,. interlock or pipe aaaociatecl with the safety injaction and sbutdOVD coolin1 syat** and which is not coveracl under 3.3.2* abna but, vbida ia required to fUDc:tion during acc:ideni coDclitiona1 may be inoperable for a period of no more than 24 hour** Prior to initiating repairs, all valve* and *
interlocka in the *Y*t* that provide th* duplicate functioa *hall ba t**tecl co damonatrate 3-29* AmncblnC Mo %1 51 Sept.mer 10, 1979 * * **** ., .. , *,, *
---*-----*---r:3--EMERGE-NC-Y-GORE--COOLIN_G SYSTE. .. , ........ ____ . * *-** g. .*.;

HPSI Pump operability shall be as follows: 2) Both HPSI Pumps shall be rendered inoperable whenev CS temperature is < 300°F unless the reactor ves 1 ad is removed. mum of 1 HPS! pump may be operable cure 300°F but <

3) One, and ly one, HPSI Pump shall be PCS tempera ra is > 350°F but < 43 whenever 4) At least one whenever PCS 5) the PCS 6) One*HPSI pump may b is subcritical and noperable when the reactor mperature is > 460°F, ction 3.3.27c are met. 7) provided the emperature*
is betve 385 °F to 430°F and is not armed, then a de cated licensed*
all be stationed in the c trol room to an inadvertent HP.SI Pump. sta and stop Charging s necessary to limit PCS pressure.
8) ty Injection Actuation System (SIAS) te ing shall not performed .while the PCS is betw'een 300°F a d 430°F.
PSI 'pump *-t.esting may be b*.:.ow 430° rovia.* d the HPSI manual disch'arge valve is closed. 3.3.3 Prior to returning to the Paver Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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.l to service after iDaintenanco, repair or replacement, the following
_conditions shall be met: a. All. pressure isolation valves listed in Table 4.3. l shall be functional as a pressure isolation except as specified in b. . Valve leakage shall not .exceed the ind_icated.
b. In the event that integrity of any pressure isolation valve specified in Tabla 4.3.1 cannot be at least two valves in each high pressure line having a non-functional valve must be in and remain in, the mode correspondinR to the isolated condttion. (l) Motor-operated valves shall be placed in the closed position and power supplies deenergized.
..
/ I I I I I I I I I . I I I I I I I I I I I /. I .I I I I I I I. I 1. I . *'* *1 '.: \ , -* > 3-30 Amendment No. JI, tel; 117 Novembiar 14, 1988.
TSP0189-0002-NL04

*-

** --**-J.-3-.-EMERGENCY CORE COOLING SYSTEM (Continued)
1) If the reactor head is installed, both HPSI pumps shall be rendered inoperable when: a. The PCS temperature*
is < 230°F, or b. Shutdown cooling isolation valves H0-3015 and H0-3016 are open * . 2) Two HPSI pumps shall be operable when the PCS temperature is > 325°F. 3) One HPSI pump may be made inoperable when the reactor is
subcritial provided the requirements of Section 3.3.2.c* are
4) HPSI pump testing may be performed when the PCS temperature is <430°F provided the HPSI pump manual discharge valve is closed.
3.3.3 Prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, .or the Cold Shutdown Condition for inore than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and testing of Specification 4.J.h has not been.accomplished in the previous 9 months, or prior to returning the check valves in Table 4.3.1 to service after maintenance, repair or replacement, the following conditions shall be met: . a. All pressure isolation valves listed'."'in .Table 4.3.1 ::;'..all be fµnctional as a isoiation device, except as specified in b. Valve leakage shall not exceed the amounts indicated
. b. In the event that integrity of any pressure isolation valve specified in Table 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.Cl)
!Motor-operated valves shall be placed in the closed position and power . supplies deenergized.
3-30 Amendment No. St, lfl, 111, TSP0889-010 l-NL04.
I I I I I I I I I I I I I -'

...!_

__
3.3 COOLING SYSTEM
. de_!!tQ_nstrate that the maximum fuel clad temperatures that* could ' occur--ove-r -th*e
__ __ well below the melting temperature of zirconium (3300°F) *.
Malfunction of th* Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injection f eatura of the
ECCS; therefore, it is disabled in the 'open' mode (by isolating the air supply) during plant operation.
This action assures
that.it will not block flow during Safety Injection.
The inadvertent closing of any one of the Safety Injection bottle isolation valves in conjunction with a LOCA has tioc been analyzed.
To provide assurance chac this will not occur. these valves are electrically lockad open by a key switch in the control room. In addition, prior to critical the valves are checked open. and then the 480 volt breakara are op.ened. Thua, a failure of a breaker and a svitch.ara required for any* of tha valves CO cloH. 1 ,,-. .r aaa , ,., le a 11*haua wwwwl1aiccl
'" I m411 additigp*
due cg tpady1rc1a1 eve pump atazts. Buch !PSI I *.
!::!iWl!;:, 'e: ..
.* temper*mra f'
438'P. *RIB tha PCS tnperatura la ! 4]U*f, -. I -. **
th* pr11eng1 "* Hht) oalVil ldaura Ehat Eba fiCI prasiUfl Utll
y / nae *xceed 1 ocnso ._,,aadtz 8 U:ud:l& Wliid 661 bf boEh &St ( I ..pimp* 'Fl lllHld . . . . . . ,; I . . ** .llL' Th* requiramant to have ... BPSI . abova <Nf*r-* I providaa added a*surance that cha effects of a LOCA occuring I under LTOP condition*
would be mitigated.
If a LOCA occur* When I the priml!.1 system 1a l*** than or equal to
M.r I the prasa\:*r*
would drop t.** *the level vberl:* lov pras*ure safety . I injection ca core dam:'.8**1
I. I I I I I I I -I BPSt pap ceatinl vi:ph ha BPSr pump manual di8char1*
v8l.ve *. * ***I cloaacl ia pamitted
ca th** cloHd valve eliminate*
th* I poaaibility of pump Hting being the* cauae of a mua addition I to the* PCS. * * ** . . *
I .Reference*
(1) (2) (3) PSil, Sect 9.10.3; FSAJl, Sec lon 6.1,
U.-PAL-OP-880121 "Calculation of Tiu for Operation to Act for BP and Bubbl*"**
20, 1988. * *3-33 I } ; . ' j. ) '*! Amand11lant No. U, JI, rt!, 117 *: _...J. -1., . Novaber 14, 1988
0 * ; * . *hti ¥/t PtS 1.r Jo 'F' o.. d 1 . J.U 0,.., * .,,,.. "'""* J/PJ/
wtl ,.. *
/,J.,"f'.J-h" NP,,,11,.;:;.,
1111cr,..,.t}nfl:t;;,,...1
1* . 1 .. . . . ' !
---L__ ____ 3. 3 EMERGENCY CORE COOLING SYSTEM -----.:=:
Basis (continued) demonstrate that the maximum fuel clad temperatures that could occur over the break size spectrum are well below the melting . temperature of zirconium (3300°F).
Malfunction of the Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injection feature of the ECCS; therefore, it is disabled in the 'open' mode (by isolating the air supply) during plant operation.
This action assures that it will not block flow during Safety Injection.
The inadvertent of any one of the Safety Injection bottle isolation valves 'in conjunction with a LOCA has not
been analyzed.
To provide assurance that this will not occur, these valves are electrically locked open by a.key switch in the control room. In addition, prior to critical the valves are checked open, and then the 480 volt breakers are opened. Thus, a failure of a breaker and a switch are required for any of the valves to. close.
Insuring bothHPSI pumps are inoperable when the PCS temperature I < 230°F or the shutdow cooling isolation valves are open I eliminates PCS mass additions due to inadvertent HPSI pump start.s. I . Both HPSI pumps starting in conjunction with a charging/letdown I imbalance may cause lOCFRSO Appendix G limits to be exceeded I when the PCS temperature is< 230°F. *When.the PCS temperature I is > 430° F, the pressurizer safety valves ensure that the PCS I pressure wi 11 not exceed lOCFRSO Appendix c, 1 hh , 1 11 Ollt or hpth "RV pumps us tut et The requirement to have both HPSI. trains ,*;erab.le ,i\bove 3-;:;°F provides added assurance that the effects of a LOCA occuring .. under LTOP conditions would be mitigated.
If a LOCA occurs when the primary system temperature is less.than or equal to 325°F, *the pressure would drop to the level where low pressure safety injection can prevent core damage.
HPSI pump testing with the HI>Sl pump manual discharge
_valve closed is permitted since the closed valve eliminates the possibility of pump testing being the cause of a. mass addition -to .. the_ I>CS. References
-. (1) FSAR, Section 9.10.3;. (2) FSAR, Section 6.1, it, U, tu, U1, TSP0889-0101-NL04 I I I

*"-* **

4.0 4.0.l 4.0.2 Surveilla.nce requirements shall be. applicable
__ reactor operating conditions associated with individual Limiting Conditions for Operation unless otherwise stated in an individual surveillance requirement.
Unless otherwise specified, each surveillance requirement shall be performed within* the specified time interval with: a. A maximum allowable extension not to exceed 25% of the surveillance intervai.
and b. A total maximum combined interval tima for any three consecutive surveillance intervals not to exceed
3.25 times the specified interval.
4.1 INSTRTJM!NTATION AND CONTROL 4.1.1 Applicability Applies to th* reactor protective system and other critical instrumentation and controls.
Objective . To specify cha minimwa frequency and type of surveillance to . be applied to critical plant instrumentation and .controls.
Specification*
Calibration, and checking of instrument rea<*or protec.tiv*
(.7stem and ingir.r .. *r*d saf egua.:,\a Ji:'YStem logic channels and miscellaneoua a.lid control* shall* be performed u specified in 4.1.1 and in Tables 4.1.1 to 4.1.3. Overpressure Protection Sz*t.ema !ach POIV shall b* demonstrated operable by: 1. Perforamac*
of a chmm*l functional teat on th* POlV actuacio11 chamlel, buc ucludin1 valv* op*ration, within 31 day8 prior to e11tertn1 a condition in vbich th* PORV i* required operable and .at leuc_ouca per 31 daya thereafter vb*? th*.PORV is required operabl*.
2. Perform.ance of a channel calibration on th* actuatio11 channel at leaat one* p*r 18 mouth ... 3. Verify1n1 th* POl.V iliolationvalve is opa11 ac leaac once per 72 hour* when th* POl.V i* being uaed for protection.
4. Teatinl in accordance vith th* in**rvica in*p*ction requirement*
for ASMI Section II, Section IWV Catesory C valves * **." Amadunc No Jt,.51 Septemb*r
10. :979 . f
' .+--__:__
..

.. ---l ---------b. The PCS vent ( s) shall b-; -veritred--co--b-e-open--at-per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the vent ( s) is being used for protection except when the "'ent pathway 19 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 .piJ IC valves are used as an alternative to venting the PCS, then verify both PORV valves and both PORV block valves are open at least once per 7 days. Basis -Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of th* functioning of an instrument or systea. Furthermore.

such failures are, in many cases. revealed by alarm or annunciator action and a check supplements this type of built-in surveillance.
Based on experience in operation of both conventional and : ' nuclear plant systems. when the plant i* in operation.-*
checking frequency of once-per-shift is deemed adequate for reactor and steam system instrumentation.
Calibration*
are performed to _insure th*
and acqui*itio11 of accurate The power range saf *ty . channela and AT pover channel a are. are calibrated daily again*t a heat balance standard to account for errors _induced by rod patterns and core_ phy*ics parameters.
Other channel* are subject only to th*'"drift" errora :induced withill th* instrumentation itself caasequ*ntly, can
tolerate longer* interval*
betvee11 calibratio11.
Proce** *Y*t*
in*trwuntatio11 error* induced by drift can b.* upected to remai11 vithill acceptable tolerance*
if recalibratioa i* performed at **ell refuel1D1 sbutdOVll interval.
Substantial calibratio11 ahifta vithill a chaiinel (****ntially a channel failure) vill be rn*al*d durina routine checking ancl_ .te*tilla proceclur***
Tlnaa, llinimwa calibratioa frequenci**
of one-per-da7 tor th* power ra1* *af*ty channel*.
and.one* each ret,..itna
butclova for th* proce** ayac .. channel*, are conaidered adequate.
The ailliaum t**tinl frequency for tho** iD8trume11t channel* coft1lectecl to th* reactor protective
Y*t.. 18 b .. ed on a11 . estimated average un*af* failure rate of 1.14 x 10-S failure/hour per ch&1U1el.
Thi* e*tiaation i* b .. ect Oil limitecl operatinl experience at conventional ad nuclear pluta. AD "wafa failure" i* def 1Decl u one vbicb negate* channel operability ancl vbich, due to it* nature, ia revealed only vbea tb* channel t* t**tecl or attempt* to re.,ond to a bonaf id* *ignal. 4-2 ..--A**A Al\A"' Aaaa*11t lfo. IJ, JI, 117, 118 Nov-.ber 15, 1981 '* ' **
,, -,....

._,..,.

__ _ ,, 4:6 .
4.6.1.
4.6.2 sA:Fnr rNJECT-(ON-AND
__ coNTAINMEMT SPRAV s'tL .........
--*------.--Applicability

Applies to the safety injection system, the containment spray system, chemical injection system and the containment cooling sy1te1D teats. Ob iecti ve To verify that the subject systems will respond promptly and perform their intended functiona, if required.

Speci fication1 Safety Iniection Syctem a. System te1t1 shall be perforlll9d at each reactor refuelina interval.
A test safety injection siga&l will be applied. to initiate operation of the sy1um. The safety injection and shutdown coolin1 system pump motor1 may be de-energized*
for this test. The sy1tea vill be considered iatisfactory if control board indication and visual ob1ervation1 indicate that all coaponenu have received the safety i.njection
signal in the proper sequence and ti1Din1 (ie, the appropriate pump breaker* shall have opened and closed, and all valve* shall have completed their_travel).
b. Both bi1b pre11ure safety injection pump1, P-66A !!!!& P-668 shall be demon1trated inoperable at least once per 12 houri whenever the t*perature of one or more of the PCS.cold le11 i1 < tee*r ual111 tbe reactor bead ia removed. . . --.So . Containment S ra S stem a. Sy1t* tHt ball be performed at eaCh reactor refuelin1 interval.
te1t 1ball be performed vitb the i1olation
-valve* in
1pra1 1uppl1 linH at the contaimaDC bl0ckad clo1ecl. peracioa of the 171t* i1 initiated by trippina the no
accuacioa inatruileatadon.
b. Ac lu e.,.ry flv* year1 the 1pra1 aoulea 1hall be
veri_f
W co be opea. * . c. 'Ill te1t will be coa1ider8d 1ati1f actory if vi1ual . _ .. _o ervatiou indicaca al 1 componentl bav*
operated tilfactorily.
--, .. I I I I I TSP1187-0218*YL04 o>-y /,,;tJo,_.
¢.<n> /,,;, /1 olv.u i'lllJ
3tJ it' -d /11/D-.301, .QA.C..
4-39 Allendaent No. si, 11, *** 111 November 14, 1981 ., '
...... . ** ,,
{J .. -----------*--------------4.6 SAFETY [NJECTrON AMO CONTArNHENT SPRAY -rr-----c--

. -4.6.l 4.6.4 4.6.5 Pumps a. 11\e safety injection pumps, shutdown cooling pumps, and containment spray pumps shall be started at intervals not to ezceed three months. Alternate manual starting between control room console and the Local breaker shall be practiced in the test program. b. Acce_ptable levels of performance shall be that the pumps start, reach their rated. shutoff head1 at miniaaim recirculatiqa flov, and operate for at least fifteen minutes. Valves Deleted .Containment Air Coolins Srste* a. b. Faergency mode automatic valve and faa operation will be checked for operability durin1 each refuelia1 sbutdova *. Each faa and valve required to function durin1 accident condition1 will be eaerciled at interval*

not to.eaceed three montb1. !!!!!. *. The safety injection 1yst* and the containment spray 1y1te* are priacip,:
plant.safety features that are r;t;1rmally . duria1 reactor operation.
*
Complete syn .. te1C1 cannot be perf ormd wha the reactor ia *
operatia1 beca11M a aafecy injection aipal cauH1 contaia.mt i1olacion aa4 a containment spray-1y1t
.. teat req11ire1 the 1y1te* to be te11porarily di1abled.
Tile .. cbod of* a11urin1 operability of th*** 1yit .. 1 i1 therefore to combine 1y1t** tHtl to be puformecl d11rin1 annual plant 1b\atdovn1, vi.th more . f requeat compoeat CHU, which cm be performed durin1. reactor operation.
Tbe.a .... ual 1y1c .. 1 te1t1 demoa1trace proper autOll&tic operacion of the:. safety injection and contaiamenc spray 1y1u... A ten ti1aal ii applied to initiate autOll&Cic aci::ion ancl. verification_
made that the camponenu receive the safety injection in the proper Hquace* Tile use demon1tratH the operation of the valve1, pump circuit breakers, and auco .. tic circuitry*
(1, 2) 4-40 Amendment Mo. St, 1J, 11, 117
t!ov .. ber 14, 1981 TSP1181-0211-Nl.04
'* -* ... '*. *
,------------------*
--*-**--*--*
,;. .. e * . ,,
SPRAY SYSTEMS TESTS (Continued) 4.6 SAFETY ----;-\,)
____ -----.c..
__ --Ba s-i:s--
____________
_ I Cl '. . *
During reactor operation, the instrumentation on to init_iate safety injection and containment spray is generally checked daily and the initiating circuits are tested * ** inonthly.
In addition, the active components (pumps and valves) are to be tested every three months to check the operation of the starting circuits and to verify that the pumps are in . . satisfactory running order. The test interval of three months is based on the judgment that more frequent testing would not increase the reliability (ie, the probability that the component would operate when required), yet more frequent test would resul_t in increased wear over a long period of time. Verifical:ion that the spray piping and nozzles are open will be made initially by a smoke test or other suitably sensitive method, and at least every five years thereafter.
Since the material is all stainless steel, normally in a dry condition, and with no plussing mechanism available", the retest every five years is considered to be more. than adequate.
Other aystema that are also important to the emergency cooling function are the SI tanks, the component cooling system, the set'vice water* system and the containment.
ait' coolui. The SI tanks are a pauive ufety featut'e *. In accordance with the specification1, the watet' volume-and pt'e11ut'e in the SI tanlu are checked pet'iodically.
The otbet' 1y1t*1 mentioned opente when the reactOt' ii in operation and by these llUnl an monitored fot' satisfactory pet'formance
csld lea tE&patatata is le11 shin 100*1 , th9r ttur of nee llHI *** nuU aauu the kppiiidli C IlSlti cu u* &c b* 1*11ededt thitifbti, bUtb pdSpl ifi fihdifid tnup111bleu=
Ref Ct'ence1,.
(1) FSA&, Section 6.1.J. * (2) . FSA&, section 6.2.3. TSP1187-0218-NL04 4-41 AIDllnd*nt No. .117 lov*ber 14, 1988 .:. ' I I I **,,
,,. ( .. .. 4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS.TESTS (Continued)
Basis * (continued)

*

During reactor operation, the--instrumentafron -wnich*-f-s-depended-----:-------------

0 on to initiate safety injection and containment spray is
generally checked daily and the initiating circuits are tested monthly. In addition, the active components (pumps and valves) are to be tested every three months to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order.* The test interval of. three months is based on the judgment 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 of time *. Verification that the spray piping and nozzles are
open will be made. initially by a smoke test or other suitably 'sensitive.
method. and at least every five years -thereafter.
' 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.
Other systems that are also important to the emergency cooling
function are the SI tanks 1 .the component cooling system, the service water system and the containment air coolers. The SI tanks are a passive safety feature. In accordance with the specifications, the water volume and pressure in the SI tanks are checked periodically.
The other systems mentioned operate when the reactor is in operation and by these means.are continuously monitored for s-atisfactory performance.
With the reactor vessel head installed when .. the PCS cold leg temperature is less.than 230°F 1 or if the shutdown cooling system isolation valves M0-3015 and M0-3016 are open, the start* of one HPSI pump could cause the Appendix G 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. TSP0889-0101-MD01-NL04 Amendment No. 117. I I I I I I

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